The MAX facility for CFD code validation
ANL has recently completed construction of a fluid dynamics test facility devised to provide validation data for CFD simulation tools used to evaluate various aspects of nuclear power plant design and safety. Experiments with the facility involve mixing air jets within a 1x1x1.7m long glass tank at atmospheric pressure. A particle image velocimetry system measures flow velocity and turbulence quantities within the tank while a high-speed infrared camera records temperatures across the tank lid. The tandem of high fidelity thermal and turbulence data is particularly useful for benchmarking transient heat transfer phenomena such as thermal striping. This paper describes the MAX facility, preliminary data obtained during shakedown tests, and the results of companion CFD calculations employing RANS-based Star-CCM+ and large eddy simulations with Nek 5000. (authors)
CFD code validation utilizing the OECD/NRC BFBT benchmark
As part of the OECD/NRC Full Size Fine Mesh Bundle Test (BFBT) Benchmark CFD and sub-channel codes are used for the prediction of void distribution and pressure drop within a BWR type fuel rod bundle. The Pennsylvania State University (PSU) participates with the commercial CFD code FLUENT 6.3 (in a co-operation with Oak Ridge National Laboratory) and the sub-channel code COBRA-TF as a thermal-hydraulic module of the coupled code system CTF/NEM. Due to the limited two-phase flow applicability of FLUENT 6.3, the current investigation is focused on the simulation of single-phase heat transfer and pressure drop for the purpose of code-to-code and code-to-data comparisons. Special exercises - Exercise 1 of Phase I and Exercise 0 of Phase II of the BFBT benchmark are utilized for an assessment of FLUENT 6.3 modeling capabilities. A special type of a BWR fuel assembly design, the so-called high burn-up assembly, was used in the measurements. The range of operating fluid conditions were as follows: pressure of 7.2 MPa; inlet temperature of 285degC; and flow rate of 20 to 70 t/h. This paper presents results obtained with FLUENT 6.3 for a low quality test, which was selected for an assessment of the single-phase heat transfer models of the code. In the test data provided there is no axial temperature distribution available. Therefore, the axial temperature distribution as predicted by the sub-channel code is used for comparison. Many steps are involved in the generation of the current CFD model. The first step is the generation of the flow domain in terms of a solid model that represents the fluid region. It defines the component of interest that represents full scale fuel rods and spacer grids in a fuel assembly. The next step consists of applying a mesh to the flow domain using commercial software GAMBIT. The third step takes place within the user's options of the commercial CFD solver FLUENT 6.3. Segregated solver is chosen as a numerical method. Methods employed to
3D CFD CONV code: validation and verification
During some years in IBRAE a set of 3D CFD modules (CONV code) for safety analysis of the operated Nuclear Power Plants (NPPs) is developing. These modules are based on the developed algorithms with small scheme diffusion, for which the discrete approximations are constructed with use of finite-volume methods and fully staggered grids. For solving of convection problem the regularized nonlinear monotonic operator-splitting scheme is developed. The Richardson iterative method with Chebyshev's set of parameters using FFT solver for Laplace's operator as pre-conditioner is applied for solving pressure equation. Such approach for solving of the elliptical equations with variable coefficients gives multiple acceleration in a comparison with a usual method of conjugate gradients. For modeling of 3D turbulent single-phase flows LES approach (commutative filters) is used. The CONV code is fully parallelized and highly effective at the high performance computers. The developed modules were validated on a series of the well known tests in a wide range of Rayleigh numbers from a range 106-1016 and Reynolds numbers from a range 103-105. The developed software has been applied to the simulation of the experiment on RASPLAV facility and of large-scale RCW test conducted in the frames of MASCA Project. As a result of numerical modeling of aforementioned experiments qualitative and quantitative agreement with experimental data was obtained including amount of the molten corium and form of the molten pool, distribution of temperature in corium, fluxes and temperatures in a test-wall. The software has been applied also to the analysis results of test L1 and joint analyses on transient molten pool thermal hydraulics in the LIVE facility in the framework of ISTC project. In this paper the examples of use of the developed software for modeling of a fuel assembly, namely, for research of a hydraulic resistance factor of a spacer are demonstrated. The calculations are carried out on a
An approach to validation of coupled CFD and system thermal-hydraulics codes
This paper discusses the development of approach and experimental facility for the validation of coupled Computational Fluid Dynamics (CFD) and System Thermal Hydraulics (STH) codes. The validation of a coupled code requires experiments which feature two way feedback between the component (CFD sub-domain) and the system (STH sub-domain). We present results of CFD analysis that are used in the development of a flexible design for the TALL-3D experimental facility. The facility consists of a lead-bismuth thermal-hydraulic loop operating in forced and natural circulation regimes with a heated pool-type 3D test section. The goal of the design is to achieve a feedback between mixing and stratification phenomena in the 3D tests section and forced / natural circulation flow conditions in the loop. Finally, we discuss the development of an experimental validation matrix for validation of coupled STH and CFD codes that considers the key physical phenomena of interest. (author)
Needs and opportunities for CFD-code validation
Smith, B.L. [Paul Scherrer Institute, Villigen (Switzerland)]|[Paul Scherrer Instiute, Wuerenlingen (Switzerland)
1996-06-01
The conceptual design for the ESS target consists of a horizontal cylinder containing a liquid metal - mercury is considered in the present study - which circulates by forced convection and carries away the waste heat generated by the spallation reactions. The protons enter the target via a beam window, which must withstand the thermal, mechanical and radiation loads to which it is subjected. For a beam power of 5MW, it is estimated that about 3.3MW of waste heat would be deposited in the target material and associated structures. it is intended to confirm, by detailed thermal-hydraulics calculations, that a convective flow of the liquid metal target material can effectively remove the waste heat. The present series of Computational Fluid Dynamics (CFD) calculations has indicated that a single-inlet Target design leads to excessive local overheating, but a multiple-inlet design, is coolable. With this option, inlet flow streams, two from the sides and one from below, merge over the target window, cooling the window itself in crossflow and carrying away the heat generated volumetrically in the mercury with a strong axial flow down the exit channel. The three intersecting streams form a complex, three-dimensional, swirling flow field in which critical heat transfer processes are taking place. In order to produce trustworthy code simulations, it is necessary that the mesh resolution is adequate for the thermal-hydraulic conditions encountered and that the physical models used by the code are appropriate to the fluid dynamic environment. The former relies on considerable user experience in the application of the code, and the latter assurance is best gained in the context of controlled benchmark activities where measured data are available. Such activities will serve to quantify the accuracy of given models and to identify potential problem area for the numerical simulation which may not be obvious from global heat and mass balance considerations.
Validation of CFD commercial codes for large diameter jet impingement flow
This paper presents a validation project on CFD code for large diameter jet (D=0.254m) impingement flow. CFD-ACE commercial software was applied in this study. The simulation results are compared with reference experimental data and evaluated in views of Stirling engine design and development. Before apply the CFD code to this study, two simulations were performed for code validation. Simulation of laminar jet impingement flow and heat transfer was performed by comparing with result of Victor and turbulent flow model simulation was performed to compare with Fitzgerald experimental results. CFD-ACE code shows very good match with the reference data. The simulations of large diameter jet impingement were performed to compare with the experimental results from Terry Simon (University of Minnesota). Two different Reynolds numbers for unidirectional flow (7600 and 17700) and two different space ratio (0.25 and 0.5) were simulated. Also oscillatory flow of same model was studied in the view of Stirling engine model. Two different oscillatory frequencies were tested and simulated for two different space ratios. The comparisons between the simulation and experimental results show good match at some crank angle of oscillatory flow. (author)
Validations of CFD Code for Density-Gradient Driven Air Ingress Stratified Flow
Air ingress into a very high temperature gas-cooled reactor (VHTR) is an important phenomena to consider because the air oxidizes the reactor core and lower plenum where the graphite structure supports the core region in the gas turbine modular helium reactor (GTMHR) design, thus jeopardizing the reactor's safety. Validating the computational fluid dynamics (CFD) code used to analyze the air ingress phenomena is therefore an essential part of the safety analysis and the ultimate computation required for licensing.
CFD code validation against stratified air-water flow experimental data
Pressurized Thermal Shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the Reactor Pressure Vessel (RPV) lifetime is the cold water Emergency Core Cooling (ECC) injection into the cold leg during a Loss of Coolant Accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the NURESIM (European Platform for Nuclear Reactor Simulations) Integrated Project as a reference two-phase problem for CFD code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mecanique des Fluides de Toulouse in 1985 [1], which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code NEPTUNECFD (developed by EDF and CEA). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against the available experimental data, and to perform code to code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag coefficient modelling. A relevant improvement of results has been achieved with 3D simulations, even if the air velocity profile was still significantly underestimated. (author)
The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes
Highlights: • Design of a heavy liquid thermal-hydraulic loop for CFD/STH code validation. • Description of the loop instrumentation and assessment of measurement error. • Experimental data from forced to natural circulation transient. - Abstract: Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper
PIV Uncertainty Methodologies for CFD Code Validation at the MIR Facility
Piyush Sabharwall; Richard Skifton; Carl Stoots; Eung Soo Kim; Thomas Conder
2013-12-01
Currently, computational fluid dynamics (CFD) is widely used in the nuclear thermal hydraulics field for design and safety analyses. To validate CFD codes, high quality multi dimensional flow field data are essential. The Matched Index of Refraction (MIR) Flow Facility at Idaho National Laboratory has a unique capability to contribute to the development of validated CFD codes through the use of Particle Image Velocimetry (PIV). The significance of the MIR facility is that it permits non intrusive velocity measurement techniques, such as PIV, through complex models without requiring probes and other instrumentation that disturb the flow. At the heart of any PIV calculation is the cross-correlation, which is used to estimate the displacement of particles in some small part of the image over the time span between two images. This image displacement is indicated by the location of the largest peak. In the MIR facility, uncertainty quantification is a challenging task due to the use of optical measurement techniques. Currently, this study is developing a reliable method to analyze uncertainty and sensitivity of the measured data and develop a computer code to automatically analyze the uncertainty/sensitivity of the measured data. The main objective of this study is to develop a well established uncertainty quantification method for the MIR Flow Facility, which consists of many complicated uncertainty factors. In this study, the uncertainty sources are resolved in depth by categorizing them into uncertainties from the MIR flow loop and PIV system (including particle motion, image distortion, and data processing). Then, each uncertainty source is mathematically modeled or adequately defined. Finally, this study will provide a method and procedure to quantify the experimental uncertainty in the MIR Flow Facility with sample test results.
A CFD validation methodology for containment code calculations of hydrogen mixing and recombination
In the frame of ANSALDO activities on containment hydrogen accident events, a simulation procedure was developed to qualify and verify the calculations performed by simplified containment computer codes through the use of a full 3-D Navier Stokes solver. The methodology aims to reduce the computational time usually associated with a general purpose CFD code complete simulation of the containment transient, limiting on the other hand the loss of accuracy typical of the use of a simplified Containment Code. This goal has been fulfilled by the development of a calculation procedure organised in several different steps able to verify the calculated transient parameters by GOTHIC3.4 (the simplified code) with specific calculations performed with CFX4.2 (the CFD code). The paper describes the main milestones of the methodology development and summarizes main results, findings, as well the possible direction of use of the performed work. (authors)
A selection of experimental test cases for the validation of CFD codes, volume 1
1994-08-01
This report presents the results of a study by Working Group 14 of the AGARD Fluid Dynamics Panel. This group was formed to establish an accessible, detailed experimental data base for the validation of Computational Fluid Dynamics (CFD) codes. The thirty nine test cases that are documented cover the subsonic, transonic, and supersonic flow regimes and five classes of geometries. Included in the five classes of geometries are: two dimensional airfoils; three dimensional wings, designed for predominantly attached flow conditions; slender bodies, typical of missile type configurations; delta wings, characterized by a conical type of vortex flow; and complex configurations, either in a geometrical sense or because of complicated flow interactions. The report is presented in two volumes. Volume 1 provides a review of the theoretical and experimental requirements, a general introduction and summary of the test cases, and recommendations for the future. Volume 2 contains detailed information on the test cases. The relevant data of all test cases has been compiled on floppy disks, which can be obtained through National Centers.
2-D Circulation Control Airfoil Benchmark Experiments Intended for CFD Code Validation
Englar, Robert J.; Jones, Gregory S.; Allan, Brian G.; Lin, Johb C.
2009-01-01
A current NASA Research Announcement (NRA) project being conducted by Georgia Tech Research Institute (GTRI) personnel and NASA collaborators includes the development of Circulation Control (CC) blown airfoils to improve subsonic aircraft high-lift and cruise performance. The emphasis of this program is the development of CC active flow control concepts for both high-lift augmentation, drag control, and cruise efficiency. A collaboration in this project includes work by NASA research engineers, whereas CFD validation and flow physics experimental research are part of NASA s systematic approach to developing design and optimization tools for CC applications to fixed-wing aircraft. The design space for CESTOL type aircraft is focusing on geometries that depend on advanced flow control technologies that include Circulation Control aerodynamics. The ability to consistently predict advanced aircraft performance requires improvements in design tools to include these advanced concepts. Validation of these tools will be based on experimental methods applied to complex flows that go beyond conventional aircraft modeling techniques. This paper focuses on recent/ongoing benchmark high-lift experiments and CFD efforts intended to provide 2-D CFD validation data sets related to NASA s Cruise Efficient Short Take Off and Landing (CESTOL) study. Both the experimental data and related CFD predictions are discussed.
During a severe accident in a Pressurized Water Reactor (PWR), the formation of a combustible gas mixture in the complex geometry of the reactor depends on the understanding of hydrogen production, the complex 3D thermal-hydraulics flow due to gas/steam injection, natural convection, heat transfer by condensation on walls and effect of mitigation devices. Numerical simulation of such flows may be performed either by Lumped Parameter (LP) or by Computational Fluid Dynamics (CFD) codes. Advantages and drawbacks of LP and CFD codes are well-known. LP codes are mainly developed for full size containment analysis but they need improvements, especially since they are not able to accurately predict the local gas mixing within the containment. CFD codes require a process of validation on well-instrumented experimental data before they can be used with a high degree of confidence. The MISTRA coupled effect test facility has been built at CEA to fulfil this validation objective: with numerous measurement points in the gaseous volume - temperature, gas concentration, velocity and turbulence - and with well controlled boundary conditions. As illustration of both experimental and simulation areas of this topic, a recent example in the use of MISTRA test data is presented for the case of the International Standard Problem ISP47. The proposed experimental work in the MISTRA facility provides essential data to fill the gaps in the modelling/validation of computational tools. (author)
related to nuclear reactor safety issues. The conference consisted of 14 technical sessions. Among the topics included were containment, advanced reactors, multiphase flows, flow in a rod bundle, fire analysis, flows in dry casks, thermal analysis, mixing flows and pressurized thermal shock (PTS). About 1/3 of the papers were concerned with two-phase flow issues and the rest were devoted to single-phase CFD validation. South Korea is a candidate to host a follow-up meeting scheduled in 2012, organized by KAERI. KAERI also volunteered to sponsor and organize the second OECD/NEA CFD benchmark exercise. In the closure meeting after the panel session discussion, the representative from the Paul Scherrer Institut (PSI) proposed to host a future workshop scheduled for 2014, and to organize and sponsor the third OECD/NEA benchmark exercise based on a stratification experiment in the PANDA facility at PSI. The great majority of participants were interested in attending a follow-up workshop within two years. Comments were made during the panel session on the content of CFD4NRS-3. Two of the comments are that experiments can provide insight into the physics, and that CFD is now an accepted analysis tool, though it is very important to follow BPGs. There was a consensus on the need to maintain the high quality of the papers. The promotion of international benchmarking exercises for CFD was strongly encouraged. Another comment suggested that such workshops should be a forum to discuss novel approaches, but that one must also keep in mind that the end users are people from the nuclear safety community. The CFD4NRS, XCFD4NRS and CFD4NRS-3 workshops have proved to be very valuable means to assess the status of CFD code capabilities and validation, to exchange experiences in CFD code applications, and to monitor future progress
Validation of CFD Codes for Parawing Geometries in Subsonic to Supersonic Flows
Cruz-Ayoroa, Juan G.; Garcia, Joseph A.; Melton, John E.
2014-01-01
Computational Fluid Dynamic studies of a rigid parawing at Mach numbers from 0.8 to 4.65 were carried out using three established inviscid, viscous and independent panel method codes. Pressure distributions along four chordwise sections of the wing were compared to experimental wind tunnel data gathered from NASA technical reports. Results show good prediction of the overall trends and magnitudes of the pressure distributions for the inviscid and viscous solvers. Pressure results for the panel method code diverge from test data at large angles of attack due to shock interaction phenomena. Trends in the flow behavior and their effect on the integrated force and moments on this type of wing are examined in detail using the inviscid CFD code results.
Modeling and validation of CFD code KIRAN3D for electron beam melting of zirconium
The validation of the computer code KIRAN3D is carried out with the physical experiments carried out using electron beam melting of zirconium ingot in cold hearth. The measured maximum surface temperature shows good agreement with the predicted results by computational analysis, when the Gaussian beam profile is used. (author)
Development, validation and application of NAFA 2D-CFD code
A 2D axi-symmetric code named NAFA (Version 1.0) is developed for studying the pipe flow under various conditions. It can handle laminar/ turbulent flows, with or without heat transfer, under sub-critical/super-critical conditions. The code solves for momentum, energy equations with standard k-ε turbulence model (with standard wall functions). It solves pipe flow subjected to 'velocity inlet', 'wall', 'axis' and 'pressure outlet' boundary conditions. It is validated for several cases by comparing its results with experimental data/analytical solutions/correlations. The code has excellent convergence characteristics as verified from fall of equation residual in each case. It has proven capability of generating mesh independent results for laminar as well as turbulent flows. The code is applied to supercritical flows. For supercritical flows, the effect of mesh size on prediction of heat transfer coefficient is studied. With grid refinement, the Y+ reduces and reaches the limiting value of 11.63. Hence the accuracy is found to increase with grid refinement. NAFA is able to qualitatively predict the effect of heat flux and operating pressure on heat transfer coefficient. The heat transfer coefficient matches well with experimental values under various conditions. (author)
A verification and validation of the new implementation of subcooled flow boiling in a CFD code
Braz Filho, Francisco A.; Ribeiro, Guilherme B.; Caldeira, Alexandre D., E-mail: fbraz@ieav.cta.br, E-mail: gbribeiro@ieav.cta.br, E-mail: alexdc@ieav.cta.br [Instituto de Estudos Avancados (IEAv), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear
2015-07-01
Subcooled flow boiling in a heated channel occurs when the liquid bulk temperature is lower than the saturation temperature and the wall temperature is higher. FLUENT computational fluid dynamics code uses Eulerian Multiphase Model to analyze this phenomenon. In FLUENT previous versions, the heat transfer correlations and the source terms of the conservation equations were added to the model using User Defined Functions (UDFs). Currently, these models are among the options of the FLUENT without the need to use UDFs. The comparison of the FLUENT calculations with experimental data for the void fraction presented a wide range of variation in the results, with values from satisfactory to poor results. There was the same problem in the previous versions. The fit factors of the FLUENT that control condensation and boiling in the system can be used to improve the results. This study showed a strong need for verification and validation of these calculations, along with a sensitivity analysis of the main parameters. (author)
A verification and validation of the new implementation of subcooled flow boiling in a CFD code
Subcooled flow boiling in a heated channel occurs when the liquid bulk temperature is lower than the saturation temperature and the wall temperature is higher. FLUENT computational fluid dynamics code uses Eulerian Multiphase Model to analyze this phenomenon. In FLUENT previous versions, the heat transfer correlations and the source terms of the conservation equations were added to the model using User Defined Functions (UDFs). Currently, these models are among the options of the FLUENT without the need to use UDFs. The comparison of the FLUENT calculations with experimental data for the void fraction presented a wide range of variation in the results, with values from satisfactory to poor results. There was the same problem in the previous versions. The fit factors of the FLUENT that control condensation and boiling in the system can be used to improve the results. This study showed a strong need for verification and validation of these calculations, along with a sensitivity analysis of the main parameters. (author)
Pena-Monferrer, C.; Miquel veyrat, A.; Munoz-Cobo, J. L.; Chiva Vicent, S.
2016-08-01
In the recent years, due, among others, the slowing down of the nuclear industry, investment in the development and validation of CFD codes, applied specifically to the problems of the nuclear industry has been seriously hampered. Thus the International Benchmark Exercise (IBE) sponsored by the OECD/NEA have been fundamental to analyze the use of CFD codes in the nuclear industry, because although these codes are mature in many fields, still exist doubts about them in critical aspects of thermohydraulic calculations, even in single-phase scenarios. The Polytechnic University of Valencia (UPV) and the Universitat Jaume I (UJI), sponsored by the Nuclear Safety Council (CSN), have actively participated in all benchmark's proposed by NEA, as in the expert meetings,. In this paper, a summary of participation in the various IBE will be held, describing the benchmark itself, the CFD model created for it, and the main conclusions. (Author)
Validation of the CFD code ANSYS CFX was performed in the frame of the experimental and analytical investigations of mixing of coolant flows with different saline concentration fulfilled in EDO 'GIDROPRESS' at 4-loop test facility modeling reactor WWER-1000 with the scale 1:5. Calculations were performed with the 3-D code complex ANSYS CFX. The objective of the analyses was the code validation with the purpose of further practical implementation. Calculations were performed for the case of experiments with the saline slug in the pump loop seal at the RCP start-up. Unsteady 3-D fields of saline concentration were calculated for the circuit of the facility. Comparison of experimental and predicted data is presented in the paper. Results of CFD analyses demonstrated very good agreement with the experimental data. (authors)
Validation of vortex code viscous models using lidar wake measurements and CFD
Branlard, Emmanuel; Machefaux, Ewan; Gaunaa, Mac;
2014-01-01
The newly implemented vortex code Omnivor coupled to the aero-servo-elastic tool hawc2 is described in this paper. Vortex wake improvements by the implementation of viscous effects are considered. Different viscous models are implemented and compared with each other. Turbulent flow fields with sh...... viscous boundaries appear more important than the modelling of viscosity in the wake. External turbulence and shear appear sufficient but their full potential flow modelling would be preferred....
Validation of the CFD code fluent by post-test calculation of a density-driven ROCOM experiment
During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems. During the course of follower core assessments, TUV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core inlet has to be determined. TUV NORD SysTec applies the CFD code FLUENT for the investigation of boron dilution events in pressurized water reactors. To affirm the FLUENT abilities for the simulation of boron dilution events, a validation against the ROCOM experiment T665521 with a density-driven coolant mixing was performed. This validation proves that FLUENT is able to appropriately simulate the effects of boron transport and dilution such as streaks of coolant with lower density in the downcomer. Deficits were identified in the simulation of fluid layering in the cold leg, which fortunately have a rather small influence on the predicted core inlet concentration. Therefore, the boron concentration in the reactor core can be determined with sufficient accuracy to solve the safety issue, regardless of the core becoming critical or not
Duduković Milorad P.
2002-01-01
Full Text Available This manuscript, based on the presentation given by one of the authors (M.P. Dudukovic at the Technological and Engineering Forum in Pančevo, May 21 2002, summarizes the use of the computer automated radioactive particle tracking (CARPT and gamma computed tomography (CT in obtaining the data needed to validate the Euler-Euler based CFD simulations for solids distribution, flow pattern and mixing in a liquid-solid riser. The riser is one of the reactors considered for acid solid catalyst promoted alkylation. It is shown that CFD calculations, validated by CARPT-CT data, show promise for scale-up and design of this novel reactor type.
Formulation, Implementation and Validation of a Two-Fluid model in a Fuel Cell CFD Code
Kunal Jain, Vernon Cole, Sanjiv Kumar and N. Vaidya
2008-11-01
more complications. A general approach would be to form a mixture continuity equation by linearly combining the phasic continuity equations using appropriate weighting factors. Analogous to mixture equation for pressure correction, a difference equation is used for the volume/phase fraction by taking the difference between the phasic continuity equations. The relative advantages of the above mentioned algorithmic variants for computing pressure correction and volume fractions are discussed and quantitatively assessed. Preliminary model validation is done for each component of the fuel cell. The two-phase transport in the channel is validated using empirical correlations. Transport in the GDL is validated against results obtained from LBM and VOF simulation techniques. The Channel-GDL interface transport will be validated against experiment and empirical correlation of droplet detachment at the interface. References [1] Y. Wang S. Basu and C.Y. Wang. Modeling two-phase flow in pem fuel cell channels. J. Power Sources, 179:603{617, 2008. [2] P. K. Sinha and C. Y. Wang. Liquid water transport in a mixed-wet gas diffusion layer of a polymer electrolyte fuel cell. Chem. Eng. Sci., 63:1081-1091, 2008. [3] Guangyu Lin and Trung Van Nguyen. A two-dimensional two-phase model of a pem fuel cell. J. Electrochem. Soc., 153(2):A372{A382, 2006. [4] T. Berning and N. Djilali. A 3d, multiphase, multicomponent model of the cathode and anode of a pem fuel cell. J. Electrochem. Soc., 150(12):A1589{A1598, 2003.
The engineering design, performance analysis and safety assessment of Generation IV heavy liquid metal cooled nuclear reactors calls for advanced and qualified numerical tools. These tools need to be qualified before used in decision making process. Computational Fluid Dynamics (CFD) codes provide detailed means for thermal-hydraulics analysis of pool-type nuclear reactors. This paper describes modeling of a forced to natural flow experiment in TALL-3D experimental facility using a commercial CFD code Star-CCM+. TALL-3D facility is 7 meters high LBE loop with two parallel hot legs and a cold leg. One of the hot legs accommodates the 3D test section, a cylindrical pool where the multi-dimensional flow conditions vary between thermal mixing and stratification depending on the mass flow rate and the power of the heater surrounding the pool. The pool outlet temperature which affects the natural convection flow rates in the system is governed by the flow structure in the pool. Therefore, in order to predict the dynamics of the TALL-3D facility it is crucial to resolve the flow inside the 3D test section. Specifically designed measurement instrumentation set-up provides steady state and transient data for calibration and validation of numerical models. The validity of the CFD model is assessed by comparing the computational results to experimental results. (author)
Highlights: • Simulation of BFBT turbine and pump transients at multiple scales. • CFD, sub-channel and system codes are used for the comparative study. • Heat transfer models are compared to identify difference between the code predictions. • All three scales predict results in good agreement to experiment. • Sub cooled boiling models are identified as field for future research. -- Abstract: The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in the validation and qualification of modern thermo hydraulic simulations tools at various scales. In the present paper, the prediction capabilities of four codes from three different scales – NEPTUNECFD as fine mesh computational fluid dynamics code, SUBCHANFLOW and COBRA-TF as sub channels codes and TRACE as system code – are assessed with respect to their two-phase flow modeling capabilities. The subject of the investigations is the well-known and widely used data base provided within the NUPEC BFBT benchmark related to BWRs. Void fraction measurements simulating a turbine and a re-circulation pump trip are provided at several axial levels of the bundle. The prediction capabilities of the codes for transient conditions with various combinations of boundary conditions are validated by comparing the code predictions with the experimental data. In addition, the physical models of the different codes are described and compared to each other in order to explain the different results and to identify areas for further improvements
Coupling CFD code with system code and neutron kinetic code
Vyskocil, Ladislav, E-mail: Ladislav.Vyskocil@ujv.cz; Macek, Jiri
2014-11-15
Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent.
Coupling CFD code with system code and neutron kinetic code
Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent
One of the RIA scenarios considered in the frame of WWER reactors safety analyses is related to boron dilution phenomenon which may cause reactivity-insertion accident. Mixture of the coolant with low boron concentration with that in the reactor allows for the mitigation of the consequences. A special experimental program was carried out in EDO 'GIDROPRESS' to investigate the mixture process of coolant flows with different boron concentration upstream the core. A saline (NaCl) solution was injected into the reactor model during the experiments. In the frame of the TACIS project R2.02/02 pre-test calculations of the velocities, pressure and salt concentration fields in EDO 'GIDROPRESS' four-loop experimental facility scaled 1:5 with regard to WWER-1000 were performed using the code CFX. The objective of this work was validation of the code ANSYS CFX in order to justify its further application for the safety analysis of the WWER type reactors. Results of CFD analyses demonstrated very good agreement with the experimental data. (authors)
Validation of NEPTUNE-CFD two-phase flow models using experimental data
Jorge Pérez Mañes; Victor Hugo Sánchez Espinoza; Sergio Chiva Vicent; Michael Böttcher; Robert Stieglitz
2014-01-01
This paper deals with the validation of the two-phase flow models of the CFD code NEPTUNEC-CFD using experimental data provided by the OECD BWR BFBT and PSBT Benchmark. Since the two-phase models of CFD codes are extensively being improved, the validation is a key step for the acceptability of such codes. The validation work is performed in the frame of the European NURISP Project and it was focused on the steady state and transient void fraction tests. The influence of different NEPTUNE-CFD ...
Containment Code Validation Matrix
and references, the synopsis also identifies the availability of the report and data, phenomena covered by the test, type of test (separate effect, combined effect or integral test), covers DBA and/or SA/BDBA conditions, range of key experimental parameters and past code validation/ benchmarks. This CCVM has identified experiments for 93% of the phenomena requiring validation. However, if only experiments suitable for CFD validation are considered, then only about half of the phenomena are covered by this CCVM. It is recommended that this work be reviewed in 5 years time to include new experiments and to attempt to close the identified experiment gaps (phenomena lacking suitable experiments for validation). (authors)
Improved interpretation and validation of CFD predictions
Popiolek, Z.; Melikov, Arsen Krikor
2004-01-01
The mean velocity in rooms predicted by CFD simulations based on RANS equations differs from the mean (in time) magnitude of the velocity, i.e. the mean speed, in rooms measured by low velocity thermal anemometers with omnidirectional sensor. This discrepancy results in incorrect thermal comfort...... assessment by the CFD predictions as well as incorrect validation of the predicted velocity field. In this paper the discrepancies are discussed and identified, and a method for estimating of the mean speed based on the CFD predictions of mean velocity and kinetic turbulence energy is suggested. The method...
In the present study, a two-dimensional computer code has been developed in FORTRAN using CFD technique, which is basically a numerical scheme. This computer code solves the Navier Stokes equations and continuity equation to find out the velocity and pressure fields within a given domain. This analysis has been done for the developed within a square cavity driven by the upper wall which has become a bench mark for testing and comparing the newly developed numerical schemes. Before to handle this task, different one-dimensional cases have been studied by CFD technique and their FORTRAN programs written. The cases studied are Couette flow, Poiseuille flow with and without using symmetric boundary condition. Finally a comparison between CFD results and analytical results has also been made. For the cavity flow the results from the developed code have been obtained for different Reynolds numbers which are finally presented in the form of velocity vectors. The comparison of the developed code results have been made with the results obtained from the share ware version of a commercially available code for Reynolds number of 10.0. The disagreement in the results quantitatively and qualitatively at some grid points of the calculation domain have been discussed and future recommendations in this regard have also been made. (author)
Perspective: Selected benchmarks from commercial CFD codes
Freitas, C.J. [Southwest Research Inst., San Antonio, TX (United States). Computational Mechanics Section
1995-06-01
This paper summarizes the results of a series of five benchmark simulations which were completed using commercial Computational Fluid Dynamics (CFD) codes. These simulations were performed by the vendors themselves, and then reported by them in ASME`s CFD Triathlon Forum and CFD Biathlon Forum. The first group of benchmarks consisted of three laminar flow problems. These were the steady, two-dimensional flow over a backward-facing step, the low Reynolds number flow around a circular cylinder, and the unsteady three-dimensional flow in a shear-driven cubical cavity. The second group of benchmarks consisted of two turbulent flow problems. These were the two-dimensional flow around a square cylinder with periodic separated flow phenomena, and the stead, three-dimensional flow in a 180-degree square bend. All simulation results were evaluated against existing experimental data nd thereby satisfied item 10 of the Journal`s policy statement for numerical accuracy. The objective of this exercise was to provide the engineering and scientific community with a common reference point for the evaluation of commercial CFD codes.
Utilizing GPUs to Accelerate Turbomachinery CFD Codes
MacCalla, Weylin; Kulkarni, Sameer
2016-01-01
GPU computing has established itself as a way to accelerate parallel codes in the high performance computing world. This work focuses on speeding up APNASA, a legacy CFD code used at NASA Glenn Research Center, while also drawing conclusions about the nature of GPU computing and the requirements to make GPGPU worthwhile on legacy codes. Rewriting and restructuring of the source code was avoided to limit the introduction of new bugs. The code was profiled and investigated for parallelization potential, then OpenACC directives were used to indicate parallel parts of the code. The use of OpenACC directives was not able to reduce the runtime of APNASA on either the NVIDIA Tesla discrete graphics card, or the AMD accelerated processing unit. Additionally, it was found that in order to justify the use of GPGPU, the amount of parallel work being done within a kernel would have to greatly exceed the work being done by any one portion of the APNASA code. It was determined that in order for an application like APNASA to be accelerated on the GPU, it should not be modular in nature, and the parallel portions of the code must contain a large portion of the code's computation time.
Suppression pool swell analysis using CFD code
A two-dimensional axi-symmetric model of suppression pool of Containment Studies Facility (CSF) along with single vent pipe was modeled to estimate the jet and hydrodynamic loads due to flow of steam air mixture during simulated loss of coolant accident (LOCA). The analysis was carried out using CFD ACE+ software with Volume of Fluid (VOF) approach. The flow velocity variation through vent pipe was estimated using in-house containment thermal hydraulic code CONTRAN, was given as input at inlet boundary condition. The transient calculations were performed for 20 seconds and suppression pool level variation, pressure loads over the floor, walls and vent pipes etc were evaluated. (author)
Validation of Francis Water Turbine CFD Simulations
Čarija, Zoran; Mrša, Zoran; Fućak, Sanjin
2008-01-01
This paper compares data from calculated and measured results covering the whole operating range for a 20 MW Francis turbine in order to validate the CFD simulation. Computed hydraulic characteristics are determined for each analyzed operating point by running numerical simulations of turbulent fluid flow through a complete Francis Turbine model using the commercial fluid flow solver Fluent. The measured hydraulic characteristics were defined by on-site measurements according to the IEC41 int...
CFD Calculations in Turbomachinery and their Validation
Váchová, J.; Louda, P.; Příhoda, Jaromír; Luxa, Martin; Šimurda, David
Praha : TechSoft Engineering s.r.o, 2015, s. 156-165. ISBN 978-80-905040-5-9. [Turbostroje 2015. Praha (CZ), 22.09.2015-24.09.2015] R&D Projects: GA TA ČR(CZ) TA03020277; GA ČR GAP101/12/1271 Institutional support: RVO:61388998 Keywords : CFD * validation * transition model * blade cascade * steamm turbine Subject RIV: BK - Fluid Dynamics
Coupling a CFD code with neutron kinetics and pin thermal models for nuclear reactor safety analyses
Highlights: • A CFD/neutron kinetics coupled code FLUENT/PK for nuclear reactor transient safety was developed. • The mathematical models and coupling methods of FLUENT/PK were described. • The code-to-code validation between FLUENT/PK and SIMMER-III was conducted. - Abstract: Most system codes are based on the one-dimensional lumped-parameter method, which is unsuitable to simulate multi-dimensional thermal-hydraulics problems. CFD method is a good tool to simulate multi-dimensional thermal-hydraulics phenomena in the nuclear reactor, which can increase the accuracy of analysis results. However, since there is no neutron kinetics model and pin thermal model in current CFD codes, the application of the CFD method in the area of nuclear reactor safety analyses is still limited. Coupling a CFD code with the neutron kinetics model (PKM) and the pin thermal model (PTM) is a good way to use CFD code to simulate multi-dimensional thermal-hydraulics problems of nuclear reactors. The motivation for this work is to develop a CFD/neutron kinetics coupled code named FLUENT/PK for nuclear reactor safety analyses by coupling the commercial CFD code named FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM). The mathematical models and the coupling method are described and the unprotected transient overpower (UTOP) accident of a liquid metal cooled fast reactor (LMFR) is chosen as an application case. As a general validation, the calculated results are used to compare with that of another multi-physics coupled code named SIMMER-III and good agreements are achieved for various characteristic parameters
CFD codes and the Onsager relations
In the last decade, papers appeared in the literature discussing a shortcoming of the basic equations of hydrodynamics: the Navier-Stokes equations do not meet the Onsager symmetry relations. Recently R. Streater wrote about the topic. The basic problem is that the solution to the Boltzmann equation f(r,v,t) depends on seven variable, nevertheless the solution of the Navier-Stokes equation yields T(r,t),v(r,t) and ?(r,t)-the temperature, velocity and density distribution, altogether ?ve functions. Clearly, the solution class of the Boltzmann equation is a broader class than the solution class of the Navier-Stokes equation. What is the importance of that question? Are the results of CFD codes questionable ones or, the contradiction can be resolved by some ignored terms of second order? (Author)
In this report some validation tests for the TESEO code are described. The TESEO code was developed at ENEA - Clementel Center in the framework of the C2RV code sequence. This code sequence produces multigroup resonance cross sections for fast reactor analysis. It consists of the codes TESEO, MC2-II, GERES, ANISN, MEDIL. The TESEO code processes basic nuclear data in ENDF-B format and produces an ultrafine group (2082 groups) cross section library for the MC2-II code. To validate the TESEO algorithms, the data produced by TESEO code were compared with the data produced by other well-tested codes which use different algorithms. No substantial differences was found between these data and the data produced by TESEO code. TESEO algorithms showed high reliability. A detailed study of TESEO calculation options was carried out. Their use and functions are shown to inform the user of the code
CFD code benchmark against the air/helium tests performed in the MISTRA facility
Highlights: • CFD code validation against the stratification and erosion experiments. • Turbulence model sensitivity was carried out to identify the best suited turbulence model. • 3-D simulations were performed. • These simulations are necessary to eventually use CFD codes for containment hydrogen distribution analysis. • Symmetric trends in the stratification have been captured. - Abstract: The behaviour of hydrogen mixing and distribution has always been an important safety issue and the hydrogen distribution studies gained importance especially after Fukushima accident. The hydrogen generated due to metal water reaction releases into the containment and may get stratified locally under accident conditions. The stratification of hydrogen may be eroded by diffusion or by other means. CFD codes are increasingly being used for hydrogen distribution analysis and need to be validated before applying it to full scale containment simulations. In this context, the CFD code FLUENT is validated against the experiment conducted in the MISTRA facility on stratification and erosion behaviour. This paper deals with the validation of the CFD code FLUENT against the experiment conducted in MISTRA facility to study the stratification behaviour. Turbulence model sensitivity was carried out to identify the best suited turbulence model
ARC Code TI: CFD Utility Software Library
National Aeronautics and Space Administration — The CFD Utility Software Library consists of nearly 30 libraries of Fortran 90 and 77 subroutines and almost 100 applications built on those libraries. Many of the...
Standard Problems for CFD Validation for NGNP - Status Report
The U.S. Department of Energy (DOE) is conducting research and development to support the resurgence of nuclear power in the United States for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The project is called the Next Generation Nuclear Plant (NGNP) Project, which is based on a Generation IV reactor concept called the very high temperature reactor (VHTR). The VHTR will be of the prismatic or pebble bed type; the former is considered herein. The VHTR will use helium as the coolant at temperatures ranging from 250 C to perhaps 1000 C. While computational fluid dynamics (CFD) has not previously been used for the safety analysis of nuclear reactors in the United States, it is being considered for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal operational and accident situations. The ''Standard Problem'' is an experimental data set that represents an important physical phenomenon or phenomena, whose selection is based on a phenomena identification and ranking table (PIRT) for the reactor in question. It will be necessary to build a database that contains a number of standard problems for use to validate CFD and systems analysis codes for the many physical problems that will need to be analyzed. The first two standard problems that have been developed for CFD validation consider flow in the lower plenum of the VHTR and bypass flow in the prismatic core. Both involve scaled models built from quartz and designed to be installed in the INL's matched index of refraction (MIR) test facility. The MIR facility employs mineral oil as the working fluid at a constant temperature. At this temperature, the index of refraction of the mineral oil is the same as that of the quartz. This provides an advantage to the
Application of CFD Code PHOENICS for simulating CYCLONE SEPARATORS
The work presents a computational fluid dynamics (CFD) calculation to investigate the flow field in a tangential inlet cyclone which is mainly used for the separation of the moisture from an air stream. Three-dimensional, steady state Eulerian simulations of the turbulent gas - droplet flow in a cyclone separator have been performed. Numerical simulation was carried out using CFD code PHOENICS for the given geometry of separators available in literature
Boiling flow simulation in Neptune-CFD and Fluent codes
This paper presents simulations of the convective boiling flow performed with NEPTUNE-CFD and FLUENT codes. The DEBORA experiments carried out at CEA Grenoble were used as an experimental data set. In these experiments, freon R12 flows upwards inside a vertical pipe. Radial profiles of the flow variables are measured at the end of the heated section. Seven DEBORA cases were selected for simulation. NEPTUNE-CFD code was used without modifications because it contains all necessary models. In FLUENT, an important part of the models has been implemented by programming in User Defined Functions. The comparison of the radial profiles of void fraction, liquid temperature, gas velocity and mean bubble diameter at the end of the heated section shows that both codes can provide reasonable results in boiling conditions. The presented work was carried out within the 6. Framework EC NURESIM project. NEPTUNE-CFD code is implemented in the NURESIM platform. (authors)
Validation process of ISIS CFD software for fire simulation
Lapuerta, C., E-mail: celine.lapuerta@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3, 13115 Saint Paul-lez-Durance (France); ETIC Laboratory, IRSN-CNRS-UAM (I,II), 5 rue Enrico Fermi, 13453 Marseille Cedex 13 (France); Suard, S., E-mail: sylvain.suard@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3, 13115 Saint Paul-lez-Durance (France); ETIC Laboratory, IRSN-CNRS-UAM (I,II), 5 rue Enrico Fermi, 13453 Marseille Cedex 13 (France); Babik, F., E-mail: fabrice.babik@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3, 13115 Saint Paul-lez-Durance (France); Rigollet, L., E-mail: laurence.rigollet@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3, 13115 Saint Paul-lez-Durance (France); ETIC Laboratory, IRSN-CNRS-UAM (I,II), 5 rue Enrico Fermi, 13453 Marseille Cedex 13 (France)
2012-12-15
Fire propagation constitutes a major safety concern in nuclear facilities. In this context, IRSN is developing a CFD code, named ISIS, dedicated to fire simulations. This software is based on a coherent set of models that can be used to describe a fire in large, mechanically ventilated compartments. The system of balance equations obtained by combining these models is discretized in time using fractional step methods, including a pressure correction technique for solving hydrodynamic equations. Discretization in space combines two techniques, each proven in the relevant context: mixed finite elements for hydrodynamic equations and finite volumes for transport equations. ISIS is currently in an advanced stage of verification and validation. The results obtained for a full-scale fire test performed at IRSN are presented.
Validation of CFD for containment jet flows including condensation
The advanced validation of a CFD code for containment applications requires the investigation of water steam in the different flow types like jets or buoyant plumes. This paper addresses therefore the simulation of two 'HYJET' experiments from the former Battelle Model Containment by CFX. These experiments involve jet releases into the multi-compartment geometry of the test facility accompanied by condensation of steam at walls and in the bulk gas. In both experiments mixtures of helium and steam are injected. Helium is used to simulate hydrogen. One experiment represents a fast jet whereas in the second test a slow helium-steam release is investigated. CFX was earlier extended by bulk and wall condensation models and is able to model all relevant phenomena observed during the experiments. The paper focuses on the simulation of the two experiments employing an identical model set-up. This provides information on how well a wider range of flowing conditions in case of a full containment simulation can be covered. Some aspects related to numerical and modelling uncertainties of CFD calculations are included in the paper by investigating different turbulence models together with the modelling errors of the differencing schemes applied. (authors)
Verification calculations as per CFD FLOWVISION code for sodium-cooled reactor plants
The paper studies the experience in application of CFD FlowVision software for analytical validation of sodium-cooled fast reactor structure components and the results of performed verification, namely: – development and implementation of new model of turbulent heat transfer in liquid sodium (LMS) in FlowVision software and model verification based on thermohydraulic characteristics studied by experiment at TEFLU test facility; – simulation of flowing and mixing of coolant with different temperatures in the upper mixing chamber of fast neutron reactor through the example of BN-600 (comparison with the results obtained at the operating reactor). Based on the analysis of the results obtained, the efficiency of CFD codes application for the considered problems is shown, and the proposals for CFD codes verification development as applied to the advanced sodium-cooled fast reactor designs are stated. (author)
Thompson, David E.
2005-01-01
Procedures and methods for veri.cation of coding algebra and for validations of models and calculations used in the aerospace computational fluid dynamics (CFD) community would be ef.cacious if used by the glacier dynamics modeling community. This paper presents some of those methods, and how they might be applied to uncertainty management supporting code veri.cation and model validation for glacier dynamics. The similarities and differences between their use in CFD analysis and the proposed application of these methods to glacier modeling are discussed. After establishing sources of uncertainty and methods for code veri.cation, the paper looks at a representative sampling of veri.cation and validation efforts that are underway in the glacier modeling community, and establishes a context for these within an overall solution quality assessment. Finally, a vision of a new information architecture and interactive scienti.c interface is introduced and advocated.
Bandini, G., E-mail: giacomino.bandini@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Polidori, M. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Gerschenfeld, A.; Pialla, D.; Li, S. [Commissariat à l’Energie Atomique (CEA) (France); Ma, W.M.; Kudinov, P.; Jeltsov, M.; Kööp, K. [Royal Institute of Technology (KTH) (Sweden); Huber, K.; Cheng, X.; Bruzzese, C.; Class, A.G.; Prill, D.P. [Karlsruhe Institute of Technology (KIT) (Germany); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Geffray, C.; Macian-Juan, R. [Technische Universität München (TUM) (Germany); Maas, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)
2015-01-15
Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.
Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes
Validation process of the ISIS CFD software for fire simulation
Fire codes are more and more used for safety analysis of nuclear power plants. In several OECD member countries, the accuracy of the calculated simulation with CFD code has to be demonstrated; this is the aim of the Verification and Validation process (V and V). In this context the French 'Institut de Radioprotection et de Surete Nucleaire' (IRSN) develops a computational software, named ISIS, dedicated to the simulation of buoyant fire in compartment mechanically ventilated. ISIS is based on the scientific computing development platform PELICANS and benefits of the practicalities for implementing methods. The code ISIS is a freeware, available at https://gforge.irsn.fr/gf/project/isis. The physical modelling used in ISIS is classic for industrial application in large compartments. The turbulence approach is based on the Reynolds-Averaged-Navier-Stokes equations, supplemented by a two-equation closure and the eddy viscosity model. The turbulent production term is adapted to cope with buoyancy effects. Combustion modelling relies on a single reaction equation. The classical eddy dissipation approach is used for the mean chemical reaction rate which means that it is controlled solely by the turbulent mixture. The Finite Volume method is employed to treat radiation exchanges. Both incompressible and low Mach number flows are dealt with. The originality of the ISIS code is its capacity to take into account the effect of ventilation on the pressure. The thermodynamic pressure and the mass flow rate for ventilation vents are related by the mass balances in the compartment and in the ventilation branch where an aeraulic resistance is taken into account. For numerical solution, a fractional step algorithm has been developed. The spatial discretization combines mixed finite element for the Navier-Stokes equation and finite volumes scheme for transport (advection-diffusion-reaction) equation in order to ensure the velocity stability and the conservation in physical range of
Validation of NEPTUNE-CFD 1.0.8 for adiabatic bubbly flow and boiling flow
The NEPTUNE-CFD code, which is based on an Eulerian two-fluid model, is developed within the framework of the NEPTUNE project, financially supported by CEA (Commissariat a l'Energie Atomique), EDF, IRSN (Institut de Radioprotection et de Surete Nucleaire) and AREVA-NP. NEPTUNE-CFD is mainly focused on Nuclear Reactor Safety applications involving two-phase flows, like two-phase Pressurized Thermal Shock (PTS) and Departure from Nucleate Boiling (DNB). Since the maturity of two-phase CFD has not reached yet the same level as single phase CFD, an important work of model development and thorough validation is needed, as stated for example in NEA/CSNI Writing Group dedicated to the 'Extension of CFD Codes to Two-Phase Flow Safety Problems' (draft6c, 2009). Many of these applications involve bubbly and boiling flows, and therefore it is essential to validate the software on such configurations. In particular, this is crucial for applications to flow in PWR fuel assemblies, including studies related to DNB. This work aims at presenting the present status of NEPTUNE-CFD validation in this area, as a step in an iterative process of improvement. To this end, this paper presents NEPTUNE-CFD code validation against four test cases based on experimental results. These data have been selected to allow separate effects validation. The adequacy of the measured quantities and the corresponding basic model of the CFD code to validate is underlined in each case. The selected test cases are the following. The Liu and Bankhoff experiment (1993) is an adiabatic air-water bubbly flow inside a vertical pipe. It allows to validate forces applied to the bubbles. The Bel F'Dhila and Simonin (1992) experiment is an adiabatic bubbly air-water flow inside a sudden pipe expansion. It allows to validate the dynamic models and turbulence. The DEBORA (CEA, 2002) and the ASU (Arizona State University, Hassan 1990) facilities provide results for boiling flows inside a vertical pipe. The working
Improvement of core effective thermal conductivity model of GAMMA+ code based on CFD analysis
Highlights: • We assessed the core effective thermal conductivity (ETC) model of GAMMA+ code. • The analytical model of GAMMA+ code was compared with the result of CFD analysis. • Effects of material property of composite and geometric configuration were studied. • The GAMMA+ model agreed with the CFD result when the fuel gap is ignored. • The GAMMA+ model was improved by the ETC model of fuel compact including fuel gap. - Abstract: The GAMMA+ code has been developed for the thermo-fluid and safety analyses of a high temperature gas-cooled reactor (HTGR). In order to calculate the core effective thermal conductivity, this code adopts a heterogeneous model derived from the Maxwell’s theory that accounts for three distinct materials in a fuel block of the reactor core. In this model, the fuel gap is neglected since the gap thickness is quite small. In addition, the configuration of the fuel block is assumed to be homogeneous, and the volume fraction and material properties of each component are taken into account. In the accident condition, the conduction and radiation are major heat transfer mechanism. Therefore, the core effective thermal conductivity model should be validated in order to estimate the heat transfer in the core appropriately. In this regard, the objective of this study is to validate the core effective thermal conductivity model of the GAMMA+ code by a computational fluid dynamics (CFD) analysis using a commercial CFD code, CFX-13. The effects of the temperature condition, material property and geometric modeling on the core effective thermal conductivity were investigated. When the fuel gap is not modeled in the CFD analysis, the result of the GAMMA+ code shows a good agreement with the CFD result. However, when the fuel gap is modeled, the GAMMA+ model overestimates the core effective thermal conductivity considerably for all cases. This is because of the increased thermal resistance by the fuel gap which is not taken into account in
Supersonic Retropropulsion CFD Validation with Ames Unitary Plan Wind Tunnel Test Data
Schauerhamer, Daniel G.; Zarchi, Kerry A.; Kleb, William L.; Edquist, Karl T.
2013-01-01
A validation study of Computational Fluid Dynamics (CFD) for Supersonic Retropropulsion (SRP) was conducted using three Navier-Stokes flow solvers (DPLR, FUN3D, and OVERFLOW). The study compared results from the CFD codes to each other and also to wind tunnel test data obtained in the NASA Ames Research Center 90 70 Unitary PlanWind Tunnel. Comparisons include surface pressure coefficient as well as unsteady plume effects, and cover a range of Mach numbers, levels of thrust, and angles of orientation. The comparisons show promising capability of CFD to simulate SRP, and best agreement with the tunnel data exists for the steadier cases of the 1-nozzle and high thrust 3-nozzle configurations.
Simulation of Jet Noise with OVERFLOW CFD Code and Kirchhoff Surface Integral
Kandula, M.; Caimi, R.; Voska, N. (Technical Monitor)
2002-01-01
An acoustic prediction capability for supersonic axisymmetric jets was developed on the basis of OVERFLOW Navier-Stokes CFD (Computational Fluid Dynamics) code of NASA Langley Research Center. Reynolds-averaged turbulent stresses in the flow field are modeled with the aid of Spalart-Allmaras one-equation turbulence model. Appropriate acoustic and outflow boundary conditions were implemented to compute time-dependent acoustic pressure in the nonlinear source-field. Based on the specification of acoustic pressure, its temporal and normal derivatives on the Kirchhoff surface, the near-field and the far-field sound pressure levels are computed via Kirchhoff surface integral, with the Kirchhoff surface chosen to enclose the nonlinear sound source region described by the CFD code. The methods are validated by a comparison of the predictions of sound pressure levels with the available data for an axisymmetric turbulent supersonic (Mach 2) perfectly expanded jet.
Alali, Abdullah
2014-02-21
The one-group interfacial area transport equation has been coupled to a wall heat flux partitioning model in the framework of two-phase Eulerian approach using the OpenFOAM CFD code for better prediction of subcooled boiling phenomena which is essential for safety analysis of nuclear reactors. The interfacial area transport equation has been modified to include the effect of bubble nucleation at the wall and condensation by subcooled liquid in the bulk that governs the non-uniform bubble size distribution.
The one-group interfacial area transport equation has been coupled to a wall heat flux partitioning model in the framework of two-phase Eulerian approach using the OpenFOAM CFD code for better prediction of subcooled boiling phenomena which is essential for safety analysis of nuclear reactors. The interfacial area transport equation has been modified to include the effect of bubble nucleation at the wall and condensation by subcooled liquid in the bulk that governs the non-uniform bubble size distribution.
Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too
Groves, Curtis E.; LLie, Marcel; Shallhorn, Paul A.
2012-01-01
There are inherent uncertainties and errors associated with using Computational Fluid Dynamics (CFD) to predict the flow field and there is no standard method for evaluating uncertainty in the CFD community. This paper describes an approach to -validate the . uncertainty in using CFD. The method will use the state of the art uncertainty analysis applying different turbulence niodels and draw conclusions on which models provide the least uncertainty and which models most accurately predict the flow of a backward facing step.
CFD validation in OECD/NEA t-junction benchmark.
Obabko, A. V.; Fischer, P. F.; Tautges, T. J.; Karabasov, S.; Goloviznin, V. M.; Zaytsev, M. A.; Chudanov, V. V.; Pervichko, V. A.; Aksenova, A. E. (Mathematics and Computer Science); (Cambridge Univ.); (Moscow Institute of Nuclar Energy Safety)
2011-08-23
When streams of rapidly moving flow merge in a T-junction, the potential arises for large oscillations at the scale of the diameter, D, with a period scaling as O(D/U), where U is the characteristic flow velocity. If the streams are of different temperatures, the oscillations result in experimental fluctuations (thermal striping) at the pipe wall in the outlet branch that can accelerate thermal-mechanical fatigue and ultimately cause pipe failure. The importance of this phenomenon has prompted the nuclear energy modeling and simulation community to establish a benchmark to test the ability of computational fluid dynamics (CFD) codes to predict thermal striping. The benchmark is based on thermal and velocity data measured in an experiment designed specifically for this purpose. Thermal striping is intrinsically unsteady and hence not accessible to steady state simulation approaches such as steady state Reynolds-averaged Navier-Stokes (RANS) models.1 Consequently, one must consider either unsteady RANS or large eddy simulation (LES). This report compares the results for three LES codes: Nek5000, developed at Argonne National Laboratory (USA), and Cabaret and Conv3D, developed at the Moscow Institute of Nuclear Energy Safety at (IBRAE) in Russia. Nek5000 is based on the spectral element method (SEM), which is a high-order weighted residual technique that combines the geometric flexibility of the finite element method (FEM) with the tensor-product efficiencies of spectral methods. Cabaret is a 'compact accurately boundary-adjusting high-resolution technique' for fluid dynamics simulation. The method is second-order accurate on nonuniform grids in space and time, and has a small dispersion error and computational stencil defined within one space-time cell. The scheme is equipped with a conservative nonlinear correction procedure based on the maximum principle. CONV3D is based on the immersed boundary method and is validated on a wide set of the experimental
Computational Fluid Dynamics (CFD) is to an increasing extent being adopted in nuclear reactor safety analyses as a tool that enables specific safety relevant phenomena occurring in the reactor coolant system to be better described. The Committee on the Safety of Nuclear Installations (CSNI), which is responsible for the activities of the OECD Nuclear Energy Agency that support advancing the technical base of the safety of nuclear installations, has in recent years conducted an important activity in the CFD area. This activity has been carried out within the scope of the CSNI working group on the analysis and management of accidents (GAMA), and has mainly focused on the formulation of user guidelines and on the assessment and verification of CFD codes. It is in this GAMA framework that a first workshop CFD4NRS was organized and held in Garching, Germany in 2006. Following the CFD4NRS workshop, this XCFD4NRS Workshop was intended to extend the forum created for numerical analysts and experimentalists to exchange information in the field of Nuclear Reactor Safety (NRS) related activities relevant to Computational Fluid Dynamics (CFD) validation, but this time with more emphasis placed on new experimental techniques and two-phase CFD applications. The purpose of the workshop was to provide a forum for numerical analysts and experimentalists to exchange information in the field of NRS-related activities relevant to CFD validation, with the objective of providing input to GAMA CFD experts to create a practical, state-of-the-art, web-based assessment matrix on the use of CFD for NRS applications. The scope of XCFD4NRS includes: - Single-phase and two-phase CFD simulations with an emphasis on validation in areas such as: boiling flows, free-surface flows, direct contact condensation and turbulent mixing. These applications should relate to NRS-relevant issues such as: pressurized thermal shocks, critical heat flux, pool heat exchangers, boron dilution, hydrogen
CFD Application in Implantable Rotary Blood Pump Design and Validation
YI Qian
2004-01-01
Implantable rotary blood pump (IRBP) has been promoted to the stage of clinical trial. This paper introduces a unique IRBP without a shaft. Instead of using thrombogenic pivots or power-drawing magnetic suspension, impeller is supported hydrodynamically when rotating, by lubrication flows in the thin spaces between itself and the pump body. To this end, the flow is very difficult to be measured using usual laboratory equipments. Therefore, computational fluid dynamics (CFD) has been applied as an important tool in the IRBP design and its validation procedure. Several CFD results such as pump performance improvement, unsteady hydraulic dynamic analysis, biocapability prediction, validation and verification (V&V), and flow visualization have been performed.
CFD Application in Implantable Rotary Blood Pump Design and Validation
YIQian
2004-01-01
Implantable rotary blood pump (IRBP) has been promoted to the stage of clinical trial. This paper introduces a unique IRBP without a.shaft. Instead of using thrombogenic pivots or power-drawing magnetic suspension, impeller is supported hydrodynamically when rotating, by lubrication flows in the thin spaces between itself and the pump body. To this end, the flow is very difficult to be measured using usual laboratory equipments. Therefore, computational fluid dynamics (CFD) has been applied as an important tool in the IRBP design and its validation procedure. Several CFD results such as pump performance improvement, unsteady hydraulic dynamic analysis, biocapability prediction, validation and verification (V&V), and flow visualization have been performed.
The paper presents results of CFD calculations of spent fuel storage pool at WWER-440 and WWER-1000 units. The calculations were performed by the Fluent 6.2 CFD code. Standard nuclear safety problems of spent fuel pools, such as keff calculation or spent fuel pool dry-out more technical problems related to spent fuel pool operation in the Czech Republic NPPs Dukovany and Temelin. Following several problems had been identified during nuclear power plant operation and shutdown procedure validation: 1. Inadequate water temperature and water level measurements; 2. Repeated cracking of pool stainless steel lining; 3. Lack of data for shutdown procedure validation. The first two items were supposed to have a common cause - significant non-uniformity of pool water temperature fields and related strong buoyancy effects. We have analysed flow patterns in spent fuel pools and temperature fields at pool walls using the Fluent CFD code to verify this assumption and to solve above-mentioned problems ( Authors)
Developing a methodology for the evaluation of results uncertainties in CFD codes
In this work the development of a methodology is studied to evaluate the uncertainty in the results of CFD codes and is compatible with the VV-20 standard Standard for Verification and Validation in CFD and Heat Transfer , developed by the Association of Mechanical Engineers ASME . Similarly, the alternatives are studied for obtaining existing uncertainty in the results to see which is the best choice from the point of view of implementation and time. We have developed two methods for calculating uncertainty of the results of a CFD code, the first method based on the use of techniques of Monte-Carlo for the propagation of uncertainty in this first method we think it is preferable to use the statistics of the order to determine the number of cases to execute the code, because this way we can always determine the confidence interval desired level of output quantities. The second type of method we have developed is based on non-intrusive polynomial chaos. (Author)
A CFD code comparison of wind turbine wakes
Laan, van der, Paul Maarten; Storey, R. C.; Sørensen, Niels N.;
2014-01-01
A comparison is made between the EllipSys3D and SnS CFD codes. Both codes are used to perform Large-Eddy Simulations (LES) of single wind turbine wakes, using the actuator disk method. The comparison shows that both LES models predict similar velocity deficits and stream-wise Reynolds-stresses for...... simulations using EllipSys3D for a test case that is based on field measurements. In these simulations, two eddy viscosity turbulence models are employed: the k- (ε) model and the k- (ε)-fp model. Where the k- (ε) model fails to predict the velocity deficit, the results of the k- (ε)-fP model show good...
Verification and Validation Studies for the LAVA CFD Solver
Moini-Yekta, Shayan; Barad, Michael F; Sozer, Emre; Brehm, Christoph; Housman, Jeffrey A.; Kiris, Cetin C.
2013-01-01
The verification and validation of the Launch Ascent and Vehicle Aerodynamics (LAVA) computational fluid dynamics (CFD) solver is presented. A modern strategy for verification and validation is described incorporating verification tests, validation benchmarks, continuous integration and version control methods for automated testing in a collaborative development environment. The purpose of the approach is to integrate the verification and validation process into the development of the solver and improve productivity. This paper uses the Method of Manufactured Solutions (MMS) for the verification of 2D Euler equations, 3D Navier-Stokes equations as well as turbulence models. A method for systematic refinement of unstructured grids is also presented. Verification using inviscid vortex propagation and flow over a flat plate is highlighted. Simulation results using laminar and turbulent flow past a NACA 0012 airfoil and ONERA M6 wing are validated against experimental and numerical data.
Phenomena such as gas stratification in an LWR containment, gas transport between containment compartments, wall condensation and hydrogen accumulation have been identified as high-ranking phenomena playing an important role in issues directly related to the safety of current LWRs and also future reactors. These phenomena are driven by buoyant high momentum injection (jets) and/or low momentum injection (plumes). For instance, mixing in the immediate vicinity of the postulated line break is mainly dominated by very high velocity efflux, while low-momentum flows are responsible for most of the transport processes within the containment. Codes with 3D capabilities, e.g. CFD codes offer the possibility of using accurate simulation models, which properly account for gas (steam, air, hydrogen, etc.) in-homogeneity and to characterize the evolution of such phenomena in complex geometries such as the LWR containment. Code assessment and validation against experimental data are needed activities for increasing the confidence in the use of the computational tools and for revealing strengths and drawbacks with respect to particular geometries, phenomena or conditions. The use of experimental data obtained in large-scale facilities, under prototypical thermal-hydraulic conditions, allows for minimizing distortion effects arising from geometrical scaling. Multi-compartments facilities allow flow transport between compartments (e.g. due to density differences induced by condensation) to be studied. Nevertheless the use of large scale facilities for generating a CFD-quality database requires from an experimental point of view a huge effort toward the upgrading of instrumentation and the use of computational tools already in the preparatory phase of the experimental program, e.g. for defining test conditions, test procedures, instrumentation needs and location of key instrumentation. The large-scale, multi-compartments PANDA facility (located at PSI in Switzerland) is one of the
Two Phase Flow Models and Numerical Methods of the Commercial CFD Codes
Bae, Sung Won; Jeong, Jae Jun; Chang, Seok Kyu; Cho, Hyung Kyu
2007-11-15
The use of commercial CFD codes extend to various field of engineering. The thermal hydraulic analysis is one of the promising engineering field of application of the CFD codes. Up to now, the main application of the commercial CFD code is focused within the single phase, single composition fluid dynamics. Nuclear thermal hydraulics, however, deals with abrupt pressure changes, high heat fluxes, and phase change heat transfer. In order to overcome the CFD limitation and to extend the capability of the nuclear thermal hydraulics analysis, the research efforts are made to collaborate the CFD and nuclear thermal hydraulics. To achieve the final goal, the current useful model and correlations used in commercial CFD codes should be reviewed and investigated. This report gives the summary information about the constitutive relationships that are used in the FLUENT, STAR-CD, and CFX. The brief information of the solution technologies are also enveloped.
Two Phase Flow Models and Numerical Methods of the Commercial CFD Codes
The use of commercial CFD codes extend to various field of engineering. The thermal hydraulic analysis is one of the promising engineering field of application of the CFD codes. Up to now, the main application of the commercial CFD code is focused within the single phase, single composition fluid dynamics. Nuclear thermal hydraulics, however, deals with abrupt pressure changes, high heat fluxes, and phase change heat transfer. In order to overcome the CFD limitation and to extend the capability of the nuclear thermal hydraulics analysis, the research efforts are made to collaborate the CFD and nuclear thermal hydraulics. To achieve the final goal, the current useful model and correlations used in commercial CFD codes should be reviewed and investigated. This report gives the summary information about the constitutive relationships that are used in the FLUENT, STAR-CD, and CFX. The brief information of the solution technologies are also enveloped
The French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) are developing a hydrogen risk analysis code (safety code) which incorporates both lumped parameter (LP) and computational fluid dynamics (CFD) formulations. In this paper we present briefly the main physical models for containment thermal-hydraulics. Validation and typical numerical results will be presented for hydrogen distribution and combustion applications in small and realistic large geometries. (authors)
Simulation and analysis of void drift using sub-channel analysis code and CFD code
Pang, Bo; Cheng, Xu; Otic, Ivan [Karlsruhe Institute of Technology (KIT) (Germany). Inst. of Fusion and Reactor Technology (IFRT)
2012-11-01
Prediction accuracy of a sub-channel analysis depends strongly on the modeling of the interchannel transverse exchange effect. Disregarding the forced mixing effects caused by extra constructive elements the natural inter-channel transverse exchange effect can be decomposed into [1] [2] [3]: turbulent mixing (TM) due to the natural eddy diffusion, diversion cross flow (DC) induced by radial pressure gradient and void drift (VD) specially under two-phase flow conditions. Among the three components, the physical mechanism of void drift is not well clarified. Previous to the time and cost demanding experimental research a systematic numerical simulation of the inter-channel exchange effect with CFD code can provide supplemental information about the physical mechanism behind the not well clarified void drift phenomena. Compared to sub-channel analysis code, CFD code solves the flow dynamic problem with a much finer mesh and in a more physical way. The inter-channel exchange terms are solved in the conservation equations rather than modeled with closure equations. Furthermore, the inter-phase exchange terms are also taken into account. A better understanding of the void drift phenomenon and a modification of the void drift models in a sub-channel analysis code basing on the CFD analysis can be achieved. In present study, both sub-channel and CFD analysis are carried out for studying the void drift in a rod bundle geometry. A model is proposed to determine the sub-channel scale void drift mass flux based on the CFD simulation results. (orig.)
The extensive international use of commercial computational fluid dynamics (CFD) codes
What are the main reasons for the extensive international success of commercial CFD codes? This is due to their ability to calculate the fine structures of the investigated processes due to their versatility, their numerical stability and that they can guarantee the proper solution in most cases. This was made possible by the constantly increasing computer power at an ever more affordable prize. Furthermore it is much more efficient to have researchers use a CFD code rather than to develop a similar code system due to the time consuming nature of this activity and the high probability of hidden coding errors. The centralized development and upgrading makes these reliable CFD codes possible and affordable. However, the CFD companies' developments are naturally concentrated on the most profitable areas, and thus, if one works in a 'non-priority' field one cannot use them. Moreover, the prize of renting CFD codes, applications to complex systems such as whole nuclear reactors and the need to teach students gives the development of self-made codes still plenty of room. But CFD codes can model detailed aspects of large systems and subroutines generated by users can be added. Since there are only a few heavily used CFD codes such as FLUENT, STAR-CD, ANSYS CFX, these are used in many countries. Also international training courses are given and the news bulletins of these codes help to spread the news on further developments. A larger number of international codes would increase the competition but would at the same time make it harder to select the most appropriate CFD code for a given problem. Examples will be presented of uses of CFD codes as more detailed system codes for the decay heat removal from reactors, the application to aerosol physics and the application to heavy metal fluids using different turbulence models. (author)
Preliminary tests of a damaged ship for CFD validation
Lee, Sungkyun; You, Ji-Myoung; Lee, Hyun-Ho; Lim, Taegu; Rhee, Shin Hyung; Rhee, Key-Pyo
2012-06-01
One of the most critical issues in naval architecture these days is the operational safety. Among many factors to be considered for higher safety level requirements, the hull stability in intact and damaged conditions is the first to ensure for both commercial and military vessels. Unlike the intact stability cases, the assessment of the damaged ship stability is very complicated physical phenomena. Therefore it is widely acknowledged that computational fluid dynamics (CFD) methods are one of most feasible approaches. In order to develop better CFD methods for damaged ship stability assessment, it is essential to perform well-designed model tests and to build a database for CFD validation. In the present study, free roll decay tests in calm water with both intact and damaged ships were performed and six degree-of-freedom (6DOF) motion responses of intact ship in regular waves were measured. Through the free roll decay tests, the effects of the flooding water on the roll decay motion of a ship were investigated. Through the model tests in regular waves, the database that provides 6DOF motion responses of intact ship was established
Development and validation of the 3-D CFD model for CANDU-6 moderator temperature predictions
A computational fluid dynamics model for predicting the moderator circulation inside the CANada Deuterium Uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the Calandria tubes. The buoyancy effect induced by internal heating is accounted for by Boussinesq approximation. The standard κ-ε turbulence model associated with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the Calandria tubes in the core region is simplified to porous media, in which an-isotropic hydraulic impedance is modeled using an empirical correlation of the frictional pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA technology. The CFD model has been successfully verified and validated against experimental data obtained in the Stern Laboratories Inc. (SLI) in Hamilton, Ontario
Extension of CFD Codes Application to Two-Phase Flow Safety Problems - Phase 3
The Writing Group 3 on the extension of CFD to two-phase flow safety problems was formed following recommendations made at the 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in Aix-en-Provence, in May 2002. Extension of CFD codes to two-phase flow is significant potentiality for the improvement of safety investigations, by giving some access to smaller scale flow processes which were not explicitly described by present tools. Using such tools as part of a safety demonstration may bring a better understanding of physical situations, more confidence in the results, and an estimation of safety margins. The increasing computer performance allows a more extensive use of 3D modelling of two-phase Thermal hydraulics with finer nodalization. However, models are not as mature as in single phase flow and a lot of work has still to be done on the physical modelling and numerical schemes in such two-phase CFD tools. The Writing Group listed and classified the NRS problems where extension of CFD to two-phase flow may bring real benefit, and classified different modelling approaches in a first report (Bestion et al., 2006). First ideas were reported about the specification and analysis of needs in terms of validation and verification. It was then suggested to focus further activity on a limited number of NRS issues with a high priority and a reasonable chance to be successful in a reasonable period of time. The WG3-step 2 was decided with the following objectives: - selection of a limited number of NRS issues having a high priority and for which two-phase CFD has a reasonable chance to be successful in a reasonable period of time; - identification of the remaining gaps in the existing approaches using two-phase CFD for each selected NRS issue; - review of the existing data base for validation of two-phase CFD application to the selected NRS problems
CFD code fluent turbulence models application. Ansaldo's prototype modeling
Among others, one of the main activities in the Nuclear Engineering and Fluid Mechanics Department of the Engineering School in Bilbao, is the study of liquid metals behavior. And for this purpose the CFD code FLUENT is being used. Currently, the code is being applied to the use of Lead-Bismuth eutectic (LBE) as the coolant of an accelerator driven system (ADS) and also as the target for a neutron source. In this paper, ANSALDO's Energy Amplifier Demonstration Facility is simulated, paying attention only on the coolant. As it will be later explained, natural convection is a very important issue, because the philosophy for safety systems in nuclear devices tends to consider passive technologies. The purpose is to avoid electrical machines like pumps, so the core should remain coolable, even if there is a blackout. To get this natural circulation, heat transfer plays a main role, and as turbulence enhances the heat transfer, it is important to choose a good turbulence model to correctly simulate this ADS's coolant system. (author)
Munoz-cobo, J. L.; Chiva, S.; Pena, C.; Vela, E.
2014-07-01
In this work the development of a methodology is studied to evaluate the uncertainty in the results of CFD codes and is compatible with the VV-20 standard Standard for Verification and Validation in CFD and Heat Transfer {sup ,} developed by the Association of Mechanical Engineers ASME . Similarly, the alternatives are studied for obtaining existing uncertainty in the results to see which is the best choice from the point of view of implementation and time. We have developed two methods for calculating uncertainty of the results of a CFD code, the first method based on the use of techniques of Monte-Carlo for the propagation of uncertainty in this first method we think it is preferable to use the statistics of the order to determine the number of cases to execute the code, because this way we can always determine the confidence interval desired level of output quantities. The second type of method we have developed is based on non-intrusive polynomial chaos. (Author)
Coupled CFD - system-code simulation of a gas cooled reactor
Yan, Yizhou; Rizwan-uddin, E-mail: yizhou.yan@shawgrp.com, E-mail: rizwan@illinois.edu [Department of Nuclear, Plasma and Radiological Engineering, University of Illinois at Urbana-Champaign, IL(United States)
2011-07-01
A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)
Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations
Recently the Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, have developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, two- or three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D. Fluent and RELAP5-3D have strengths that complement one another. CFD codes, such as Fluent, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. On the other hand, RELAP5-3D was developed to analyze the behavior of two-phase systems that could be modeled in one-dimension. Empirical relationships were used where first-principle physics were not well developed. Both Fluent and RELAP5-3D are exemplary in their areas of specialization. The differences between Fluent and RELAP5 fundamentally stem from their field equations. This study focuses on the differences between the momentum equation representations in the two codes (the continuity equation formulations are equivalent for single phase flow). First the differences between the momentum equations are summarized. Next the effect of the differences in the momentum equations are examined by comparing the results obtained using both codes to study the same problem, i.e., fully-developed turbulent pipe flow. Finally, conclusions regarding the significance of the differences are given. (author)
CFD simulation analysis and validation for CPR1000 pressurized water reactor
Background: With the rapid growth in the non-nuclear area for industrial use of Computational fluid dynamics (CFD) which has been accompanied by dramatically enhanced computing power, the application of CFD methods to problems relating to Nuclear Reactor Safety (NRS) is rapidly accelerating. Existing research data have shown that CFD methods could predict accurately the pressure field and the flow repartition in reactor lower plenum. But simulations for the full domain of the reactor have not been reported so far. Purpose: The aim is to determine the capabilities of the codes to model accurately the physical phenomena which occur in the full reactor vessel. Methods: The flow field of the CPR1000 reactor which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX. The pressure loss in reactor pressure vessel, the hydraulic loads of guide tubes and support columns, and the bypass flow of head dome were obtained by calculations for the full domain of the reactor. The results were validated by comparing with the determined reference value of the operating nuclear plant (LingAo nuclear plant), and the transient simulation was conducted in order to better understand the flow in reactor pressure vessel. Results: It was shown that the predicted pressure loss with CFD code was slightly different with the determined value (10% relative deviation for the total pressure loss), the hydraulic loads were less than the determined value with maximum relative deviation 50%, and bypass flow of head dome was approximately the same with determined value. Conclusion: This analysis practice predicts accurately the physical phenomena which occur in the full reactor vessel, and can be taken as a guidance for the nuclear plant design development and improve our understanding of reactor flow phenomena. (authors)
Morghi, Youssef; Mesquita, Amir Zacarias; Santos, Andre Augusto Campagnole dos; Vasconcelos, Victor, E-mail: ymo@cdtn.br, E-mail: amir@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2015-07-01
For the experimental study on the air/water countercurrent flow limitation in Nuclear Reactors, were built at CDTN an acrylic test sections with the same geometric shape of 'hot leg' of a Pressurized Water Reactor (PWR). The hydraulic circuit is designed to be used with air and water at pressures near to atmospheric and ambient temperature. Due to the complexity of the CCFL experimental, the numerical simulation has been used. The aim of the numerical simulations is the validation of experimental data. It is a global trend, the use of computational fluid dynamics (CFD) modeling and prediction of physical phenomena related to heat transfer in nuclear reactors. The most used CFD codes are: FLUENT®, STAR- CD®, Open Foam® and CFX®. In CFD, closure models are required that must be validated, especially if they are to be applied to nuclear reactor safety. The Thermal- Hydraulics Laboratory of CDTN offers computing infrastructure and license to use commercial code CFX®. This article describes a review about CCFL and the use of CFD for numerical simulation of this phenomenal for Nuclear Rector. (author)
For the experimental study on the air/water countercurrent flow limitation in Nuclear Reactors, were built at CDTN an acrylic test sections with the same geometric shape of 'hot leg' of a Pressurized Water Reactor (PWR). The hydraulic circuit is designed to be used with air and water at pressures near to atmospheric and ambient temperature. Due to the complexity of the CCFL experimental, the numerical simulation has been used. The aim of the numerical simulations is the validation of experimental data. It is a global trend, the use of computational fluid dynamics (CFD) modeling and prediction of physical phenomena related to heat transfer in nuclear reactors. The most used CFD codes are: FLUENT®, STAR- CD®, Open Foam® and CFX®. In CFD, closure models are required that must be validated, especially if they are to be applied to nuclear reactor safety. The Thermal- Hydraulics Laboratory of CDTN offers computing infrastructure and license to use commercial code CFX®. This article describes a review about CCFL and the use of CFD for numerical simulation of this phenomenal for Nuclear Rector. (author)
The purpose of the meeting was to develop an Action Plan on the application of CFD to nuclear reactor safety (NRS) problems. This would require to work towards technical consensus on 'Best Practice Guidelines' for CFD use in nuclear reactor safety and on an assessment methodology (validation matrix and associated validation guide) adapted to nuclear reactor safety problems, and to better identify the needs for additional investigations in this field. The meeting was open to organisations interested in participating in the definition of this programme. In order to identify clearly what is available and what is needed, the work should review the following areas: - Identification and classification of the main NRS problems for which CFD has brought or may bring pertinent and useful information. - Inventory of existing CFD methods applicable to NRS problems. - Analysis of specific aspects of these NRS against the existing assessment basis of CFD methods and/or codes. - Inventory of existing CFD guidelines. - Analysis of specific aspects of NRS problems against available guidelines for CFD methods. An Action Plan should be developed in order to: - Adapt/complete existing CFD guidelines and develop reference guidelines for NRS applications. - Define assessment matrices and assessment methodologies suited to NRS applications. - Specify needs regarding additional assessment and developments. - Organise workshops, computational benchmark exercises, ISPs, etc
Preliminary study of coupling CFD code FLUENT and system code RELAP5
Highlights: • System code RELAP5/MOD3.1 is coupled with CFD code FLUENT through DLL and UDF. • Transient water flow in a simple straight tube is tested using the coupled tool. • Simulation of Edwards’ pipe blowdown experiment using the coupled tool is conducted. • Coupled analysis of a more comprehensive thermal–hydraulic system is performed. - Abstract: The present paper discusses a coupling strategy of the 3D (three-dimensional) computational fluid dynamics (CFD) code ANSYS-FLUENT with the best estimate 1D (one-dimensional) thermal–hydraulic system code RELAP5/MOD3.1. Preliminarily, by using DLL (Dynamic Link Library) technology and FLUENT UDF (User Defined Functions), an explicit coupling method expected to be able to support the analysis of multi-purpose thermal–hydraulic phenomena in nuclear reactor systems has been developed. Calculations for two test cases using the coupled FLUENT/RELAP5 code have been carried out to test and demonstrate the coupling capability: (i) the first one consisting of single-phase water transient flow in a square straight tube with well controlled mass flow rates; (ii) the second one illustrating the process of single-phase water flow in a system including two closed loops and one vessel, on which loss of loop water flow due to pump trip and increase of loop water temperature are studied. Both reasonable 1D systematic behaviors and 3D distribution information are naturally obtained for the test cases. Besides, a study of a highly transient experiment problem, i.e. Edwards–O’Brien pipe blowdown problem, has been performed by using the coupled FLUENT/RELAP5 code. The results are compared with standalone RELAP5 calculation and available experimental data, which shows the coupled FLUENT/RELAP5 code’s acceptable potential for the capability of analyzing either simple single-phase or complex two-phase flow problem
Validation of CFD Simulations of Cerebral Aneurysms With Implication of Geometric Variations
Hoi, Yiemeng; Woodward, Scott H.; Kim, Minsuok; Taulbee, Dale B.; Meng, Hui
2006-01-01
Computational fluid dynamics (CFD) simulations using medical-image-based anatomical vascular geometry are now gaining clinical relevance. This study aimed at validating the CFD methodology for studying cerebral aneurysms by using particle image velocimetry (PIV) measurements, with a focus on the effects of small geometric variations in aneurysm models on the flow dynamics obtained with CFD. Method of Approach. An experimental phantom was fabricated out of silicone elastomer to best mimic a sp...
Extending the capabilities of CFD codes to assess ash related problems
Kær, Søren Knudsen; Rosendahl, Lasse Aistrup; Baxter, B. B.
2004-01-01
This paper discusses the application of FLUENT? in theanalysis of grate-fired biomass boilers. A short description of theconcept used to model fuel conversion on the grate and the couplingto the CFD code is offered. The development and implementation ofa CFD-based deposition model is presented in...... the reminder of thepaper. The growth of deposits on furnace walls and super heatertubes is treated including the impact on heat transfer rates determinedby the CFD code. Based on the commercial CFD code FLUENT?,the overall model is fully implemented through the User DefinedFunctions. The model is...... configured entirely through a graphical userinterface integrated in the standard FLUENT? interface. The modelconsiders fine and coarse mode ash deposition and stickingmechanisms for the complete deposit growth, as well as an influenceon the local boundary conditions for heat transfer due to thermalresistance...
Application of CFD code for simulation of an inclined snow chute flow
R K Aggarwal
2013-03-01
Full Text Available In this paper, 2-D simulation of a 61 m long inclined snow chute flow and its interaction with a catch dam type obstacle has been carried out at Dhundhi field research station near Manali, Himachal Pradesh (India using a commercially available computational fluid dynamics (CFD code ANSYS Fluent. Eulerian non-granular multiphase model was chosen to model the snow flow in the surrounding atmospheric air domain. Both air and snow were assumed as laminar and incompressible fluids. User defined functions(UDF were written for the computation of bi-viscous Bingham fluid viscosity and wall shear stress of snow to account for the slip at the interface between the flowing snow and the stationary snow chute surface. Using the proposed CFD model, the velocity, dynamic pressure and debris deposition were simulatedfor flowing snow mass in the chute. Experiments were performed on the snow chute to validate the simulated results. On comparison, the simulated results were found in good agreement with the experimental results.
A validation process for CFD use in building physics – Study of the different length scales
Barbason, Mathieu; Reiter, Sigrid
2011-01-01
Due to growing environmental concerns, Computational Fluid Dynamics (CFD) is more and more used in building physics. Until today, the research community has validated separately several cases but there is no global validation process for this method. The aim of this paper is to provide a way for new users to develop and improve their CFD skills. This paper deals with the different geometry scales involved in building physics. Experimental results are available and will assess the ability of C...
A proposed framework for computational fluid dynamics code calibration/validation
Oberkampf, W.L.
1993-12-31
The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ``calibrated code,`` ``validated code,`` and a ``validation experiment`` is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance.
A proposed framework for computational fluid dynamics code calibration/validation
The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ''calibrated code,'' ''validated code,'' and a ''validation experiment'' is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance
Simulation of natural circulation in a rectangular loop using CFD code PHOENICS
Kumar, M.; Borghain, A.; Maheshwari, N.K.; Vijayan, P.K. [Bhabha Atomic Reseach Centre, Trombay, Mumbai (India). Reactor Engineering Div.
2011-05-15
Single phase natural circulation in a rectangular loop is simulated using the PHOENICS code, a general purpose Computational Fluid Dynamics (CFD) code. The rectangular loop, having different operating power levels, has been modeled with the help of the Multiple Block Fine Grid Embedment (MBFGE) technique. The Co-located Co-variant Method (CCM) is used to simulate this loop in PHOENICS. Extensive experimental and CFD studies have been conducted on single phase natural circulation in a rectangular loop. The paper presents the results of three-dimensional CFD analysis for the prediction of steady state behavior in a rectangular loop and its comparison with experimental data. The results of code prediction and readily available experimental data show good agreement. (orig.)
A CFD Validation of Fire Dynamics Simulator for Corner Fire
Pavan K. Sharma
2010-12-01
Full Text Available A computational study has been carried out for predicting the behaviour of a corner fire source for a reported experiment using a field model based code Fire Dynamics Simulator (FDS. Time dependent temperature is predicted along with the resulting changes in the plume structure. The flux falling on the wall was also observed. The analysis has been carried out with the correct value of the grid size based on earlier experiences and also by performing a grid sensitivity study. The predicted temperatures of the two scenarios at two points by the current analysis are in very good agreement with the earlier reported experimental data and numerical prediction. The studies have extended the utility of field model based tools to model the particular separate effect phenomenon like corner for one such situation and validate against experimental data. The present study have several applications in such as room fires, hydrogen transport in nuclear reactor containment, natural convection in building flows etc. The present approach uses the advanced Large Eddy Simulation (LES based CFD turbulence model. The paper presents brief description of the code FDS, details of the computational model along with the discussions on the results obtained under these studies. The validated CFD based procedure has been used for solving various problems enclosure fire, ventilated fire and open fire from nuclear industry which are however not included in the present paper.
CFD validation for flyash particle classification in hydrocyclones
K. Udaya Bhaskar; Y. Rama Murthy; N. Ramakrishnan; J.K. Srivastava; Supriya Sarkar; Vimal Kumar [Regional Research Laboratory (CSIR), Bhopal (India)
2007-03-15
The investigation pertains to establishing a simulation methodology for understanding the flyash classification characteristics of a 76 and 50 mm diameter hydrocyclone where the work was carried out using commercially available CFD software. Comparative results on the simulated and experimental water throughput, split values are presented. Results indicted that there is a good match in water split between the experimental and simulated values with error values below 10% at different hydrocyclone designs. Further a discussion is made on the flow features at comparable ratio of cyclone diameter to spigot opening in the 76 and 50 mm designs. Classification of flyash particulates is simulated through discrete phase modeling using particles injection technique and the simulated results are further validated with suitably performed experiments. With 50 mm diameter hydrocyclone, reasonable predictions are observed at 9.4 mm spigot opening. Considerable deviation in particle distribution points with This hydrocyclone is observed at narrowest spigot diameter of 3.2 mm. The simulated values of d{sub 50} in case of 50 mm diameter hydrocyclone are 8 and 10 {mu}m at 9.4 and 3.2 mm diameter spigot openings. Better predictions are obtained with 76 mm diameter hydrocyclone at both 10 and 15 mm diameter spigot openings. Similarly, the simulated d{sub 50} values are 14 and 20 {mu}m at 15 and 10 mm diameter hydrocyclones. Possible reasons for deviations in the results relating the spigot opening, solids concentration at the underflow and in turn role of slurry viscosity on the air core diameter are proposed.
Development and validation of a CFD model predicting the backfill process of a nuclear waste gallery
Gopala, Vinay Ramohalli, E-mail: gopala@nrg.eu [Nuclear Research and consultancy Group (NRG), P.O. Box 25, 1755 ZG Petten (Netherlands); Lycklama a Nijeholt, Jan-Aiso [Nuclear Research and consultancy Group (NRG), P.O. Box 25, 1755 ZG Petten (Netherlands); Bakker, Paul [Van Hattum en Blankevoort, Woerden (Netherlands); Haverkate, Benno [Nuclear Research and consultancy Group (NRG), P.O. Box 25, 1755 ZG Petten (Netherlands)
2011-07-15
Research highlights: > This work presents the CFD simulation of the backfill process of Supercontainers with nuclear waste emplaced in a disposal gallery. > The cement-based material used for backfill is grout and the flow of grout is modelled as a Bingham fluid. > The model is verified against an analytical solution and validated against the flowability tests for concrete. > Comparison between backfill plexiglas experiment and simulation shows a distinct difference in the filling pattern. > The numerical model needs to be further developed to include segregation effects and thixotropic behavior of grout. - Abstract: Nuclear waste material may be stored in underground tunnels for long term storage. The example treated in this article is based on the current Belgian disposal concept for High-Level Waste (HLW), in which the nuclear waste material is packed in concrete shielded packages, called Supercontainers, which are inserted into these tunnels. After placement of the packages in the underground tunnels, the remaining voids between the packages and the tunnel lining is filled-up with a cement-based material called grout in order to encase the stored containers into the underground spacing. This encasement of the stored containers inside the tunnels is known as the backfill process. A good backfill process is necessary to stabilize the waste gallery against ground settlements. A numerical model to simulate the backfill process can help to improve and optimize the process by ensuring a homogeneous filling with no air voids and also optimization of the injection positions to achieve a homogeneous filling. The objective of the present work is to develop such a numerical code that can predict the backfill process well and validate the model against the available experiments and analytical solutions. In the present work the rheology of Grout is modelled as a Bingham fluid which is implemented in OpenFOAM - a finite volume-based open source computational fluid dynamics
Development and validation of a CFD model predicting the backfill process of a nuclear waste gallery
Research highlights: → This work presents the CFD simulation of the backfill process of Supercontainers with nuclear waste emplaced in a disposal gallery. → The cement-based material used for backfill is grout and the flow of grout is modelled as a Bingham fluid. → The model is verified against an analytical solution and validated against the flowability tests for concrete. → Comparison between backfill plexiglas experiment and simulation shows a distinct difference in the filling pattern. → The numerical model needs to be further developed to include segregation effects and thixotropic behavior of grout. - Abstract: Nuclear waste material may be stored in underground tunnels for long term storage. The example treated in this article is based on the current Belgian disposal concept for High-Level Waste (HLW), in which the nuclear waste material is packed in concrete shielded packages, called Supercontainers, which are inserted into these tunnels. After placement of the packages in the underground tunnels, the remaining voids between the packages and the tunnel lining is filled-up with a cement-based material called grout in order to encase the stored containers into the underground spacing. This encasement of the stored containers inside the tunnels is known as the backfill process. A good backfill process is necessary to stabilize the waste gallery against ground settlements. A numerical model to simulate the backfill process can help to improve and optimize the process by ensuring a homogeneous filling with no air voids and also optimization of the injection positions to achieve a homogeneous filling. The objective of the present work is to develop such a numerical code that can predict the backfill process well and validate the model against the available experiments and analytical solutions. In the present work the rheology of Grout is modelled as a Bingham fluid which is implemented in OpenFOAM - a finite volume-based open source computational fluid
On application of CFD codes to problems of nuclear reactor safety
The 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in May 2002 at Aix-en-Province, France, recommended formation of writing groups to report the need of guidelines for use and assessment of CFD in single-phase nuclear reactor safety problems, and on recommended extensions to CFD codes to meet the needs of two-phase problems in nuclear reactor safety. This recommendations was supported also by Working Group on the Analysis and Management of Accidents and led to formation oaf three Writing Groups. The first writing Group prepared a summary of existing best practice guidelines for single phase CFD analysis and made a recommendation on the need for nuclear reactor safety specific guidelines. The second Writing Group selected those nuclear reactor safety applications for which understanding requires or is significantly enhanced by single-phase CFD analysis, and proposed a methodology for establishing assesment matrices relevant to nuclear reactor safety applications. The third writing group performed a classification of nuclear reactor safety problems where extension of CFD to two-phase flow may bring real benefit, a classification of different modeling approaches, and specification and analysis of needs in terms of physical and numerical assessments. This presentation provides a review of these activities with the most important conclusions and recommendations (Authors)
Ferreira, E.; Alves, E.; Ferreira, R. M. L.
2012-04-01
Sediment deposition by continuous turbidity currents may affect eco-environmental river dynamics in natural reservoirs and hinder the maneuverability of bottom discharge gates in dam reservoirs. In recent years, innovative techniques have been proposed to enforce the deposition of turbidity further upstream in the reservoir (and away from the dam), namely, the use of solid and permeable obstacles such as water jet screens , geotextile screens, etc.. The main objective of this study is to validate a computational fluid dynamics (CFD) code applied to the simulation of the interaction between a turbidity current and a passive retention system, designed to induce sediment deposition. To accomplish the proposed objective, laboratory tests were conducted where a simple obstacle configuration was subjected to the passage of currents with different initial sediment concentrations. The experimental data was used to build benchmark cases to validate the 3D CFD software ANSYS-CFX. Sensitivity tests of mesh design, turbulence models and discretization requirements were performed. The validation consisted in comparing experimental and numerical results, involving instantaneous and time-averaged sediment concentrations and velocities. In general, a good agreement between the numerical and the experimental values is achieved when: i) realistic outlet conditions are specified, ii) channel roughness is properly calibrated, iii) two equation k - ɛ models are employed iv) a fine mesh is employed near the bottom boundary. Acknowledgements This study was funded by the Portuguese Foundation for Science and Technology through the project PTDC/ECM/099485/2008. The first author thanks the assistance of Professor Moitinho de Almeida from ICIST and to all members of the project and of the Fluvial Hydraulics group of CEHIDRO.
Validation of Neptune-CFD Module with Data of a Plunging Water Jet Entering a Free Surface
Galassi, M.C.; D' Auria, F. [Univ Pisa, DIMNP, Pisa, (Italy); Bestion, D.; Morel, C.; Pouvreau, J. [CEA, DEN DER SSTH, Grenoble, (France)
2009-07-01
This work presents a validation of NEPTUNE-CFD against plunging water jet experiments by Iguchi et al., with sensitivity tests to turbulence modeling. NEPTUNE-CFD is the thermal-hydraulic two-phase computational fluid dynamics tool of NURESIM (European Platform for Nuclear Reactor Simulations) and is designed to simulate two-phase flow in situations encountered in nuclear power plants. Iguchi et al.'s flow configuration shares common physical features with the emergency core cooling injection in a pressurized water reactor uncovered cold leg during a small-break loss-of-coolant accident. This work contributes to the validation of the NEPTUNE-CFD code capability to predict the turbulence below a free surface produced by a plunging jet. In the experiment, the water was injected vertically, down a straight circular pipe into a cylindrical vessel containing water. Mean velocity and turbulent fluctuations were measured below the jet at several depths below the free surface. The influence of several models on code predictions was investigated, and both standard and modified turbulence models were tested. A single-phase jet case was also simulated and compared with both measurements and two-phase calculations, to investigate bubble entrainment influence on turbulence prediction. The calculated mean velocity field was always in quite good agreement with the experimental data, while the turbulence intensity was generally good with some underestimation far from the jet axis region. (authors)
Optimization of an industrial heat exchanger using an open-source CFD code
The objective of the present study is to develop an optimized heat pipe exchanger used to improve the energy efficiency in building ventilation systems. The optimized design is based on a validated model used inside a numerical plan built on a design of experiments statistical procedure. The numerical model, built using the open-source package OpenFOAM, is validated through experimental measurements done on a small scale heat pipe industrial exchanger. The results from the open source model are also compared to the numerical predictions obtained from a commercial code. Modelling results show good agreement with experiment measurements, thus showing the great potential of the model as a tool for heat pipe engineering design. The results are analysed in terms of efficiency for different configurations. - Highlights: •Development of solvers using open-source CFD package OpenFOAM. •Optimization of air exchangers using a test bench. •Calculation of pressure drop and heat transfer coefficient. •We increased thermal efficiency and a very good performance was obtained
Thermal hydraulic system codes have been extensively developed by the nuclear industry, research institutes and technical safety organizations with the goal to improve the design and safety of nuclear installations. A large number of these simulation tools are based on the lumped parameter theory. Such programs use networks consisting of 1D cells, where mass, momentum and energy equations are solved for each fluid phase and balanced over each node of the network. System codes are extensively validated against experiments and provide reliable results at low computational cost. Lump parameter programs use simplifications in the mathematical models describing the simulated systems. Balance equations for mass, momentum and energy for two phases are obtained by averaging of the local basic flow equations in the space. As a result, mean values for relevant physical parameters which in reality are spatially distributed fields are calculated. However, since relevant reactor fluid flow and heat transfer phenomena are 3D in nature, 1D system codes have limitations on their application for specific nuclear reactor safety (NRS) problems with pronounced 3D phenomena like boron dilution, pressurized thermal shock and main steam line break. Modern computational fluid dynamics (CFD) codes are capable to predict fluid flow behavior in complex geometries and can provide detailed distribution of the physical parameters in the space. Unfortunately, CFD simulations require very high computation time and hence full representation of the primary circuit of a PWR is currently not feasible. In order to overcome the deficiencies of CFD and system codes, different approaches are used by the scientists dealing with complex fluid flows. (orig.)
Papukchiev, Angel; Lerchl, Georg [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany)
2010-05-15
Thermal hydraulic system codes have been extensively developed by the nuclear industry, research institutes and technical safety organizations with the goal to improve the design and safety of nuclear installations. A large number of these simulation tools are based on the lumped parameter theory. Such programs use networks consisting of 1D cells, where mass, momentum and energy equations are solved for each fluid phase and balanced over each node of the network. System codes are extensively validated against experiments and provide reliable results at low computational cost. Lump parameter programs use simplifications in the mathematical models describing the simulated systems. Balance equations for mass, momentum and energy for two phases are obtained by averaging of the local basic flow equations in the space. As a result, mean values for relevant physical parameters which in reality are spatially distributed fields are calculated. However, since relevant reactor fluid flow and heat transfer phenomena are 3D in nature, 1D system codes have limitations on their application for specific nuclear reactor safety (NRS) problems with pronounced 3D phenomena like boron dilution, pressurized thermal shock and main steam line break. Modern computational fluid dynamics (CFD) codes are capable to predict fluid flow behavior in complex geometries and can provide detailed distribution of the physical parameters in the space. Unfortunately, CFD simulations require very high computation time and hence full representation of the primary circuit of a PWR is currently not feasible. In order to overcome the deficiencies of CFD and system codes, different approaches are used by the scientists dealing with complex fluid flows. (orig.)
Validation of Boundary Conditions for CFD Simulations on Ventilated Rooms
Topp, Claus; Jensen, Rasmus Lund; Pedersen, D.N.;
2001-01-01
The application of Computational Fluid Dynamics (CFD) for ventilation research and design of ventilation systems has increased during the recent years. This paper provides an investigation of direct description of boundary conditions for a complex inlet diffuser and a heated surface. A series of ...
Application of CFD Codes in Nuclear Reactor Safety Analysis
T. Höhne
2010-01-01
Full Text Available Computational Fluid Dynamics (CFD is increasingly being used in nuclear reactor safety (NRS analyses as a tool that enables safety relevant phenomena occurring in the reactor coolant system to be described in more detail. Numerical investigations on single phase coolant mixing in Pressurised Water Reactors (PWR have been performed at the FZD for almost a decade. The work is aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. For the experimental investigation of horizontal two phase flows, different non pressurized channels and the TOPFLOW Hot Leg model in a pressure chamber was build and simulated with ANSYS CFX. In a common project between the University of Applied Sciences Zittau/Görlitz and FZD the behaviour of insulation material released by a LOCA released into the containment and might compromise the long term emergency cooling systems is investigated. Moreover, the actual capability of CFD is shown to contribute to fuel rod bundle design with a good CHF performance.
Hardman, R. R.; Mahan, J. R.; Smith, M. H.; Gelhausen, P. A.; Van Dalsem, W. R.
1991-01-01
The need for a validation technique for computational fluid dynamics (CFD) codes in STOVL applications has led to research efforts to apply infrared thermal imaging techniques to visualize gaseous flow fields. Specifically, a heated, free-jet test facility was constructed. The gaseous flow field of the jet exhaust was characterized using an infrared imaging technique in the 2 to 5.6 micron wavelength band as well as conventional pitot tube and thermocouple methods. These infrared images are compared to computer-generated images using the equations of radiative exchange based on the temperature distribution in the jet exhaust measured with the thermocouple traverses. Temperature and velocity measurement techniques, infrared imaging, and the computer model of the infrared imaging technique are presented and discussed. From the study, it is concluded that infrared imaging techniques coupled with the radiative exchange equations applied to CFD models are a valid method to qualitatively verify CFD codes used in STOVL applications.
Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The entire collection of papers is provided in this report
This paper presents an unstructured mesh based multi-physics interface implemented in the Serpent 2 Monte Carlo code, for the purpose of coupling the neutronics solution to component-scale thermal hydraulics calculations, such as computational fluid dynamics (CFD). The work continues the development of a multi-physics coupling scheme, which relies on the separation of state-point information from the geometry input, and the capability to handle temperature and density distributions by a rejection sampling algorithm. The new interface type is demonstrated by a simplified molten-salt reactor test case, using a thermal hydraulics solution provided by the CFD solver in OpenFOAM. (author)
Implementation of CFD module in the KORSAR thermal-hydraulic system code
Yudov, Yury V.; Danilov, Ilia G.; Chepilko, Stepan S. [Alexandrov Research Inst. of Technology (NITI), Sosnovy Bor (Russian Federation)
2015-09-15
The Russian KORSAR/GP (hereinafter KORSAR) computer code was developed by a joint team from Alexandrov NITI and OKB ''Gidropress'' for VVER safety analysis and certified by the Rostechnadzor of Russia in 2009. The code functionality is based on a 1D two-fluid model for calculation of two-phase flows. A 3D CFD module in the KORSAR computer code is being developed by Alexandrov NITI for representing 3D effects in the downcomer and lower plenum during asymmetrical loop operation. The CFD module uses Cartesian grid method with cut cell approach. The paper presents a numerical algorithm for coupling 1D and 3D thermal- hydraulic modules in the KORSAR code. The combined pressure field is calculated by the multigrid method. The performance efficiency of the algorithm for coupling 1D and 3D modules was demonstrated by solving the benchmark problem of mixing cold and hot flows in a T-junction.
Acceleration of a CFD Code with a GPU
Dennis C. Jespersen
2010-01-01
Full Text Available The Computational Fluid Dynamics code OVERFLOW includes as one of its solver options an algorithm which is a fairly small piece of code but which accounts for a significant portion of the total computational time. This paper studies some of the issues in accelerating this piece of code by using a Graphics Processing Unit (GPU. The algorithm needs to be modified to be suitable for a GPU and attention needs to be given to 64-bit and 32-bit arithmetic. Interestingly, the work done for the GPU produced ideas for accelerating the CPU code and led to significant speedup on the CPU.
Hypersonic Intake Starting Characteristics–A CFD Validation Study
Soumyajit Saha; Debasis Chakraborty
2012-01-01
Numerical simulation of hypersonic intake starting characteristics is presented. Three dimensional RANS equations are solved alongwith SST turbulence model using commercial computational fluid dynamics (CFD) software. Wall pressure distribution and intake performance parameters are found to match well with experimental data for different free stream Mach number in the range of 3-8. The unstarting of the intake is traced from the sudden drop of mass capture ratio. Wall condition (adiabatic or ...
Validation of a FLUENT CFD model for hydrogen distribution in a containment
Highlights: ► NRG developed a CFD model to simulate the hydrogen distribution in the containment during a severe accident. ► The containment model is validated for the formation and break-up of a stable hydrogen-rich layer. ► Guidelines are obtained on mesh resolution, near-wall treatment and turbulence modeling. - Abstract: Hydrogen may be released into the containment atmosphere of a nuclear power plant during a severe accident. Locally, high hydrogen concentrations may be reached that can possibly cause fast deflagration or even detonation and put the integrity of the containment at risk. The distribution and mixing of hydrogen is, therefore, an important safety issue for nuclear power plants. Computational fluid dynamics (CFD) codes can be applied to predict the hydrogen distribution in the containment within the course of a hypothetical severe accident and get an estimate of the local hydrogen concentration in the various zones of the containment. In this way the risk associated with the hydrogen safety issue can be determined, and safety related measurements and procedures could be assessed. In order to further validate the CFD containment model of NRG in the context of hydrogen distribution in the containment of a nuclear power plant, the HM-2 test performed in the German THAI (thermal-hydraulics, hydrogen, aerosols and iodine) facility is selected. In the first phase of the HM-2 test a stratified hydrogen-rich light gas layer was established in the upper part of the THAI containment. In the second phase steam was injected at a lower position. This induced a rising plume that gradually dissolved the stratified hydrogen-rich layer from below. Phenomena that are expected in severe accidents, like natural convection, turbulent mixing, condensation, heat transfer and distribution in different compartments, are simulated in this hypothetical severe accident scenario. The hydrogen distribution and associated physical phenomena monitored during the HM-2 test
Rong, Li; Nielsen, Peter Vilhelm; Bjerg, Bjarne;
2016-01-01
scale pig barns was simulated to show the procedures of validating a CFD simulation in livestock buildings. After summarizing the guideline and/or best practice for CFD modeling, the authors addressed the issues related to numerical methods and the governing equations, which were limited to RANS models......, simulating domain etc. This information is particularly important for the readers to evaluate the quality of the CFD simulation results.......Computational Fluid Dynamics (CFD) is increasingly used to study airflow around and in livestock buildings, to develop technologies to mitigate emissions and to predict the contaminant dispersion from livestock buildings. In this paper, an example of air flow distribution in a room with two full...
Validation of 3-D CFD Model of Tritium Transport in the Atmosphere
When solving 3-D problems for the atmospheric impurity transport in the bounded area, it is essential for the atmospheric dynamics to be correctly computed taking into account the actual terrain topography and environments specified by the boundary conditions. Such conditions as turbulence, convection, condensation and moisture evaporation processes, etc. are to be also taken into account as well as the interaction processes among impurities (gases, aerosols), atmosphere and the Earth's surface.3-D computational fluid dynamics model(CFD) developed on the basis of SRP hydrodynamic code was used to simulate tritium plume evolution and tritium transport in atmosphere under the area with relatively complex topography. SRP code is based on the continuum motion equations (Navier-Stockes equations) and thermodynamic relations taking into account specific features of atmospheric flows and complex topography and is designed to use on PC-type computers.The model has been validated using experimental release of tritium with specified source term and meteorology. Due to low release height above the underlying surface a fine grid was used in the vertical direction near the underlying surface. HT and HTO/H2O vertical fluxes were taken into account. Evolution of HT and HTO activities at 2 sampling locations along the plume axe were available for model-experiment inter-comparison. The modeling results of HT and HTO activities in the air during plume travel are in satisfactory agreement with observed values
Aeschliman, D. P.; Oberkampf, W. L.; Blottner, F. G.
Verification, calibration, and validation (VCV) of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. The exact manner in which code VCV activities are planned and conducted, however, is critically important. It is suggested that the way in which code validation, in particular, is often conducted--by comparison to published experimental data obtained for other purposes--is in general difficult and unsatisfactory, and that a different approach is required. This paper describes a proposed methodology for CFD code VCV that meets the technical requirements and is philosophically consistent with code development needs. The proposed methodology stresses teamwork and cooperation between code developers and experimentalists throughout the VCV process, and takes advantage of certain synergisms between CFD and experiment. A novel approach to uncertainty analysis is described which can both distinguish between and quantify various types of experimental error, and whose attributes are used to help define an appropriate experimental design for code VCV experiments. The methodology is demonstrated with an example of laminar, hypersonic, near perfect gas, 3-dimensional flow over a sliced sphere/cone of varying geometrical complexity.
Simulation of two-phase flows in vertical tubes with the CFD code FLUBOX
The Computational Fluid Dynamics (CFD) code FLUBOX is developed at GRS for the multidimensional simulation of two-phase flows. The single-pressure two-fluid model is used as basis of the simulation. A basic mathematical property of the two-fluid model of FLUBOX is the hyperbolic character of the convection. The numerical solution methods of FLUBOX make explicit use of the hyperbolic structure of the coefficient matrices. The simulation of two-phase flow phenomena needs, apart from the conservation equations for each phase, an additional transport equation for the interfacial area concentration. The concentration of the interfacial area is one of the key parameters for the modelling of interfacial friction forces and interfacial transfer terms. A new transport equation for the interfacial area concentration is in development. It describes the dynamic change of the interfacial area concentration due to mass exchange and a force balance at the phase boundary. Results from FLUBOX calculations for different experiments of two-phase flows in vertical tubes are presented as part of the validation. (authors)
Christian F. Janßen
2015-07-01
Full Text Available This contribution is dedicated to demonstrating the high potential and manifold applications of state-of-the-art computational fluid dynamics (CFD tools for free-surface flows in civil and environmental engineering. All simulations were performed with the academic research code ELBE (efficient lattice boltzmann environment, http://www.tuhh.de/elbe. The ELBE code follows the supercomputing-on-the-desktop paradigm and is especially designed for local supercomputing, without tedious accesses to supercomputers. ELBE uses graphics processing units (GPU to accelerate the computations and can be used in a single GPU-equipped workstation of, e.g., a design engineer. The code has been successfully validated in very different fields, mostly related to naval architecture and mechanical engineering. In this contribution, we give an overview of past and present applications with practical relevance for civil engineers. The presented applications are grouped into three major categories: (i tsunami simulations, considering wave propagation, wave runup, inundation and debris flows; (ii dam break simulations; and (iii numerical wave tanks for the calculation of hydrodynamic loads on fixed and moving bodies. This broad range of applications in combination with accurate numerical results and very competitive times to solution demonstrates that modern CFD tools in general, and the ELBE code in particular, can be a helpful design tool for civil and environmental engineers.
The interfacial area transport equation for the subcooled boiling flow was developed with a mechanistic model for the wall boiling source term. It included the bubble lift-off diameter model and lift-off frequency reduction factor model. To implement the model, the two-phase flow CFD code was developed, which was named as EAGLE (Elaborated Analysis of Gas-Liquid Evolution). The developed model and EAGLE code was validated the experimental data of SUBO and SNU facilities. The computational analysis revealed that the interfacial area transport equation with the bubble lift-off diameter model agreed well with the experimental results. It presents that the source term for the wall nucleation enhanced the prediction capability for a multi-dimensional behavior of void fraction or interfacial area concentration
The interfacial area transport equation for a subcooled boiling flow is developed with a mechanistic model for the wall boiling source term. It includes the bubble lift-off diameter model and the lift-off frequency reduction factor model. Those models take into account a bubble's sliding on the heated wall after a departure from the nucleate site and the coalescences of sliding bubbles. To implement the model, the two-phase flow CFD code was developed, which is named as EAGLE (Elaborated Analysis of Gas-Liquid Evolution). The developed model and EAGLE code are validated by the experimental data of SUBO (Subcooled Boiling) facility. The computational analysis reveals that the interfacial area transport equation with the bubble lift-off diameter model agrees well with the experimental results. It presents that the source term for the wall nucleation enhanced the prediction capability for a multidimensional behavior of void fraction or interfacial area concentration. (authors)
Hypersonic Intake Starting Characteristics–A CFD Validation Study
Soumyajit Saha
2012-05-01
Full Text Available Numerical simulation of hypersonic intake starting characteristics is presented. Three dimensional RANS equations are solved alongwith SST turbulence model using commercial computational fluid dynamics (CFD software. Wall pressure distribution and intake performance parameters are found to match well with experimental data for different free stream Mach number in the range of 3-8. The unstarting of the intake is traced from the sudden drop of mass capture ratio. Wall condition (adiabatic or isothermal is seen to have pronounced effect in estimating the performance parameters in the intake. The computed unstarting Mach number is seen to be higher for adiabatic condition compared to isothermal condition. For unstarting case, large separation bubble is seen near the entrance of the intake, which is responsible for expulsion of the shock system out of the intake.Defence Science Journal, 2012, 62(1, pp.147-152, DOI:http://dx.doi.org/10.14429/dsj.62.1340
Application of CFD code for simulation of an inclined snow chute flow
Aggarwal, R K; Amod Kumar
2013-01-01
In this paper, 2-D simulation of a 61 m long inclined snow chute flow and its interaction with a catch dam type obstacle has been carried out at Dhundhi field research station near Manali, Himachal Pradesh (India) using a commercially available computational fluid dynamics (CFD) code ANSYS Fluent. Eulerian non-granular multiphase model was chosen to model the snow flow in the surrounding atmospheric air domain. Both air and snow were assumed as laminar and incompressible fluids. User defined ...
CFD modelling and validation of wall condensation in the presence of non-condensable gases
Zschaeck, G., E-mail: guillermo.zschaeck@ansys.com [ANSYS Germany GmbH, Staudenfeldweg 12, Otterfing 83624 (Germany); Frank, T. [ANSYS Germany GmbH, Staudenfeldweg 12, Otterfing 83624 (Germany); Burns, A.D. [ANSYS UK Ltd, 97 Milton Park, Abingdon, Oxfordshire OX14 4RY (United Kingdom)
2014-11-15
Highlights: • A wall condensation model was implemented and validated in ANSYS CFX. • Condensation rate is assumed to be controlled by the concentration boundary layer. • Validation was done using two laboratory scale experiments. • CFD calculations show good agreement with experimental data. - Abstract: The aim of this paper is to present and validate a mathematical model implemented in ANSYS CFD for the simulation of wall condensation in the presence of non-condensable substances. The model employs a mass sink at isothermal walls or conjugate heat transfer (CHT) domain interfaces where condensation takes place. The model was validated using the data reported by Ambrosini et al. (2008) and Kuhn et al. (1997)
CFD modelling and validation of wall condensation in the presence of non-condensable gases
Highlights: • A wall condensation model was implemented and validated in ANSYS CFX. • Condensation rate is assumed to be controlled by the concentration boundary layer. • Validation was done using two laboratory scale experiments. • CFD calculations show good agreement with experimental data. - Abstract: The aim of this paper is to present and validate a mathematical model implemented in ANSYS CFD for the simulation of wall condensation in the presence of non-condensable substances. The model employs a mass sink at isothermal walls or conjugate heat transfer (CHT) domain interfaces where condensation takes place. The model was validated using the data reported by Ambrosini et al. (2008) and Kuhn et al. (1997)
Rocket-Based Combined Cycle Engine Technology Development: Inlet CFD Validation and Application
DeBonis, J. R.; Yungster, S.
1996-01-01
A CFD methodology has been developed for inlet analyses of Rocket-Based Combined Cycle (RBCC) Engines. A full Navier-Stokes analysis code, NPARC, was used in conjunction with pre- and post-processing tools to obtain a complete description of the flow field and integrated inlet performance. This methodology was developed and validated using results from a subscale test of the inlet to a RBCC 'Strut-Jet' engine performed in the NASA Lewis 1 x 1 ft. supersonic wind tunnel. Results obtained from this study include analyses at flight Mach numbers of 5 and 6 for super-critical operating conditions. These results showed excellent agreement with experimental data. The analysis tools were also used to obtain pre-test performance and operability predictions for the RBCC demonstrator engine planned for testing in the NASA Lewis Hypersonic Test Facility. This analysis calculated the baseline fuel-off internal force of the engine which is needed to determine the net thrust with fuel on.
Pollutant Emission Validation of a Heavy-Duty Gas Turbine Burner by CFD Modeling
Roberto Meloni
2013-10-01
Full Text Available 3D numerical combustion simulation in a can burner fed with methane was carried out in order to evaluate pollutant emissions and the temperature field. As a case study, the General Electric Frame 6001B system was considered. The numerical investigation has been performed using the CFD code named ACE+ Multiphysics (by Esi-Group. The model was validated against the experimental data provided by Cofely GDF SUEZ and related to a real power plant. To completely investigate the stability of the model, several operating conditions were taken into account, at both nominal and partial load. In particular, the influence on emissions of some important parameters, such as air temperature at compressor intake and steam to fuel mass ratio, have been evaluated. The flamelet model and Zeldovich’s mechanism were employed for combustion modeling and NOx emissions, respectively. With regard to CO estimation, an innovative approach was used to compute the Rizk and Mongia relationship through a user-defined function. Numerical results showed good agreement with experimental data in most of the cases: the best results were obtained in the NOx prediction, while unburned fuel was slightly overestimated.
Assessment of CFD Codes for Nuclear Reactor Safety Problems - Revision 2
Following recommendations made at an 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety (NRS) Problems', held in Aix-en-Provence, France, 15-16 May, 2002, and a follow-up meeting 'Use of Computational Fluid Dynamics (CFD) Codes for Safety Analysis of Reactor Systems including Containment', which took place in Pisa on 11-14 Nov., 2002, a CSNI action plan was drawn up which resulted in the creation of three Writing Groups, with mandates to perform the following tasks: (1) Provide a set of guidelines for the application of CFD to NRS problems; (2) Evaluate the existing CFD assessment bases, and identify gaps that need to be filled; (3) Summarise the extensions needed to CFD codes for application to two-phase NRS problems. Work began early in 2003. In the case of Writing Group 2 (WG2), a preliminary report was submitted to Working Group on the Analysis and Management of Accidents (WGAMA) in September 2004 that scoped the work needed to be carried out to fulfil its mandate, and made recommendations on how to achieve the objective. A similar procedure was followed by the other two groups, and in January 2005 all three groups were reformed to carry out their respective tasks. In the case of WG2, this resulted in the issue of a CSNI report (NEA/CSNI/R(2007)13), issued in January 2008, describing the work undertaken. The writing group met on average twice per year during the period March 2005 to May 2007, and coordinated activities strongly with the sister groups WG1 (Best Practice Guidelines) and WG3 (Multiphase Extensions). The resulting document prepared at the end of this time still represents the core of the present revised version, though updates have been made as new material has become available. After some introductory remarks, Chapter 3 lists twenty-three (23) NRS issues for which it is considered that the application of CFD would bring real benefits
Two-Phase Flow Simulations for PTS Investigation by Means of Neptune_CFD Code
Fabio Moretti; Maria Cristina Galassi; Pierre Coste; Christophe Morel
2009-01-01
Two-dimensional axisymmetric simulations of pressurized thermal shock (PTS) phenomena through Neptune_CFD module are presented aiming at two-phase models validation against experimental data. Because of PTS complexity, only some thermal-hydraulic aspects were considered. Two different flow configurations were studied, occurring when emergency core cooling (ECC) water is injected in an uncovered cold leg of a pressurized water reactor (PWR)Ã¢Â€Â”a plunging water jet entering a free surface, an...
Barrera, J.
2011-07-01
This article explores a Jet Pump in reactor type BWR-3 using the CFD STAR-CCM +, aiming to compare the various options presenting the code and analyze its impact on the quality of the results, compared with the theoretical value of design.
Verification of the CFD code FLUENT by post test calculation of ROCOM experiments
Full text of publication follows: The TUV NORD e.V. is an independent Technical Support Organisation (TSO) performing safety assessments in almost every field of technology. In nuclear safety the TUV can look back on more than 40 years of experience. In the last years in Germany PWR safety analyses were focussed on boron dilution events with the potential of reactivity transients. The possibility of coolant with a low boron concentration collected in localized areas of the reactor coolant system (RCS) can be caused by injection of coolant with less boron content from interfacing systems (external dilution) as well as separation of borated reactor coolant into highly concentrated and diluted fractions (inherent dilution). Inherent dilution can e.g. occur after reflux-condenser heat transfer after a small break loss of coolant accident (SBLOCA) with a limited operability of the emergency core cooling (ECC) systems. The TUV Nord e.V. was charged by German supervisory authorities with the assessment of the safety analyses presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The simulation of boron dilution and transport processes in PWR reactor coolant systems (RCS) and especially reactor pressure vessels (RPV) requires the application of computational fluid dynamic (CFD) codes. At present the validation of these codes is performed by post test calculations of boron dilution experiments e.g. Rossendorf Coolant Mixing Model (ROCOM). They were chosen by TUV Nord e.V. for validation of FLUENT, because of the excellent experimental data base, especially the high spatial and temporal resolution measurements of boron concentration distribution with wire mesh sensors. The ROCOM facility was built at the Forschungszentrum Rossendorf e.V. near Dresden in linear scale of 1:5 for the investigation of coolant mixing in a wide range of PWR flow conditions. ROCOM is a
Production Level CFD Code Acceleration for Hybrid Many-Core Architectures
Duffy, Austen C.; Hammond, Dana P.; Nielsen, Eric J.
2012-01-01
In this work, a novel graphics processing unit (GPU) distributed sharing model for hybrid many-core architectures is introduced and employed in the acceleration of a production-level computational fluid dynamics (CFD) code. The latest generation graphics hardware allows multiple processor cores to simultaneously share a single GPU through concurrent kernel execution. This feature has allowed the NASA FUN3D code to be accelerated in parallel with up to four processor cores sharing a single GPU. For codes to scale and fully use resources on these and the next generation machines, codes will need to employ some type of GPU sharing model, as presented in this work. Findings include the effects of GPU sharing on overall performance. A discussion of the inherent challenges that parallel unstructured CFD codes face in accelerator-based computing environments is included, with considerations for future generation architectures. This work was completed by the author in August 2010, and reflects the analysis and results of the time.
Integration of CFD codes and advanced combustion models for quantitative burnout determination
Javier Pallares; Inmaculada Arauzo; Alan Williams [University of Zaragoza, Zaragoza (Spain). Centre of Research for Energy Resources and Consumption (CIRCE)
2007-10-15
CFD codes and advanced kinetics combustion models are extensively used to predict coal burnout in large utility boilers. Modelling approaches based on CFD codes can accurately solve the fluid dynamics equations involved in the problem but this is usually achieved by including simple combustion models. On the other hand, advanced kinetics combustion models can give a detailed description of the coal combustion behaviour by using a simplified description of the flow field, this usually being obtained from a zone-method approach. Both approximations describe correctly general trends on coal burnout, but fail to predict quantitative values. In this paper a new methodology which takes advantage of both approximations is described. In the first instance CFD solutions were obtained of the combustion conditions in the furnace in the Lamarmora power plant (ASM Brescia, Italy) for a number of different conditions and for three coals. Then, these furnace conditions were used as inputs for a more detailed chemical combustion model to predict coal burnout. In this, devolatilization was modelled using a commercial macromolecular network pyrolysis model (FG-DVC). For char oxidation an intrinsic reactivity approach including thermal annealing, ash inhibition and maceral effects, was used. Results from the simulations were compared against plant experimental values, showing a reasonable agreement in trends and quantitative values. 28 refs., 4 figs., 4 tabs.
RELIABLE VALIDATION BASED ON OPTICAL FLOW VISUALIZATION FOR CFD SIMULATIONS
姜宗林
2003-01-01
A reliable validation based on the optical flow visualization for numerical simulations of complex flowfields is addressed in this paper.Several test cases,including two-dimensional,axisymmetric and three-dimensional flowfields,were presented to demonstrate the effectiveness of the validation and gain credibility of numerical solutions of complex flowfields.In the validation,images of these flowfields were constructed from numerical results based on the principle of the optical flow visualization,and compared directly with experimental interferograms.Because both experimental and numerical results are of identical physical representation,the agreement between them can be evaluated effectively by examining flow structures as well as checking discrepancies in density.The study shows that the reliable validation can be achieved by using the direct comparison between numerical and experiment results without any loss of accuracy in either of them.
Batet, L.; Mas de les Valls, E.; Sedano, L. A.
2012-07-01
In the context of regenerating sheaths for fusion reactors, the CFD simulations of liquid metal channels (ML) are essential to know the phenomenology and obtain relevant information for design as: ML thermal gain, to know the thermal efficiency of the component, existence of hot spots, to define the materials to use, existence of flow inversion, etc. Apart from design parameters there are others, bridge parameter, required as inputs into system code. In this work shown GREENER/T4F capabilities for obtaining both parameters with a CFD tool based on open source OpenFOAM.
Validation of Hydrodynamic Load Models Using CFD for the OC4-DeepCwind Semisubmersible: Preprint
Benitz, M. A.; Schmidt, D. P.; Lackner, M. A.; Stewart, G. M.; Jonkman, J.; Robertson, A.
2015-03-01
Computational fluid dynamics (CFD) simulations were carried out on the OC4-DeepCwind semi-submersible to obtain a better understanding of how to set hydrodynamic coefficients for the structure when using an engineering tool such as FAST to model the system. The focus here was on the drag behavior and the effects of the free-surface, free-ends and multi-member arrangement of the semi-submersible structure. These effects are investigated through code-to-code comparisons and flow visualizations. The implications on mean load predictions from engineering tools are addressed. The work presented here suggests that selection of drag coefficients should take into consideration a variety of geometric factors. Furthermore, CFD simulations demonstrate large time-varying loads due to vortex shedding, which FAST's hydrodynamic module, HydroDyn, does not model. The implications of these oscillatory loads on the fatigue life needs to be addressed.
Prototype coupling of the CFD code ANSYS CFX with the 3D neutron kinetic core model DYN3D
Analyses of postulated reactivity initiated accidents in nuclear reactors are carried out using 3D neutron kinetic core models. The feedback is usually calculated using 1D thermal hydraulic models for channel flow, partly with the possibility of cross flow between these channels. A different possibility is the use of subchannel codes for the determination of the feedback. The code DYN3D developed at Forschungszentrum Dresden-Rossendorf is an example for a 3D neutron kinetic core model. In its basic version, the code contains models for the solution of the 3D neutron diffusion equation in two energy groups for fuel assemblies with rectangular and hexagonal cross section. Recently the code was extended to an arbitrary number of energy groups. Further, a simplified transport approximation for the flux calculation was implemented for fuel assemblies with quadratic cross section. The CFD code ANSYS CFX is the reference CFD code of the German CFD Network in Nuclear Reactor Safety. One of the goals of the co-operation inside this network is the development of CFD software for the simulation of multi-dimensional flows in reactor cooling systems. This includes the coupling of the CFD code ANSYS CFX with the 3D neutron kinetic core model DYN3D. (orig.)
Development of a CFD code TFC2D for numerical analysis of turbulent flow
A computational fluid dynamics (CFD) code TFC2D(Turbulent Flow Calculator for 2-Dimension) was developed to perform a numerical analysis of the two-dimensional turbulent flow using the various turbulent models and differencing schemes. The TFC2D code uses a finite volume approach on the staggered grid in either the Cartesian or the cylindrical coordinate system. The SIMPLER algorithm is used to solve the pressure field in association with the continuity equation. The typical high Reynolds number and low Reynolds number turbulence models can be optionally chosen to analyze the turbulent flows. The power-law differencing scheme is also used to discretize the convection term. The numerical analyses of the turbulent flow in plane channel, circular pipe and sudden-expansion pipe were performed to verify the TFC2D code. The TFC2D predictions of the mean flow velocity and the turbulence showed a reasonable agreement with the experimental results. TFC2D could be therefore used to perform a numerical analysis of various turbulent flows and to develop a CFD code for the turbulent flow in rod bundle in the future
West, Jeff; Yang, H. Q.
2014-01-01
There are many instances involving liquid/gas interfaces and their dynamics in the design of liquid engine powered rockets such as the Space Launch System (SLS). Some examples of these applications are: Propellant tank draining and slosh, subcritical condition injector analysis for gas generators, preburners and thrust chambers, water deluge mitigation for launch induced environments and even solid rocket motor liquid slag dynamics. Commercially available CFD programs simulating gas/liquid interfaces using the Volume of Fluid approach are currently limited in their parallel scalability. In 2010 for instance, an internal NASA/MSFC review of three commercial tools revealed that parallel scalability was seriously compromised at 8 cpus and no additional speedup was possible after 32 cpus. Other non-interface CFD applications at the time were demonstrating useful parallel scalability up to 4,096 processors or more. Based on this review, NASA/MSFC initiated an effort to implement a Volume of Fluid implementation within the unstructured mesh, pressure-based algorithm CFD program, Loci-STREAM. After verification was achieved by comparing results to the commercial CFD program CFD-Ace+, and validation by direct comparison with data, Loci-STREAM-VoF is now the production CFD tool for propellant slosh force and slosh damping rate simulations at NASA/MSFC. On these applications, good parallel scalability has been demonstrated for problems sizes of tens of millions of cells and thousands of cpu cores. Ongoing efforts are focused on the application of Loci-STREAM-VoF to predict the transient flow patterns of water on the SLS Mobile Launch Platform in order to support the phasing of water for launch environment mitigation so that vehicle determinantal effects are not realized.
In order to demonstrate the accuracy of predictions in a turbulent mixed convection regime in which both inertia and buoyancy force compete with each other, we found out that assessments done using a single-dimensional system code with a recently updated heat transfer package have shown that this approach cannot give a reasonable prediction of the wall temperature in a case involving strong heating, where the regime falls into turbulent mixed convection regime. It has been known that the main reason of this deficiency comes from the degraded heat transfer in turbulent mixed convection regime, which is below that of convective heat transfer during turbulent forced convection. We investigated two mechanisms that cause this deterioration in convective heat transfer influenced by buoyancy: (1) modification of turbulence, also known as the direct (structural) effect, through the buoyancy-induced production of turbulent kinetic energy: and (2) an indirect (external) effect that occurs through modification of the mean flow. We investigated the Launder-Sharma model of turbulence whether it can appropriately represent the mechanisms causing the degraded heat transfer in Computational Fluid Dynamics (CFD). We found out that this model can capture low Re effects such that a non-equilibrium turbulent boundary layer in turbulent mixed convection regime can be resolved. The model was verified and validated extensively initially with the commercial CFD code, Fluent with a user application package known as the User Defined Function (UDF). The results from this implementation were compared to a set of data that included (1) an experimental data commonly accepted as a standardized problem to verify a turbulent flow, (2) the results from a Direct Numerical Simulation (DNS) in a turbulent forced and mixed convection regime, (3) empirical correlations regarding the friction coefficient and the non-dimensional heat transfer coefficient, the Nusselt number for a turbulent forced
Hardman, Robert R.
1990-01-01
The need for a validation technique for computational fluid dynamics (CFD) codes in STOVL applications has led to research efforts to apply infrared thermal imaging techniques to visualize gaseous flow fields. Specifically, a heated, free-jet test facility was constructed. The gaseous flow field of the jet exhaust was characterized using an infrared imaging technique in the 2 to 5.6Î¼m wavelength band as well as conventional pitot tube and thermocouple methods. These infrared i...
ESCADRE Code Development and Validation -AN OVERVIEW-
The ESCADRE code system (Ensemble de Systems de Codes d'Analyse d'accidents Des Reactors a Eau) is tool designed to help in evaluating the response of nuclear power plants during hypothetical severe accidents. It is an integral code, built with simple engineering models describing major phenomena involved in the accidental sequences; its main objective is to compute the whole sequence, starting from the core uncover right up to the release of fission products outside the plant containment. In the last few years, ESCADRE has been extensively used in France and Eastern countries such as Russia, Hungary, Slovakia, Bulgaria, China, etc...and also modified to match the Russian PWR (WWER) reactors. Since this, ESCADRE has been significantly improved to cope with the needs of French Probabilistic Safety Analysis level 2, which require extensive calculations involving numerous scenarios and parametric studies. A new release, ESCADRE mod 1.1 has been thus developed and is currently used in France and will be soon available for foreign countries. In a first part, new features of ESCADRE mod 1.1 are presented; on the modeling point of view (for example improvement in the core degradation phenomenology description, consideration of Direct Containment Heating phenomena...) and on the mode of use (improved coupling between ESCADRE modules, safety systems management, consideration of events occurring during the accident). A brief description of the new environment of ESCADRE, making this code much more user friendly, is also provided. Second part of the presentation concerns the ESCADRE validation program. The validation is supported by both French and foreign experimental programs. A validation test matrix is presented, showing the experiments used so far for the validation (only the tests for which a validation work has been achieved and documented are mentioned). This validation effort is still in progress. As an illustration, some of the results of this validation work are
45 CFR 162.1011 - Valid code sets.
2010-10-01
... 45 Public Welfare 1 2010-10-01 2010-10-01 false Valid code sets. 162.1011 Section 162.1011 Public... ADMINISTRATIVE REQUIREMENTS Code Sets § 162.1011 Valid code sets. Each code set is valid within the dates specified by the organization responsible for maintaining that code set....
Base Flow Model Validation Project
National Aeronautics and Space Administration — The innovation is the systematic "building-block" validation of CFD/turbulence models employing a GUI driven CFD code (RPFM) and existing as well as new data sets...
Validation of the reactor dynamics code TRAB
The one-dimensional reactor dynamics code TRAB (Transient Analysis code for BWRs) developed at VTT was originally designed for BWR analyses, but it can in its present version be used for various modelling purposes. The core model of TRAB can be used separately for LWR calculations. For PWR modelling the core model of TRAB has been coupled to circuit model SMABRE to form the SMATRA code. The versatile modelling capabilities of TRAB have been utilized also in analyses of e.g. the heating reactor SECURE and the RBMK-type reactor (Chernobyl). The report summarizes the extensive validation of TRAB. TRAB has been validated with benchmark problems, comparative calculations against independent analyses, analyses of start-up experiments of nuclear power plants and real plant transients. Comparative RBMES type reactor calculations have been made against Soviet simulations and the initial power excursion of the Chernobyl reactor accident has also been calculated with TRAB
Validation of High-Resolution CFD Method for Slosh Damping Extraction of Baffled Tanks
Yang, H. Q.; West, Jeff
2016-01-01
Determination of slosh damping is a very challenging task as there is no analytical solution. The damping physics involve the vorticity dissipation which requires the full solution of the nonlinear Navier-Stokes equations. As a result, previous investigations and knowledge were mainly carried out by extensive experimental studies. A Volume-Of-Fluid (VOF) based CFD program developed at NASA MSFC was applied to extract slosh damping in a baffled tank from the first principle. First, experimental data using water with subscale smooth wall tank were used as the baseline validation. CFD simulation was demonstrated to be capable of accurately predicting natural frequency and very low damping value from the smooth wall tank at different fill levels. The damping due to a ring baffle at different liquid fill levels from barrel section and into the upper dome was then investigated to understand the slosh damping physics due to the presence of a ring baffle. Based on this study, the Root-Mean-Square error of our CFD simulation in estimating slosh damping was less than 4.8%, and the maximum error was less than 8.5%. Scalability of subscale baffled tank test using water was investigated using the validated CFD tool, and it was found that unlike the smooth wall case, slosh damping with baffle is almost independent of the working fluid and it is reasonable to apply water test data to the full scale LOX tank when the damping from baffle is dominant. On the other hand, for the smooth wall, the damping value must be scaled according to the Reynolds number. Comparison of experimental data, CFD, with the classical and modified Miles equations for upper dome was made, and the limitations of these semi-empirical equations were identified.
A Supersonic Argon/Air Coaxial Jet Experiment for Computational Fluid Dynamics Code Validation
Clifton, Chandler W.; Cutler, Andrew D.
2007-01-01
A non-reacting experiment is described in which data has been acquired for the validation of CFD codes used to design high-speed air-breathing engines. A coaxial jet-nozzle has been designed to produce pressure-matched exit flows of Mach 1.8 at 1 atm in both a center jet of argon and a coflow jet of air, creating a supersonic, incompressible mixing layer. The flowfield was surveyed using total temperature, gas composition, and Pitot probes. The data set was compared to CFD code predictions made using Vulcan, a structured grid Navier-Stokes code, as well as to data from a previous experiment in which a He-O2 mixture was used instead of argon in the center jet of the same coaxial jet assembly. Comparison of experimental data from the argon flowfield and its computational prediction shows that the CFD produces an accurate solution for most of the measured flowfield. However, the CFD prediction deviates from the experimental data in the region downstream of x/D = 4, underpredicting the mixing-layer growth rate.
Validation of CFD-models for natural convection, heat transfer and turbulence phenomena
Natural convection, heat transfer and turbulence phenomena play an important role for the distribution of steam and hydrogen in a reactor containment in the case of a severe accident. These phenomena have influence on all important aspects of an accident scenario, on transport processes, mixing of steam, hydrogen and air, the flammability and combustibility of the air/H2/steam-mixture, the temperature distribution and on the containment pressure. In cooperation with other institutions the GRS adapts and validates the CFX code developed by ANSYS for containment applications. To simulate convection and turbulence phenomena in an accident scenario in a reactor containment the simulation tools and models have to be validated with experimental data. For the validation of CFX two experiments performed at the THAI test facility were simulated (TH-18 and TH-21). THAI is a down-scaled containment facility operated at Becker Technologies, Eschborn, Germany, which was designed to investigate thermal hydraulic processes. The main component is a steel vessel with a height of 9.2 m and a cross-section of 3.2 m. The THAI facility could be divided into different subsections by an inner cylinder and different steel plates. The TH-18 experiment was designed for the validation of CFD models for mass transfer and turbulence. In the inner cylinder a fan was installed which produces a circular flow field in the THAI vessel. At different positions in the THAI vessel the velocity of the flow field was measured by PIV (Particle Image Velocimetry) and LDA (Laser Doppler Anemometer). The TH-21 experiment was designed for the investigation of heat transfer and natural convection phenomena. For this purpose the walls of the THAI vessel were heated differentially. The lower vessel wall was heated up to 120 deg.C and the upper vessel wall was cooled down to 46 deg.C. This differential heating induced a natural convection process in the THAI vessel. Pressure, temperature and flow velocity were
CFD Validation Experiment of a Mach 2.5 Axisymmetric Shock-Wave/Boundary-Layer Interaction
Davis, David Owen
2015-01-01
Preliminary results of an experimental investigation of a Mach 2.5 two-dimensional axisymmetric shock-wave/ boundary-layer interaction (SWBLI) are presented. The purpose of the investigation is to create a SWBLI dataset specifically for CFD validation purposes. Presented herein are the details of the facility and preliminary measurements characterizing the facility and interaction region. These results will serve to define the region of interest where more detailed mean and turbulence measurements will be made.
Bavière, R., E-mail: roland.baviere@cea.fr; Tauveron, N., E-mail: nicolas.tauveron@cea.fr; Perdu, F., E-mail: fabien.perdu@cea.fr; Garré, E., E-mail: emile.garre@cea.fr; Li, S., E-mail: simon.li@cea.fr
2014-10-01
Highlights: • A system/CFD coupling methodology for thermal-hydraulics analysis. • Application of the model to the Phénix Reactor Natural Circulation Test. • Validation of the methodology against experimental data. - Abstract: The natural circulation test (NCT) was conducted in the Phénix prototype French 580 MWth sodium fast reactor (SFR) in 2009. The main goal of the Phénix NCT is to validate system- and CFD-codes with respect to the establishment of natural circulation in the primary system of a pool type SFR. The present paper describes the calculation of the NCT by coupling the 3D computational fluid dynamics (CFD) code TRIO{sub U} with the best estimate thermal hydraulic system code CATHARE. The coupling methodology and the modeling at the system and at the CFD scales are first presented. A validation of the coupling methodology based on a coupled CATHARE/CATHARE calculation compared to the standard CATHARE predictions is then proposed. In a second step, the results of the TRIO{sub U}/CATHARE calculation are compared both to the available experimental data and to the results of a CATHARE alone computation. These comparisons highlight the effectiveness of coupling CFD- and system-codes for the analysis of plant transients where three-dimensional phenomena play an important role.
Highlights: • A system/CFD coupling methodology for thermal-hydraulics analysis. • Application of the model to the Phénix Reactor Natural Circulation Test. • Validation of the methodology against experimental data. - Abstract: The natural circulation test (NCT) was conducted in the Phénix prototype French 580 MWth sodium fast reactor (SFR) in 2009. The main goal of the Phénix NCT is to validate system- and CFD-codes with respect to the establishment of natural circulation in the primary system of a pool type SFR. The present paper describes the calculation of the NCT by coupling the 3D computational fluid dynamics (CFD) code TRIOU with the best estimate thermal hydraulic system code CATHARE. The coupling methodology and the modeling at the system and at the CFD scales are first presented. A validation of the coupling methodology based on a coupled CATHARE/CATHARE calculation compared to the standard CATHARE predictions is then proposed. In a second step, the results of the TRIOU/CATHARE calculation are compared both to the available experimental data and to the results of a CATHARE alone computation. These comparisons highlight the effectiveness of coupling CFD- and system-codes for the analysis of plant transients where three-dimensional phenomena play an important role
Numerical modelling of pressure suppression pools with CFD and FEM codes
Paettikangas, T.; Niemi, J.; Timperi, A. (VTT Technical Research Centre of Finland (Finland))
2011-06-15
Experiments on large-break loss-of-coolant accident for BWR is modeled with computational fluid (CFD) dynamics and finite element calculations. In the CFD calculations, the direct-contact condensation in the pressure suppression pool is studied. The heat transfer in the liquid phase is modeled with the Hughes-Duffey correlation based on the surface renewal model. The heat transfer is proportional to the square root of the turbulence kinetic energy. The condensation models are implemented with user-defined functions in the Euler-Euler two-phase model of the Fluent 12.1 CFD code. The rapid collapse of a large steam bubble and the resulting pressure source is studied analytically and numerically. Pressure source obtained from simplified calculations is used for studying the structural effects and FSI in a realistic BWR containment. The collapse results in volume acceleration, which induces pressure loads on the pool walls. In the case of a spherical bubble, the velocity term of the volume acceleration is responsible of the largest pressure load. As the amount of air in the bubble is decreased, the peak pressure increases. However, when the water compressibility is accounted for, the finite speed of sound becomes a limiting factor. (Author)
Observations on CFD Verification and Validation from the AIAA Drag Prediction Workshops
Morrison, Joseph H.; Kleb, Bil; Vassberg, John C.
2014-01-01
The authors provide observations from the AIAA Drag Prediction Workshops that have spanned over a decade and from a recent validation experiment at NASA Langley. These workshops provide an assessment of the predictive capability of forces and moments, focused on drag, for transonic transports. It is very difficult to manage the consistency of results in a workshop setting to perform verification and validation at the scientific level, but it may be sufficient to assess it at the level of practice. Observations thus far: 1) due to simplifications in the workshop test cases, wind tunnel data are not necessarily the “correct” results that CFD should match, 2) an average of core CFD data are not necessarily a better estimate of the true solution as it is merely an average of other solutions and has many coupled sources of variation, 3) outlier solutions should be investigated and understood, and 4) the DPW series does not have the systematic build up and definition on both the computational and experimental side that is required for detailed verification and validation. Several observations regarding the importance of the grid, effects of physical modeling, benefits of open forums, and guidance for validation experiments are discussed. The increased variation in results when predicting regions of flow separation and increased variation due to interaction effects, e.g., fuselage and horizontal tail, point out the need for validation data sets for these important flow phenomena. Experiences with a recent validation experiment at NASA Langley are included to provide guidance on validation experiments.
Validation of TNXY code with reference problems
In this paper, the validation process for TNXY code, as well as the rference problems used in the same (Wagner and Benchmark 14 problems) are described. TNXY code is based on a polynomial type nodal method known as RTN-0. Several numerical results obtained with such code and others frequently illustrated in the literature related with numerical calculus for nuclear reactors are presented. Tests were done with different size meshes and different SN approximations. Several conclusions based on comparisons among different results obtained, as well as the present state of the already mentioned code and its almost inmediate applications to fuel assemblies as the used in the nuclear reactor of Laguna Verde are given. (Author)
Development of a Prototype Lattice Boltzmann Code for CFD of Fusion Systems.
Pattison, Martin J; Premnath, Kannan N; Banerjee, Sanjoy; Dwivedi, Vinay
2007-02-26
Designs of proposed fusion reactors, such as the ITER project, typically involve the use of liquid metals as coolants in components such as heat exchangers, which are generally subjected to strong magnetic fields. These fields induce electric currents in the fluids, resulting in magnetohydrodynamic (MHD) forces which have important effects on the flow. The objective of this SBIR project was to develop computational techniques based on recently developed lattice Boltzmann techniques for the simulation of these MHD flows and implement them in a computational fluid dynamics (CFD) code for the study of fluid flow systems encountered in fusion engineering. The code developed during this project, solves the lattice Boltzmann equation, which is a kinetic equation whose behaviour represents fluid motion. This is in contrast to most CFD codes which are based on finite difference/finite volume based solvers. The lattice Boltzmann method (LBM) is a relatively new approach which has a number of advantages compared with more conventional methods such as the SIMPLE or projection method algorithms that involve direct solution of the Navier-Stokes equations. These are that the LBM is very well suited to parallel processing, with almost linear scaling even for very large numbers of processors. Unlike other methods, the LBM does not require solution of a Poisson pressure equation leading to a relatively fast execution time. A particularly attractive property of the LBM is that it can handle flows in complex geometries very easily. It can use simple rectangular grids throughout the computational domain -- generation of a body-fitted grid is not required. A recent advance in the LBM is the introduction of the multiple relaxation time (MRT) model; the implementation of this model greatly enhanced the numerical stability when used in lieu of the single relaxation time model, with only a small increase in computer time. Parallel processing was implemented using MPI and demonstrated the
Validation Report for ISAAC Computer Code
A fully integrated severe accident code ISAAC was developed to simulate the accident scenarios that could lead to a severe core damage and eventually to the containment failure in CANDU reactors. Three ways of validation were adopted in this report. The first approach is to show the ISAAC results for the typical severe core damage sequences. In general, the ISAAC computer code shows the reasonable results in terms of the thermal hydraulic behavior as well as fission product transport from the PHTS to the containment. As the second step, the ISAAC results are compared against those from CATHENA and MAAP4-CANDU. In spite of the modeling differences, the overall trend is similar to each other. Especially, the major severe accident phenomena and the accident progression are similar to MAAP4-CANDU, though ISAAC predicts the accident progression faster. Finally ISAAC results are compared with the experimental data. The ISAAC models provide a good agreement with the measured data. Still more efforts are needed to validate the code by the code-to-code comparison and the comparison against the experimental data available
To justify the use of a commercial Computational Fluid Dynamics (CFD) code for a CANDU fuel channel analysis, especially for the radiation heat transfer dominant conditions, the CFX-10 code is tested against three benchmark problems which were used for the validation of a radiation heat transfer in the CANDU analysis code, a CATHENA. These three benchmark problems are representative of the CANDU fuel channel configurations from a simple geometry to a whole fuel channel geometry. For the solutions of the benchmark problems, the temperature or the net radiation heat flux boundary conditions are prescribed for each radiating surface to determine the radiation heat transfer rate or the surface temperature, respectively by using the network method. The Discrete Transfer Model (DTM) is used for the CFX-10 radiation model and its calculation results are compared with the solutions of the benchmark problems. The CFX-10 results for the three benchmark problems are in close agreement with those solutions, so it is concluded that the CFX-10 with a DTM radiation model can be applied to the CANDU fuel channel analysis where a surface radiation heat transfer is a dominant mode of the heat transfer. (author)
Application of CFD techniques toward the validation of nonlinear aerodynamic models
Schiff, L. B.; Katz, J.
1985-01-01
Applications of computational fluid dynamics (CFD) methods to determine the regimes of applicability of nonlinear models describing the unsteady aerodynamic responses to aircraft flight motions are described. The potential advantages of computational methods over experimental methods are discussed and the concepts underlying mathematical modeling are reviewed. The economic and conceptual advantages of the modeling procedure over coupled, simultaneous solutions of the gas dynamic equations and the vehicle's kinematic equations of motion are discussed. The modeling approach, when valid, eliminates the need for costly repetitive computation of flow field solutions. For the test cases considered, the aerodynamic modeling approach is shown to be valid.
The aim of the project was the qualification of CFD codes for steam-water flows with phase transfer. While CFD methods for single-phase flows are already widely used for industrial applications, a corresponding use for two-phase flows is only at the beginning due to the complex structure of the interface and the related interactions between the phases. For the further development and validation of appropriate closure models, experimental data with high spatial and temporal resolution are required. Such data were obtained at the TOPFLOW test facility of HZDR by combination of experiments at realistic parameters for the nuclear reactor safety (large scales, high pressures and temperatures) with innovative measuring techniques. The wire-mesh sensor technology, which provides detailed information on the structure of the interface, was applied in adiabatic air-water experiments as well as in condensation and pressure relief experiments in a large DN200 pipe. As the result of the project, extensive databases with high quality are available. The technology for the fast X-ray tomography, which allows measurements without influencing the flow, was further developed and successfully applied in a first test series. High-resolution data were also obtained from experiments in a model of the hot leg of a pressurized water reactor for different flow situations, including counter-current flow limitation. For the corresponding steam-water experiments conducted at pressures of up to 5 MPa, the newly developed pressure tank technology was successfully used for the first time. For the qualification of CFD codes for two-phase flows the Inhomogeneous MUSIG model was extended in co.operation with ANSYS to consider phase transfer and validated on the basis of the above mentioned TOPFLOW experiments. In addition, improvements were achieved e.g. for turbulence modelling in bubbly flows and simulations were done to validate models for bubble forces and bubble coalescence and breakup. A
The thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is developed at Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs. In order to extend the simulation capabilities of the 1D system code ATHLET, different approaches are applied at GRS to enable multidimensional thermal-hydraulic representation of relevant primary circuit geometries. One of the current major strategies at the technical safety organization is the coupling of ATHLET with the commercial 3D Computational Fluid Dynamics (CFD) software package ANSYS CFX. This code is a general purpose CFD software program that combines an advanced solver with powerful pre- and post-processing capabilities. It is an efficient tool for simulating the behavior of systems involving fluid flow, heat transfer, and other related physical processes. In the frame of the German CFD Network on Nuclear Reactor Safety, GRS and ANSYS Germany developed a general computer interface for the coupling of both codes. This paper focuses on the methodology and the challenges related to the coupling process. A great number of simulations including test cases with closed loop configurations have been carried out to evaluate and improve the performance of the coupled code system. Selected results of the 1D-3D thermal-hydraulic calculations are presented and analyzed. Preliminary comparative calculations with CFX-ATHLET and ATHLET stand alone showed very good agreement. Nevertheless, an extensive validation of the developed coupled code is planned. Finally, the optimization potential of the coupling methodology is discussed. (author)
Evaporation over sump surface in containment studies: code validation on TOSQAN tests
During the course of a severe accident in a Nuclear Power Plant, water can be collected in the sump containment through steam condensation on walls and spray systems activation. The objective of this paper is to present code validation on evaporative sump tests performed on the TOSQAN facility. The ASTEC-CPA code is used as a lumped-parameter code and specific user-defined-functions are developed for the TONUS-CFD code. The tests are air-steam tests, as well as tests with other non-condensable gases (He, CO2 and SF6) under steady and transient conditions. The results show a good agreement between codes and experiments, indicating a good behaviour of the sump models in both codes. (author)
Experimental verification of CFD and thermal hydraulics codes by quantitative flow visualisation
Complex flow fields are encountered in many reactor components and processes. Measurement and analysis of various flow parameters are very important for optimal design, experimental determination of safety margins and verification of CFD and thermal hydraulics codes. Development of image capture hardware and digital image processing technique in Particle Image Velocimetry (PIV) has made possible to map complex flow fields instantaneously at thousands of points with very high temporal and spatial resolution. PIV is a non intrusive and very flexible technique. In this technique using synchronized operation of laser and CCD camera, seeded flow is illuminated by pulsing laser sheet and images of seeded particles are recorded on CCD camera. The displacement of the particles is measured in the plane of the image and used to determine the velocity of the flow. Image plane is divided into small interrogation regions. Velocity vectors are calculated with the help of cross correlated images obtained from two time exposures. This paper describes 2D PIV System used, flow mapping and verification of CFD codes for pipe flow, submerged jet, thermal stratification in water pool and Fluidic Flow Control Device (FFCD) proposed to be used in advanced accumulator of Emergency Core Cooling System (ECCS). (author)
Investigation of natural circulation two-phase flow behaviour in header manifold using CFD code
The three-dimensional (3-D), multiphase, computational fluid dynamic (CFD) code FLUENT is used to simulated two-phase flow behaviour in a CANDU header manifold under low (natural circulation) flow conditions. This behaviour was previously inferred from experimental data. The CFD simulations reported here are being used to support these inferences and to obtain a better understanding of phase distribution in the header manifold. The simulations seem to show that the vapor-water mixture models in the FLUENT code do not capture properly phase separation in the header and proper phase branching at the header-feeder connections that have been observed in experiments at low flows. The simulations using discrete-phase model in FLUENT, which tracks the pathlines of the individual vapor bubbles in the water continuum phase, show interesting, complicated and, in some cases, unexpected bubble trajectories from the point of injection of the bubbles at a feeder connection to the other parts of the header and other feeder connections. These simulations have the potential of providing needed insight into the vapor-phase behaviour in the header and may be useful in accident analyses. (author)
Yang, H. Q.; West, Jeff
2016-01-01
Propellant slosh is a potential source of disturbance critical to the stability of space vehicles. The slosh dynamics are typically represented by a mechanical model of a spring-mass-damper. This mechanical model is then included in the equation of motion of the entire vehicle for Guidance, Navigation and Control analysis. A Volume-Of-Fluid (VOF) based Computational Fluid Dynamics (CFD) program developed at MSFC was applied to extract slosh damping in the baffled tank from the first principle. First the experimental data using water with sub-scale smooth wall tank were used as the baseline validation. It is demonstrated that CFD can indeed accurately predict low damping values from the smooth wall at different fill levels. The damping due to a ring baffles at different depths from the free surface was then simulated, and fairly good agreement with experimental measurement was observed. Comparison with an empirical correlation of Miles equation is also made.
Qualification of the CFD code TRIO-U for full scale nuclear reactor applications
Numerical and experimental research on nuclear safety is in the end dedicated to understand, on a plant scale, the fundamental physical phenomena which are associated to specific accident scenarios. Hence, the results derived from single effect experiments or reduced scale analysis have to be extrapolated to plant scale whereas plant scale experiments should be evaluated with respect to their applicability to the physics of the specific scenario. For several years, IRSN and CEA have used Computational Fluid Dynamics (CFD) codes for detailed nuclear safety analyses on plant scale. The paper presents a procedure which has been used to qualify the Trio-U code for the prediction of the boron concentration at the core inlet of a French Pressurized Water Reactor (PWR) in accidental conditions (inherent dilution problem) 1. A ROCOM experiment as well as an UPTF Tram-C3 experiment has been used for this purpose. (authors)
ALEGRA -- code validation: Experiments and simulations
Chhabildas, L.C.; Konrad, C.H.; Mosher, D.A.; Reinhart, W.D; Duggins, B.D.; Rodeman, R.; Trucano, T.G.; Summers, R.M.; Peery, J.S.
1998-03-16
In this study, the authors are providing an experimental test bed for validating features of the ALEGRA code over a broad range of strain rates with overlapping diagnostics that encompass the multiple responses. A unique feature of the Arbitrary Lagrangian Eulerian Grid for Research Applications (ALEGRA) code is that it allows simultaneous computational treatment, within one code, of a wide range of strain-rates varying from hydrodynamic to structural conditions. This range encompasses strain rates characteristic of shock-wave propagation (10{sup 7}/s) and those characteristic of structural response (10{sup 2}/s). Most previous code validation experimental studies, however, have been restricted to simulating or investigating a single strain-rate regime. What is new and different in this investigation is that the authors have performed well-instrumented experiments which capture features relevant to both hydrodynamic and structural response in a single experiment. Aluminum was chosen for use in this study because it is a well characterized material--its EOS and constitutive material properties are well defined over a wide range of loading rates. The current experiments span strain rate regimes of over 10{sup 7}/s to less than 10{sup 2}/s in a single experiment. The input conditions are extremely well defined. Velocity interferometers are used to record the high strain-rate response, while low strain rate data were collected using strain gauges.
The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal-hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coefficient, ...), critical position of control rods, reactivity insertion aspects, .... For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, ...) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal-hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark
Recommendation for maximum allowable mesh size for plant combustion analyses with CFD codes
Highlights: ► Used mesh size has to be small enough to resolve all pressure waves relevant for the structural response analyses. ► Maximum allowable mesh size for a combustion pressure load calculation decreases with increasing relevant natural frequency of the structure. ► Maximum allowable mesh size for a combustion pressure load calculation increases with increasing of the speed of the sound in the gas mixture. ► Maximum allowable mesh size can be calculated from the developed analytical formula. - Abstract: The selection of the maximum allowable mesh size for a fluid dynamic calculation with Computational Fluid Dynamic (CFD) codes is essential for the reliability of the results assuming suitable physical and numerical models are used. Calculations with CFD codes are necessary for the assessment of the consequences of pressure loads on containment structures due to possible hydrogen combustion in nuclear power plants in a severe accident and on piping system due to pressure wave propagation in case of a pipe break accident or fast closing of a valve in a pipe with forced flow. CFD simulations of the transport and distribution of the released hydrogen/steam as well as the possible combustion during the transient in the containment require an appropriate mesh size to resolve the relevant phenomena and loads. The determination of the mesh size has to take into account: •adequate delineation of the containment geometry for accurate hydrogen distribution calculations, •sufficient conservative resolution of the combustion phenomena for the determination of pressure wave propagation and pressure loads, •no loss of pressure wave loads with relevant frequencies for the structural response analysis of the containment during the combustion calculation. In this paper, it is found that the accuracy of the calculated pressure wave associated with its frequency depends on the mesh size and a simple and easily useable analytical formula for the determination of
Recommendation for maximum allowable mesh size for plant combustion analyses with CFD codes
Movahed-Shariat-Panahi, M.A., E-mail: Mohammad-Ali.Movahed@areva.com [AREVA GmbH Offenbach (Germany)
2012-12-15
Highlights: Black-Right-Pointing-Pointer Used mesh size has to be small enough to resolve all pressure waves relevant for the structural response analyses. Black-Right-Pointing-Pointer Maximum allowable mesh size for a combustion pressure load calculation decreases with increasing relevant natural frequency of the structure. Black-Right-Pointing-Pointer Maximum allowable mesh size for a combustion pressure load calculation increases with increasing of the speed of the sound in the gas mixture. Black-Right-Pointing-Pointer Maximum allowable mesh size can be calculated from the developed analytical formula. - Abstract: The selection of the maximum allowable mesh size for a fluid dynamic calculation with Computational Fluid Dynamic (CFD) codes is essential for the reliability of the results assuming suitable physical and numerical models are used. Calculations with CFD codes are necessary for the assessment of the consequences of pressure loads on containment structures due to possible hydrogen combustion in nuclear power plants in a severe accident and on piping system due to pressure wave propagation in case of a pipe break accident or fast closing of a valve in a pipe with forced flow. CFD simulations of the transport and distribution of the released hydrogen/steam as well as the possible combustion during the transient in the containment require an appropriate mesh size to resolve the relevant phenomena and loads. The determination of the mesh size has to take into account: Bullet adequate delineation of the containment geometry for accurate hydrogen distribution calculations, Bullet sufficient conservative resolution of the combustion phenomena for the determination of pressure wave propagation and pressure loads, Bullet no loss of pressure wave loads with relevant frequencies for the structural response analysis of the containment during the combustion calculation. In this paper, it is found that the accuracy of the calculated pressure wave associated with its
The paper presents a HPC (High Performance Computing) calculation of a conjugate heat transfer simulation in fuel assembly as those found in liquid metal coolant fast reactors. The wire spacers, helically wound along each pin axis, generate a strong secondary flow pattern in opposition to smooth pins. Assemblies with a range of pins going from 7 to 271 have been simulated, 271 pins corresponding to the industrial case. Both the fluid domain, as well as the solid part, are detailed leading to large meshes. The fluid is handled by the CFD code Code-Saturne using 98 million cells, while the solid domain is taken care of thanks to the thermal code SYRTHES on meshes up to 240 million cells. Both codes are fully parallelized and run on cluster with hundreds of processors. Simulations allow access to the temperature field in nominal conditions and degraded situations. (authors)
Nelson, Edward L.
1992-01-01
The increased use of infrared imaging as a flow visualization technique and as a validation technique for computational fluid dynamics (CFD) codes has led to an in-depth study of infrared band models. The ability to create fast and accurate images of airframe and plume infrared emissions often depends on the complexity of the band model. An infrared band model code has been created based largely on the band model published in NASA SP-3080, Handbook of Infrared Radiation from Combustion Gases....
To provide a high degree of confidence in the results predicted by the FLUENT CFD computer code for safety evaluation of storing Chernobyl spent nuclear fuel (SNF) in double-walled canisters (DWC) a full scale prototype DWC was manufactured and tested at the Holtec Manufacturing Division in Turtle Creek, PA. The DWC was instrumented and fuel heat simulated by inserting electrically heated rods in storage cells under two extreme heat distribution scenarios: Core heated test wherein the heat is applied to the innermost storage cells and Peripherally heated test wherein the heat is applied to the outermost storage cells. The heater tubes, storage cells, DWC shell and lid were instrumented to measure and record temperatures during the testing. To validate the FLUENT CFD code the thermal tests were simulated on FLUENT by constructing geometrically accurate 3D model of the DWC with all internals significant to mimic the thermal-hydraulic state in the DWC. These included heated rods, fuel tubes, support plates and the DWC shell. The test measurements and FLUENT simulations were evaluated and the predictability of the FLUENT CFD code for safety evaluation of fuel storage in double-walled canisters confirmed. (authors)
Validation and Analysis of Forward Osmosis CFD Model in Complex 3D Geometries
Lars Yde
2012-11-01
Full Text Available In forward osmosis (FO, an osmotic pressure gradient generated across a semi-permeable membrane is used to generate water transport from a dilute feed solution into a concentrated draw solution. This principle has shown great promise in the areas of water purification, wastewater treatment, seawater desalination and power generation. To ease optimization and increase understanding of membrane systems, it is desirable to have a comprehensive model that allows for easy investigation of all the major parameters in the separation process. Here we present experimental validation of a computational fluid dynamics (CFD model developed to simulate FO experiments with asymmetric membranes. Simulations are compared with experimental results obtained from using two distinctly different complex three-dimensional membrane chambers. It is found that the CFD model accurately describes the solute separation process and water permeation through membranes under various flow conditions. It is furthermore demonstrated how the CFD model can be used to optimize membrane geometry in such as way as to promote the mass transfer.
Examples of unsteady CFD validation system response quantities in a cylinder array
Highlights: ► Unsteady validation of k − ω and DES models using high-speed PIV. ► Local validation SRQs, include frequency spectra, autocorrelations, and correlations, while global SRQs include rms values and distributions. ► The DES-NB model predicts good approximations for most unsteady validation SRQs. ► The k − ω and DES-B models predict oscillatory flow with amplitudes larger than the experiment. ► All CFD models are capable of accurately predicting global validation SRQs, such as the minor loss factor. - Abstract: A validation study for several CFD models of the time-varying flow through a confined bank of cylinders is presented. The geometry is cylinders arranged on equilateral triangles with pitch to diameter ratio of 1.7 to represent a scaled subsection of the lower plenum of a high temperature reactor. Time-resolved Particle Image Velocimetry (PIV) measurements, coupled with time-varying pressure measurements along the facility walls, are compared to the Unsteady Reynolds-Averaged Navier–Stokes (URANS) k − ω model and two variations of a Detached Eddy Simulation (DES) model. Spatial and temporal validation system response quantities (SRQs) on both the local and global scales were used for validation. The DES model accurately predicted frequencies present in the pressure along the walls next to the cylinders in the first and the last cylinder, yet predicted other dominant frequencies in the remaining cylinders that were not found in the experiment. As expected, the temporal behavior of the DES was generally far superior to that of the URANS model. A grid convergence study shows typical global quantities (such as pressure losses) converge well while temporal quantities converge poorly for the same grids.
Validation of High-Fidelity CFD Simulations for Rocket Injector Design
Tucker, P. Kevin; Menon, Suresh; Merkle, Charles L.; Oefelein, Joseph C.; Yang, Vigor
2008-01-01
Computational fluid dynamics (CFD) has the potential to improve the historical rocket injector design process by evaluating the sensitivity of performance and injector-driven thermal environments to the details of the injector geometry and key operational parameters. Methodical verification and validation efforts on a range of coaxial injector elements have shown the current production CFD capability must be improved in order to quantitatively impact the injector design process. This paper documents the status of a focused effort to compare and understand the predictive capabilities and computational requirements of a range of CFD methodologies on a set of single element injector model problems. The steady Reynolds-Average Navier-Stokes (RANS), unsteady Reynolds-Average Navier-Stokes (URANS) and three different approaches using the Large Eddy Simulation (LES) technique were used to simulate the initial model problem, a single element coaxial injector using gaseous oxygen and gaseous hydrogen propellants. While one high-fidelity LES result matches the experimental combustion chamber wall heat flux very well, there is no monotonic convergence to the data with increasing computational tool fidelity. Systematic evaluation of key flow field regions such as the flame zone, the head end recirculation zone and the downstream near wall zone has shed significant, though as of yet incomplete, light on the complex, underlying causes for the performance level of each technique. 1 Aerospace Engineer and Combustion CFD Team Leader, MS ER42, NASA MSFC, AL 35812, Senior Member, AIAA. 2 Professor and Director, Computational Combustion Laboratory, School of Aerospace Engineering, 270 Ferst Dr., Atlanta, GA 30332, Associate Fellow, AIAA. 3 Reilly Professor of Engineering, School of Mechanical Engineering, 585 Purdue Mall, West Lafayette, IN 47907, Fellow, AIAA. 4 Principal Member of Technical Staff, Combustion Research Facility, 7011 East Avenue, MS9051, Livermore, CA 94550, Associate
Development of a Common Research Model for Applied CFD Validation Studies
Vassberg, John C.; Dehaan, Mark A.; Rivers, S. Melissa; Wahls, Richard A.
2008-01-01
The development of a wing/body/nacelle/pylon/horizontal-tail configuration for a common research model is presented, with focus on the aerodynamic design of the wing. Here, a contemporary transonic supercritical wing design is developed with aerodynamic characteristics that are well behaved and of high performance for configurations with and without the nacelle/pylon group. The horizontal tail is robustly designed for dive Mach number conditions and is suitably sized for typical stability and control requirements. The fuselage is representative of a wide/body commercial transport aircraft; it includes a wing-body fairing, as well as a scrubbing seal for the horizontal tail. The nacelle is a single-cowl, high by-pass-ratio, flow-through design with an exit area sized to achieve a natural unforced mass-flow-ratio typical of commercial aircraft engines at cruise. The simplicity of this un-bifurcated nacelle geometry will facilitate grid generation efforts of subsequent CFD validation exercises. Detailed aerodynamic performance data has been generated for this model; however, this information is presented in such a manner as to not bias CFD predictions planned for the fourth AIAA CFD Drag Prediction Workshop, which incorporates this common research model into its blind test cases. The CFD results presented include wing pressure distributions with and without the nacelle/pylon, ML/D trend lines, and drag-divergence curves; the design point for the wing/body configuration is within 1% of its max-ML/D. Plans to test the common research model in the National Transonic Facility and the Ames 11-ft wind tunnels are also discussed.
Validation and analysis of forward osmosis CFD model in complex 3D geometries
Gruber, Mathias F.; Johnson, Carl J.; Tang, Chuyang;
2012-01-01
, seawater desalination and power generation. To ease optimization and increase understanding of membrane systems, it is desirable to have a comprehensive model that allows for easy investigation of all the major parameters in the separation process. Here we present experimental validation of a computational...... separation process and water permeation through membranes under various flow conditions. It is furthermore demonstrated how the CFD model can be used to optimize membrane geometry in such as way as to promote the mass transfer. © 2012 by the authors; licensee MDPI, Basel, Switzerland....
CFD Analyses for Water-Air Flow With the Euler-Euler Two-Phase Model in the Fluent4 CFD Code
calculation results were adjusted for a good agreement with the experimental data. The analysis results were very valuable for designing the final water/steam facility for final CHF tests. The validation against data from the air-water experiments proved that the present CFD codes approach to the state where they can be used for simulating such two-phase experiments, where the fraction of both phases is essential and the flow is strongly affected by the density differences. It is still too early to predict, if the CFD calculation of the 1:1 scale critical heat flux experiments is successful, could the result be used for formulating a new type of a critical heat flux correlation, where the effects of CRD's on the flow patterns and gap dimensions are model parameters. (authors)
Further code validation of Technical Specification Bases
Personnel from the Applied Physics Group (APG) and from the Reactor Technology Section devised the K-14 reactor restart test program. The primary purpose of that program was to acquire data which could be used to further validate the computer codes which supported the Technical Specification Bases. The test program was highly successful and a large amount of useful data was obtained. DOE has now requested the schedule and plan for the additional code validation activities. A response to DOE was made in which an outline of WSRC's (Westinghouse Savannah River Co.) planned activities was presented, but specifics were not discussed. This memorandum is intended to provide additional details on these activities. The WSRC plan consists of five major activities. The first activity is to catalog and document the measured data. The second activity is to investigate the radial and axial neutron flux distributions. The results from this task will complement the analyses being performed in the final three activities: steady state reactivity effects, transient reactivity effects, and planned an unplanned shutdowns. Work on these activities is scheduled to be completed by the end of the fiscal year 1993
Nuclear data to support computer code validation
The rate of plutonium disposition will be a key parameter in determining the degree of success of the Fissile Materials Disposition Program. Estimates of the disposition rate are dependent on neutronics calculations. To ensure that these calculations are accurate, the codes and data should be validated against applicable experimental measurements. Further, before mixed-oxide (MOX) fuel can be fabricated and loaded into a reactor, the fuel vendors, fabricators, fuel transporters, reactor owners and operators, regulatory authorities, and the Department of Energy (DOE) must accept the validity of design calculations. This report presents sources of neutronics measurements that have potential application for validating reactor physics (predicting the power distribution in the reactor core), predicting the spent fuel isotopic content, predicting the decay heat generation rate, certifying criticality safety of fuel cycle facilities, and ensuring adequate radiation protection at the fuel cycle facilities and the reactor. The U.S. in-reactor experience with MOX fuel is first presented, followed by information related to other aspects of the MOX fuel performance information that is valuable to this program, but the data base remains largely proprietary. Thus, this information is not reported here. It is expected that the selected consortium will make the necessary arrangements to procure or have access to the requisite information
Anusonti-Inthra, Phuriwat
2010-01-01
This paper presents validations of a novel rotorcraft analysis that coupled Computational Fluid Dynamics (CFD), Computational Structural Dynamics (CSD), and Particle Vortex Transport Method (PVTM) methodologies. The CSD with associated vehicle trim analysis is used to calculate blade deformations and trim parameters. The near body CFD analysis is employed to provide detailed near body flow field information which is used to obtain high-fidelity blade aerodynamic loadings. The far field wake dominated region is simulated using the PVTM analysis which provides accurate prediction of the evolution of the rotor wake released from the near body CFD domains. A loose coupling methodology between the CSD and CFD/PVTM modules are used with appropriate information exchange amongst the CSD/CFD/PVTM modules. The coupled CSD/CFD/PVTM methodology is used to simulate various rotorcraft flight conditions (i.e. hover, transition, and high speed flights), and the results are compared with several sets of experimental data. For the hover condition, the results are compared with hover data for the HART II rotor tested at DLR Institute of Flight Systems, Germany. For the forward flight conditions, the results are validated with the UH-60A flight test data.
Experimental and numerical approach to validate pressure loss predictability of a commercial code
Experimental and numerical works to validate a commercial CFD code for predicting the pressure loss of a PWR grid spacer were presented. The experimental data was obtained for full size spacer mockups with different inclination of mixing-vanes. The pressure loss in the complex configuration of spacers arises from a several hydrodynamic effects included in the flow. Only the experimental result therefore was not enough to provide detailed data for validating the turbulence model used in the CFD code. To this end this study used a large eddy simulation (LES) to look at the hydrodynamic effects. The result of the LES indicated that the flow field around the spacer included a large-scale unsteadiness and an undeveloped turbulent flow. Turbulence models based on a developed turbulent flow were theoretically inapplicable to these flows. The commercial code with the standard high Reynolds number k-ε model with the law of the wall however successfully reproduced the trend of the measurement. This suggests that a large-scale unsteadiness and an undeveloped turbulent flow are not dominant for the pressure loss. It is noted that commercial codes should be applied to the flows where dominant physics is clarified. (authors)
On the role of code comparisons in verification and validation.
Oberkampf, William Louis; Trucano, Timothy Guy; Pilch, Martin M.
2003-08-01
This report presents a perspective on the role of code comparison activities in verification and validation. We formally define the act of code comparison as the Code Comparison Principle (CCP) and investigate its application in both verification and validation. One of our primary conclusions is that the use of code comparisons for validation is improper and dangerous. We also conclude that while code comparisons may be argued to provide a beneficial component in code verification activities, there are higher quality code verification tasks that should take precedence. Finally, we provide a process for application of the CCP that we believe is minimal for achieving benefit in verification processes.
Numerical modeling of immiscible two-phase flow in micro-models using a commercial CFD code
Crandall, Dustin; Ahmadia, Goodarz; Smith, Duane H.
2009-01-01
Off-the-shelf CFD software is being used to analyze everything from flow over airplanes to lab-on-a-chip designs. So, how accurately can two-phase immiscible flow be modeled flowing through some small-scale models of porous media? We evaluate the capability of the CFD code FLUENT{trademark} to model immiscible flow in micro-scale, bench-top stereolithography models. By comparing the flow results to experimental models we show that accurate 3D modeling is possible.
Although the heat transfer problem of pressurized supercritical water (SCW) flows in around tube has been studied for decades, the subject is still considerably of interest nowadays. This is partly because of the expanded investigation of using SCW for nuclear engineering applications like SCWR which is generation IV reactor and promising advanced nuclear systems because of their high thermal efficiency(i.e., about 45% as opposed to about 33% efficiency for current light water reactors LWRs) and considerable plant simplification. Literature survey shows that heat transfer coefficient (HTC) is sharply enhanced near the pseudo critical temperature. As the heat flux increases, the peak of the HTC decreases. When the heat flux reaches to some high values, heat transfer deterioration (HTD) occurs. CFD code with various turbulence models are being used to evaluate HTC. Modeling of Yamagata's experiment has been carried out for evaluation of HTC using CFD code FLUENT with standard kε turbulence model, nonequilibrium wall function,viscous heating, full buoyancy effect and including wall roughness effect.In this paper model constants for standard kε model have been derived. In the Yamagata experiment, investigations were made for HTC to supercritical water flowing vertically upward in vertical tubes of 10 and 7.5mm internal diameter, at pressures 22.6, 24.5 and 29.5 MPa, bulk temperature from 230 to 540 oC, heat flux 233, 465, 698 and 930kW/m2 and mass flux 1200 kg/m2.s. Two dimensional axisymmetry grid generation has been done using GAMBIT. Inbuilt boundary conditions in the FLUENT are invoked for mass flow rate at inlet,pressure outlet at the outlet of the tube and wall at the cylindrical surface where heat flux is given. Thermo-physical properties are taken from the (IAPWSIF97) and piecewise linear variation are given in the FLUENT for 30 temperature points. Bulk fluid temperature is obtained using user defined function. HTC are obtained based on heat flux, surface
Leber, A. [WTI Wissenschaftlich-Technische-Ingenieurberatung GmbH (Germany); Graf, W. [GNS Gesellschaft fuer Nuklear-Service mbH (Germany); Hueggenberg, R. [GNB Gesellschaft fuer Nuklear-Behaelter mbH (Germany)
2004-07-01
With respect to the transport of casks for radioactive material, the proof of the safe heat removal can be accomplished by validated calculation methods. The boundary conditions for thermal tests for type B packages are specified in the ADR based on the regulations defined by the International Atomic Energy Agency. The varying boundary conditions under transport or storage conditions are based on the varying thermal conditions true for different cask types. In most cases the cask will be transported in lying position under a cover (e.g. canopy or tarpaulin) and stored in standing position in an array with other casks. The main heat transport mechanisms are natural convection and thermal radiation. The cover or the storage building are furnished with vents that create an air flow, which will improve the natural convection. Depending on the thermal boundary conditions, the cask design and the heat power, about 50 - 95% of the heat power will be removed from the finned cask surface by natural convection. Consequently the convection by air flow is the main heat transport mechanism. The air flow can be approximated with analytical methods by solving the integral heat and flow balances for the domain. In a stationary state the overpressure due the buoyancy and the pressure loss in the flow resistances are equal. Based on the air flow, the relevant temperatures of the cask can be calculated in an iterative process. Due to the fast development of numerical calculation methods and computer hardware, the use of Computational- Fluid-Dynamics(CFD) calculations plays an important role. CFD-calculations are based on solving the equations of conservation (Navier-Stokes equations) using a finite element mesh or a finite volume mesh of the model. For a finned cask lying under a cover, where the main contributing element for heat removal is natural convection in combination with the thermal radiation, a CFD-calculation can be the most appropriate method. Common CFD-Codes are FLUENT
With respect to the transport of casks for radioactive material, the proof of the safe heat removal can be accomplished by validated calculation methods. The boundary conditions for thermal tests for type B packages are specified in the ADR based on the regulations defined by the International Atomic Energy Agency. The varying boundary conditions under transport or storage conditions are based on the varying thermal conditions true for different cask types. In most cases the cask will be transported in lying position under a cover (e.g. canopy or tarpaulin) and stored in standing position in an array with other casks. The main heat transport mechanisms are natural convection and thermal radiation. The cover or the storage building are furnished with vents that create an air flow, which will improve the natural convection. Depending on the thermal boundary conditions, the cask design and the heat power, about 50 - 95% of the heat power will be removed from the finned cask surface by natural convection. Consequently the convection by air flow is the main heat transport mechanism. The air flow can be approximated with analytical methods by solving the integral heat and flow balances for the domain. In a stationary state the overpressure due the buoyancy and the pressure loss in the flow resistances are equal. Based on the air flow, the relevant temperatures of the cask can be calculated in an iterative process. Due to the fast development of numerical calculation methods and computer hardware, the use of Computational- Fluid-Dynamics(CFD) calculations plays an important role. CFD-calculations are based on solving the equations of conservation (Navier-Stokes equations) using a finite element mesh or a finite volume mesh of the model. For a finned cask lying under a cover, where the main contributing element for heat removal is natural convection in combination with the thermal radiation, a CFD-calculation can be the most appropriate method. Common CFD-Codes are FLUENT
Coupled CFD/CSD Analysis of an Active-Twist Rotor in a Wind Tunnel with Experimental Validation
Massey, Steven J.; Kreshock, Andrew R.; Sekula, Martin K.
2015-01-01
An unsteady Reynolds averaged Navier-Stokes analysis loosely coupled with a comprehensive rotorcraft code is presented for a second-generation active-twist rotor. High fidelity Navier-Stokes results for three configurations: an isolated rotor, a rotor with fuselage, and a rotor with fuselage mounted in a wind tunnel, are compared to lifting-line theory based comprehensive rotorcraft code calculations and wind tunnel data. Results indicate that CFD/CSD predictions of flapwise bending moments are in good agreement with wind tunnel measurements for configurations with a fuselage, and that modeling the wind tunnel environment does not significantly enhance computed results. Actuated rotor results for the rotor with fuselage configuration are also validated for predictions of vibratory blade loads and fixed-system vibratory loads. Varying levels of agreement with wind tunnel measurements are observed for blade vibratory loads, depending on the load component (flap, lag, or torsion) and the harmonic being examined. Predicted trends in fixed-system vibratory loads are in good agreement with wind tunnel measurements.
This paper gives an overview of the advances in development and use of computational fluid dynamics (CFD) models and codes for industrial, particularly multiphase processing applications. Experimental needs for validation and improvement of CFD models and soft wares are highlighted. Integration of advanced CFD modelling with radioisotopes or tracer techniques as a complementary technology for future research and industrial applications is discussed. (author)
Hadade, Ioan; di Mare, Luca
2016-08-01
Modern multicore and manycore processors exhibit multiple levels of parallelism through a wide range of architectural features such as SIMD for data parallel execution or threads for core parallelism. The exploitation of multi-level parallelism is therefore crucial for achieving superior performance on current and future processors. This paper presents the performance tuning of a multiblock CFD solver on Intel SandyBridge and Haswell multicore CPUs and the Intel Xeon Phi Knights Corner coprocessor. Code optimisations have been applied on two computational kernels exhibiting different computational patterns: the update of flow variables and the evaluation of the Roe numerical fluxes. We discuss at great length the code transformations required for achieving efficient SIMD computations for both kernels across the selected devices including SIMD shuffles and transpositions for flux stencil computations and global memory transformations. Core parallelism is expressed through threading based on a number of domain decomposition techniques together with optimisations pertaining to alleviating NUMA effects found in multi-socket compute nodes. Results are correlated with the Roofline performance model in order to assert their efficiency for each distinct architecture. We report significant speedups for single thread execution across both kernels: 2-5X on the multicore CPUs and 14-23X on the Xeon Phi coprocessor. Computations at full node and chip concurrency deliver a factor of three speedup on the multicore processors and up to 24X on the Xeon Phi manycore coprocessor.
Study of supercritical carbon dioxide natural circulation by the use of CFD codes
Molfese, E.; Ambrosini, W.; Forgione, N., E-mail: w.ambrosini@ing.unipi.it, E-mail: n.forgione@ing.unipi.it [Univ. of Pisa, Dipartimento di Ingegneria Meccanica Nucleare e della Produzione (Italy); Vijayan, P.K.; Sharma, M., E-mail: vijayanp@barc.gov.in, E-mail: manishs@barc.gov.in [Bhabha Atomic Research Centre, Reactor Engineering Div., Mumbai (India)
2011-07-01
In this paper, experiments on natural circulation of CO{sub 2}, previously performed at the Bhabha Atomic Research Centre (BARC), are addressed by the use of the FLUENT and the STAR-CCM+ CFD codes. The experiments were carried out in an experimental facility installed at the Reactor Engineering Division of BARC in Mumbai, consisting in a uniform diameter (13.88 mm ID & 21.34 mm OD) rectangular loop (SCNCL) with different orientations of heater and cooler, which can operate with either supercritical water and supercritical carbon dioxide. The tests with carbon dioxide were performed at different power levels, at the supercritical pressures of 8.6 and 9.1 MPa. The steady-state characteristics of the loop were obtained for the horizontal heater and the horizontal cooler configuration (HHHC) and for the horizontal heater and vertical cooler one (HHVC). Unstable behaviour was observed only for the HHHC configuration. The FLUENT and the STAR-CCM+ codes were adopted for reproducing the observed behaviour of the experimental loop in the HHHC configuration. Steady-state as well as transient analyses were performed to be compared with the observed behaviour of the loop. (author)
Study of supercritical carbon dioxide natural circulation by the use of CFD codes
In this paper, experiments on natural circulation of CO2, previously performed at the Bhabha Atomic Research Centre (BARC), are addressed by the use of the FLUENT and the STAR-CCM+ CFD codes. The experiments were carried out in an experimental facility installed at the Reactor Engineering Division of BARC in Mumbai, consisting in a uniform diameter (13.88 mm ID & 21.34 mm OD) rectangular loop (SCNCL) with different orientations of heater and cooler, which can operate with either supercritical water and supercritical carbon dioxide. The tests with carbon dioxide were performed at different power levels, at the supercritical pressures of 8.6 and 9.1 MPa. The steady-state characteristics of the loop were obtained for the horizontal heater and the horizontal cooler configuration (HHHC) and for the horizontal heater and vertical cooler one (HHVC). Unstable behaviour was observed only for the HHHC configuration. The FLUENT and the STAR-CCM+ codes were adopted for reproducing the observed behaviour of the experimental loop in the HHHC configuration. Steady-state as well as transient analyses were performed to be compared with the observed behaviour of the loop. (author)
Radiation Coupling with the FUN3D Unstructured-Grid CFD Code
Wood, William A.
2012-01-01
The HARA radiation code is fully-coupled to the FUN3D unstructured-grid CFD code for the purpose of simulating high-energy hypersonic flows. The radiation energy source terms and surface heat transfer, under the tangent slab approximation, are included within the fluid dynamic ow solver. The Fire II flight test, at the Mach-31 1643-second trajectory point, is used as a demonstration case. Comparisons are made with an existing structured-grid capability, the LAURA/HARA coupling. The radiative surface heat transfer rates from the present approach match the benchmark values within 6%. Although radiation coupling is the focus of the present work, convective surface heat transfer rates are also reported, and are seen to vary depending upon the choice of mesh connectivity and FUN3D ux reconstruction algorithm. On a tetrahedral-element mesh the convective heating matches the benchmark at the stagnation point, but under-predicts by 15% on the Fire II shoulder. Conversely, on a mixed-element mesh the convective heating over-predicts at the stagnation point by 20%, but matches the benchmark away from the stagnation region.
Application of TONUS V2006 and FLUENT 6.2.16 CFD codes to ENACCEF hydrogen combustion tests
Three ENACCEF hydrogen combustion tests have been simulated for code validation purposes using the TONUS V2006 and FLUENT 6.2.16 CFD software. The test series investigated deflagration in a uniform hydrogen concentration, in a concentration that decreases and in a concentration that increases along the height of the facility. In the TONUS calculations the CREBCOM combustion model and k-e turbulence model with Eddy Break-Up (EBU) reaction kinetics have been used. In the FLUENT calculations only the k-e and EBU models have been used and the simulation results are compared to each other and the test results. TONUS CREBCOM results of the uniform mixture case obtained by 3D model of the facility are qualitatively in a reasonable agreement with the test results. In the increasing concentration case the flame speeds are exaggerated while in the decreasing concentration case the flame acceleration is underestimated. Further evaluation of the model parameters is suggested for non-homogenous mixtures. Generally, the EBU calculations by FLUENT show similar pressures and flame speed profiles with slightly higher maximum speeds as the CREBCOM cases. Also the FLUENT results are in a relatively good agreement with the test results. (orig.)
Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)
Validating CFD Models of Multiphase Mixing in the Waste Treatment Plant at the Hanford Site
The Columbia River in Washington State is threatened by the radioactive legacy of the cold war. Two hundred thousand cubic meters (fifty-three million US gallons) of radioactive waste is stored in 177 underground tanks (60% of the Nation's radioactive waste). A vast complex of waste treatment facilities is being built to convert this waste into stable glass (vitrification). The waste in these underground tanks is a combination of sludge, slurry, and liquid. The waste will be transported to a pre-treatment facility where it will be processed before vitrification. It is necessary to keep the solids in suspension during processing. The mixing devices selected for this task are known as pulse-jet mixers (PJMs). PJMs cyclically empty and refill with the contents of the vessel to keep it mixed. The transient operation of the PJMs has been proven successful in a number of applications, but needs additional evaluation to be proven effective for the slurries and requirements at the Waste Treatment Plant (WTP). Computational fluid dynamic (CFD) models of mixing vessels have been developed to demonstrate the ability of the PJMs to meet mixing criteria. Experimental studies have been performed to validate these models. These tests show good agreement with the transient multiphase CFD models developed for this engineering challenge. (authors)
CFD Modeling of Free-Piston Stirling Engines
Ibrahim, Mounir B.; Zhang, Zhi-Guo; Tew, Roy C., Jr.; Gedeon, David; Simon, Terrence W.
2001-01-01
NASA Glenn Research Center (GRC) is funding Cleveland State University (CSU) to develop a reliable Computational Fluid Dynamics (CFD) code that can predict engine performance with the goal of significant improvements in accuracy when compared to one-dimensional (1-D) design code predictions. The funding also includes conducting code validation experiments at both the University of Minnesota (UMN) and CSU. In this paper a brief description of the work-in-progress is provided in the two areas (CFD and Experiments). Also, previous test results are compared with computational data obtained using (1) a 2-D CFD code obtained from Dr. Georg Scheuerer and further developed at CSU and (2) a multidimensional commercial code CFD-ACE+. The test data and computational results are for (1) a gas spring and (2) a single piston/cylinder with attached annular heat exchanger. The comparisons among the codes are discussed. The paper also discusses plans for conducting code validation experiments at CSU and UMN.
Martin-Valdepenas, J.M.; Jimenez, M.A.; Martin-Fuertes, F. [Universidad Politecnica de Madrid, Department of Nuclear Engineering, Jose Gutierrez Abascal, 2, E-28006, Madrid (Spain); Benitez, J.A. Fernandez [Universidad Politecnica de Madrid, Departamento de Ingenieria Energetica y Fluidomecanica, Jose Gutierrez Abascal, 2, E-28006, Madrid (Spain)
2005-09-01
Several film condensation models in presence of non-condensable gases are presented. They have been implemented in a CFD code and compared with experimental data. The aim was to improve the code for simulating the gas mixing process in large containment buildings involving steam. The models based on correlation are more robust and simpler, but they work badly out of their experimental conditions. The mechanistic models, based on the diffusion layer theory, work well in numerous conditions but the algorithm are more complicated. Moreover, they run badly when the convective heat transfer is not well predicted by the code. (orig.)
Christian F. Janßen; Dennis Mierke; Micha Überrück; Silke Gralher; Thomas Rung
2015-01-01
This contribution is dedicated to demonstrating the high potential and manifold applications of state-of-the-art computational fluid dynamics (CFD) tools for free-surface flows in civil and environmental engineering. All simulations were performed with the academic research code ELBE (efficient lattice boltzmann environment, http://www.tuhh.de/elbe). The ELBE code follows the supercomputing-on-the-desktop paradigm and is especially designed for local supercomputing, without tedious accesses t...
Simulation of boiling flow experiments close to CHF with the NEPTUNE-CFD code
A three-dimensional two-fluid code NEPTUNECFD has been validated against the ASU (Arizona State University) [1] and DEBORA [2, 3] boiling flow experiments. Nucleate boiling processes in the subcooled flow boiling regime have been studied on ASU experiments. Within this scope a new wall function is implemented in the NEPTUNECFD V1.0.6 code to improve the prediction of flow parameters in the boiling boundary layer. The capability of the code to predict boiling flow regime close to critical heat flux (CHF) conditions has been assessed on selected DEBORA experiments. It was shown that the code is able to predict wall temperature excursion and a sharp void fraction increase near the heated wall, which are characteristic phenomena for CHF conditions. (author)
Porous Media Approach of a CFD Code to Analyze a PWR Component with Tube or Rod Bundles
This paper presents a strategy to innovate CFD code into a PWR component analysis code. A porous media approach is adapted to two-fluid model and conductor model, and a pack of constitutive relations to close the numerical model into component analysis code. The separate verification calculations on open media, conductor model and porous media approach are introduced. Based on the CUPID code, the component analysis code has been developed. For porous media model, constitutive correlations of a two-phase flow regime map, interfacial area, interfacial heat and mass transfer, interfacial drag, wall friction, wall heat transfer and heat partitioning in flows through tube or rod bundles are added. Separate calculations were also conducted to verify the developed code
The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain
Study of the distribution of steam plumes in the PANDA facility using CFD code
Highlights: • The standard k–ε model has been verified for gas plume simulation in the large-scale volume. • The k–kl–ω model has been improved for gas plume simulations. • The sensitivity analyses about the computational mesh, time step, Froude numbers have been carried out. - Abstract: During a postulated severe accident in light water reactor, a large amount of steam is injected into containment through the break. This would lead to the increases of pressure and temperature, and consequently threaten the integrity of the containment. In this study the light gas (saturated steam) distribution in a large-scale multi-compartment volume is simulated by using CFD code. Several turbulence models, including the standard k–ε model, the k–kl–ω model, the transitional SST model, and the improved k–kl–ω model with considering buoyancy effect are used for the simulation. The results show that both the standard k–ε model and the improved k–kl–ω model with considering the buoyancy effect can get good results comparing to the experimental results. The improved k–kl–ω model can get much better than the original k–kl–ω model without considering the buoyancy effect for predicting the steam distribution in vessels, and some characteristics in concerned region are predicted well. The sensitivity analyses about the computational mesh, time step, Froude numbers are also carried out
Software verification and validation plan for the GWSCREEN code
The purpose of this Software Verification and Validation Plan (SVVP) is to prescribe steps necessary to verify and validate the GWSCREEN code, version 2.0 to Quality Level B standards. GWSCREEN output is to be verified and validated by comparison with hand calculations, and by output from other Quality Level B computer codes. Verification and validation will also entail performing static and dynamic tests on the code using several analysis tools. This approach is consistent with guidance in the ANSI/ANS-10.4-1987, open-quotes Guidelines for Verification and Validation of Scientific and Engineering Computer Programs for the Nuclear Industry.close quotes
Dannemand, Mark; Fan, Jianhua; Furbo, Simon;
2014-01-01
Computational Fluid Dynamics (CFD) model. The CFD calculated temperatures are compared to measured temperatures internally in the box to validate the CFD model. Four cases are investigated; heating the test module with the sodium acetate water mixture in solid phase from ambient temperature to 52˚C; heating the...... the crystallization, ending at ambient temperature with the sodium acetate water mixture in solid phase. Comparisons have shown reasonable good agreement between experimental measurements and theoretical simulation results for the investigated scenarios....
Relating system-to-CFD coupled code analyses to theoretical framework of a multi-scale method
Cadinu, F.; Kozlowski, T.; Dinh, T.N. [Royal Institute of Technology, Div. of Nuclear Power Safety, Stockholm (Sweden)
2007-07-01
Over past decades, analyses of transient processes and accidents in a nuclear power plant have been performed, to a significant extent and with a great success, by means of so called system codes, e.g. RELAP5, CATHARE, ATHLET codes. These computer codes, based on a multi-fluid model of two-phase flow, provide an effective, one-dimensional description of the coolant thermal-hydraulics in the reactor system. For some components in the system, wherever needed, the effect of multi-dimensional flow is accounted for through approximate models. The later are derived from scaled experiments conducted for selected accident scenarios. Increasingly, however, we have to deal with newer and ever more complex accident scenarios. In some such cases the system codes fail to serve as simulation vehicle, largely due to its deficient treatment of multi-dimensional flow (in e.g. downcomer, lower plenum). A possible way of improvement is to use the techniques of Computational Fluid Dynamics (CFD). Based on solving Navier-Stokes equations, CFD codes have been developed and used, broadly, to perform analysis of multi-dimensional flow, dominantly in non-nuclear industry and for single-phase flow applications. It is clear that CFD simulations can not substitute system codes but just complement them. Given the intrinsic multi-scale nature of this problem, we propose to relate it to the more general field of research on multi-scale simulations. Even though multi-scale methods are developed on case-by-case basis, the need for a unified framework brought to the development of the heterogeneous multi-scale method (HMM)
CFD simulation of a burner for syngas characterization and experimental validation
Fantozzi, Francesco; Desideri, Umberto [University of Perugia (Italy). Dept. of Industrial Engineering], Emails: fanto@unipg.it, umberto.desideri@unipg.it; D' Amico, Michele [University of Perugia (Italy). Dept. of Energetic Engineering], E-mail: damico@crbnet.it
2009-07-01
Biomass and waste are distributed and renewable energy sources that may contribute effectively to sustainability if used on a small and micro scale. This requires the transformation through efficient technologies (gasification, pyrolysis and anaerobic digestion) into a suitable gaseous fuel to use in small internal combustion engines and gas turbines. The characterization of biomass derived syngas during combustion is therefore a key issue to improve the performance of small scale integrated plants because synthesis gas show significant differences with respect to Natural Gas (mixture of gases, low calorific value, hydrogen content, tar and particulate content) that may turn into ignition problems, combustion instabilities, difficulties in emission control and fouling. To this aim a burner for syngas combustion and LHV measurement through mass and energy balance was realized and connected to the rotary-kiln laboratory scale pyrolyzer at the Department of Industrial Engineering of the University of Perugia. A computational fluid dynamics (CFD) simulation of the burner was carried out considering the combustion of propane to investigate temperature and pressure distribution, heat transmission and distribution of the combustion products and by products. The simulation was carried out using the CFD program Star-CD. Before the simulation a geometrical model of the burner was built and the volume of model was subdivided in cells. A sensibility analysis of cells was carried out to estimate the approximation degree of the model. Experimental data about combustion emission were carried out with the propane combustion in the burner, the comparison between numerical results and experimental data was studied to validate the simulation for future works involved with the combustion of treated or raw (syngas with tar) syngas obtained from pyrolysis process. (author)
Highlights: • Models for large interfaces in two-phase CFD were developed for PTS. • The COSI experiment is used for NEPTUNECFD integral validation. • COSI is a PWR cold leg scaled 1/100 for volume. • Fifty runs are calculated, covering a large range of flow configurations. • The CFD predicting capability is analysed using global and local measurements. - Abstract: In the context of the Pressurized Water Reactors (PWR) life duration safety studies, some models were developed to address the Pressurized Thermal Shock (PTS) from the two-phase CFD angle, dealing with interfaces much larger than cells size and with direct contact condensation. Such models were implemented in NEPTUNECFD, a 3D transient Eulerian two-fluid model. The COSI experiment is used for its integral validation. It represents a cold leg scaled 1/100 for volume and power from a 900 MW PWR under a large range of LOCA PTS conditions. In this study, the CFD is evaluated in the whole range of parameters and flow configurations covered by the experiment. In a first step, a single choice of mesh and CFD models parameters is fixed and justified. In a second step, fifty runs are calculated. The CFD predicting capability is analysed, comparing the liquid temperature and the total condensation rate with the experiment, discussing their dependency on the inlet cold liquid rate, on the liquid level in the cold leg and on the difference between co-current and counter-current runs. It is shown that NEPTUNECFD 1.0.8 calculates with a fair agreement a large range of flow configurations related to ECCS injection and steam condensation
Validation and verification plan for safety and PRA codes
This report discusses a verification and validation (V ampersand V) plan for computer codes used for safety analysis and probabilistic risk assessment calculations. The present plan fulfills the commitments by Westinghouse Savannah River Company (WSRC) to the Department of Energy Savannah River Office (DOE-SRO) to bring the essential safety analysis and probabilistic risk assessment codes in compliance with verification and validation requirements
A validated CFD model to predict O₂ and CO₂ transfer within hollow fiber membrane oxygenators.
Hormes, Marcus; Borchardt, Ralf; Mager, Ilona; Rode, Thomas Schmitz; Behr, Marek; Steinseifer, Ulrich
2011-03-01
Hollow fiber oxygenators provide gas exchange to and from the blood during heart surgery or lung recovery. Minimal fiber surface area and optimal gas exchange rate may be achieved by optimization of hollow fiber shape and orientation (1). In this study, a modified CFD model is developed and validated with a specially developed micro membrane oxygenator (MicroMox). The MicroMox was designed in such a way that fiber arrangement and bundle geometry are highly reproducible and potential flow channeling is avoided, which is important for the validation. Its small size (V(Fluid)=0.04 mL) allows the simulation of the entire bundle of 120 fibers. A non-Newtonian blood model was used as simulation fluid. Physical solubility and chemical bond of O₂ and CO₂ in blood was represented by the numerical model. Constant oxygen partial pressure at the pores of the fibers and a steady state flow field was used to calculate the mass transport. In order to resolve the entire MicroMox fiber bundle, the mass transport was simulated for symmetric geometry sections in flow direction. In vitro validation was achieved by measurements of the gas transfer rates of the MicroMox. All measurements were performed according to DIN EN 12022 (2) using porcine blood. The numerical simulation of the mass transfer showed good agreement with the experimental data for different mass flows and constant inlet partial pressures. Good agreement could be achieved for two different fiber configurations. Thus, it was possible to establish a validated model for the prediction of gas exchange in hollow fiber oxygenators. PMID:21462147
QM-400 CFD 自然对流模型研究及验证%Research and Validation on CFD Natural Convection Model of QM-400
左巧林; 干富军; 朱丽兵
2016-01-01
The spent fuel dry storage facility named QM-400 module for Third Qinshan Nuclear Power Co.Ltd.(TQNPC)is the first commercial dry storage facility in opera-tion in China.The heat transfer in QM-400 mainly consists of natural convention,con-duction,conjugate heat transfer and radiation,etc.The decay heat of each fuel basket was calculated accurately at typical surrounding temperature.Mesh sensitivity analysis was performed using commercial computational fluid dynamics (CFD)code FLUENT 14.0. A set of CFD simulation models on natural convection of QM-400 were developed.The results show that the distributions of the pressure and temperature on the cylinder sur-face meet the rules of natural convection.Good agreements are achieved between the simulated temperature and the measured temperature at the measured points and the simulated temperature trend varying with surrounding temperature agree well with the measured trend,which demonstrates the correctness of the calculation method of natural convection in this paper.This work can be the reference of the further CFD simulation on temperature distributions of dry storage facility without thermal insulation panels.%秦山第三核电厂乏燃料干式贮存模块 QM-400是我国第一座投入商业运行的干式贮存设施，模块内的热量交换主要包括自然对流、热传导、耦合传热和辐射换热等。本文精确计算了典型环境温度下每个燃料篮的衰变热，运用商用计算流体动力学(CFD)软件 FLUENT 14.0开展了网格敏感性分析，并建立了 QM-400存储模块的自然对流 CFD 分析模型。结果表明，模块顶面、侧面以及贮存筒表面压力和温度分布符合自然对流规律，计算的测点温度与现场的实测温度符合良好，测点温度随环境温度的变化趋势也与实测趋势符合良好，证明了建立的 CFD 自然对流计算方法的正确性。本文结果为后续采用CFD 方法进行取消绝热板后的温度场计算奠定了基础。
Thermal-hydraulic analysis of water-water heat exchanger under low flow conditions using CFD code
In order to establish the evaluation method of the local heat transfer in the intermediate heat exchanger (IHX) for a fast breeder reactor, a CFD analysis method has been applied to a heat exchanger with the primary and secondary water-loops. Analyses were conducted under the forced circulation and natural circulation conditions. For the forced circulation experiment with the Reynolds number at 104, a quasi-steady state condition is analyzed. For the natural circulation experiment, an analysis is also conducted for a quasi-steady state condition where the Reynolds number is approximately 102. The calculated heat transfer coefficients are converted into the Nu numbers and compared with the experimental results. Good agreement is obtained between the analytical results and the test results. Temperature distributions by the calculation results with the 1-dimensional NETFLOW++ code and CFD code are compared with the test results. For the natural circulation condition, it is clarified that there is almost no temperature distribution in radial direction, and the temperature is distributed only in axial direction. The flow on the primary-side seems to be rectified by the group of the heat transfer tubes and the turbulence is suppressed. For the forced circulation condition, the flow on the primary-side of the heat exchanger is stabilized also. The present CFD evaluation method can be applied to the IHX of the fast reactor with complex flow system. (author)
Study of the distribution of steam plumes in the PANDA facility using CFD code
Guo, Shuanshuan [School of Physics and Engineering, Sun Yat-sen University, Guangzhou (China); Cai, Jiejin, E-mail: chiven77@hotmail.com [Sino-French Institute of Nuclear Engineering & Technology, Sun Yat-sen University, Guangzhou (China); Zhang, Huiyong [China Nuclear Power Technology Research Institute, Shenzhen 518026 (China); Yin, Huaqiang; Yang, Xingtuan [Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)
2015-08-15
Highlights: • The standard k–ε model has been verified for gas plume simulation in the large-scale volume. • The k–k{sub l}–ω model has been improved for gas plume simulations. • The sensitivity analyses about the computational mesh, time step, Froude numbers have been carried out. - Abstract: During a postulated severe accident in light water reactor, a large amount of steam is injected into containment through the break. This would lead to the increases of pressure and temperature, and consequently threaten the integrity of the containment. In this study the light gas (saturated steam) distribution in a large-scale multi-compartment volume is simulated by using CFD code. Several turbulence models, including the standard k–ε model, the k–k{sub l}–ω model, the transitional SST model, and the improved k–k{sub l}–ω model with considering buoyancy effect are used for the simulation. The results show that both the standard k–ε model and the improved k–k{sub l}–ω model with considering the buoyancy effect can get good results comparing to the experimental results. The improved k–k{sub l}–ω model can get much better than the original k–k{sub l}–ω model without considering the buoyancy effect for predicting the steam distribution in vessels, and some characteristics in concerned region are predicted well. The sensitivity analyses about the computational mesh, time step, Froude numbers are also carried out.
Ma, Baoshun; Ruwet, Vincent; Corieri, Patricia; Theunissen, Raf; Riethmuller, Michel; Darquenne, Chantal
2009-05-01
Accurate modeling of air flow and aerosol transport in the alveolated airways is essential for quantitative predictions of pulmonary aerosol deposition. However, experimental validation of such modeling studies has been scarce. The objective of this study is to validate CFD predictions of flow field and particle trajectory with experiments within a scaled-up model of alveolated airways. Steady flow (Re = 0.13) of silicone oil was captured by particle image velocimetry (PIV), and the trajectories of 0.5 mm and 1.2 mm spherical iron beads (representing 0.7 to 14.6 mum aerosol in vivo) were obtained by particle tracking velocimetry (PTV). At twelve selected cross sections, the velocity profiles obtained by CFD matched well with those by PIV (within 1.7% on average). The CFD predicted trajectories also matched well with PTV experiments. These results showed that air flow and aerosol transport in models of human alveolated airways can be simulated by CFD techniques with reasonable accuracy. PMID:20161301
Computer codes validation for conditions of core voiding
Void generation during a Loss of Coolant Accident (LOCA) in a core of a CANDU reactor is of specific importance because of its strong coupling with reactor neutronics. The use of dynamic behaviour and computer code capability to predict void generation accurately in the temporal and spatial domain of the reactor core is fundamental for the determination of CANDU safety. The Canadian industry has used the RD-14M test facilities for its code validation. The validation exercises for the Canadian computer codes TUF and CATHENA were performed some years ago. Recently, the CNSC has gained access to the USNRC computer code TRACE. This has provided an opportunity to explore the use of this code in CANDU related applications. As a part of regulatory assessment and resolving identified Generic Issues (GI), and in an effort to build independent thermal hydraulic computer codes assessment capability within the CNSC, preliminary validation exercises were performed using the TRACE computer code for an evaluation of the void generation phenomena. The paper presents a preliminary assessment of the TRACE computer code for an RD-14M channel voiding test. It is also a validation exercise of void generation for the TRACE computer code. The accuracy of the obtained results is discussed and compared with previous validation assessments that were done using the CATHENA and TUF codes. (author)
Convective heat transfer at subchannel in vertical cylinder arranged is very useful in many engineering application, include the design and operation of heat exchanger, steam generator and nuclear reactor safety. It is important to learn characteristic of fluid flow in subchannel before learn convective heat transfer in subchannel. In this research, theoretical study of flow characteristic in subchannel has been carried out by using CFD code. The subchannel is square arrangement and consist of nine cylinder heater with 2.54 cm diameter and P/D ratio of 1.5. For the inlet velocity are 0.01 m/s, 0.02 m/s and 0.03 m/s, the result of CFD analysis indicated that fully developed region is formed at 0.2 m below the reference axis. The velocity of coolant in the center of subchannel is faster than in the edge of subchannel. (author)
Qualification of the CFD code TrioU for full scale reactor applications
The article presents a procedure to qualify the TrioU code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem). The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the TrioU code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result
A fuel performance code TRUST VIc and its validation
This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs
Using the RELAP5-3D advanced systems analysis code with commercial and advanced CFD software
The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. A demonstration calculation is presented. Finally, future tasks and plans are outlined. (author)
Coupling the RELAP5-3d advanced systems analysis code with commercial and advanced CFD software
The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. Finally, future tasks and plans are outlined. (author)
Validation of impinging jet models to be used in CANDU calandria vessel CFD simulations
The knowledge of the external wall temperature distributions on calandria tubes is of major concern in nuclear safety analysis. One of the models used by the Canadian industry consists in replacing the calandria by an equivalent porous media with appropriate anisotropic hydraulic resistances. This technique has the advantage to treat a non-connected domain as an equivalent quasi-continuous media; however, it cannot provide information about local velocity variations. Within the framework of the present study, a full-scale modeling of the moderator using a Computational Fluid Dynamic code (FLUENT) is underway. The use of a 2D model have shown that the geometry of calandria nozzles have a strong effect on the flow distribution. Some authors suggest to model the flow at the entrance of the calandria as successive flow circulation through a portion of a straight pipe, a curved pipe, and a circular nozzle placed in front of an impinging plate, and to use the results as input data in full-scale calculations. Obtaining these data requires large computational resources before performing complete flow simulations, while they do necessarily represent neither the real geometry nor the actual flow conditions. Therefore, the present study is aimed to find appropriate water-jet modeling approaches that can help in improving moderator circulation simulations. In particular, the principal interest consists in finding a semi-analytical nozzle model that can be used as a constitutive relationship in a CFD code. This approach will contribute both to increase the number of meshes in the calandria vessel as well as to decrease the computational time. (author)
European validation of the integral code ASTEC (EVITA)
The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code Accident Source Term Evaluation Code (ASTEC) to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management (SAM) measures, and compare results with reference codes. The basis version V0 of ASTEC - commonly developed and basically validated by GRS and IRSN - was made available in late 2000 for the EVITA partners on their individual platforms. Users' training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A 'hot line' assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:-for reactor coolant system two-phase simplified thermal hydraulics (five-equation approach) during both front-end and core degradation phases;-for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Other main achievements of the project are up to now:-the EVITA validation matrix focused on main risk issues;-first validation results on PACTEL, STORM and EREC tests;-first plant applications on German 1300 PWR and VVER-440/V230
Validations of CFD against detailed velocity and pressure measurements in water turbine runner flow
Nilsson, H.; Davidson, L.
2003-03-01
This work compares CFD results with experimental results of the flow in two different kinds of water turbine runners. The runners studied are the GAMM Francis runner and the Hölleforsen Kaplan runner. The GAMM Francis runner was used as a test case in the 1989 GAMM Workshop on 3D Computation of Incompressible Internal Flows where the geometry and detailed best efficiency measurements were made available. In addition to the best efficiency measurements, four off-design operating condition measurements are used for the comparisons in this work. The Hölleforsen Kaplan runner was used at the 1999 Turbine 99 and 2001 Turbine 99 - II workshops on draft tube flow, where detailed measurements made after the runner were used as inlet boundary conditions for the draft tube computations. The measurements are used here to validate computations of the flow in the runner.The computations are made in a single runner blade passage where the inlet boundary conditions are obtained from an extrapolation of detailed measurements (GAMM) or from separate guide vane computations (Hölleforsen). The steady flow in a rotating co-ordinate system is computed. The effects of turbulence are modelled by a low-Reynolds number k- turbulence model, which removes some of the assumptions of the commonly used wall function approach and brings the computations one step further.
Mach number validation of a new zonal CFD method (ZAP2D) for airfoil simulations
Strash, Daniel J.; Summa, Michael; Yoo, Sungyul
1991-01-01
A closed-loop overlapped velocity coupling procedure has been utilized to combine a two-dimensional potential-flow panel code and a Navier-Stokes code. The fully coupled two-zone code (ZAP2D) has been used to compute the flow past a NACA 0012 airfoil at Mach numbers ranging from 0.3 to 0.84 near the two-dimensional airfoil C(lmax) point for a Reynolds number of 3 million. For these cases, the grid domain size can be reduced to 3 chord lengths with less than 3-percent loss in accuracy for freestream Mach numbers through 0.8. Earlier validation work with ZAP2D has demonstrated a reduction in the required Navier-Stokes computation time by a factor of 4 for subsonic Mach numbers. For this more challenging condition of high lift and Mach number, the saving in CPU time is reduced to a factor of 2.
CFD Validation Experiment of a Mach 2.5 Axisymmetric Shock-Wave Boundary-Layer Interaction
Davis, David O.
2015-01-01
Preliminary results of an experimental investigation of a Mach 2.5 two-dimensional axisymmetric shock-wave/boundary-layer interaction (SWBLI) are presented. The purpose of the investigation is to create a SWBLI dataset specifically for CFD validation purposes. Presented herein are the details of the facility and preliminary measurements characterizing the facility and interaction region. The results will serve to define the region of interest where more detailed mean and turbulence measurements will be made.
Highlights: • A DES of a turbocharger compressor working at peak pressure point is performed. • In-duct pressure signals are measured in a steady flow rig with 3-sensor arrays. • Pressure spectra comparison is performed as a validation for the numerical model. • A suitable comparison methodology is developed, relying on pressure decomposition. • Whoosh noise at outlet duct is detected in experimental and numerical spectra. - Abstract: Centrifugal compressors working in the surge side of the map generate a broadband noise in the range of 1–3 kHz, named as whoosh noise. This noise is perceived at strongly downsized engines operating at particular conditions (full load, tip-in and tip-out maneuvers). A 3-dimensional CFD model of a centrifugal compressor is built to analyze fluid phenomena related to whoosh noise. A detached eddy simulation is performed with the compressor operating at the peak pressure point of 160 krpm. A steady flow rig mounted on an anechoic chamber is used to obtain experimental measurements as a means of validation for the numerical model. In-duct pressure signals are obtained in addition to standard averaged global variables. The numerical simulation provides global variables showing excellent agreement with experimental measurements. Pressure spectra comparison is performed to assess noise prediction capability of numerical model. The influence of the type and position of the virtual pressure probes is evaluated. Pressure decomposition is required by the simulations to obtain meaningful spectra. Different techniques for obtaining pressure components are analyzed. At the simulated conditions, a broadband noise in 1–3 kHz frequency band is detected in the experimental measurements. This whoosh noise is also captured by the numerical model
D. Bestion
2009-01-01
Full Text Available The NURESIM Project of the 6th European Framework Program initiated the development of a new-generation common European Standard Software Platform for nuclear reactor simulation. The thermal-hydraulic subproject aims at improving the understanding and the predictive capabilities of the simulation tools for key two-phase flow thermal-hydraulic processes such as the critical heat flux (CHF. As part of a multi-scale analysis of reactor thermal-hydraulics, a two-phase CFD tool is developed to allow zooming on local processes. Current industrial methods for CHF mainly use the sub-channel analysis and empirical CHF correlations based on large scale experiments having the real geometry of a reactor assembly. Two-phase CFD is used here for understanding some boiling flow processes, for helping new fuel assembly design, and for developing better CHF predictions in both PWR and BWR. This paper presents a review of experimental data which can be used for validation of the two-phase CFD application to CHF investigations. The phenomenology of DNB and Dry-Out are detailed identifying all basic flow processes which require a specific modeling in CFD tool. The resulting modeling program of work is given and the current state-of-the-art of the modeling within the NURESIM project is presented.
CFD analysis of core melt spreading on the reactor cavity floor using ANSYS CFX code
Highlights: ► Spreading of core melt on nuclear reactor cavity is calculated using ANSYS CFX. ► Thermal radiation and viscosity of liquid–solid mixture of the melt are modeled. ► The code is validated with FARO and VULCANO spreading experiments. ► Calculation of a full-scale cavity shows the spreading completes within a minute. - Abstract: In the very unlikely event of a severe reactor accident involving core melt and reactor pressure vessel failure, it is important to provide an accident management strategy that would allow the molten core material to cool down, resolidify and bring the core debris to a coolable state for Light Water Reactors (LWRs). One approach to achieve a coolable state is to quench the core melt after its relocation from the reactor pressure vessel into the reactor cavity. This approach typically requires a large cavity floor area on which a large amount of core melt spreads well and forms a shallow melt thickness for small thermal resistance across the melt pool. Spreading of high temperature (∼3000 K), low superheat (∼200 K) core melt over a wide cavity floor has been a key question to the success of the ex-vessel core coolability. A computational model for the melt spreading requires a multiphase treatment of liquid melt, solidified melt, and air. Also solidification and thermal radiation physics should be included. This paper reports the approach and computational model development to simulate core melt spreading on the reactor cavity using ANSYS-CFX code. Solidification and thermal radiation heat transfer were modeled in the code and analyses of the FARO and VULCANO spreading experiments have been carried out to check the validity of the model. The calculation of 100 tons of core melt spreading over the full scale reactor cavity (6 m × 16 m) showed that the melt spread was completed within a minute.
European Validation of the Integral Code ASTEC (EVITA)
The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes. The basis version V0 of ASTEC (Accident Source Term Evaluation Code)-commonly developed and basically validated by GRS and IRSN-was made available in late 2000 for the EVITA partners on their individual platforms. Users' training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A 'hot line' assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:- for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases; - for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users' tools. As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme
Simulation codes and the impact of validation/uncertainty requirements
Several of the OECD/CSNI members have adapted a proposed methodology for code validation and uncertainty assessment. Although the validation process adapted by members has a high degree of commonality, the uncertainty assessment processes selected are more variable, ranaing from subjective to formal. This paper describes the validation and uncertainty assessment process, the sources of uncertainty, methods of reducing uncertainty, and methods of assessing uncertainty.Examples are presented from the Ontario Hydro application of the validation methodology and uncertainty assessment to the system thermal hydraulics discipline and the TUF (1) system thermal hydraulics code. (author)
The objective of this work is to study the flow and heat transfer for water under super-critical conditions. Two dimensional (axi-symmetric) CFD simulation is performed for this purpose using an in-house developed code named NAFA. The flow is computed for vertically upward as well as downward orientations. Further, for each orientation, wide range of heat flux is considered. It is found that for downward flow, heat transfer coefficient is higher than that for upward flow, other conditions remaining same. The heat transfer characteristics are found to be dependent on the pipe outlet temperature with reference to pseudo-critical temperature. (author)
Counter current flow limitation CCFL is one of the phenomena that incorporate complex two-phase flows, including the existence of numerous flow patterns simultaneously, a complicated gas/liquid interface, and interfacial momentum transfer. Such a complexity makes it one of the challenging two-phase flow configurations for CFD validation. Numerous experimental investigations were carried out in recent years to enlarge the existing knowledge about this phenomenon. However, most of those investigations were carried out either in small-diameter geometry, or in a non-realistic geometry (rectangular cross section instead of a circular pipe). A review of experimental investigations shows that the scale and geometry have a large impact upon CCFL. In order to provide a better understanding of this phenomenon in a real PWR hot-leg geometry, and at a relatively large-diameter and scale, a test facility was constructed for this purpose. The facility consists of a reactor vessel simulator, a hot-leg geometry pipe (with 190 mm inner diameter), and a steam generator simulator. The facility represents a ∼1/3.9 scale of a PWR geometry and is completely made of transparent material allowing detailed optical observations. Experimental investigations were carried out at atmospheric pressure using distilled water and air. High-speed recording was implemented to acquire high-quality images of the air/water interface for experimental analysis and CFD validations. CCFL mechanisms, flow patterns, and the limits of the onset of CCFL and deflooding were experimentally identified. Current measurements are compared against previous investigations showing diverse effects of scale and geometry upon results. CFD simulations of two representative experimental cases were carried out and validated against the experimentally acquired air/water interface and the pressure difference between the reactor vessel and the steam generator. The CFD simulations shows the required improvements of this
Validation uncertainty of MATRA code for subchannel void distributions
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
Validation uncertainty of MATRA code for subchannel void distributions
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
Computer code validation by high temperature chemistry
At least five of the computer codes utilized in analysis of severe fuel damage-type events are directly dependent upon or can be verified by high temperature chemistry. These codes are ORIGEN, CORSOR, CORCON, VICTORIA, and VANESA. With the exemption of CORCON and VANESA, it is necessary that verification experiments be performed on real irradiated fuel. For ORIGEN, the familiar knudsen effusion cell is the best choice and a small piece of known mass and known burn-up is selected and volatilized completely into the mass spectrometer. The mass spectrometer is used in the integral mode to integrate the entire signal from preselected radionuclides, and from this integrated signal the total mass of the respective nuclides can be determined. For CORSOR and VICTORIA, experiments with flowing high pressure hydrogen/steam must flow over the irradiated fuel and then enter the mass spectrometer. For these experiments, a high pressure-high temperature molecular beam inlet must be employed. Finally, in support of VANESA-CORCON, the very highest temperature and molten fuels must be contained and analyzed. Results from all types of experiments will be discussed and their applicability to present and future code development will also be covered
Befrui, Bizhan A.
1995-01-01
This viewgraph presentation discusses the following: STAR-CD computational features; STAR-CD turbulence models; common features of industrial complex flows; industry-specific CFD development requirements; applications and experiences of industrial complex flows, including flow in rotating disc cavities, diffusion hole film cooling, internal blade cooling, and external car aerodynamics; and conclusions on turbulence modeling needs.
Performance of the OVERFLOW-MLP and LAURA-MLP CFD Codes on the NASA Ames 512 CPU Origin System
Taft, James R.
2000-01-01
The shared memory Multi-Level Parallelism (MLP) technique, developed last year at NASA Ames has been very successful in dramatically improving the performance of important NASA CFD codes. This new and very simple parallel programming technique was first inserted into the OVERFLOW production CFD code in FY 1998. The OVERFLOW-MLP code's parallel performance scaled linearly to 256 CPUs on the NASA Ames 256 CPU Origin 2000 system (steger). Overall performance exceeded 20.1 GFLOP/s, or about 4.5x the performance of a dedicated 16 CPU C90 system. All of this was achieved without any major modification to the original vector based code. The OVERFLOW-MLP code is now in production on the inhouse Origin systems as well as being used offsite at commercial aerospace companies. Partially as a result of this work, NASA Ames has purchased a new 512 CPU Origin 2000 system to further test the limits of parallel performance for NASA codes of interest. This paper presents the performance obtained from the latest optimization efforts on this machine for the LAURA-MLP and OVERFLOW-MLP codes. The Langley Aerothermodynamics Upwind Relaxation Algorithm (LAURA) code is a key simulation tool in the development of the next generation shuttle, interplanetary reentry vehicles, and nearly all "X" plane development. This code sustains about 4-5 GFLOP/s on a dedicated 16 CPU C90. At this rate, expected workloads would require over 100 C90 CPU years of computing over the next few calendar years. It is not feasible to expect that this would be affordable or available to the user community. Dramatic performance gains on cheaper systems are needed. This code is expected to be perhaps the largest consumer of NASA Ames compute cycles per run in the coming year.The OVERFLOW CFD code is extensively used in the government and commercial aerospace communities to evaluate new aircraft designs. It is one of the largest consumers of NASA supercomputing cycles and large simulations of highly resolved full
An Improved FFR Design with a Ventilation Fan: CFD Simulation and Validation.
Zhang, Xiaotie; Li, Hui; Shen, Shengnan; Rao, Yu; Chen, Feng
2016-01-01
This article presents an improved Filtering Facepiece Respirator (FFR) designed to increase the comfort of wearers during low-moderate work. The improved FFR aims to lower the deadspace temperature and CO2 level by an active ventilation fan. The reversing modeling is used to build the 3D geometric model of this FFR; the Computational Fluid Dynamics (CFD) simulation is then introduced to investigate the flow field. Based on the simulation result, the ventilation fan of the improved FFR can fit the flow field well when placed in the proper blowing orientation; streamlines from this fan show a cup-shape distribution and are perfectly matched to the shape of the FFR and human face when the fan blowing inward. In the deadspace of the improved FFR, the CO2 volume fraction is controlled by the optimized flow field. In addition, an experimental prototype of the improved FFR has been tested to validate the simulation. A wireless temperature sensor is used to detect the temperature variation inside the prototype FFR, deadspace temperature is lowered by 2 K compared to the normal FFR without a fan. An infrared camera (IRC) method is used to elucidate the temperature distribution on the prototype FFR's outside surface and the wearer's face, surface temperature is lowered notably. Both inside and outside temperature results from the simulation are in agreement with experimental results. Therefore, adding an inward-blowing fan on the outer surface of an N95 FFR is a feasible approach to reducing the deadspace CO2 concentration and improve temperature comfort. PMID:27454123
An Improved FFR Design with a Ventilation Fan: CFD Simulation and Validation
Zhang, Xiaotie; Li, Hui; Shen, Shengnan; Rao, Yu; Chen, Feng
2016-01-01
This article presents an improved Filtering Facepiece Respirator (FFR) designed to increase the comfort of wearers during low-moderate work. The improved FFR aims to lower the deadspace temperature and CO2 level by an active ventilation fan. The reversing modeling is used to build the 3D geometric model of this FFR; the Computational Fluid Dynamics (CFD) simulation is then introduced to investigate the flow field. Based on the simulation result, the ventilation fan of the improved FFR can fit the flow field well when placed in the proper blowing orientation; streamlines from this fan show a cup-shape distribution and are perfectly matched to the shape of the FFR and human face when the fan blowing inward. In the deadspace of the improved FFR, the CO2 volume fraction is controlled by the optimized flow field. In addition, an experimental prototype of the improved FFR has been tested to validate the simulation. A wireless temperature sensor is used to detect the temperature variation inside the prototype FFR, deadspace temperature is lowered by 2 K compared to the normal FFR without a fan. An infrared camera (IRC) method is used to elucidate the temperature distribution on the prototype FFR's outside surface and the wearer's face, surface temperature is lowered notably. Both inside and outside temperature results from the simulation are in agreement with experimental results. Therefore, adding an inward-blowing fan on the outer surface of an N95 FFR is a feasible approach to reducing the deadspace CO2 concentration and improve temperature comfort. PMID:27454123
Barrera matalla, J. E.; Hernandez Gomez, J.; Riverala Gurruchaga, J.
2012-07-01
Irradiated fuel has become an object of interest in the industry by the importance of ensuring its safety during long periods of storage time. New containers, stores, methods and codes will be used to ensure a suitable cooling and residual heat removal, and secure the safety of fuel elements in dry storage. The codes CFD (Computational Fluid Dynamics) have great potential to help in design of containers and stores, improving thermal-hydraulic performance and the extraction of heat generated.
In the framework of the 5th EU-FWP project ECORA the capabilities of CFD software packages for simulating flows in the containment of nuclear reactors was evaluated. Four codes were assessed using two basic tests in the PANDA facility addressing the transport of gases in a multi-compartment geometry. The assessment included a first attempt to use Best Practice Guidelines (BPGs) for the analysis of long, large-scale, transient problems. Due to the large computational overhead of the analysis, the BPGs could not fully be applied. It was thus concluded that the application of the BPGs to full containment analysis is out of reach with the currently available computer power. On the other hand, CFD codes used with a sufficiently detailed mesh seem to be capable to give reliable answers on issues relevant for containment simulation using standard two-equation turbulence models. Development on turbulence models is constantly ongoing. If it turns out that advanced (and more computationally intensive) turbulence models may not be needed, the use of the BPGs for 'certified' simulations could become feasible within a relatively short time
In the framework of the 5. EU-FWP project ECORA the capabilities of CFD software packages for simulating flows in the containment of nuclear reactors was evaluated. Four codes were assessed using two basic tests in the PANDA facility addressing the transport of gases in a multi-compartment geometry. The assessment included a first attempt to use Best Practice Guidelines (BPG) to the analysis of long, large-scale, transient problems. Due to the large computational overhead of the analysis, the BPGs could not fully be applied. It was thus concluded that the application of the BPGs to full containment analysis is out of reach with the currently available computer power. On the other hand, CFD codes used with a sufficiently detailed mesh seem to be capable to give reliable answers on issues relevant for containment simulation using standard two-equation turbulence models. Development on turbulence models is constantly ongoing. If it turns out that advanced (and more computationally intensive) turbulence models may not be needed, the use of the BPG for 'certified' simulations could become feasible within a relatively short time. (authors)
This paper reports the preliminary studies carried out with the CFD (computational fluid dynamics) code TrioU to study the natural gas circulation that may flow in the primary circuit of a pressurized water reactor during a high-pressure severe accident scenario. Two types of 3-dimensional simulations have been performed on one loop using a LES (large eddy simulations) approach. In the first type of calculations, the gas flow in the hot leg has been investigated with a simplified representation of the reactor vessel and the Steam Generator (SG) tubes. Structured and unstructured meshing have been tested on the full-scale geometry with and without radiative heat transfer modelling between walls and gas. The second type of calculations deals with the gas circulation in the SG. The first results show a good agreement with the available experimental data and provide some confidence in the TrioU code to simulate complex natural flows. (authors)
Twenty four years after the Three Mile Island Accident, Hydrogen risk remains a safety issue for current and future Pressurized Water Reactors (PWR). The formation of a combustible gas mixture in the complex geometry of a reactor containment depends on the understanding of hydrogen production, complex 3D flow due to gas/steam injection, natural convection, heat transfer by condensation on walls and effect of mitigation devices. Lumped parameter safety codes mainly developed for full containment analysis are not able to accurately predict the local gas mixing within the containment. 3D CFD codes are required but a thorough validation process on well-instrumented experimental data is necessary before they can be used with a high degree of confidence. The MISTRA coupled effect test facility has been recently built at CEA to fulfill these objectives: numerous measurement points in the gaseous volume (temperature and gas concentration) and the use of Laser technology (L.D.V. and P.I.V.) provide suitable experimental data for code validation. The in-house CEA-IRSN CAST3M/TONUS code is developed and validated against experimental data provided by this facility. Some of these tests have been proposed to the international community for code benchmarking (MICOCO benchmark and OECD/ISP47 exercise). Finally, extrapolation to global containment scale requires the validation of the code on more complex flow patterns and a detailed investigation of scaling effects. These two items will be the guidelines of future MISTRA tests
To model Generation IV reactor systems in detail, INEEL is currently developing a new thermal hydraulic analysis tool coupling RELAP5-3D / ATHENA and the computational fluid dynamics (CFD) software, Fluent. One of the first steps in this endeavor is extensive validation and verification (V and V) of Fluent for various situations of interest, such as the abrupt expansion of a gas entering a gas-cooled reactor core. Fluent results were compared to validation data provided by Baughn, et al. on turbulent air flow through an axisymmetric pipe expansion with constant wall heat flux [1] and uniform wall temperature [2]. Fluent peak Nusselt numbers varied 25% from validation data ell outside experimental uncertainties of 5%. However, non-peak Nusselt numbers varied only 10% from validation data and fully-developed Nusselt numbers were in good agreement with widely-accepted empirical relations such as the Dittus-Boelter Correlation. (authors)
Richard W. Johnson; Hugh M. McIlroy
2010-08-01
The U. S. Department of Energy (DOE) is supporting the development of a next generation nuclear plant (NGNP), which will be based on a very high temperature reactor (VHTR) design. The VHTR is a single-phase helium-cooled reactor wherein the helium will be heated initially to 750 °C and later to temperatures approaching 1000 °C. The high temperatures are desired to increase reactor efficiency and to provide a heat source for the manufacture of hydrogen and other applications. While computational fluid dynamics (CFD) has not been used in the past to design or license nuclear reactors in the U. S., it is expected that CFD will be used in the design and safety analysis of forthcoming designs. This is partly because of the maturity of CFD and partly because detailed information is desired of the flow and heat transfer inside the reactor to avoid hot spots and other conditions that might compromise reactor safety. Numerical computations of turbulent flow should be validated against experimental data for flow conditions that contain some or all of the physics expected in the thermal fluid machinery of interest. To this end, a scaled model of a narrow slice of the lower plenum of the prismatic VHTR was constructed and installed in the Idaho National Laboratory’s (INL) matched index of refraction (MIR) test facility and data were taken. The data were then studied and compared to CFD calculations to help determine their suitability for validation data. One of the main findings was that the inlet data, which were measured and controlled by calibrated mass flow rotameters and were also measured using detailed stereo particle image velocimetry (PIV) showed considerable discrepancies in mass flow rate between the two methods. The other finding was that a randomly unstable recirculation zone occurs in the flow. This instability has a very significant effect on the flow field in the vicinity of the inlet jets. Because its time scale is long and because it is apparently a
Overview of CFD Validation Experiments for Circulation Control Applications at NASA
Jones, G. S.; Lin, J. C.; Allan, B. G.; Milholen, W. E.; Rumsey, C. L.; Swanson, R. C.
2008-01-01
Circulation control is a viable active flow control approach that can be used to meet the NASA Subsonic Fixed Wing project s Cruise Efficient Short Take Off and Landing goals. Currently, circulation control systems are primarily designed using empirical methods. However, large uncertainty in our ability to predict circulation control performance has led to the development of advanced CFD methods. This paper provides an overview of a systematic approach to developing CFD tools for basic and advanced circulation control applications. This four-step approach includes "Unit", "Benchmar", "Subsystem", and "Complete System" experiments. The paper emphasizes the ongoing and planned 2-D and 3-D physics orientated experiments with corresponding CFD efforts. Sample data are used to highlight the challenges involved in conducting circulation control computations and experiments.
Verification and Validation of a Multi-Scale Code, CUPID/MARS
This paper presents a coupling of MARS with the CUPID code, and application to ROCOM TEST 1.1. The multi-scale analysis method can be either an 'explicitly (weak)-coupled' or an 'implicitly (strong)-coupled'. In this paper, a CFD scale code, CUPID, has been coupled with a system scale code, MARS, implicitly by solving the pressure equations of the two codes simultaneously. This method has an advantage over the explicitly-coupled method in numerical stability and calculation time for an analysis of a transient two-phase flow where the boundary condition changes during the calculation. A simulation of forced convection flows in a straight channel shows that total mass is conserved at the interface cells of the MARS and CUPID in both single and two phase flows. A calculation set of the manometer flow oscillation shows that the CUPID/MARS multi-scale coupling is successful in two-way direction in both two and single phase flows. The validation calculation for the Rossendorf Core Mixing ROCOM test, where the pressure vessel is calculated by the CUPID and the other components such as hot and cold legs are simulated by the MARS, shows that the multi-scale approach is cheap and convenient
Validation of CFD Simulation for Ammonia Emissions from an Equeous Solution
Rong, Li; Elhadidib, Basman; Khalifa, Ezzat;
2011-01-01
as boundary condition for CFD prediction of ammonia emission. The accuracy of CFD simulation depends on many factors. In this study, the effects of appropriate geometry model, inlet turbulent parameters and three turbulence models (low-Reynolds number k–ε model, renormalization group k–ε model and...... current HLC models generally over-predict the ammonia emissions from aqueous solution in this study whereas VLE gives better agreement between simulated and measured results. A linear relation is observed between ammonia mass transfer coefficient obtained from the VLE relation and those from HLC models....