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Sample records for cfd code validation

  1. The MAX facility for CFD code validation

    International Nuclear Information System (INIS)

    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)

  2. CFD code validation utilizing the OECD/NRC BFBT benchmark

    International Nuclear Information System (INIS)

    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

  3. 3D CFD CONV code: validation and verification

    International Nuclear Information System (INIS)

    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

  4. An approach to validation of coupled CFD and system thermal-hydraulics codes

    International Nuclear Information System (INIS)

    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)

  5. Needs and opportunities for CFD-code validation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Validation of CFD commercial codes for large diameter jet impingement flow

    International Nuclear Information System (INIS)

    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)

  7. Validations of CFD Code for Density-Gradient Driven Air Ingress Stratified Flow

    International Nuclear Information System (INIS)

    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.

  8. CFD code validation against stratified air-water flow experimental data

    International Nuclear Information System (INIS)

    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)

  9. The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

    International Nuclear Information System (INIS)

    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

  10. PIV Uncertainty Methodologies for CFD Code Validation at the MIR Facility

    Energy Technology Data Exchange (ETDEWEB)

    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.

  11. A CFD validation methodology for containment code calculations of hydrogen mixing and recombination

    International Nuclear Information System (INIS)

    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)

  12. A selection of experimental test cases for the validation of CFD codes, volume 1

    Science.gov (United States)

    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.

  13. 2-D Circulation Control Airfoil Benchmark Experiments Intended for CFD Code Validation

    Science.gov (United States)

    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.

  14. MISTRA facility for containment lumped parameter and CFD codes validation. Example of the International Standard Problem ISP47

    International Nuclear Information System (INIS)

    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)

  15. Computational Fluid Dynamics (CFD) for Nuclear Reactor Safety Applications - Workshop Proceedings, CFD4NRS-3 - Experimental Validation and Application of CFD and CMFD Codes to Nuclear Reactor Safety Issues

    International Nuclear Information System (INIS)

    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

  16. Validation of CFD Codes for Parawing Geometries in Subsonic to Supersonic Flows

    Science.gov (United States)

    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.

  17. Modeling and validation of CFD code KIRAN3D for electron beam melting of zirconium

    International Nuclear Information System (INIS)

    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)

  18. Development, validation and application of NAFA 2D-CFD code

    International Nuclear Information System (INIS)

    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)

  19. A verification and validation of the new implementation of subcooled flow boiling in a CFD code

    Energy Technology Data Exchange (ETDEWEB)

    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)

  20. A verification and validation of the new implementation of subcooled flow boiling in a CFD code

    International Nuclear Information System (INIS)

    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)

  1. OECD/NEA International Benchmark exercises: Validation of CFD codes applied nuclear industry; OECD/NEA internatiion Benchmark exercices: La validacion de los codigos CFD aplicados a la industria nuclear

    Energy Technology Data Exchange (ETDEWEB)

    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)

  2. Validation of CFD code ANSYS CFX against experiments with saline slug mixing performed at the Gidropress 4-Loop WWER-1000 test facility

    International Nuclear Information System (INIS)

    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)

  3. Validation of vortex code viscous models using lidar wake measurements and CFD

    DEFF Research Database (Denmark)

    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....

  4. Validation of the CFD code fluent by post-test calculation of a density-driven ROCOM experiment

    International Nuclear Information System (INIS)

    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

  5. Experimental validation of computational fluid dynamic codes (CFD for liquid-solid risers in clean alkylation processes

    Directory of Open Access Journals (Sweden)

    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.

  6. Formulation, Implementation and Validation of a Two-Fluid model in a Fuel Cell CFD Code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Validation of a CFD code Star-CCM+ for liquid lead-bismuth eutectic thermal-hydraulics using TALL-3D experiment

    International Nuclear Information System (INIS)

    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)

  8. Validation and comparison of two-phase flow modeling capabilities of CFD, sub channel and system codes by means of post-test calculations of BFBT transient tests

    International Nuclear Information System (INIS)

    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

  9. Coupling CFD code with system code and neutron kinetic code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Coupling CFD code with system code and neutron kinetic code

    International Nuclear Information System (INIS)

    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

  11. Validation of CFD code ANSYS CFX against experiments with asymmetric saline injection performed at the Gidropress 4-Loop WWER-1000 test facility

    International Nuclear Information System (INIS)

    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)

  12. Validation of NEPTUNE-CFD two-phase flow models using experimental data

    OpenAIRE

    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 ...

  13. Containment Code Validation Matrix

    International Nuclear Information System (INIS)

    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)

  14. Improved interpretation and validation of CFD predictions

    DEFF Research Database (Denmark)

    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...

  15. Development of 2-d cfd code

    International Nuclear Information System (INIS)

    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)

  16. Perspective: Selected benchmarks from commercial CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Utilizing GPUs to Accelerate Turbomachinery CFD Codes

    Science.gov (United States)

    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.

  18. Suppression pool swell analysis using CFD code

    International Nuclear Information System (INIS)

    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)

  19. Validation of Francis Water Turbine CFD Simulations

    OpenAIRE

    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...

  20. CFD Calculations in Turbomachinery and their Validation

    Czech Academy of Sciences Publication Activity Database

    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

  1. Coupling a CFD code with neutron kinetics and pin thermal models for nuclear reactor safety analyses

    International Nuclear Information System (INIS)

    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

  2. CFD codes and the Onsager relations

    International Nuclear Information System (INIS)

    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)

  3. Teseo code validation

    International Nuclear Information System (INIS)

    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

  4. CFD code benchmark against the air/helium tests performed in the MISTRA facility

    International Nuclear Information System (INIS)

    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

  5. ARC Code TI: CFD Utility Software Library

    Data.gov (United States)

    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...

  6. Standard Problems for CFD Validation for NGNP - Status Report

    International Nuclear Information System (INIS)

    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

  7. Application of CFD Code PHOENICS for simulating CYCLONE SEPARATORS

    International Nuclear Information System (INIS)

    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

  8. Boiling flow simulation in Neptune-CFD and Fluent codes

    International Nuclear Information System (INIS)

    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)

  9. Validation process of ISIS CFD software for fire simulation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Validation of CFD for containment jet flows including condensation

    International Nuclear Information System (INIS)

    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)

  11. Verification calculations as per CFD FLOWVISION code for sodium-cooled reactor plants

    International Nuclear Information System (INIS)

    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)

  12. Verification, Validation, and Solution Quality in Computational Physics: CFD Methods Applied to Ice Sheet Physics

    Science.gov (United States)

    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.

  13. Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

    International Nuclear Information System (INIS)

    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

  15. Validation process of the ISIS CFD software for fire simulation

    International Nuclear Information System (INIS)

    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

  16. Validation of NEPTUNE-CFD 1.0.8 for adiabatic bubbly flow and boiling flow

    International Nuclear Information System (INIS)

    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

  17. Improvement of core effective thermal conductivity model of GAMMA+ code based on CFD analysis

    International Nuclear Information System (INIS)

    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

  18. Supersonic Retropropulsion CFD Validation with Ames Unitary Plan Wind Tunnel Test Data

    Science.gov (United States)

    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.

  19. Simulation of Jet Noise with OVERFLOW CFD Code and Kirchhoff Surface Integral

    Science.gov (United States)

    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.

  20. Development and validation of a new solver based on the interfacial area transport equation for the numerical simulation of sub-cooled boiling with OpenFOAM CFD code for nuclear safety applications

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Development and validation of a new solver based on the interfacial area transport equation for the numerical simulation of sub-cooled boiling with OpenFOAM CFD code for nuclear safety applications

    International Nuclear Information System (INIS)

    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.

  2. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    International Nuclear Information System (INIS)

    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

  3. Comprehensive Approach to Verification and Validation of CFD Simulations Applied to Backward Facing Step-Application of CFD Uncertainty Analysis

    Science.gov (United States)

    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.

  4. CFD validation in OECD/NEA t-junction benchmark.

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. Computational Fluid Dynamics (CFD) in Nuclear Reactor Safety (NRS) - Proceedings of the workshop on Experiments and CFD Code Application to Nuclear Reactor Safety (XCFD4NRS)

    International Nuclear Information System (INIS)

    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

  6. CFD Application in Implantable Rotary Blood Pump Design and Validation

    Institute of Scientific and Technical Information of China (English)

    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.

  7. CFD Application in Implantable Rotary Blood Pump Design and Validation

    Institute of Scientific and Technical Information of China (English)

    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.

  8. Multidimensional modelling of temperature distribution in spent fuel pools of WWER-1000 and WWER-440 using Fluent CFD code

    International Nuclear Information System (INIS)

    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)

  9. Developing a methodology for the evaluation of results uncertainties in CFD codes

    International Nuclear Information System (INIS)

    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)

  10. A CFD code comparison of wind turbine wakes

    DEFF Research Database (Denmark)

    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...

  11. Verification and Validation Studies for the LAVA CFD Solver

    Science.gov (United States)

    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.

  12. Toward a CFD-quality (CFD-grade) database addressing LWR containment phenomena Codes to Nuclear Reactor Safety Issues

    International Nuclear Information System (INIS)

    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

  13. Two Phase Flow Models and Numerical Methods of the Commercial CFD Codes

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Two Phase Flow Models and Numerical Methods of the Commercial CFD Codes

    International Nuclear Information System (INIS)

    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

  15. Containment Thermal-Hydraulic Simulations With an LP-CFD Approach: Qualification Matrix of the Tonus Code

    International Nuclear Information System (INIS)

    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)

  16. Simulation and analysis of void drift using sub-channel analysis code and CFD code

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  17. The extensive international use of commercial computational fluid dynamics (CFD) codes

    International Nuclear Information System (INIS)

    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)

  18. Preliminary tests of a damaged ship for CFD validation

    Science.gov (United States)

    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

  19. Development and validation of the 3-D CFD model for CANDU-6 moderator temperature predictions

    International Nuclear Information System (INIS)

    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

  20. Extension of CFD Codes Application to Two-Phase Flow Safety Problems - Phase 3

    International Nuclear Information System (INIS)

    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

  1. CFD code fluent turbulence models application. Ansaldo's prototype modeling

    International Nuclear Information System (INIS)

    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)

  2. Developing a methodology for the evaluation of results uncertainties in CFD codes; Desarrollo de una Metodologia para la Evaluacion de Incertidumbres en los Resultados de Codigos de CFD

    Energy Technology Data Exchange (ETDEWEB)

    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)

  3. Coupled CFD - system-code simulation of a gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    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)

  4. Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations

    International Nuclear Information System (INIS)

    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)

  5. CFD simulation analysis and validation for CPR1000 pressurized water reactor

    International Nuclear Information System (INIS)

    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)

  6. The use of CFD code for numerical simulation study on the air/water countercurrent flow limitation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    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)

  7. The use of CFD code for numerical simulation study on the air/water countercurrent flow limitation in nuclear reactors

    International Nuclear Information System (INIS)

    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)

  8. Proceedings of the exploratory meeting of experts to define an action plan on the application of computational fluid dynamics (CFD) codes to nuclear reactor safety problems

    International Nuclear Information System (INIS)

    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

  9. Preliminary study of coupling CFD code FLUENT and system code RELAP5

    International Nuclear Information System (INIS)

    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

  10. Validation of CFD Simulations of Cerebral Aneurysms With Implication of Geometric Variations

    OpenAIRE

    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...

  11. Extending the capabilities of CFD codes to assess ash related problems

    DEFF Research Database (Denmark)

    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...

  12. Application of CFD code for simulation of an inclined snow chute flow

    Directory of Open Access Journals (Sweden)

    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.

  13. A validation process for CFD use in building physics – Study of the different length scales

    OpenAIRE

    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...

  14. A proposed framework for computational fluid dynamics code calibration/validation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. A proposed framework for computational fluid dynamics code calibration/validation

    International Nuclear Information System (INIS)

    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

  16. Simulation of natural circulation in a rectangular loop using CFD code PHOENICS

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  17. A CFD Validation of Fire Dynamics Simulator for ‎Corner Fire ‎

    Directory of Open Access Journals (Sweden)

    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. ‎

  18. CFD validation for flyash particle classification in hydrocyclones

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Development and validation of a CFD model predicting the backfill process of a nuclear waste gallery

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. Development and validation of a CFD model predicting the backfill process of a nuclear waste gallery

    International Nuclear Information System (INIS)

    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

  1. On application of CFD codes to problems of nuclear reactor safety

    International Nuclear Information System (INIS)

    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)

  2. Mitigation of turbidity currents in reservoirs with passive retention systems: validation of CFD modeling

    Science.gov (United States)

    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.

  3. Validation of Neptune-CFD Module with Data of a Plunging Water Jet Entering a Free Surface

    Energy Technology Data Exchange (ETDEWEB)

    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)

  4. Optimization of an industrial heat exchanger using an open-source CFD code

    International Nuclear Information System (INIS)

    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

  5. Development of a coupling methodology for the system code ATHLET and the advanced 3D CFD tool ANSYS CFX

    International Nuclear Information System (INIS)

    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.)

  6. Development of a coupling methodology for the system code ATHLET and the advanced 3D CFD tool ANSYS CFX

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  7. Validation of Boundary Conditions for CFD Simulations on Ventilated Rooms

    DEFF Research Database (Denmark)

    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 ...

  8. Application of CFD Codes in Nuclear Reactor Safety Analysis

    Directory of Open Access Journals (Sweden)

    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.

  9. Infrared imaging - A validation technique for computational fluid dynamics codes used in STOVL applications

    Science.gov (United States)

    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.

  10. Use of computational fluid dynamics (CFD) codes for safety analysis of nuclear reactor systems, including containment

    International Nuclear Information System (INIS)

    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

  11. Unstructured mesh based multi-physics interface for CFD code coupling in the Serpent 2 Monte Carlo code

    International Nuclear Information System (INIS)

    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)

  12. Implementation of CFD module in the KORSAR thermal-hydraulic system code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Acceleration of a CFD Code with a GPU

    Directory of Open Access Journals (Sweden)

    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.

  14. Hypersonic Intake Starting Characteristics–A CFD Validation Study

    OpenAIRE

    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 ...

  15. Validation of a FLUENT CFD model for hydrogen distribution in a containment

    International Nuclear Information System (INIS)

    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

  16. Summary of best guidelines and validation of CFD modeling in livestock buildings to ensure prediction quality

    DEFF Research Database (Denmark)

    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...

  17. Validation of 3-D CFD Model of Tritium Transport in the Atmosphere

    International Nuclear Information System (INIS)

    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

  18. A proposed methodology for computational fluid dynamics code verification, calibration, and validation

    Science.gov (United States)

    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.

  19. Simulation of two-phase flows in vertical tubes with the CFD code FLUBOX

    International Nuclear Information System (INIS)

    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)

  20. Validation of the GPU-Accelerated CFD Solver ELBE for Free Surface Flow Problems in Civil and Environmental Engineering

    Directory of Open Access Journals (Sweden)

    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.

  1. Development of CFD Code for Subcooled Boiling Two-Phase Flow with Modeling of the Interfacial Area Transport Equation

    International Nuclear Information System (INIS)

    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

  2. Development of two-phase flow CFD code (EAGLE) with interfacial area transport equation for analysis of subcooled boiling flow

    International Nuclear Information System (INIS)

    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)

  3. Hypersonic Intake Starting Characteristics–A CFD Validation Study

    Directory of Open Access Journals (Sweden)

    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

  4. Application of CFD code for simulation of an inclined snow chute flow

    OpenAIRE

    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 ...

  5. CFD modelling and validation of wall condensation in the presence of non-condensable gases

    Energy Technology Data Exchange (ETDEWEB)

    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)

  6. CFD modelling and validation of wall condensation in the presence of non-condensable gases

    International Nuclear Information System (INIS)

    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)

  7. Rocket-Based Combined Cycle Engine Technology Development: Inlet CFD Validation and Application

    Science.gov (United States)

    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.

  8. Pollutant Emission Validation of a Heavy-Duty Gas Turbine Burner by CFD Modeling

    Directory of Open Access Journals (Sweden)

    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.

  9. Assessment of CFD Codes for Nuclear Reactor Safety Problems - Revision 2

    International Nuclear Information System (INIS)

    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

  10. Two-Phase Flow Simulations for PTS Investigation by Means of Neptune_CFD Code

    OpenAIRE

    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)—a plunging water jet entering a free surface, an...

  11. Simulation of a Jet Pump with the code of CFD STAR-CCM +.; Simulacion de una Jet Pump con el codigo de CFD STAR-CCM+.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Verification of the CFD code FLUENT by post test calculation of ROCOM experiments

    International Nuclear Information System (INIS)

    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

  13. Production Level CFD Code Acceleration for Hybrid Many-Core Architectures

    Science.gov (United States)

    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.

  14. Integration of CFD codes and advanced combustion models for quantitative burnout determination

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. RELIABLE VALIDATION BASED ON OPTICAL FLOW VISUALIZATION FOR CFD SIMULATIONS

    Institute of Scientific and Technical Information of China (English)

    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.

  16. MHD for fusion: parameters bridge between CFD tools and system codes; MHD para fusion: parametros puente entre herramientas CFD y codigos de sistema

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Validation of Hydrodynamic Load Models Using CFD for the OC4-DeepCwind Semisubmersible: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Prototype coupling of the CFD code ANSYS CFX with the 3D neutron kinetic core model DYN3D

    International Nuclear Information System (INIS)

    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.)

  19. Development of a CFD code TFC2D for numerical analysis of turbulent flow

    International Nuclear Information System (INIS)

    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

  20. Development, Verification and Validation of Parallel, Scalable Volume of Fluid CFD Program for Propulsion Applications

    Science.gov (United States)

    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.

  1. Development of a system code with CFD capability for analyzing turbulent mixed convection in gas-cooled reactors

    International Nuclear Information System (INIS)

    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

  2. Infrared imaging :a proposed validation technique for computational fluid dynamics codes used in STOVL applications

    OpenAIRE

    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 5.6μm wavelength band as well as conventional pitot tube and thermocouple methods. These infrared i...

  3. ESCADRE Code Development and Validation -AN OVERVIEW-

    International Nuclear Information System (INIS)

    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

  4. 45 CFR 162.1011 - Valid code sets.

    Science.gov (United States)

    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....

  5. Base Flow Model Validation Project

    Data.gov (United States)

    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...

  6. Validation of the reactor dynamics code TRAB

    International Nuclear Information System (INIS)

    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

  7. Validation of High-Resolution CFD Method for Slosh Damping Extraction of Baffled Tanks

    Science.gov (United States)

    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.

  8. A Supersonic Argon/Air Coaxial Jet Experiment for Computational Fluid Dynamics Code Validation

    Science.gov (United States)

    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.

  9. Validation of CFD-models for natural convection, heat transfer and turbulence phenomena

    International Nuclear Information System (INIS)

    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

  10. CFD Validation Experiment of a Mach 2.5 Axisymmetric Shock-Wave/Boundary-Layer Interaction

    Science.gov (United States)

    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.

  11. A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIO{sub U}. Preliminary validation on the Phénix Reactor Natural Circulation Test

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. A first system/CFD coupled simulation of a complete nuclear reactor transient using CATHARE2 and TRIOU. Preliminary validation on the Phénix Reactor Natural Circulation Test

    International Nuclear Information System (INIS)

    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

  13. Numerical modelling of pressure suppression pools with CFD and FEM codes

    Energy Technology Data Exchange (ETDEWEB)

    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)

  14. Observations on CFD Verification and Validation from the AIAA Drag Prediction Workshops

    Science.gov (United States)

    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.

  15. Validation of TNXY code with reference problems

    International Nuclear Information System (INIS)

    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)

  16. Development of a Prototype Lattice Boltzmann Code for CFD of Fusion Systems.

    Energy Technology Data Exchange (ETDEWEB)

    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

  17. Validation Report for ISAAC Computer Code

    International Nuclear Information System (INIS)

    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

  18. Benchmark calculations of a radiation heat transfer for a CANDU fuel channel analysis using the CFD code

    International Nuclear Information System (INIS)

    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)

  19. Application of CFD techniques toward the validation of nonlinear aerodynamic models

    Science.gov (United States)

    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.

  20. TOPFLOW-experiments, development and validation of CFD models for steam-water flows with phase transfer. Final report

    International Nuclear Information System (INIS)

    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

  1. Extension of the simulation capabilities of the 1D system code ATHLET by coupling with the 3D CFD software package ANSYS CFX

    International Nuclear Information System (INIS)

    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)

  2. Evaporation over sump surface in containment studies: code validation on TOSQAN tests

    International Nuclear Information System (INIS)

    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)

  3. Experimental verification of CFD and thermal hydraulics codes by quantitative flow visualisation

    International Nuclear Information System (INIS)

    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)

  4. Investigation of natural circulation two-phase flow behaviour in header manifold using CFD code

    International Nuclear Information System (INIS)

    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)

  5. Validation of High-Resolution CFD Method for Slosh Damping Extraction of Baffled Cryogenic Propellant Tanks

    Science.gov (United States)

    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.

  6. Qualification of the CFD code TRIO-U for full scale nuclear reactor applications

    International Nuclear Information System (INIS)

    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)

  7. ALEGRA -- code validation: Experiments and simulations

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal-hydraulic analyzes of gas cooled reactors

    International Nuclear Information System (INIS)

    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

  9. Recommendation for maximum allowable mesh size for plant combustion analyses with CFD codes

    International Nuclear Information System (INIS)

    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

  10. Recommendation for maximum allowable mesh size for plant combustion analyses with CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    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

  11. Conjugate heat transfer study of a wire spacer SFR fuel assembly thanks to the thermal code SYRTHES and the CFD code Code-Saturne

    International Nuclear Information System (INIS)

    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)

  12. Development of an infrared gaseous radiation band model based on NASA SP-3080 for computational fluid dynamic code validation applications

    OpenAIRE

    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....

  13. Validation of the FLUENT CFD Computer Program by Thermal Testing of a Full Scale Double-Walled Prototype Canister for Storing Chernobyl Fuel

    International Nuclear Information System (INIS)

    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)

  14. Validation and Analysis of Forward Osmosis CFD Model in Complex 3D Geometries

    Directory of Open Access Journals (Sweden)

    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.

  15. Examples of unsteady CFD validation system response quantities in a cylinder array

    International Nuclear Information System (INIS)

    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.

  16. Validation of High-Fidelity CFD Simulations for Rocket Injector Design

    Science.gov (United States)

    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

  17. Development of a Common Research Model for Applied CFD Validation Studies

    Science.gov (United States)

    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.

  18. Validation and analysis of forward osmosis CFD model in complex 3D geometries

    DEFF Research Database (Denmark)

    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....

  19. CFD Analyses for Water-Air Flow With the Euler-Euler Two-Phase Model in the Fluent4 CFD Code

    International Nuclear Information System (INIS)

    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)

  20. Further code validation of Technical Specification Bases

    International Nuclear Information System (INIS)

    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

  1. Nuclear data to support computer code validation

    International Nuclear Information System (INIS)

    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

  2. Validations of Coupled CSD/CFD and Particle Vortex Transport Method for Rotorcraft Applications: Hover, Transition, and High Speed Flights

    Science.gov (United States)

    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.

  3. Experimental and numerical approach to validate pressure loss predictability of a commercial code

    International Nuclear Information System (INIS)

    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)

  4. On the role of code comparisons in verification and validation.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Numerical modeling of immiscible two-phase flow in micro-models using a commercial CFD code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Evaluation of heat transfer coefficient of supercritical water flowing inside a tube using CFD code fluent

    International Nuclear Information System (INIS)

    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

  7. Validation of CFD-methods to predict heat transfer and temperatures during the transport and storage of casks under a cover

    Energy Technology Data Exchange (ETDEWEB)

    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

  8. Validation of CFD-methods to predict heat transfer and temperatures during the transport and storage of casks under a cover

    International Nuclear Information System (INIS)

    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

  9. Coupled CFD/CSD Analysis of an Active-Twist Rotor in a Wind Tunnel with Experimental Validation

    Science.gov (United States)

    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.

  10. Advanced CFD and radiotracer techniques - a complementary technology - for industrial multiphase applications

    International Nuclear Information System (INIS)

    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)

  11. Modern multicore and manycore architectures: Modelling, optimisation and benchmarking a multiblock CFD code

    Science.gov (United States)

    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.

  12. Study of supercritical carbon dioxide natural circulation by the use of CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    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)

  13. Study of supercritical carbon dioxide natural circulation by the use of CFD codes

    International Nuclear Information System (INIS)

    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)

  14. Radiation Coupling with the FUN3D Unstructured-Grid CFD Code

    Science.gov (United States)

    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.

  15. Application of TONUS V2006 and FLUENT 6.2.16 CFD codes to ENACCEF hydrogen combustion tests

    International Nuclear Information System (INIS)

    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.)

  16. Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

    International Nuclear Information System (INIS)

    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)

  17. Validating CFD Models of Multiphase Mixing in the Waste Treatment Plant at the Hanford Site

    International Nuclear Information System (INIS)

    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)

  18. CFD Modeling of Free-Piston Stirling Engines

    Science.gov (United States)

    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.

  19. Comparison of film condensation models in presence of non-condensable gases implemented in a CFD Code

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  20. Validation of the GPU-Accelerated CFD Solver ELBE for Free Surface Flow Problems in Civil and Environmental Engineering

    OpenAIRE

    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...

  1. Simulation of boiling flow experiments close to CHF with the NEPTUNE-CFD code

    International Nuclear Information System (INIS)

    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)

  2. Porous Media Approach of a CFD Code to Analyze a PWR Component with Tube or Rod Bundles

    International Nuclear Information System (INIS)

    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

  3. Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

    International Nuclear Information System (INIS)

    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

  4. Study of the distribution of steam plumes in the PANDA facility using CFD code

    International Nuclear Information System (INIS)

    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

  5. Software verification and validation plan for the GWSCREEN code

    International Nuclear Information System (INIS)

    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

  6. Validation of a CFD model simulating charge and discharge of a small heat storage test module based on a sodium acetate water mixture

    DEFF Research Database (Denmark)

    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....

  7. Relating system-to-CFD coupled code analyses to theoretical framework of a multi-scale method

    Energy Technology Data Exchange (ETDEWEB)

    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)

  8. CFD simulation of a burner for syngas characterization and experimental validation

    Energy Technology Data Exchange (ETDEWEB)

    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)

  9. Two-phase CFD PTS validation in an extended range of thermohydraulics conditions covered by the COSI experiment

    International Nuclear Information System (INIS)

    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

  10. Validation and verification plan for safety and PRA codes

    International Nuclear Information System (INIS)

    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

  11. A validated CFD model to predict O₂ and CO₂ transfer within hollow fiber membrane oxygenators.

    Science.gov (United States)

    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

  12. 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

    Institute of Scientific and Technical Information of China (English)

    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。

  13. Thermal-hydraulic analysis of water-water heat exchanger under low flow conditions using CFD code

    International Nuclear Information System (INIS)

    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)

  14. Study of the distribution of steam plumes in the PANDA facility using CFD code

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. CFD Simulation and Experimental Validation of Fluid Flow and Particle Transport in a Model of Alveolated Airways.

    Science.gov (United States)

    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

  16. Computer codes validation for conditions of core voiding

    International Nuclear Information System (INIS)

    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)

  17. Study of Convective Heat Transfer at Vertical Cylinder Arranged in Nuclear Reactor or Heat Exchanger Using CFD Code

    International Nuclear Information System (INIS)

    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)

  18. Qualification of the CFD code TrioU for full scale reactor applications

    International Nuclear Information System (INIS)

    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

  19. A fuel performance code TRUST VIc and its validation

    International Nuclear Information System (INIS)

    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

  20. Using the RELAP5-3D advanced systems analysis code with commercial and advanced CFD software

    International Nuclear Information System (INIS)

    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)

  1. Coupling the RELAP5-3d advanced systems analysis code with commercial and advanced CFD software

    International Nuclear Information System (INIS)

    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)

  2. Validation of impinging jet models to be used in CANDU calandria vessel CFD simulations

    International Nuclear Information System (INIS)

    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)

  3. European validation of the integral code ASTEC (EVITA)

    International Nuclear Information System (INIS)

    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

  4. Validations of CFD against detailed velocity and pressure measurements in water turbine runner flow

    Science.gov (United States)

    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.

  5. Mach number validation of a new zonal CFD method (ZAP2D) for airfoil simulations

    Science.gov (United States)

    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.

  6. CFD Validation Experiment of a Mach 2.5 Axisymmetric Shock-Wave Boundary-Layer Interaction

    Science.gov (United States)

    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.

  7. Methodology for experimental validation of a CFD model for predicting noise generation in centrifugal compressors

    International Nuclear Information System (INIS)

    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

  8. Review of Available Data for Validation of Nuresim Two-Phase CFD Software Applied to CHF Investigations

    Directory of Open Access Journals (Sweden)

    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.

  9. CFD analysis of core melt spreading on the reactor cavity floor using ANSYS CFX code

    International Nuclear Information System (INIS)

    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.

  10. European Validation of the Integral Code ASTEC (EVITA)

    International Nuclear Information System (INIS)

    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

  11. Simulation codes and the impact of validation/uncertainty requirements

    International Nuclear Information System (INIS)

    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)

  12. Computational study of flow and heat transfer for water under supercritical conditions in a vertical pipe using NAFA CFD code

    International Nuclear Information System (INIS)

    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)

  13. Experimental investigation and CFD validation of countercurrent flow limitation (CCFL) in a large-diameter hot-leg PWR geometry

    International Nuclear Information System (INIS)

    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

  14. Validation uncertainty of MATRA code for subchannel void distributions

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  15. Validation uncertainty of MATRA code for subchannel void distributions

    International Nuclear Information System (INIS)

    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

  16. Computer code validation by high temperature chemistry

    International Nuclear Information System (INIS)

    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

  17. Turbulence modeling needs of commercial CFD codes: Complex flows in the aerospace and automotive industries

    Science.gov (United States)

    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.

  18. Performance of the OVERFLOW-MLP and LAURA-MLP CFD Codes on the NASA Ames 512 CPU Origin System

    Science.gov (United States)

    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

  19. An Improved FFR Design with a Ventilation Fan: CFD Simulation and Validation.

    Science.gov (United States)

    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

  20. An Improved FFR Design with a Ventilation Fan: CFD Simulation and Validation

    Science.gov (United States)

    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

  1. Simulation of the heat transfer of a irradiated fuel storage container with code CFD STAR- CCM+; Simulacion de la transferencia de calor de un contenedor de almacenamiento de combustible irradiado con el codigo CFD STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. A benchmark exercise on the use of CFD codes for containment issues using best practice guidelines: A computational challenge

    International Nuclear Information System (INIS)

    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

  3. A benchmark exercise on the use of CFD codes for containment issues using best practice guidelines: a computational challenge

    International Nuclear Information System (INIS)

    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)

  4. Detailed thermalhydraulic analysis of induced break severe accidents using the massively parallel CFD code TrioU/Priceles

    International Nuclear Information System (INIS)

    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)

  5. On the use of the MISTRA coupled effect test facility for the validation of containment thermal-hydraulics codes

    International Nuclear Information System (INIS)

    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

  6. One Validation Case of the CFD Software Fluent: Part of the Development Effort of a New Reactor Analysis Tool

    International Nuclear Information System (INIS)

    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)

  7. CFD SIMULATION OF PROPOSED VALIDATION DATA FOR A FLOW PROBLEM RECONFIGURED TO ELIMINATE AN UNDESIRABLE FLOW INSTABILITY

    Energy Technology Data Exchange (ETDEWEB)

    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

  8. Overview of CFD Validation Experiments for Circulation Control Applications at NASA

    Science.gov (United States)

    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.

  9. Verification and Validation of a Multi-Scale Code, CUPID/MARS

    International Nuclear Information System (INIS)

    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

  10. Validation of CFD Simulation for Ammonia Emissions from an Equeous Solution

    DEFF Research Database (Denmark)

    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....

  11. Validation of Neptune CFD two phase flow models using the OECD/NRC BFBT benchmark database

    International Nuclear Information System (INIS)

    In this work the flow within a fuel assembly of a boiling water reactor was modeled using NEPTUNE-CFD. The most important parameters to define the flow like the incipient boiling condition, the heat flux partitioning and the heat transfer models are identified and tested against experimental data from BFBT bundle test. Different heat transfer models are applied for the water/steam interface. Additionally the heat conduction is solved for the insulator and cladding of the heater rods by coupling NEPTUNE-CFD with the SYRTHES package. The calculated average void fractions are in good agreement with the experimental data and the areas for future improvements are identified. (author)

  12. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  13. WSRC approach to validation of criticality safety computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Finch, D.R.; Mincey, J.F.

    1991-12-31

    Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (K{sub eff}) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope {sup 236}U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed.

  14. WSRC approach to validation of criticality safety computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Finch, D.R.; Mincey, J.F.

    1991-01-01

    Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (K{sub eff}) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope {sup 236}U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed.

  15. WSRC approach to validation of criticality safety computer codes

    International Nuclear Information System (INIS)

    Recent hardware and operating system changes at Westinghouse Savannah River Site (WSRC) have necessitated review of the validation for JOSHUA criticality safety computer codes. As part of the planning for this effort, a policy for validation of JOSHUA and other criticality safety codes has been developed. This policy will be illustrated with the steps being taken at WSRC. The objective in validating a specific computational method is to reliably correlate its calculated neutron multiplication factor (Keff) with known values over a well-defined set of neutronic conditions. Said another way, such correlations should be: (1) repeatable; (2) demonstrated with defined confidence; and (3) identify the range of neutronic conditions (area of applicability) for which the correlations are valid. The general approach to validation of computational methods at WSRC must encompass a large number of diverse types of fissile material processes in different operations. Special problems are presented in validating computational methods when very few experiments are available (such as for enriched uranium systems with principal second isotope 236U). To cover all process conditions at WSRC, a broad validation approach has been used. Broad validation is based upon calculation of many experiments to span all possible ranges of reflection, nuclide concentrations, moderation ratios, etc. Narrow validation, in comparison, relies on calculations of a few experiments very near anticipated worst-case process conditions. The methods and problems of broad validation are discussed

  16. Validation of Numerical Codes to Compute Tsunami Runup And Inundation

    Science.gov (United States)

    Velioğlu, Deniz; Cevdet Yalçıner, Ahmet; Kian, Rozita; Zaytsev, Andrey

    2015-04-01

    FLOW 3D and NAMI DANCE are two numerical codes which can be applied to analysis of flow and motion of long waves. Flow 3D simulates linear and nonlinear propagating surface waves as well as irregular waves including long waves. NAMI DANCE uses finite difference computational method to solve nonlinear shallow water equations (NSWE) in long wave problems, specifically tsunamis. Both codes can be applied to tsunami simulations and visualization of long waves. Both codes are capable of solving flooding problems. However, FLOW 3D is designed mainly to solve flooding problem from land and NAMI DANCE is designed to solve flooding problem from the sea. These numerical codes are applied to some benchmark problems for validation and verification. One useful benchmark problem is the runup of solitary waves which is investigated analytically and experimentally by Synolakis (1987). Since 1970s, solitary waves have commonly been used to model tsunamis especially in experimental and numerical studies. In this respect, a benchmark problem on runup of solitary waves is a relevant choice to assess the capability and validity of the numerical codes on amplification of tsunamis. In this study both codes have been tested, compared and validated by applying to the analytical benchmark problem of solitary wave runup on a sloping beach. Comparison of the results showed that both codes are in good agreement with the analytical and experimental results and thus can be proposed to be used in inundation of long waves and tsunami hazard analysis.

  17. Study of the stratification of temperature in the cold branch of a PWR with ANSYS CFD-CFX Code

    International Nuclear Information System (INIS)

    This article presents the results of the complete analysis of the OECD ROSA project using the ANSYS CFX Test 1-1. The objective of this study is the validation of the code for the simulation of cases in which the thrust forces play an important role together with the use of different models of turbulence that has implemented the code ANSYS CFX.

  18. Results from the First Validation Phase of CAP code

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yeon Joon; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul [FNC Tech., SNU, Seoul (Korea, Republic of); Ha, Sang Jun; Choi, Hoon [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The second stage of Safety Analysis Code Development for Nuclear Power Plants was lunched on Apirl, 2010 and is scheduled to be through 2012, of which the scope of work shall cover from code validation to licensing preparation. As a part of this project, CAP(Containment Analysis Package) will follow the same procedures. CAP's validation works are organized hieratically into four validation steps using; 1) Fundamental phenomena. 2) Principal phenomena (mixing and transport) and components in containment. 3) Demonstration test by small, middle, large facilities and International Standard Problems. 4) Comparison with other containment codes such as GOTHIC or COMTEMPT. In addition, collecting the experimental data related to containment phenomena and then constructing the database is one of the major works during the second stage as a part of this project. From the validation process of fundamental phenomenon, it could be expected that the current capability and the future improvements of CAP code will be revealed. For this purpose, simple but significant problems, which have the exact analytical solution, were selected and calculated for validation of fundamental phenomena. In this paper, some results of validation problems for the selected fundamental phenomena will be summarized and discussed briefly

  19. COCOSYS: Status of development and validation of the German containment code system

    International Nuclear Information System (INIS)

    simulation of the chemistry inside the core melt to calculate the release of gaseous components and fission products. The overall concept of the COCOSYS system has turned out to be suitable for parallel calculation of different processes and including further detailed models. First attempt for the connection with the CFD code CFX4.1 have been made. External codes like ATHLET for reactor circuit thermal hydraulics, LAVA for melt spreading, DET3D for denotative hydrogen combustion and the industrial CFD code CFX are connected with COCOSYS. COCOSYS is subject to an ongoing internal and external validation process. At present this validation process is mainly based on tests being performed in the German ThAI facility. Experiments to be performed in ThAI dealing with hydrogen combustion, recombiner behaviour and aerosol and iodine issues are currently offered to the community as an OECD project. Examples given for the successful validation are the participation in the OECD/NEA ISP-47 and the benchmark for the CCI-2 test in the frame of the OECD-MCCI project. E. g. COCOSYS has been used in licensing procedure performed for the installation of catalytic recombiners in German nuclear power plants. Variation of the boundary conditions have underlined the need of detailed nodalization of the containment and the need of comprehensive simulation of system components (like doors, ventilation systems, rupture discs), having an influence on the overall gas distribution and on local effect. In the future further improvements and model extensions like pyrolysis processes, direct containment heating sand the combined use with CFD models will be performed. (author)

  20. CFD modelling and wind tunnel validation of airflow through plant canopies using 3D canopy architecture

    International Nuclear Information System (INIS)

    The efficiency of pesticide application to agricultural fields and the resulting environmental contamination highly depend on atmospheric airflow. A computational fluid dynamics (CFD) modelling of airflow within plant canopies using 3D canopy architecture was developed to understand the effect of the canopy to airflow. The model average air velocity was validated using experimental results in a wind tunnel with two artificial model trees of 24 cm height. Mean air velocities and their root mean square (RMS) values were measured on a vertical plane upstream and downstream sides of the trees in the tunnel using 2D hotwire anemometer after imposing a uniform air velocity of 10 m s-1 at the inlet. 3D virtual canopy geometries of the artificial trees were modelled and introduced into a computational fluid domain whereby airflow through the trees was simulated using Reynolds-Averaged Navier-Stokes (RANS) equations and k-ε turbulence model. There was good agreement of the average longitudinal velocity, U between the measurements and the simulation results with relative errors less than 2% for upstream and 8% for downstream sides of the trees. The accuracy of the model prediction for turbulence kinetic energy k and turbulence intensity I was acceptable within the tree height when using a roughness length (y0 = 0.02 mm) for the surface roughness of the tree branches and by applying a source model in a porous sub-domain created around the trees. The approach was applied for full scale orchard trees in the atmospheric boundary layer (ABL) and was compared with previous approaches and works. The simulation in the ABL was made using two groups of full scale orchard trees; short (h = 3 m) with wider branching and long (h = 4 m) with narrow branching. This comparison showed good qualitative agreements on the vertical profiles of U with small local differences as expected due to the spatial disparities in tree architecture. This work was able to show airflow within and above the

  1. Validation of the Large Interface Method of NEPTUNE{sub C}FD 1.0.8 for Pressurized Thermal Shock (PTS) applications

    Energy Technology Data Exchange (ETDEWEB)

    Coste, P., E-mail: pierre.coste@cea.fr [CEA, DEN, DER/SSTH, F-38054 Grenoble (France); Lavieville, J. [Electricite de France, Chatou (France); Pouvreau, J. [CEA, DEN, DER/SSTH, F-38054 Grenoble (France); Baudry, C.; Guingo, M.; Douce, A. [Electricite de France, Chatou (France)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer The two-phase Pressurized Thermal Shock (PTS) is a key thermohydraulics issue for PWR safety. Black-Right-Pointing-Pointer The dynamic and condensation models are firstly validated separately. Black-Right-Pointing-Pointer Then the global validation is done with the COSI experiment. Black-Right-Pointing-Pointer All the calculations performed with the same set of models both in the Large Interface Method and in the k-{epsilon} approach for turbulence substantiate the application of the tool to PTS. - Abstract: NEPTUNE{sub C}FD is a code based on a 3D transient Eulerian two-fluid model. One of the main application targets is the two-phase Pressurized Thermal Shock (PTS), which is related to PWR Reactor Pressure Vessel (RPV) lifetime safety studies, when sub-cooled water from Emergency Core Cooling (ECC) system is injected into the possibly uncovered cold leg and penetrates into the RPV downcomer. Five experiments were selected for the validation, a selection reviewed by a panel of European experts. The dynamic models are validated with a co-current smooth and wavy air-water stratified flow in a rectangular channel with detailed measurements of turbulence and velocities. The condensation models are validated with a co-current smooth and wavy steam-water stratified flow in a rectangular channel with measurements of the steam flow rates. The dynamic models are validated in the situation of a jet impinging a pool free surface with two experiments dealing with a water jet impingement on a water pool free surface in air environment. Finally, all the models involved in the reactor conditions are validated with the COSI experiment. The calculations are done with the same set of Large Interface Method models and a RANS (k-{epsilon}) approach for turbulence. They substantiate the application of the tool to PTS studies.

  2. Advances in the development and validation of CFD-BWR, a two-phase computational fluid dynamics model for the simulation of flow and heat transfer in boiling water reactors

    International Nuclear Information System (INIS)

    This paper presents recent advances in the validation of an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel bundles. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. We have described the model development strategy that has been adopted by the development team for the prediction of boiling flow regimes in a BWR fuel bundle. This strategy includes the use of local flow topology maps and flow topology specific phenomenological models. The paper reviews the key boiling phenomenological models and focuses on recent results of experiment analyses for the validation of two-phase BWR phenomena models including cladding-to-coolant heat transfer and Critical Heat Flux experiments and the BWR Full-size Assembly Boiling Test (BFBT). The two-phase flow models implemented in the CFD-BWR code can be grouped into three broad categories: models describing the vapor generation at the heated cladding surface, models describing the interactions between the vapor and the liquid coolant, and models describing the heat transfer between the fuel pin and the two-phase coolant. These models have been described and will be briefly reviewed. The boiling model used in the second generation of the CFD-BWR code includes a local flow topology map which allows the cell-by-cell selection of the local flow topology. Local flow topologies can range from a bubbly flow topology where the continuous phase is liquid, to a transition flow topology, to a droplet flow topology where the continuous phase is vapor, depending primarily on the local void fraction. The models describing the cladding-to-coolant heat transfer and the interplay between these models and the local flow topology are important in Critical Heat Flux (CHF) analyses, and will

  3. Validation of cfd and simplified models with experimental data for multiphase flow in bends

    NARCIS (Netherlands)

    Nennie, E.D.; Belfroid, S.P.C.; O'Mahoney, T.S.D.

    2013-01-01

    In this paper details of the measurement results of the forces on the bends in a 4" setup are compared to two models. The first model is a simple analytical model and is used to estimate the forces. In the second model, CFD is used. In the experiments only resulting forces, including upstream and do

  4. EURISOL-DS Multi-MW Target: Experimental program associated to validation of CFD simulations of the mercury loop

    CERN Document Server

    Blumenfeld, Laure; Kadi, Yacine; Samec, Karel; Lindroos, Mats

    At the core of the Eurisol project facility, the neutron source produces spallation neutrons from a proton beam impacting dense liquid. The liquid circulates at high speed inside the source, a closed vessel with beam windows.This technical note summarises the needed of the hydraulic METEX 1 and METEX 2 data tests to contribute to validate CFD turbulent simulation of liquid metal with the LES model and FEM structural model as well as a-dimensional analysis of Laser Dopplet Velocimetry for cavitation measurements.

  5. Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Escalante, Javier Jimenez; Espinoza, Victor Sanchez

    2015-07-15

    Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes.

  6. Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests

    International Nuclear Information System (INIS)

    Highlights: • Simulation of BFBT-BWR steady-state and transient tests with ATHLET. • Validation of thermal-hydraulic models based on pressure drops and void fraction measurements. • TRACE system code is used for the comparative study. • Predictions result in a good agreement with the experiments. • Discrepancies are smaller or comparable with respect to the measurements uncertainty. - Abstract: Validation and qualification of thermal-hydraulic system codes based on separate effect tests are essential for the reliability of numerical tools when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, sub-channel and system codes. In this paper, the capabilities of the thermal-hydraulic code ATHLET are assessed based on the experimental results provided within the NUPEC BFBT benchmark related to key Boiling Water Reactors (BWR) phenomena. Void fraction and pressure drops measurements in the BFBT bundle performed under steady-state and transient conditions which are representative for e.g. turbine trip and recirculation pump trip events, are compared with the numerical results of ATHLET. The comparison of code predictions with the BFBT data has shown good agreement given the experimental uncertainty and the results are consistent with the trends obtained with similar thermal-hydraulic codes

  7. The best estimate codes applied to VVER calculations validation methodology

    International Nuclear Information System (INIS)

    The best estimate thermal hydraulic codes for PWRs and BWRs accident analysis were elaborated in 80-th years. The main goal of the best estimate codes (BEC) calculations was to obtain the real picture of the reactor facilities parameters changes during transient and accident regimes. This codes were based on 5--6 equations mathematical models taking into account separate flows of water and vapor. Validation of the best estimate thermal hydraulic codes applied VVER calculations is the main objective of the article. The concept of CSNI test matrices application to the codes validation is presented. The ways of the matrices improvement for this purpose is outlined. Taking into account deficiency of the operating integral test facilities a special attention is focused on the original separate effects experimental data obtained in Ukraine and which was not included to the CSNI matrices. To facilitate application of the obtained data to the codes validation the numerical experiment method was elaborated. The example of the method application is described

  8. Angle-of-attack validation of a new zonal CFD method for airfoil simulations

    Science.gov (United States)

    Yoo, Sungyul; Summa, J. Michael; Strash, Daniel J.

    1990-01-01

    The angle-of-attack validation of a new concept suggested by Summa (1990) for coupling potential and viscous flow methods has been investigated for two-dimensional airfoil simulations. The fully coupled potential/Navier-Stokes code, ZAP2D (Zonal Aerodynamics Program 2D), has been used to compute the flow field around an NACA 0012 airfoil for a range of angles of attack up to stall at a Mach number of 0.3 and a Reynolds number of 3 million. ZAP2D calculation for various domain sizes from 25 to 0.12 chord lengths are compared with the ARC2D large domain solution as well as with experimental data.

  9. Development and validation of a nodal code for core calculation

    International Nuclear Information System (INIS)

    The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

  10. Validation of SOCRAT-BN code on rod bundle experiments

    International Nuclear Information System (INIS)

    SOCRAT-BN code is developed for the analysis of design and beyond design basis accidents at NPPs with liquid sodium as a coolant. To simulate the behavior of the coolant in the reactor core heat transfer and friction in rod bundle geometry are required to consider. The code SOCRAT-BN uses specialized closing relations to simulate rod bundles. The article describes the validation of the code SOCRAT-BN on experiments with fuel rod imitators in the triangular geometry with a wire-wound. (author)

  11. Development and validation of thermal hydraulic code in rolling motion

    International Nuclear Information System (INIS)

    The RELAP5/MOD3.3 code was modified by adding a module calculating the effect of rolling motion and introducing new flow and heat transfer models. The thermal hydraulic code in rolling motion was developed. The experimental data were used to validate the theoretical models and calculation results. It is shown that the new flow and heat transfer models can correctly calculate the frictional resistance and heat transfer coefficients in rolling motion. The developed thermal hydraulic code can be used to simulate the thermal hydraulic system in rolling motion. (authors)

  12. Validation of multipoint kinetics model against 3D Trikin Code

    International Nuclear Information System (INIS)

    Validation of multipoint kinetics formulation for RELAP5 code has been carried out against 3D TRIKIN code. Core behavior of an asymmetric reactivity transient has been simulated through artificial tuning of lattice constants in 3D code. Individual node normalized reactivity has been conserved and power estimates from multipoint model have been compared with 3D simulation. It has been observed that localized peak power estimates from multipoint simulation are on higher side and therefore are conservative in nature. Improvements in multipoint formulation in regards to evolving coupling coefficients and involving more number of nodes can help in improving its accuracy to some extent. (author)

  13. A program to validate computer codes for container impact analysis

    International Nuclear Information System (INIS)

    The detailed analysis of containers during impacts to assess either margins to failure or the consequences of different design strategies, requires the use of sophisticated computer codes to model the interactions of the various structural components. The combination of plastic deformation, impact and sliding at interfaces and dynamic loading effects provides a severe test of both the skill of the analyst and the robustness of the computer codes. A program of experiments has been under way at Winfrith since 1987 using extensively instrumented models to provide data for the validation of such codes. Three finite element codes, DYNA3D, HONDO-II and ABAQUS, were selected as suitable tools to cover the range of conditions expected in typical impacts. The impact orientation, velocity and instrumentation locations for the experiments are specified by pre-test calculations using these codes. Post-test analyses using the actual impact orientation and velocities are carried out as necessary if significant discrepancies are found

  14. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  15. CFD Simulation of Thermal-Hydraulic Benchmark V1000CT-2 Using ANSYS CFX

    OpenAIRE

    Thomas Höhne

    2009-01-01

    Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD) codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam gene...

  16. Investigating the effect of crevice flow on internal combustion engines using a new simple crevice model implemented in a CFD code

    International Nuclear Information System (INIS)

    A theoretical investigation is conducted to examine the way the crevice regions affect the mean cylinder pressure, the in-cylinder temperature, and the velocity field of internal combustion engines running at motoring conditions. For the calculation of the wall heat flux, a wall heat transfer formulation developed by the authors is used, while for the simulation of the crevices and the blow-by a newly developed simplified simulation model is presented herein. These sub-models are incorporated into an in-house Computational Fluid Dynamics (CFD) code. The main advantage of the new crevice model is that it can be applied in cases where no detailed information of the ring-pack configuration is available, which is important as this information is rarely known or may have been altered during the engine's life. Thus, an adequate estimation of the blow-by effect on the cylinder pressure can be drawn. To validate the new model, the measured in-cylinder pressure traces of a diesel engine, located at the authors' laboratory, running under motoring conditions at four engine speeds were used as reference, together with measured velocity profiles and turbulence data of a motored spark-ignition engine. Comparing the predicted and measured cylinder pressure traces of the diesel engine for all cases examined, it is observed that by incorporating the new crevice sub-model into the in-house CFD code, significant improvements on the predictive accuracy of the model is obtained. The calculated cylinder pressure traces almost coincide with the measured ones, thus avoiding the use of any calibration constants as would have been the case with the crevice effect omitted. Concerning the radial and swirl velocity profiles and the turbulent kinetic energy measured in the spark-ignition engine, the validation process revealed that the developed crevice model has a minor influence on the aforementioned parameters. The theoretical study has been extended by investigating in the same spark

  17. In-core fuel management code package validation for PWRs

    International Nuclear Information System (INIS)

    In the framework of its reactor physics activities conducted within its nuclear power programme, the IAEA has long provided its Member States with a forum for the exchange of technical information on in-core fuel management. This TECDOC discusses in-core fuel management code package validation for PWRs. 43 refs, figs and tabs

  18. The Mistra experiment for field containment code validation first results

    International Nuclear Information System (INIS)

    The MISTRA facility is a large scale experiment, designed for the purpose of thermal-hydraulics multi-D codes validation. A short description of the facility, the set up of the instrumentation and the test program are presented. Then, the first experimental results, studying helium injection in the containment and their calculations are detailed. (author)

  19. Development, use, and validation of the CFD tool FLACS for hydrogen safety studies

    OpenAIRE

    Middha, Prankul

    2010-01-01

    Computational Fluid Dynamics (CFD) calculations for gas explosion safety have been widely used for doing risk assessments within the oil and gas industry for more than a decade. Based on predicted consequences of a range of potential accident scenarios a risk level is predicted. The development of applications using hydrogen as a clean energy carrier has accelerated in recent years, and hydrogen may be used widely in future. Due to the very high reactivity of hydrogen, safe han...

  20. Validation of CFD models for microscale nanoprecipitation reactor using μ-PIV and confocal μ-LIF

    Science.gov (United States)

    Shi, Yanxiang; Olsen, Michael G.; Fox, Rodney O.

    2011-11-01

    Over the past a few decades, computational fluid dynamics (CFD) models have become more and more important in the process of reactor design in chemical engineering. Compared to experimental methods, they can provide comprehensive information on the flow field as well as other fields, such as concentration. However, they also need to be validated against experimental data to ensure the accuracy. In this work, the micro-scale particle image velocimetry (μ-PIV) is employed in conjunction with the confocal-base micro-scale laser induced fluorescence (μ-LIF) to specifically validate CFD models for use in microscale nanoprecipitation reactor. The former is for the velocity field measurement and the latter gives us the mixture fraction information. Both RANS and LES are used to simulate the field flow. For RANS, a DQMOM-IEM micromixing model is used to predict the mixture fraction field while only a scalar transport equation is solved in the LES simulations. Comparisons between simulation results and experimental data show that RANS might not be the right tool for such reactors. LES, on the other hand, gives reasonably satisfactory predictions.

  1. Auxiliary ventilation in mining roadways driven with roadheaders: Validated CFD modelling of dust behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Torano, J.; Torno, S.; Menendez, M.; Gent, M. [University of Oviedo, Asturias (Spain)

    2010-01-15

    The production of dust when driving mining roadways can affect workers health. In addition, there is a decrease in productivity since Mine Safety regulations establish a reduction in the working time depending on the quartz content and dust concentration in the atmosphere. One of the gate roadways of the longwall named E4-S, belonging to the underground coal mine Carbonar SA located in Northern Spain, is being driven by an AM50 roadheader machine. The mined coal has a high coal dust content. This paper presents a study of dust behaviour in two auxiliary ventilation systems by Computational Fluid Dynamics (CFD) models, taking into account the influence of time. The accuracy of these CFD models was assessed by airflow velocity and respirable dust concentration measurements taken in six points of six roadway cross-sections of the mentioned operating coal mine. It is concluded that these models predicted the airflow and dust behaviour at the working face, where the dust source is located, and in different roadways cross-sections behind the working face. As a result, CFD models allow optimization of the auxiliary ventilation system used, avoiding the important deficiencies when it is calculated by conventional methods.

  2. Lawrence Livermore National Laboratory Probabilistic Seismic Hazard Codes Validation

    Energy Technology Data Exchange (ETDEWEB)

    Savy, J B

    2003-02-08

    Probabilistic Seismic Hazard Analysis (PSHA) is a methodology that estimates the likelihood that various levels of earthquake-caused ground motion will be exceeded at a given location in a given future time-period. LLNL has been developing the methodology and codes in support of the Nuclear Regulatory Commission (NRC) needs for reviews of site licensing of nuclear power plants, since 1978. A number of existing computer codes have been validated and still can lead to ranges of hazard estimates in some cases. Until now, the seismic hazard community had not agreed on any specific method for evaluation of these codes. The Earthquake Engineering Research Institute (EERI) and the Pacific Engineering Earthquake Research (PEER) center organized an exercise in testing of existing codes with the aim of developing a series of standard tests that future developers could use to evaluate and calibrate their own codes. Seven code developers participated in the exercise, on a voluntary basis. Lawrence Livermore National laboratory participated with some support from the NRC. The final product of the study will include a series of criteria for judging of the validity of the results provided by a computer code. This EERI/PEER project was first planned to be completed by June of 2003. As the group neared completion of the tests, the managing team decided that new tests were necessary. As a result, the present report documents only the work performed to this point. It demonstrates that the computer codes developed by LLNL perform all calculations correctly and as intended. Differences exist between the results of the codes tested, that are attributed to a series of assumptions, on the parameters and models, that the developers had to make. The managing team is planning a new series of tests to help in reaching a consensus on these assumptions.

  3. High-resolution two-phase flow measurement techniques for the generation of experimental data for CFD code qualification

    International Nuclear Information System (INIS)

    Computational fluid dynamics simulations for two-phase flows are important in different fields of engineering and science. Since two-phase flows are inherently complex, also CFD modeling development requires special attention. The validation of model implementation and the derivation of physics based models for momentum, heat, and mass transfer in two-phase flow require experiments with generation of high-resolution measurement data. This, however, is a great challenge, since most standard flow measurement tools used in single phase flow situations, are not suited for multiphase flows. In this article we report on advanced imaging and measuring methods for two-phase flow experiments, which have been extensively used in the recent past to conduct experiments for two-phase flows at the Helmholtz-Zentrum Dresden-Rossendorf. In particular the application of wire-mesh sensors, ultrafast X-ray tomography, gamma ray tomography and positron emission tomography will be introduced and discussed. (orig.)

  4. Validation of IRBURN calculation code system through burnup benchmark analysis

    International Nuclear Information System (INIS)

    Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.

  5. A Comprehensive Validation Approach Using The RAVEN Code

    Energy Technology Data Exchange (ETDEWEB)

    Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J; Rinaldi, Ivan; Giannetti, Fabio; Caruso, Gianfranco

    2015-06-01

    The RAVEN computer code , developed at the Idaho National Laboratory, is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. RAVEN is a multi-purpose probabilistic and uncertainty quantification platform, capable to communicate with any system code. A natural extension of the RAVEN capabilities is the imple- mentation of an integrated validation methodology, involving several different metrics, that represent an evolution of the methods currently used in the field. The state-of-art vali- dation approaches use neither exploration of the input space through sampling strategies, nor a comprehensive variety of metrics needed to interpret the code responses, with respect experimental data. The RAVEN code allows to address both these lacks. In the following sections, the employed methodology, and its application to the newer developed thermal-hydraulic code RELAP-7, is reported.The validation approach has been applied on an integral effect experiment, representing natu- ral circulation, based on the activities performed by EG&G Idaho. Four different experiment configurations have been considered and nodalized.

  6. VALIDATION OF THE JRC TSUNAMI PROPAGATION AND INUNDATION CODES

    Directory of Open Access Journals (Sweden)

    N. Zamora

    2014-07-01

    Full Text Available In the last years several numerical codes have been developed to analyse tsunami waves. Most of these codes use a finite difference numerical approach giving good results for tsunami wave propagation, but with limitations in modelling inundation processes. The HyFlux2 model has been developed to simulate inundation scenario due to dam break, flash flood and tsunami-wave run-up. The model solves the conservative form of the two-dimensional shallow water equations using a finite volume method. The implementation of a shoreline-tracking method provides reliable results. HyFlux2 robustness has been tested using several tsunami events. The main aim of this study is code validation by means of comparing different code results with available measurements. Another objective of the study is to evaluate how the different fault models could generate different results that should be considered for coastal planning. Several simulations have been performed to compare HyFlux2 code with SWAN-JRC code and the TUNAMI-N2. HyFlux2 has been validated taking advantage of the extensive seismic, geodetic measurements and post-tsunami field surveys performed after the Nias March 28th tsunami. Although more detailed shallow bathymetry is needed to assess the inundation, diverse results in the wave heights have been revealed when comparing the different fault mechanism. Many challenges still exist for tsunami researchers especially when concern to early warning systems as shown in this Nias March 28th tsunami.

  7. Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

    International Nuclear Information System (INIS)

    Highlights: ► In this study, helium supplying system was constructed. ► Preparation for heat transfer experiment in KO TBM condition using helium supplying system was progressed. ► To get more applicable results, test matrix was made to cover the condition for KO TBM. ► Using CFD code; CFX 11, validation and modification for system code GAMMA was performed. -- Abstract: By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m2 using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11

  8. The Fast Scattering Code (FSC): Validation Studies and Program Guidelines

    Science.gov (United States)

    Tinetti, Ana F.; Dunn, Mark H.

    2011-01-01

    The Fast Scattering Code (FSC) is a frequency domain noise prediction program developed at the NASA Langley Research Center (LaRC) to simulate the acoustic field produced by the interaction of known, time harmonic incident sound with bodies of arbitrary shape and surface impedance immersed in a potential flow. The code uses the equivalent source method (ESM) to solve an exterior 3-D Helmholtz boundary value problem (BVP) by expanding the scattered acoustic pressure field into a series of point sources distributed on a fictitious surface placed inside the actual scatterer. This work provides additional code validation studies and illustrates the range of code parameters that produce accurate results with minimal computational costs. Systematic noise prediction studies are presented in which monopole generated incident sound is scattered by simple geometric shapes - spheres (acoustically hard and soft surfaces), oblate spheroids, flat disk, and flat plates with various edge topologies. Comparisons between FSC simulations and analytical results and experimental data are presented.

  9. Validation of post-dryout phenomena for the space code

    International Nuclear Information System (INIS)

    SPACE code which is based on a multi-dimensional two-fluid, three-field model is under development for the licensing calculation of pressurized water reactors. Unlike other major best-estimate nuclear reactor system analysis codes that have been developed based on a two-fluid six equation model, the field equations of SPACE code incorporates a dispersed liquid field in addition to vapor and continuous liquid fields. This model features a set of nine equations of mass, energy and momentum conservation. A dispersed liquid field is expected to be important in annular-mist and post-dryout conditions since a dispersed liquid field behaves differently with a continuous liquid field. This is the major reason to incorporate a dispersed liquid field as an additional liquid field. As a part of the validation effort of SPACE code, FLECHT-SEASET reflood problems have been assessed and are presented in this paper. (author)

  10. Development and validation of I-activation analysis code

    International Nuclear Information System (INIS)

    I-Activation Analysis Code (IAAC) is a nuclear depletion code which solves coupled Bateman equations for radioactive-transmutation and growth-decay system for large numbers of isotopes to get time evolution of decay products and nuclear activity. It is currently being developed primarily for neutron activation and radiation waste analysis, as a part of the code development activities. The code functions by separating long and short-lived isotopes and then uses the well-known matrix exponential method to quickly solve a large system of coupled, linear, first-order ordinary differential equations with constant coefficients for long-lived isotopes. This method allows a faster treatment of complex decay and transmutation schemes. The short-lived isotopes are solved using approximated decay-chain method. FENDL 3.0 neutron activation files are used for data library. Separate set of code modules are designed to read, decode, convert and condense the continuous-energy ACE formatted data into 175 VITAMIN-J energy groups. The new compiled library that includes half-lives and neutron absorption cross sections is then used as input source for nuclear data. The code is readily suitable for calculations pertaining to nuclear transmutation, activation and decay studies in mainly fusion applications and activation analyses. Details of the code and its primary validation performed for various test cases and material compositions, largely related to current ITER project specific neutronic and radiation analyses will be presented. The nuclear activity calculations are validated against FISPACT, available under EASY code system. (author)

  11. CFD analysis of poison injection in AHWR calandria

    International Nuclear Information System (INIS)

    The present work intends to give details of design and performance validation of SDS-2. The performance is evaluated on the basis of dispersion of poison in calandria in a given period of time. Location of injection tube and injection holes, size of jet hole and number of holes are some of the design parameters which greatly affect dispersion of poison in calandria. A Computational Fluid Dynamic (CFD) study for axial and radial injection of poison was carried out using open source CFD code OpenFOAM. CFD benchmarking was done using experiments performed by Johari (Johari et al. 1997) to identify suitable turbulence model for this problem. An experimental facility simulating poison injection in moderator in presence of calandria tubes was used to further validate the CFD model is shown in the paper. CFD analysis was carried out for axial as well as radial injection for AHWR geometry. CFD analysis using OpenFOAM has been carried out to study high pressure poison injection for single jet of Shut Down System - 2 (SDS- 2) of Advanced Heavy Water Reactor (AHWR) for various design options. CFD model used in analysis have been validated with experimental data available in literature as well as experiments performed for AHWR specific geometry. Various turbulence models are tested and their adequacy for such flow problems has been established. The CFD model is then used to simulate poison injection for two design options for AHWR and their performance is compared. (author)

  12. System verification and validation report for the TMAD code

    International Nuclear Information System (INIS)

    This document serves as the Verification and Validation Report for the TMAD code system, which includes the TMAD code and the LIBMAKR Code. The TMAD code was commissioned to facilitate the interpretation of moisture probe measurements in the Hanford Site waste tanks. In principle, the code is an interpolation routine that acts over a library of benchmark data based on two independent variables, typically anomaly size and moisture content. Two additional variables, anomaly type and detector type, can also be considered independent variables, but no interpolation is done over them. The dependent variable is detector response. The intent is to provide the code with measured detector responses from two or more detectors. The code will then interrogate (and interpolate upon) the benchmark data library and find the anomaly-type/anomaly-size/moisture-content combination that provides the closest match to the measured data. The primary purpose of this document is to provide the results of the system testing and the conclusions based thereon. The results of the testing process are documented in the body of the report. Appendix A gives the test plan, including test procedures, used in conducting the tests. Appendix B lists the input data required to conduct the tests, and Appendices C and 0 list the numerical results of the tests

  13. Validation of OPERA3D PCMI Analysis Code

    International Nuclear Information System (INIS)

    This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel

  14. Development of the KIVA-2 CFD code for rocket propulsion applications

    Science.gov (United States)

    Shannon, Robert V., Jr.; Murray, Alvin L.

    1992-07-01

    The KIVA-2 code, originally developed to solve computational fluid dynamics problems in internal combustion engines, has been developed to solve rocket propulsion type flows. The objective of the work was to develop a code such that both liquid and solid particle motion could be simulated for arbitrary geometry and high speed as well as low speed reacting flows. Modification to the original code include: incorporating independently specific supersonic and subsonic inflows and outflows; symmetric as well as periodic boundary conditions; and the capability to use generalized single or multi-specie thermodynamic data and transport coefficients allowing the user to specify arbitrary wall temperature/heat flux distributions. This code has been shown to successfully solve rocket propulsion flows as well as flows with entrained particles for several different rocket nozzles.

  15. Experimental validation of CASTEM code for buckling problems

    International Nuclear Information System (INIS)

    For validating the buckling analysis capability of CASTEM code which is used for the buckling design of Prototype Fast Breeder Reactor (PFBR) vessels, a few experiments have been carried out at Indira Gandhi Centre for Atomic Research (IGCAR)in Kalpakkam. Experiments were conducted on aluminium cylindrical shells under axial compression and stainless steel cylindrical shells under external pressure and transverse shear loading. This paper presents the results of experimental and associated theoretical buckling studies performed using the code INCA. (author). 3 refs., 9 figs., 3 tabs

  16. Input model of a VVER 440/213 fuel assembly and CFD calculations by the FLUENT code

    International Nuclear Information System (INIS)

    The preparation of the final version of the computation network of a VVER 440/213 fuel assembly by the GAMBIT code is described. Qualified estimates of the size of the networks (numbers of cells) are made especially in dependence on the network density in the transverse section. Some sensitivity analyses performed on smaller geometries, devoted to the effect of computation network density on turbulence modelling and to the effect of the limiting layer thickness on the temperature and flow rate fields are described. Of importance was the analysis of spacer grid replacement by means of the porous media function with regard to savings in the number of computational cells. The CFD calculations in the FLUENT code, performed first for smaller test problems and subsequently for large problems describing 1/6 to 1/2 of the fuel assembly, are described. Analyzed are the coolant distribution within the assembly after passing the bottom supporting plate and the effect of the spacer grids on coolant distribution and on heat transfer

  17. Simulation of atmosphere mixing and stratification in the ThAI experimental facility with a CFD code

    International Nuclear Information System (INIS)

    The CFD code CFX4.4 was used to simulate an experiment in the ThAI facility, which was designed for investigation of thermal-hydraulic processes during a severe accident inside a Light Water Reactor containment. In the considered experiment, air was initially present in the vessel, and helium and steam were injected during different phases of the experiment at various mass flow rates and at different locations. In the performed work, the 1st and 2nd phase of the considered experiment were simulated. The main purpose was to reproduce the non-homogeneous temperature and species concentration distributions in the ThAI experimental facility. A three-dimensional model of the ThAI vessel for the CFX4.4 code was developed. The flow in the simulation domain was modelled as single-phase. Steam condensation on vessel walls was modelled as a sink of mass and energy using a correlation that was originally developed for an integral approach. A simple model of bulk phase change was also included. The calculated time dependent data together with temperature, concentrations and velocity distributions at the end of each phase are compared to experimental results. (author)

  18. Investigation of heat transfer and distribution in the core of Ghana Research Reactor-1 (GHARR-1) using STAR-CCM+ CFD Code

    International Nuclear Information System (INIS)

    In the present work, STAR-CCM+ CFD code was used to investigate steady state thermal hydraulic parameters in the core of Ghana Research Reactor-1 (GHARR-1). The core was segmented into 21 axial segments. 3D-CAD parametric solid modeler embedded in STAR-CCM+ was used to model the geometry. The geometry was discretized by the use of appropriate meshing models. GHARR-1 operating conditions were set as boundary conditions for the STAR-CCM+ simulation conducted. Heat flux specific to individual axial segment computed based on segment power peaking factors and surface area was applied at the wall of the flow channel. For each power level, mass flow rate and temperature were imposed as boundary conditions at the inlet. Standard k-ε turbulence model was adopted for the solution of the transported variables namely turbulent kinetic energy and its dissipation rate. The results obtained were validated with experimental data from GHARR-1 operation and observed to be in appreciable agreement. The plots of the evaluated flow parameters show that the heat applied at the surface of the flow channel is efficiently transferred to the bulk of the fluid. In addition, effective distribution of temperature in the domain was observed. With effective heat transfer coupled with uniform heat distribution, it could be stated that cooling of GHARR-1 fuel which is needed for safety operation of the facility is assured. (au)

  19. Input model of a VVER 440/213 fuel assembly for the CFD computational code FLUENT

    International Nuclear Information System (INIS)

    The preparation of the input data and computation network for FLUENT 6.1 CFD (computational fluid dynamics) calculations by using the GAMBIT preprocessor is described. The input data for the thermal hydraulic calculation and the general issue of network creation - nodalization by using GAMBIT are highlighted. Creation of the particular computation network for the given fuel assembly geometry is described in detail. Attention was paid to the approach to the complex parts of the assembly, the inlet section in particular. The flow simulation in the fuel channel was analyzed. Solutions with lower numbers of channels and various degree of complexity were developed. The effect of the various solutions on the accuracy and time of calculation was investigated. The results were used to create the computation network of the whole assembly. In view of the complexity and volume of the network, the issue was discussed of how to find a suitable approach enabling test analyses to be performed on available hardware using available software

  20. Sensitivity analysis of CFD code FLUENT-12 for supercritical water in vertical bare tubes

    International Nuclear Information System (INIS)

    The ability to use FLUENT 12 or other CFD software to accurately model supercritical water flow through various geometries in diabatic conditions is integral to research involving coal-fired power plants as well as Supercritical Water-cooled Reactors (SCWR). The cost and risk associated with constructing supercritical water test loops are far too great to use in a university setting. Previous work has shown that FLUENT 12, specifically realizable k-ε model, can reasonably predict the bulk and wall temperature distributions of externally heated vertical bare tubes for cases with relatively low heat and mass fluxes. However, sizeable errors were observed for other cases, often those which involved large heat fluxes that produce deteriorated heat transfer (DHT) regimes. The goal of this research is to gain a more complete understanding of how FLUENT 12 models supercritical water cases and where errors can be expected to occur. One control case is selected where expected changes in bulk and wall temperatures occur and they match empirical correlations' predictions, and the operating parameters are varied individually to gauge their effect on FLUENT's solution. The model used is the realizable k-ε, and the parameters altered are inlet pressure, mass flux, heat flux, and inlet temperature. (author)

  1. Tracer experimental techniques for CFD model verification and validation in sugar crystallizer

    International Nuclear Information System (INIS)

    In the framework of the CRP improvement of the experimental design for RTD tests at a pilot crystallizer was performed. A new approach for RTD studies in non-Newtonian fluids for flow patterns characterization at the pilot crystallizer was carried out. Batch mixing process was tested and the homogenization time for massecuite fluid close to seven hours proved that the crystallizers with relative low residence time that the Cuban sugar industry is trying to develop will achieve low exhaustion efficiency with significant sugar losses. The flow simulation was carried out by CFD Flotran in ANSYS 5.4 package. The possibility of RTD prediction on the basis of numerical solution of transport equations for fluid dynamics in a more simplified geometry of the crystallizer and molasses as fluid using transient analysis of temperature pulse spreading is discussed. Finally, taking in account that in complex flow structure, the most general Stimulus Response Method (SRM) based in point source response (PSR) is more suitable for CFD verification than the ambiguous, in some cases, RTD function, the design and preparation of a multipurpose relative long live 99Mo/99mTc point source device has been developed and special PC-program has been prepared for the interpolation of detector responses function. (author)

  2. Numerical simulation and validation of helicopter blade-vortex interaction using coupled CFD/CSD and three levels of aerodynamic modeling

    Science.gov (United States)

    Amiraux, Mathieu

    Rotorcraft Blade-Vortex Interaction (BVI) remains one of the most challenging flow phenomenon to simulate numerically. Over the past decade, the HART-II rotor test and its extensive experimental dataset has been a major database for validation of CFD codes. Its strong BVI signature, with high levels of intrusive noise and vibrations, makes it a difficult test for computational methods. The main challenge is to accurately capture and preserve the vortices which interact with the rotor, while predicting correct blade deformations and loading. This doctoral dissertation presents the application of a coupled CFD/CSD methodology to the problem of helicopter BVI and compares three levels of fidelity for aerodynamic modeling: a hybrid lifting-line/free-wake (wake coupling) method, with modified compressible unsteady model; a hybrid URANS/free-wake method; and a URANS-based wake capturing method, using multiple overset meshes to capture the entire flow field. To further increase numerical correlation, three helicopter fuselage models are implemented in the framework. The first is a high resolution 3D GPU panel code; the second is an immersed boundary based method, with 3D elliptic grid adaption; the last one uses a body-fitted, curvilinear fuselage mesh. The main contribution of this work is the implementation and systematic comparison of multiple numerical methods to perform BVI modeling. The trade-offs between solution accuracy and computational cost are highlighted for the different approaches. Various improvements have been made to each code to enhance physical fidelity, while advanced technologies, such as GPU computing, have been employed to increase efficiency. The resulting numerical setup covers all aspects of the simulation creating a truly multi-fidelity and multi-physics framework. Overall, the wake capturing approach showed the best BVI phasing correlation and good blade deflection predictions, with slightly under-predicted aerodynamic loading magnitudes

  3. Academic validation of multi-phase flow codes

    International Nuclear Information System (INIS)

    Transport equations solved in the multiphase codes are more complex than those in single phase codes. The origin of this difference is mainly due to the complex interactions between the different length and time scales involved in the different phenomena. It is therefore difficult to estimate the potentialities of the numerics in multiphase codes without the help of computational results. For the validation, simple situations relevant for two phase phenomena are chosen. They must be as simple as possible, mainly with single effects and related to the scope of codes devoted to safety analysis. Cases in which one physical phenomenon is uppermost are preferred. Heuristic criteria which must be fulfilled by the computational results are then defined on the basis of our experience in multiphase flow computational analysis. They are illustrated with results from benchmarks. The way each criterion is fulfilled in a given calculation and its relative importance are underlined. This approach enables to separate the numerical and physical part of the results. The results allow to exhibit a correct behavior of the numerical method of the SIMMER-IV code, developed jointly by JAEA (Japan), FzK (Germany), and CEA (France). More generally, the approach is convenient to investigate the potentialities of new developed codes. It allows to avoid unwanted effects like 'black box' ones, when unexpected results are obtained in non-academic situations.

  4. Validation of system codes for plant application on selected experiments

    Energy Technology Data Exchange (ETDEWEB)

    Koch, Marco K.; Risken, Tobias; Agethen, Kathrin; Bratfisch, Christoph [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2016-05-15

    For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.

  5. Validation of system codes for plant application on selected experiments

    International Nuclear Information System (INIS)

    For decades, the Reactor Simulation and Safety Group at Ruhr-Universitaet Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.

  6. Codes for NPP severe accident simulation: development, validation and applications

    International Nuclear Information System (INIS)

    The software tools that describe various safety aspects of NPP with VVER reactor have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN). Functionally, the codes can be divided into two groups: the calculation codes that describe separate elements of NPP equipment and/or a group of processes and integrated software systems that allow solving the tasks of the NPP safety assessment in coupled formulation. In particular, IBRAE RAN in cooperation with the nuclear industry organizations has developed the integrated software package SOCRAT designed to analyze the behavior of NPP with VVER at various stages of beyond-design-basis accidents, including the stages of reactor core degradation and long-term melt retention in a core catcher. The general information about development, validation and applications of SOCRAT code is presented and discussed in the paper. (author)

  7. Hybrid mesh finite volume CFD code for studying heat transfer in a forward-facing step

    International Nuclear Information System (INIS)

    Computational fluid dynamics (CFD) methods employ two types of grid: structured and unstructured. Developing the solver and data structures for a finite-volume solver is easier than for unstructured grids. But real-life problems are too complicated to be fitted flexibly by structured grids. Therefore, unstructured grids are widely used for solving real-life problems. However, using only one type of unstructured element consumes a lot of computational time because the number of elements cannot be controlled. Hence, a hybrid grid that contains mixed elements, such as the use of hexahedral elements along with tetrahedral and pyramidal elements, gives the user control over the number of elements in the domain, and thus only the domain that requires a finer grid is meshed finer and not the entire domain. This work aims to develop such a finite-volume hybrid grid solver capable of handling turbulence flows and conjugate heat transfer. It has been extended to solving flow involving separation and subsequent reattachment occurring due to sudden expansion or contraction. A significant effect of mixing high- and low-enthalpy fluid occurs in the reattached regions of these devices. This makes the study of the backward-facing and forward-facing step with heat transfer an important field of research. The problem of the forward-facing step with conjugate heat transfer was taken up and solved for turbulence flow using a two-equation model of k-ω. The variation in the flow profile and heat transfer behavior has been studied with the variation in Re and solid to fluid thermal conductivity ratios. The results for the variation in local Nusselt number, interface temperature and skin friction factor are presented.

  8. Technique for Calculating Solution Derivatives With Respect to Geometry Parameters in a CFD Code

    Science.gov (United States)

    Mathur, Sanjay

    2011-01-01

    A solution has been developed to the challenges of computation of derivatives with respect to geometry, which is not straightforward because these are not typically direct inputs to the computational fluid dynamics (CFD) solver. To overcome these issues, a procedure has been devised that can be used without having access to the mesh generator, while still being applicable to all types of meshes. The basic approach is inspired by the mesh motion algorithms used to deform the interior mesh nodes in a smooth manner when the surface nodes, for example, are in a fluid structure interaction problem. The general idea is to model the mesh edges and nodes as constituting a spring-mass system. Changes to boundary node locations are propagated to interior nodes by allowing them to assume their new equilibrium positions, for instance, one where the forces on each node are in balance. The main advantage of the technique is that it is independent of the volumetric mesh generator, and can be applied to structured, unstructured, single- and multi-block meshes. It essentially reduces the problem down to defining the surface mesh node derivatives with respect to the geometry parameters of interest. For analytical geometries, this is quite straightforward. In the more general case, one would need to be able to interrogate the underlying parametric CAD (computer aided design) model and to evaluate the derivatives either analytically, or by a finite difference technique. Because the technique is based on a partial differential equation (PDE), it is applicable not only to forward mode problems (where derivatives of all the output quantities are computed with respect to a single input), but it could also be extended to the adjoint problem, either by using an analytical adjoint of the PDE or a discrete analog.

  9. The role of CFD computer analyses in hydrogen safety management

    International Nuclear Information System (INIS)

    The risks of hydrogen release and combustion during a severe accident in a light water reactor have attracted considerable attention after the Fukushima accident in Japan. Reliable computer analyses are needed for the optimal design of hydrogen mitigation systems, like e.g. passive autocatalytic recombiners (PARs), and for the assessment of the associated residual risk of hydrogen combustion. Traditionally, so-called Lumped Parameter (LP) computer codes are being used for these purposes. In the last decade, significant progress has been made in the development, validation, and application of more detailed, three-dimensional Computational Fluid Dynamics (CFD) simulations for hydrogen safety analyses. The objective of the current paper is to address the following questions: - When are CFD computer analyses needed complementary to the traditional LP code analyses for hydrogen safety management? - What is the validation status of the CFD computer code for hydrogen distribution, mitigation, and combustion analyses? - Can CFD computer analyses nowadays be executed in practical and reliable way for full scale containments? The validation status and reliability of CFD code simulations will be illustrated by validation analyses performed for experiments executed in the PANDA, THAI, and ENACCEF facilities. (authors)

  10. Validation of the MORET 5 code for criticality safety applications

    International Nuclear Information System (INIS)

    The MORET-5 Monte Carlo code includes 2 calculation routes: a multi-group route based on cross-sections calculated from various cell codes such a APOLLO2, DRAGON4 or SCALE, and a continuous energy calculation route. The validation of the MORET-5 code is done through the comparison between the calculated benchmark k(eff) and the experimental benchmark k(eff). If the discrepancy between these 2 k(eff) is higher than the combined standard deviation of the benchmark uncertainty and the Monte Carlo standard deviation, a bias can be identified. The criticality experimental validation database is made up of 2255 benchmarks. Concerning the multi-group approach, the present work deals only with the APOLLO2 - MORET-5 route. The APOLLO2 cell code uses a 281 energy-group structure library based on JEFF3.1. Preliminary analyses have shown that the continuous energy route using JEFF3.1 or ENDF/B-VII.0 libraries are in good agreement with the experimental k(eff) in the majority of cases. Regarding the APOLLO2 - MORET-5 calculation route, some improvements are still needed, especially for what concerns the multi-group treatment

  11. Study of selected turbulent models for supercritical water heat transfer in vertical bare tubes using CFD code FLUENT-12

    International Nuclear Information System (INIS)

    Many of the available empirical correlations existing today cannot predict closely the effects of the heat transfer phenomena within the pseudocritical region, and some do not predict well enough heat transfer coefficients even outside of this region. In this study, the Computational Fluid Dynamics (CFD) code FLUENT-12 is used with associated software such as Gambit and NIST REFPROP to predict the Heat Transfer Coefficient and corresponding wall temperature profiles inside circular tubes cooled with SuperCritical Water (SCW), and to compare them with experimental data and various empirical correlations. In this paper, a numerical study of heat transfer to SCW flowing upwards in vertical bare tubes using the CFD-code FLUENT-12 is presented for comparison to 1-D models. A large dataset was collected within conditions similar to those of proposed SuperCritical Water-cooled Reactors (SCWRs) at the Institute for Physics and Power Engineering in Obninsk, Russia. This dataset includes 80 runs in a 4-m long, 10-mm ID vertical bare tube within a wide range of operating parameters including pressure at about 24 MPa, inlet temperatures from 320 to 350°C, mass flux ranges from 200 to 1500 kg/m2s and heat fluxes up to 1250 kW/m2. Wall and bulk-fluid temperatures measured along the 4-m heated length test section were below, at, or above the pseudocritical point. Further analysis of the individual heat-transfer regimes was conducted using an axisymmetric 2-D model of a tube with 10,000 nodes along the heated length. Wall temperatures and heat transfer coefficients were analysed for 1-m sections at a time to select the best model for each region (below, within and beyond the pseudocritical region), and to neutralize effects of the rest of the tube on that region. Two turbulent models were used in the process: k-ε and k-ω, with many variations in the sub-model parameters such as viscous heating, thermal effects, and low-Reynolds number correction. The results show a good fit

  12. CFD评估框架的进展及其最佳实践: QNET-CFD知识库%The Development of a Framework for CFD Validation and Best Practice: The QNET-CFD Knowledge Base

    Institute of Scientific and Technical Information of China (English)

    Charles HIRSCH

    2006-01-01

    QNET-CFD is a thematic network on quality and trust for the industrial applications of Computational Fluid Dynamics (CFD), developed under the European Union R&D program. The main objectives of QNET-CFD were to collect CFD and experimental data in a systematic and quality controlled way and to set the basis for a consistent Knowledge Base in support of CFD guidance and validation. The QNET-CFD activity was organized around six Thematic Areas (TAs) covering the following industry sectors: external aerodynamics; combustion & heat transfer; chemical process, thermal hydraulics and nuclear safety; civil construction & HVAC; environment; turbomachinery internal flows. The main outcome of the QNET-CFD actions is the Knowledge Base (KB) with contains in a user oriented interface, extensive experimental and CFD data for a large number of test cases subdivided into 53 Application Challenges (AC) and 43 Underlying Flow Regimes (UFR). The KB contains, in addition to state-of-the-art reviews for each of the six thematic areas, Best Practice Advice (BPA) in the use of CFD for most of AC. This is considered as a significant contribution form the QNET-CFD activities and it is expected that the level of the thrust and quality in CFD will hereby be improved.%QNET-CFD是欧盟R&D项目开发的一个主题网络,该网络讨论了计算流体力学(CFD)供工业应用应具有的品质和可信度,其主要目的是以系统和质量控制的方式收集CFD和实验数据,建立两者相容的知识库(KB),以此为基础,支持对CFD进行指导和评估.QNET-CFD围绕覆盖下述工业部门的6个主题(TA)组成:外流空气动力学; 燃烧和传热; 化学过程、热水力学和核安全; 土木建筑和HVAC; 环境; 涡轮机内流.其主要成果是建立了具有面向用户界面及丰富的实验和CFD数据的知识库,这些数据来自于大量实验数据,分为53种应用挑战(AC)和43种基本流动状态(UFR).除对上述6个主题领域中每一个的科学发

  13. Criticality Safety Code Validation with LWBR’s SB Cores

    Energy Technology Data Exchange (ETDEWEB)

    Putman, Valerie Lee

    2003-01-01

    The first set of critical experiments from the Shippingport Light Water Breeder Reactor Program included eight, simple geometry critical cores built with 233UO2-ZrO2, 235UO2-ZrO2, ThO2, and ThO2-233UO2 nuclear materials. These cores are evaluated, described, and modeled to provide benchmarks and validation information for INEEL criticality safety calculation methodology. In addition to consistency with INEEL methodology, benchmark development and nuclear data are consistent with International Criticality Safety Benchmark Evaluation Project methodology.Section 1 of this report introduces the experiments and the reason they are useful for validating some INEEL criticality safety calculations. Section 2 provides detailed experiment descriptions based on currently available experiment reports. Section 3 identifies criticality safety validation requirement sources and summarizes requirements that most affect this report. Section 4 identifies relevant hand calculation and computer code calculation methodologies used in the experiment evaluation, benchmark development, and validation calculations. Section 5 provides a detailed experiment evaluation. This section identifies resolutions for currently unavailable and discrepant information. Section 5 also reports calculated experiment uncertainty effects. Section 6 describes the developed benchmarks. Section 6 includes calculated sensitivities to various benchmark features and parameters. Section 7 summarizes validation results. Appendices describe various assumptions and their bases, list experimenter calculations results for items that were independently calculated for this validation work, report other information gathered and developed by SCIENTEC personnel while evaluating these same experiments, and list benchmark sample input and miscellaneous supplementary data.

  14. Issues in the validation of CFD modelling of semi-solid metal forming

    International Nuclear Information System (INIS)

    Modelling of die filling during semi-solid metal processing (thixoforming) places particular demands on the CFD package being used. Not only are the velocities of the metal slurry in the die very high, the viscosity is too. Furthermore, the viscosity changes with shear rate (i.e. with changes in cross sectional area of the region the slurry travels through) and with time, as the injected material is thixotropic. The CFD software therefore requires good free surface tracking, accurate implicit solutions of the flow equations (as the CPU times for explicit solutions at high viscosities are impractical) and a model that adequately describes the slurry thixotropy. Finally, reliable, experimentally determined viscosity data are required. This paper describes the experiments on tin-lead and aluminium alloy slurries using compressive tests and rotating cylinder viscometry, followed by modelling using FLOW-3D. This package is known for its ability to track free surfaces accurately. Compressive tests allow rapid changes in shear rate to be imparted to the slurry, without wall slip, while the simple geometry of the viscometer makes it possible to compare analytical and numerical solutions. It is shown that the implicit viscous solver in its original form can reproduce the general trends found in the compressive and viscometry tests. However, sharp changes in shear rate lead to overestimation of pressure gradients in the slurry, making it difficult to separate these effects from those due to thixotropic breakdown. In order to achieve this separation, it is necessary to implement a more accurate implicit solver, which is currently under development. (author)

  15. Development of steam separator performance analysis code and its validation. 3. Validation of code for carryover

    International Nuclear Information System (INIS)

    The purpose of the resent study is to develop computer program models in the separator to analyze the behavior of droplets (carryover) which are ejected from the separator on the basis of physical understanding of phenomena. Firstly, the behavior of steam, liquid film, liquid droplets, and void in the separator are calculated using basic conservation equations coupled with empirical correlations derived from the mock-up test. The behavior of droplets, droplets ejected from the separator are analyzed by the following method. Steam velocity outside the separator was calculated with the versatile 3-dimensional flow analysis code in advance, and the trajectory of each liquid droplet was calculated with the Monte Carlo method. Wetness fraction around the rated condition of the ATR-type separator was calculated with the developed computer program, and compared with the measured data outside the separator. Good agreement was obtained between calculated and measured results. (author)

  16. Integration of CFD into systems analysis codes for modeling thermal stratification during SFR transients

    International Nuclear Information System (INIS)

    The whole-plant systems analysis code SAS4A/SASSYS-1 has been coupled with a computational fluid dynamics code to assess the impact of high-fidelity simulations on safety-related performance for a sodium-cooled fast reactor (SFR). With the coupled capability, it is possible to identify critical safety-related phenomenon that cannot be resolved accurately with existing tools. In this work, the impact of coupling is demonstrated by evaluating plenum thermal stratification during a protected loss of flow transient. Stratification is shown to significantly alter core temperatures and flows predicted during natural circulation conditions. Significant temperature and flow impacts were also observed in the secondary coolant system, suggesting that resolving thermal stratification has far-reaching impacts on the whole plant. (author)

  17. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  18. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD-BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt; Steven P. Antal; Michael Z. Podowski

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  19. CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow

    International Nuclear Information System (INIS)

    number of semi-empirical data, and instead, mathematical relationships are used taking into account the various physical phenomena as well the interactions that occur among them, such as heat transfer between the fluid and the solid walls condensation of water vapor on the walls, the turbulent effects in areas of restricted passage, etc. Taking into account these advantages, this study presents a qualitative and quantitative comparison between the CFD codes OpenFOAM and Gas-Flow related to the transport phenomena of Hydrogen and other gases in the primary containment of a BWR reactor. Gas-Flow is a code of commercial license that is well validated, developed in Germany to analyze the transport of gases in nuclear reactor containments. On the other hand, OpenFOAM is an open source CFD code offering several solvers for different phenomena assessments, in this work, the reacting Foam solver is used because it has a strong similarity to the intended application of Hydrogen transport. In this thesis the results obtained using the reacting Foam solver of OpenFOAM for the calculation of transport of Hydrogen are compared with the results of the Gas-Flow code in order to assess if it is feasible to use the open source code OpenFOAM in the case of Hydrogen transport in primary containment of a BWR reactor. Some differences in the qualitative and quantitative results from both codes were found, the differences (with a maximum error rate of 4%) in the quantitative results were found are small and are considered more than acceptable for this type of analysis, moreover, these differences are mainly attributed to the transport models used, mainly because OpenFOAM uses a homogeneous mixture model and Gas-Flow a heterogeneous one. Implementing appropriate solvers in codes like OpenFOAM has the goal to develop own tools that are applicable to the transport of Hydrogen in the primary containment of a BWR reactor and thus, to gain some independence while not relying on commercial codes

  20. In-core fuel management code package validation for WWERs

    International Nuclear Information System (INIS)

    Under its in-core fuel management activities, the IAEA has set up a co-ordinated research programme (CRP) on In-core Fuel Management Code Package Validation for LWRs. The objective of this CRP was to obtain well defined cases for verifying the in-core fuel management code packages, and further to improve the in-core fuel management capabilities. This CRP was performed in three reactor types, namely PWR, BWR and WWER, because of the significant differences in core layout and core management of these reactor types. Accordingly, in the framework of the WWER part, benchmarks were developed for the 440 and 1000 MWe WWER reactors. These benchmarks are appropriate to validate WWER fuel management computer code packages. The benchmarks were outlined at a Consultancy in 1988 in Ljubljana and further specified at a Research Co-ordination Meeting/Technical Committee Meeting (RCM/TCM), held in Rez in 1991. In Chapter 2 of the report, the reactor and components descriptions are given and measured data from operating plants are obtained for specifying benchmark problems. A brief description of the methods and calculating procedures by each participants to analyse the WWER reactors is presented in Chapter 3. In Chapter 4 the results of the analyses and comparisons with the measured data are included. 51 refs, figs, tabs

  1. BIOTC: An open-source CFD code for simulating biomass fast pyrolysis

    Science.gov (United States)

    Xiong, Qingang; Aramideh, Soroush; Passalacqua, Alberto; Kong, Song-Charng

    2014-06-01

    The BIOTC code is a computer program that combines a multi-fluid model for multiphase hydrodynamics and global chemical kinetics for chemical reactions to simulate fast pyrolysis of biomass at reactor scale. The object-oriented characteristic of BIOTC makes it easy for researchers to insert their own sub-models, while the user-friendly interface provides users a friendly environment as in commercial software. A laboratory-scale bubbling fluidized bed reactor for biomass fast pyrolysis was simulated using BIOTC to demonstrate its capability.

  2. CFD and Ventilation Research

    DEFF Research Database (Denmark)

    Li, Y.; Nielsen, Peter V.

    2011-01-01

    There has been a rapid growth of scientific literature on the application of computational fluid dynamics (CFD) in the research of ventilation and indoor air science. With a 1000–10,000 times increase in computer hardware capability in the past 20 years, CFD has become an integral part of...... scientific research and engineering development of complex air distribution and ventilation systems in buildings. This review discusses the major and specific challenges of CFD in terms of turbulence modelling, numerical approximation, and boundary conditions relevant to building ventilation. We emphasize...... the growing need for CFD verification and validation, suggest on-going needs for analytical and experimental methods to support the numerical solutions, and discuss the growing capacity of CFD in opening up new research areas. We suggest that CFD has not become a replacement for experiment and...

  3. COCOSYS: Status of development and validation of the German containment code system

    International Nuclear Information System (INIS)

    For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants. COCOSYS is composed of three main modules, which are separate executable files. These modules are covering thermal hydraulics including hydrogen combustion, fission products mainly aerosols and iodine behaviour, and corium behaviour with molten corium concrete interaction. The communication between these modules is realized via PVM (parallel virtual machine). COCOSYS is subject to an ongoing internal and external validation process. At present this validation process is mainly based on tests being performed in the German ThAI facility. Experiments to be performed in ThAI dealing with hydrogen combustion, recombiner behaviour and aerosol and iodine issues are currently subject of the just started OECD-THAI project. Examples given for the successful validation are the participation in the OECD/NEA ISP-47 and the benchmark for the CCI-2 test in the frame of the OECD-MCCI project. For example COCOSYS has been used in licensing procedure performed for the installation of catalytic recombiners in German nuclear power plants. At present COCOSYS is in use for the licensing process of the new Finnish EPR plant on

  4. Validation of the ATHLET system code on integral experimental facilities

    International Nuclear Information System (INIS)

    Full text of publication follows: In the Czech Republic, there are 4 VVER440 units and 2 VVER 1000 units in operation. For safety analyses, advanced BE computer codes of the type RELAP, ATHLET and CATHARE etc. developed for the Western types of PWRs are used. Up to now, methodology of application of these codes to licensing analyses assumed conservative boundary and initial conditions which requires a number of sensitivity analyses. One of the most important part of the licensing procedure is demonstration of right function of the computer code, that is, its validation on experimental facilities for the given type of reactor, in our case, for the VVER reactors. For example, the system computer code ATHLET is systematically verified in the NRI for a number of years on available integral experimental facilities PMK (VVER440/213), ISB (VVER1000) and also on experimental facility PKL (PWR). In the presented paper, the performed verifications are summarized with description of dominant phenomena corresponding to the analysed initiating events. One case is selected for verification on an integral facility and for comparison with the same process on a real NPP. The possibility of statistical evaluation of comparison of computation with experiment is also discussed. (authors)

  5. CATHENA code validation with Wolsong 4 plant commissioning test data

    International Nuclear Information System (INIS)

    The turbine trip test at 100% full power performed during Wolsong 4 commissioning test period is simulated with CATHENA code to validate its application for plant transient analysis. The purpose of the turbine trip test is to check the correct actuation of condenser steam discharge valves and atmospheric steam discharge valves without opening the main steam safety valve and to show the plant can be maintained in the poison prevent mode. The turbine trip test at 100% full power shows that the capacity of steam discharge valves is enough to prevent the MSSV opening while maintaining the plant in the poison prevent mode. The CATHENA simulation results show very good agreement with plant test data. Therefore, it is concluded that the CATHENA modeling for various plant systems including control programs are correct and the CATHENA code is appropriate to simulate CANDU plant transients

  6. The role of CFD computer analyses in hydrogen safety management

    Energy Technology Data Exchange (ETDEWEB)

    Komen, Ed M.J.; Visser, Dirk C.; Roelofs, Ferry [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Te Lintelo, Jos G.T. [N.V. Elekticiteits-Productiemaatschappij Zuid-Nederland EPZ, Borssele (Netherlands)

    2015-11-15

    The risks of hydrogen release and combustion during a severe accident in a light water reactor have attracted considerable attention after the Fukushima accident in Japan. Reliable computer analyses are needed for the optimal design of hydrogen mitigation systems. In the last decade, significant progress has been made in the development, validation, and application of more detailed, three-dimensional Computational Fluid Dynamics (CFD) simulations for hydrogen safety analyses. The validation status and reliability of CFD code simulations will be illustrated by validation analyses performed for experiments executed in the PANDA, THAI, and ENACCEF facilities.

  7. Active Aerodynamic Load Reduction on a Rotorcraft Fuselage With Rotor Effects: A CFD Validation Effort

    Science.gov (United States)

    Allan, Brian G.; Schaeffler, Norman W.; Jenkins, Luther N.; Yao, Chung-Sheng; Wong, Oliver D.; Tanner, Philip E.

    2015-01-01

    A rotorcraft fuselage is typically designed with an emphasis on operational functionality with aerodynamic efficiency being of secondary importance. This results in a significant amount of drag during high-speed forward flight that can be a limiting factor for future high-speed rotorcraft designs. To enable higher speed flight, while maintaining a functional fuselage design (i.e., a large rear cargo ramp door), the NASA Rotary Wing Project has conducted both experimental and computational investigations to assess active flow control as an enabling technology for fuselage drag reduction. This paper will evaluate numerical simulations of a flow control system on a generic rotorcraft fuselage with a rotor in forward flight using OVERFLOW, a structured mesh Reynolds-averaged Navier-Stokes flow solver developed at NASA. The results are compared to fuselage forces, surface pressures, and PN flow field data obtained in a wind tunnel experiment conducted at the NASA Langley 14-by 22-Foot Subsonic Tunnel where significant drag and download reductions were demonstrated using flow control. This comparison showed that the Reynolds-averaged Navier-Stokes flow solver was unable to predict the fuselage forces and pressure measurements on the ramp for the baseline and flow control cases. While the CFD was able to capture the flow features, it was unable to accurately predict the performance of the flow control.

  8. Calculations to an IAHR-benchmark test using the CFD-code CFX-4

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E.

    1998-10-01

    The calculation concerns a test, which was defined as a benchmark for 3-D codes by the working group of advanced nuclear reactor types of IAHR (International Association of Hydraulic Research). The test is well documented and detailed measuring results are available. The test aims at the investigation of phenomena, which are important for heat removal at natural circulation conditions in a nuclear reactor. The task for the calculation was the modelling of the forced flow field of a single phase incompressible fluid with consideration of heat transfer and influence of gravity. These phenomena are typical also for other industrial processes. The importance of correct modelling of these phenomena also for other applications is a motivation for performing these calculations. (orig.)

  9. Shared Memory Parallelization of an Implicit ADI-type CFD Code

    Science.gov (United States)

    Hauser, Th.; Huang, P. G.

    1999-01-01

    A parallelization study designed for ADI-type algorithms is presented using the OpenMP specification for shared-memory multiprocessor programming. Details of optimizations specifically addressed to cache-based computer architectures are described and performance measurements for the single and multiprocessor implementation are summarized. The paper demonstrates that optimization of memory access on a cache-based computer architecture controls the performance of the computational algorithm. A hybrid MPI/OpenMP approach is proposed for clusters of shared memory machines to further enhance the parallel performance. The method is applied to develop a new LES/DNS code, named LESTool. A preliminary DNS calculation of a fully developed channel flow at a Reynolds number of 180, Re(sub tau) = 180, has shown good agreement with existing data.

  10. New pre-coded food record form validation

    OpenAIRE

    Víctor Manuel Rodríguez; Ana Elbusto-Cabello; Mireia Alberdi-Albeniz; Amaia De la Presa-Donado; Francisco Gómez-Pérez de Mendiola; Maria Puy Portillo-Baquedano; Itziar Churruca-Ortega

    2014-01-01

    Introduction: For some research fields, simple and accurate food intake quantification tools are needed. The aim of the present work was to design a new self-administered and pre-coded food intake record form and assess its reliability and validity when quantifying the food intake of adult population, in terms of food or food-groups portions.Material and Methods: First of all, a new food-record form was designed, which included food usually consumed and which sought to be easy-to-use, short, ...

  11. Simulator validation of calculation code in REDNET upgrade system

    International Nuclear Information System (INIS)

    The reactor data network (REDNET) is a computer-based data acquisition, display and archival system which acquires data from the National Research Universal (NRU) reactor's 'fuelled sites', and several experimental loop facilities in support of CANDU technology development (e.g., fuel, fuel behaviour, and materials research programs). The system supports the processing of data collected for subsequent display at the respective experimental facilities, and in the NRU control room. REDNET was installed in the 1980s based on the 1970s computer technology. The computer hardware is obsolete and spare parts are either extremely hard to find or are now unavailable. The Upgrade system is intended to replace the REDNET and eliminate the risk of losing the data acquisition of important experimental data needed in support of the CANDU Fuel Development Program. An important goal of the Upgrade system is to improve the accuracy in the measurement and calculation of thermal power. Calculations in REDNET are performed in FORTRAN code with some in-house macros. The same calculations are re-implemented in the Upgrade system in structured-text and function-block languages. To ensure that there is no deviation or loss of accuracy in the calculations of the Upgrade system compared to those in REDNET, software validation is performed on calculation code in the Upgrade system. The validation consists of a two-stage and three-point check (at ∼0%, 50% and ∼100% signal level) process for every data type and data point in the Upgrade system. This paper presents the purpose, the major tools and process, and the results of the validation. It is concluded, based on the validation results, that the Upgrade system achieves at least the same, and in many cases better, accuracy in all the calculations. (author)

  12. NEAMS Experimental Support for Code Validation, INL FY2009

    Energy Technology Data Exchange (ETDEWEB)

    G. Youinou; G. Palmiotti; M. Salvatore; C. Rabiti

    2009-09-01

    The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Whereas the Verification part of the process does not rely on experiment, the Validation part, on the contrary, necessitates as many relevant and precise experimental data as possible to make sure the models reproduce reality as closely as possible. Hence, this report presents a limited selection of experimental data that could be used to validate the codes devoted mainly to Fast Neutron Reactor calculations in the US. Emphasis has been put on existing data for thermal-hydraulics, fuel and reactor physics. The principles of a new “smart” experiment that could be used to improve our knowledge of neutron cross-sections are presented as well. In short, it consists in irradiating a few milligrams of actinides and analyzing the results with Accelerator Mass Spectroscopy to infer the neutron cross-sections. Finally, the wealth of experimental data relevant to Fast Neutron Reactors in the US should not be taken for granted and efforts should be put on saving these 30-40 years old data and on making sure they are validation-worthy, i.e. that the experimental conditions and uncertainties are well documented.

  13. NEAMS Experimental Support for Code Validation, INL FY2009

    International Nuclear Information System (INIS)

    The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V and V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Whereas the Verification part of the process does not rely on experiment, the Validation part, on the contrary, necessitates as many relevant and precise experimental data as possible to make sure the models reproduce reality as closely as possible. Hence, this report presents a limited selection of experimental data that could be used to validate the codes devoted mainly to Fast Neutron Reactor calculations in the US. Emphasis has been put on existing data for thermal-hydraulics, fuel and reactor physics. The principles of a new 'smart' experiment that could be used to improve our knowledge of neutron cross-sections are presented as well. In short, it consists in irradiating a few milligrams of actinides and analyzing the results with Accelerator Mass Spectroscopy to infer the neutron cross-sections. Finally, the wealth of experimental data relevant to Fast Neutron Reactors in the US should not be taken for granted and efforts should be put on saving these 30-40 years old data and on making sure they are validation-worthy, i.e. that the experimental conditions and uncertainties are well documented.

  14. Guide to Using the WIND Toolkit Validation Code

    Energy Technology Data Exchange (ETDEWEB)

    Lieberman-Cribbin, W.; Draxl, C.; Clifton, A.

    2014-12-01

    In response to the U.S. Department of Energy's goal of using 20% wind energy by 2030, the Wind Integration National Dataset (WIND) Toolkit was created to provide information on wind speed, wind direction, temperature, surface air pressure, and air density on more than 126,000 locations across the United States from 2007 to 2013. The numerical weather prediction model output, gridded at 2-km and at a 5-minute resolution, was further converted to detail the wind power production time series of existing and potential wind facility sites. For users of the dataset it is important that the information presented in the WIND Toolkit is accurate and that errors are known, as then corrective steps can be taken. Therefore, we provide validation code written in R that will be made public to provide users with tools to validate data of their own locations. Validation is based on statistical analyses of wind speed, using error metrics such as bias, root-mean-square error, centered root-mean-square error, mean absolute error, and percent error. Plots of diurnal cycles, annual cycles, wind roses, histograms of wind speed, and quantile-quantile plots are created to visualize how well observational data compares to model data. Ideally, validation will confirm beneficial locations to utilize wind energy and encourage regional wind integration studies using the WIND Toolkit.

  15. Dynamic flow analysis using an OpenFOAM based CFD tool: Validation of Turbulence Intensity in a testing site

    Directory of Open Access Journals (Sweden)

    Casella Livio

    2014-01-01

    Full Text Available The presenting paper investigates on the validation of the turbulence intensity (TI modeled by a CFD tool. Six meteorological masts, equipped with cup anemometers, have been used for the purpose. Three different turbulence closure schemes, which are the SST k-omega and the k-epsilon in two different configurations, have been tested. The flow analysis shows a qualitative agreement between measurements and models, which are capable to simulate the turning of the wind towards South when it comes from SSE. Furthermore, the simulations predict a zone of high turbulence in the northern part of the site that is confirmed by the local measurements. The scores for TI have been quantified by considering the observed directional frequencies in the validation analysis. For the testing site, the SST k-omega scheme achieves the best performance when using the TI definition which is representative of the longitudinal fluctuations of the velocity vector, against the other one, which considers the fluctuation of the horizontal vector. Lastly, the model errors have been used to correct the simulated values using two approaches; the analysis shows that, for the presented case, these correction methods do not always improve the accuracy of the simulations.

  16. Beta Testing of CFD Code for the Analysis of Combustion Systems

    Science.gov (United States)

    Yee, Emma; Wey, Thomas

    2015-01-01

    A preliminary version of OpenNCC was tested to assess its accuracy in generating steady-state temperature fields for combustion systems at atmospheric conditions using three-dimensional tetrahedral meshes. Meshes were generated from a CAD model of a single-element lean-direct injection combustor, and the latest version of OpenNCC was used to calculate combustor temperature fields. OpenNCC was shown to be capable of generating sustainable reacting flames using a tetrahedral mesh, and the subsequent results were compared to experimental results. While nonreacting flow results closely matched experimental results, a significant discrepancy was present between the code's reacting flow results and experimental results. When wide air circulation regions with high velocities were present in the model, this appeared to create inaccurately high temperature fields. Conversely, low recirculation velocities caused low temperature profiles. These observations will aid in future modification of OpenNCC reacting flow input parameters to improve the accuracy of calculated temperature fields.

  17. Testing the validity of the ray-tracing code GYOTO

    CERN Document Server

    Grould, Marion; Perrin, Guy

    2016-01-01

    In the next few years, the near-infrared interferometer GRAVITY will be able to observe the Galactic center. Astrometric data will be obtained with an anticipated accuracy of 10 $\\mu$as. To analyze these future data, we have developed a code called GYOTO to compute orbits and images. We want to assess the validity and accuracy of GYOTO in a variety of contexts, in particular for stellar astrometry in the Galactic center. Furthermore, we want to tackle and complete a study made on the astrometric displacements that are due to lensing effects of a star of the central parsec with GYOTO. We first validate GYOTO in the weak-deflection limit (WDL) by studying primary caustics and primary critical curves obtained for a Kerr black hole. We compare GYOTO results to available analytical approximations and estimate GYOTO errors using an intrinsic estimator. In the strong-deflection limit (SDL), we choose to compare null geodesics computed by GYOTO and the ray-tracing code named Geokerr. Finally, we use GYOTO to estimate...

  18. Validation of the ATHLET system code on integral experimental facilities

    International Nuclear Information System (INIS)

    The paper presents a procedure applicable for the ATHLET code validation on the integral experimental facility PSB-VVER-1000 for 11% LOCA (Loss of Coolant Accident). Validation was performed using Fast Fourier Transformation. Even if the experimental facility model was relatively simple, the obtained results were encouraging and confirmed capability of the computer code to model, with a sufficient accuracy, the LOCA type analyses. The paper also describes a statistical method used to verify range and selection of the parameters for the best estimate LOCA analysis. Dependence of the uncertainty analysis results (non-parametric method using Wilk's formula) on the number of computation runs, as well as their sensitivity to uncertainty of the individual input parameters. It was concluded: -) that the experimental results are always within the uncertainty range of the calculation output values, -) that the uncertainty range is strongly dependent on the properties and number of input parameters, -) that the raise in the number of runs increases the range of results (maximum and minimum) and decreases the upper tolerance limit (not to a significant degree) and -) that sensitivity (dependence of results on the input parameters) changes with the number of runs. The parameters verified on the integral experimental facility were applied for the uncertainty analysis of a real power plant and the obtained results were very satisfactory

  19. Validation and Performance Comparison of Numerical Codes for Tsunami Inundation

    Science.gov (United States)

    Velioglu, D.; Kian, R.; Yalciner, A. C.; Zaytsev, A.

    2015-12-01

    In inundation zones, tsunami motion turns from wave motion to flow of water. Modelling of this phenomenon is a complex problem since there are many parameters affecting the tsunami flow. In this respect, the performance of numerical codes that analyze tsunami inundation patterns becomes important. The computation of water surface elevation is not sufficient for proper analysis of tsunami behaviour in shallow water zones and on land and hence for the development of mitigation strategies. Velocity and velocity patterns are also crucial parameters and have to be computed at the highest accuracy. There are numerous numerical codes to be used for simulating tsunami inundation. In this study, FLOW 3D and NAMI DANCE codes are selected for validation and performance comparison. Flow 3D simulates linear and nonlinear propagating surface waves as well as long waves by solving three-dimensional Navier-Stokes (3D-NS) equations. FLOW 3D is used specificaly for flood problems. NAMI DANCE uses finite difference computational method to solve linear and nonlinear forms of shallow water equations (NSWE) in long wave problems, specifically tsunamis. In this study, these codes are validated and their performances are compared using two benchmark problems which are discussed in 2015 National Tsunami Hazard Mitigation Program (NTHMP) Annual meeting in Portland, USA. One of the problems is an experiment of a single long-period wave propagating up a piecewise linear slope and onto a small-scale model of the town of Seaside, Oregon. Other benchmark problem is an experiment of a single solitary wave propagating up a triangular shaped shelf with an island feature located at the offshore point of the shelf. The computed water surface elevation and velocity data are compared with the measured data. The comparisons showed that both codes are in fairly good agreement with each other and benchmark data. All results are presented with discussions and comparisons. The research leading to these

  20. Evaluation of a combustion model for the simulation of hydrogen spark-ignition engines using a CFD code

    Energy Technology Data Exchange (ETDEWEB)

    Rakopoulos, C.D.; Kosmadakis, G.M. [Internal Combustion Engines Laboratory, Thermal Engineering Department, School of Mechanical Engineering, National Technical University of Athens, 9 Heroon Polytechniou St., Zografou Campus, 15780 Athens (Greece); Pariotis, E.G. [Laboratory of Naval Propulsion Systems, Section of Naval Architecture and Marine Engineering, Department of Naval Sciences, Hellenic Naval Academy, End of Hatzikyriakou Ave., Hatzikyriakio, 18539 Piraeus (Greece)

    2010-11-15

    The present work deals with the evaluation of a combustion model that has been developed, in order to simulate the power cycle of hydrogen spark-ignition engines. The motivation for the development of such a model is to obtain a simple combustion model with few calibration constants, applicable to a wide range of engine configurations, incorporated in an in-house CFD code using the RNG k-{epsilon} turbulence model. The calculated cylinder pressure traces, gross heat release rate diagrams and exhaust nitric oxide (NO) emissions are compared with the corresponding measured ones at various engine loads. The engine used is a Cooperative Fuel Research (CFR) engine fueled with hydrogen, operating at a constant engine speed of 600 rpm. This model is composed of various sub-models used for the simulation of combustion of conventional fuels in SI engines; it has been adjusted in the current study specifically for hydrogen combustion. The basic sub-model incorporated for the calculation of the reaction rates is the characteristic conversion time-scale method, meaning that a time-scale is used depending on the laminar conversion time and the turbulent mixing time, which dictates to what extent the combustible gas has reached its chemical equilibrium during a predefined time step. Also, the laminar and turbulent combustion velocity is used to track the flame development within the combustion chamber, using two correlations for the laminar flame speed and the Zimont/Lipatnikov approach for the modeling of the turbulent flame speed, whereas the (NO) emissions are calculated according to the Zeldovich mechanism. From the evaluation conducted, it is revealed that by using the developed hydrogen combustion model and after adjustment of the unique model calibration constant, there is an adequate agreement with measured data (regarding performance and emissions) for the investigated conditions. However, there are a few more issues to be resolved dealing mainly with the ignition

  1. Further validation of the KARATE-440 code system

    International Nuclear Information System (INIS)

    In the last years several projects aiming at introduction of new VVER-440 fuel types and resulting in more economic fuel cycles were initiated. Increased average enrichment, modification of the lattice pitch and fuels diameter, profiled enrichment, application of burnable absorber, modification of the absorber assembly coupler part could lead to higher burnup and maximum allowed reactor power. The above fuel modifications and the upgraded regimes requiring more accurate calculations have necessitated the further development and validation of the KARATE code system: application of new, more accurate nuclear data, corresponding renewal of the multigroup libraries and the parametrized few group constants. For the further validation the operational data of PAKS (Hungary) NPP and zero reactor measurements were used. The global calculations, where burnup dependent node-wise power distributions, critical boron concentrations, reactivity coefficients of the core etc. are determined, have been validated against operational data of PAKS NPP. Measured critical boron concentrations, reactivity coefficients, radial temperature and axial self power detector signal distributions were used for the comparison for 14 new cycles of each four units. The fine mesh diffusion calculations were benchmarked by using the measurements of ZR-ZR-6 critical facility lattices containing Gd burnable poison (Authors)

  2. Extending a serial 3D two-phase CFD code to parallel execution over MPI by using the PETSc library for domain decomposition

    CERN Document Server

    Ervik, Åsmund; Müller, Bernhard

    2014-01-01

    To leverage the last two decades' transition in High-Performance Computing (HPC) towards clusters of compute nodes bound together with fast interconnects, a modern scalable CFD code must be able to efficiently distribute work amongst several nodes using the Message Passing Interface (MPI). MPI can enable very large simulations running on very large clusters, but it is necessary that the bulk of the CFD code be written with MPI in mind, an obstacle to parallelizing an existing serial code. In this work we present the results of extending an existing two-phase 3D Navier-Stokes solver, which was completely serial, to a parallel execution model using MPI. The 3D Navier-Stokes equations for two immiscible incompressible fluids are solved by the continuum surface force method, while the location of the interface is determined by the level-set method. We employ the Portable Extensible Toolkit for Scientific Computing (PETSc) for domain decomposition (DD) in a framework where only a fraction of the code needs to be a...

  3. A CFD Validation of Fire Dynamics Simulator for ‎Corner Fire ‎

    OpenAIRE

    Pavan K Sharma; Gera‎ R.K. Singh

    2010-01-01

    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 pr...

  4. Spent reactor fuel benchmark composition data for code validation

    International Nuclear Information System (INIS)

    To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays are being obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Destructive assay data are being obtained from representative reactor fuels having experienced irradiation exposures up to about 55 GWD/MTM. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of the same fuel rod and represent radiation exposures of about 27, 37, and 44 GWD/MTM. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input

  5. Code validation and scaling of the LOBI BL-30 experiment

    International Nuclear Information System (INIS)

    Integral test facilities (ITFs) are one of the main tools for the validation of best-estimate thermalhydraulic system codes. The experimental data are also of great value when compared to the experiment scaled-conditions in a full Nuclear Power Plant (NPP). The LOBI-MOD2 was a single plus a triple loop (simulated by one loop) test facility electrically heated to simulate a 1300 MWe Pressurized Water Reactor (PWR). The scaling factor was 712 for the core power, volume and mass flow. Primary and secondary sides contained all the main active elements. It was located and operated at the European Commission Joint Research Centre (JRC) of Ispra, Italy. Experimental data of tests performed at the facility are available on-line through the JRC STRESA web platform for code validation purposes. The paper is focused in the simulation (with RELAP5 Mod-3.3) of the LOBI BL-30 experiment - a 5% small break Loss Of Coolant Accident (SB-LOCA) in the cold leg - the prediction of the general timing of events and primary system and secondary system quantities trends appear to be in good agreement with experimental data. The exercise continued with the simulation (with RELAP5 Mod 3.3) of the scaling of the LOBI BL-30 experiment to the Spanish reactor ASCO-2, a 3-loops 2940.6 MWth Westinghouse PWR. We have obtained good agreement with experimental data and have explained the reason of some discrepancies in the accumulator behaviour. To complete and compare the results, for the same NPP two other calculations have been performed: a non-scaled reference calculation of a 5% SB-LOCA in the cold leg with all the typical NPP safety systems connected and a SB-LOCA calculation with only the break area scaled and all the typical NPP safety systems connected. Both calculations have led to a safe end of the transient

  6. CFD based numerical modules for safety analysis at NPPs: validation and verification

    International Nuclear Information System (INIS)

    In the 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 sequence of condensed grids with an amount of nodes from a range 107 - 108, for which the convergence was obtained. Moreover, the attention of the paper is focused on validation and verification of software with usage of such tests as: full turbulent flow of water in a round pipe and backward-facing step (BFS) flow

  7. Optimization and Validation of the Developed Uranium Isotopic Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    γ-ray spectroscopy is a representative non-destructive assay for nuclear material, and less time-consuming and less expensive than the destructive analysis method. The destructive technique is more precise than NDA technique, however, there is some correction algorithm which can improve the performance of γ-spectroscopy. For this reason, an analysis code for uranium isotopic analysis is developed by Applied Nuclear Physics Group in Seoul National University. Overlapped γ- and x-ray peaks in the 89-101 keV X{sub α}-region are fitted with Gaussian and Lorentzian distribution peak functions, tail and background functions. In this study, optimizations for the full-energy peak efficiency calibration and fitting parameters of peak tail and background are performed, and validated with 24 hour acquisition of CRM uranium samples. The optimization of peak tail and background parameters are performed with the validation by using CRM uranium samples. The analysis performance is improved in HEU samples, but more optimization of fitting parameters is required in LEU sample analysis. In the future, the optimization research about the fitting parameters with various type of uranium samples will be performed. {sup 234}U isotopic analysis algorithms and correction algorithms (coincidence effect, self-attenuation effect) will be developed.

  8. INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    Glenn McCreery; Hugh McIlroy

    2007-09-01

    Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related

  9. INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

    International Nuclear Information System (INIS)

    Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V and V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a 'benchmark' database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related

  10. CFD Simulation and Experimental Validation of Fluid Flow in Pre-distributor

    Institute of Scientific and Technical Information of China (English)

    张吕鸿; 高国华; 隋红; 李洪; 李鑫钢

    2011-01-01

    Liquid distributor is a very import intemal for distillation columns. Pre-distributor is usually set on the top of distributor for initial distribution. Fluid flow in pre-distributor is a complex system of variable mass flow with many orifices and sub-branches. Consequently, the two phase modeling of pre-distributors was carried out andthe homogeneous model with free surface model was applied. The numerical method was validated by comparing with experimental data. Using the simulated results for different pre-distributors, the impacts of inflow rate, location and orientation uoon the outflow distribution were investigated. Furthermore, influences of the outflow distribution for pre-distributor on liquid uniformity in trough were also analyzed, The conclusions can De aaoptea for me structural design of liquid distributor and pre-distributor of large scale.

  11. Study of heat transfer in circular tubes with supercritical fluid by the STAR-CCM+ CFD code

    International Nuclear Information System (INIS)

    Supercritical Cooled-Water Reactor (SCWR) which is planned to be deployed by 2030 derives its concept from Light Water Reactors (Boiling Water Reactor (BWR), Pressurized Water Reactor (PWR) and Fossil Fired Coal Plant but with a simpler design. Due to the strong variations of density at supercritical pressure, the SCWR is likely to inherit some of the issues related to the LWR’s in terms of heat transfer (e.g. thermal crisis). This research was undertaken in order to better understand the phenomena of heat transfer as applied to SCWR and also to test the applicability of Reynolds-Average Navier-Stokes (STARCCM+ CFD code). Kim’s et al., (2005) data which employs supercritical CO2 as a simulant of water at 8 MPa was used to test the applicability. The computational simulation by STAR-CCM+ on the prediction of a 2-D axisymmetric heat transfer of carbon dioxide at supercritical pressure flowing upward through heated cross-section of a circular tube was performed with six (6) low-Reynolds number models; κ -epsilon AKN, EB, standard low- Re and V2F with two κ -ω turbulence models; SST and standard Wilcox with low y+ wall treatment. The results of heat fluxes of 20, 23, 30 and 40 kW/m2 and mass flux of 314 kg/m2s were compared to the experimental data of Kim et al, (2005). The Standard low Reynolds turbulence models were seen to have better capabilities to predict the heat transfer behaviour of supercritical CO2 as observed in the experiment. The κ -ω models did not perform favourably in the prediction of heat transfer deterioration. The V2F turbulence model performed better than the other models quantitatively when compared to the experimental data. The results of the simulation has been found to be able to reproduce the general features exhibited in the experimental data even though they over predicted the observed heat transfer deterioration both quantitatively and qualitatively. (author)

  12. Validation of NASA Thermal Ice Protection Computer Codes. Part 3; The Validation of Antice

    Science.gov (United States)

    Al-Khalil, Kamel M.; Horvath, Charles; Miller, Dean R.; Wright, William B.

    2001-01-01

    An experimental program was generated by the Icing Technology Branch at NASA Glenn Research Center to validate two ice protection simulation codes: (1) LEWICE/Thermal for transient electrothermal de-icing and anti-icing simulations, and (2) ANTICE for steady state hot gas and electrothermal anti-icing simulations. An electrothermal ice protection system was designed and constructed integral to a 36 inch chord NACA0012 airfoil. The model was fully instrumented with thermo-couples, RTD'S, and heat flux gages. Tests were conducted at several icing environmental conditions during a two week period at the NASA Glenn Icing Research Tunnel. Experimental results of running-wet and evaporative cases were compared to the ANTICE computer code predictions and are presented in this paper.

  13. FLUENT code CFD calculations of the mixing in the downcomer and lower plenum of the Dukovany reactor after SG1 steam outlet header rupture at full power

    International Nuclear Information System (INIS)

    The mixing in the downcomer and lower plenum of the Dukovany reactor after SG1 steam outlet header rupture at full power was simulated. The calculation was performed by means of the CFD FLUENT code. The computation domain includes the bent parts of all 6 cold legs, downcomer, and lower plenum. The boundary and initial conditions were taken from RELAP5-3D calculations. The study was aimed at finding how the temperature field at the core inlet develops during the initial 50 s of the transient and comparing it with the results from RELAP5-3D calculations

  14. EXPERIMENTAL VALIDATION AND CFD PREDICTION OF NATURAL CONVECTION IN A VERTICAL PIPE USING FLUENT 14.5 SOLVER

    OpenAIRE

    Farooque Azam; Dr. Jitendra Kumar; Arham Javed

    2016-01-01

    In this study, an experimental and CFD prediction of natural convection heat transfer in a vertical cylinder (pipe) is done by FLUENT 14.5 solver. Rosseland radiation model available in ANASYS CFD FLUENT 14.5 software will be investigate to find their usefulness for different parameters and cases as well as their limitation. Simulations will be performed with systematic parameterization to investigate the effect of changing models and parameters. Governing equations are solved using a finite ...

  15. CFD Simulation Of Air-Flow Over A „Quarter-Circular” Object Valided By Experimental Measurement

    OpenAIRE

    Králik Juraj; Hubová Oľga; Konečná Lenka

    2015-01-01

    A Computer-Fluid-Dynamic (CFD) simulation of air-flow around quarter-circular object using commercial software ANSYS Fluent was used to study iteration of building to air-flow. Several, well know transient turbulence models were used and results were compared to experimental measurement of this object in Boundary Layer Wind Tunnel (BLWT) of Slovak University of Technology (SUT) in Bratislava. Main focus of this article is to compare pressure values from CFD in three different elevations, whic...

  16. A new neutronics analysis code system for fast reactors and validation

    International Nuclear Information System (INIS)

    A new neutronics analysis code system has been developed for detailed analysis of fast reactor cores. The code system is composed of a calculation code of effective cross sections, an assembly calculation code based on the method of characteristics, and a full core transport/diffusion calculation code. The validity of the code system is investigated by applying it to the prototype fast reactor Monju, and by comparing the calculation results with measured ones. (author)

  17. Electrical capacitance tomography (ECT) and gamma radiation meter for comparison with and validation and tuning of computational fluid dynamics (CFD) modeling of multiphase flow

    International Nuclear Information System (INIS)

    The electrical capacitance tomographic (ECT) approach is increasingly seen as attractive for measurement and control applications in the process industries. Recently, there is increased interest in using the tomographic details from ECT for comparing with and validating and tuning CFD models of multiphase flow. Collaboration with researchers working in the field of computational fluid dynamics (CFD) modeling of multiphase flows gives valuable information for both groups of researchers in the field of ECT and CFD. By studying the ECT tomograms of multiphase flows under carefully monitored inflow conditions of the different media and by obtaining the capacitance values, C(i, j, t) with i = 1…N, j = 1, 2,…N and i ≠ j obtained from ECT modules with N electrodes, it is shown how the interface heights in a pipe with stratified flow of oil and air can be fruitfully compared to the values of those obtained from ECT and gamma radiation meter (GRM) for improving CFD modeling. Monitored inflow conditions in this study are flow rates of air, water and oil into a pipe which can be positioned at varying inclinations to the horizontal, thus emulating the pipelines laid in subsea installations. It is found that ECT-based tomograms show most of the features seen in the GRM-based visualizations with nearly one-to-one correspondence to interface heights obtained from these two methods, albeit some anomalies at the pipe wall. However, there are some interesting features the ECT manages to capture: features which the GRM or the CFD modeling apparently do not show, possibly due to parameters not defined in the inputs to the CFD model or much slower response of the GRM. Results presented in this paper indicate that a combination of ECT and GRM and preferably with other modalities with enhanced data fusion and analysis combined with CFD modeling can help to improve the modeling, measurement and control of multiphase flow in the oil and gas industries and in the process industries

  18. Single-Shot Scalar-Triplet Measurements in High-Pressure Swirl-Stabilized Flames for Combustion Code Validation

    Science.gov (United States)

    Kojima, Jun; Nguyen, Quang-Viet

    2007-01-01

    In support of NASA ARMD's code validation project, we have made significant progress by providing the first quantitative single-shot multi-scalar data from a turbulent elevated-pressure (5 atm), swirl-stabilized, lean direct injection (LDI) type research burner operating on CH4-air using a spatially-resolved pulsed-laser spontaneous Raman diagnostic technique. The Raman diagnostics apparatus and data analysis that we present here were developed over the past 6 years at Glenn Research Center. From the Raman scattering data, we produce spatially-mapped probability density functions (PDFs) of the instantaneous temperature, determined using a newly developed low-resolution effective rotational bandwidth (ERB) technique. The measured 3-scalar (triplet) correlations, between temperature, CH4, and O2 concentrations, as well as their PDF s, also provide a high-level of detail into the nature and extent of the turbulent mixing process and its impact on chemical reactions in a realistic gas turbine injector flame at elevated pressures. The multi-scalar triplet data set presented here provides a good validation case for CFD combustion codes to simulate by providing both average and statistical values for the 3 measured scalars.

  19. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  20. Fuel assembly simulations using LRGR-CFD and CGCFD

    International Nuclear Information System (INIS)

    In addition to the traditional fuel assembly simulation approaches using system codes, subchannel codes or porous medium approaches, as well as detailed CFD simulations to analyze single sub channels, a Low Resolution Geometry Resolving (LRGR) CFD approach and a Coarse-Grid-CFD (CGCFD) approach is taken. Both methods are based on a low resolution mesh that allows the capture of large and medium scale flow features such as recirculation zones, which cannot be reproduced by the system codes, subchannel codes and porous media approaches. The LRGR approach allows for instance fine-tuning the porous parameters which are important input for a porous medium approach. However, it should be noted that the prediction of detailed flow features such as secondary flows is not feasible. Using this approach, the consequences of flow blockages for detection possibilities and cladding temperatures can be discussed. Within the Coarse-Grid CFD approach a subgrid model (SGM) accounts for sub grid volumetric forces which are derived from validated CFD simulations. The volumetric forces take account of the non resolved physics due to the coarse mesh. The CGCFD approach with SGM can be applied to simulate complete fuel assemblies or even complete cores capturing the unique features of the complex flow induced by the fuel assembly geometry and its spacers. In such a case, grids with a very low grid resolution are employed. The current paper discusses and presents both, the CGCFD and the LRGR approaches. (author)

  1. The role of CFD combustion modeling in hydrogen safety management – III: Validation based on homogeneous hydrogen–air–diluent experiments

    International Nuclear Information System (INIS)

    Highlights: • A CFD based method proposed in the previous article is used for the simulation of the effect of CO2–He dilution on hydrogen deflagration. • A theoretical study is presented to verify whether CO2–He diluent can be used as a replacement for H2O as diluent. • CFD model used for the validation work is described. • TFC combustion model results are in good agreement with large-scale homogeneous hydrogen–air–CO2–He experiments. - Abstract: Large quantities of hydrogen can be generated and released into the containment during a severe accident in a PWR. The generated hydrogen, when mixed with air, can lead to hydrogen combustion. The dynamic pressure loads resulting from hydrogen combustion can be detrimental to the structural integrity of the reactor safety systems and the reactor containment. Therefore, accurate prediction of these pressure loads is an important safety issue. In our previous article, a CFD based method to determine these pressure loads was presented. This CFD method is based on the application of a turbulent flame speed closure combustion model. The method was validated against three uniform hydrogen–air deflagration experiments with different blockage ratio performed in the ENACCEF facility. It was concluded that the maximum pressures were predicted within 13% accuracy, while the rate of pressure rise dp/dt was predicted within about 30%. The eigen frequencies of the residual pressure wave phenomena were predicted within a few %. In the present article, we perform additional validation of the CFD based method against three uniform hydrogen–air–CO2–He deflagration experiments with three different concentrations of the CO2–He diluent. The trends of decrease in the flame velocity, the intermediate peak pressure, the rate of pressure rise dp/dt, and the maximum value of the mean pressure with an increase in the CO2–He dilution are captured well in the simulations. From the presented validation analyses, it can be

  2. CAST3M/ARCTURUS: a coupled heat transfer / CFD code for thermal-hydraulic analyses of gas cooled reactors

    International Nuclear Information System (INIS)

    , heat exchangers...) are highly recommended. Nevertheless, in case of a total loss of station service power, the safety demonstration of the concept should be guaranteed by natural circulation decay heat removal. This could be performed by keeping a relatively high back-up pressure for pure He natural convection and also by heavy gas injection. So, it also necessary to compute the mixing of different gases and the on-set of natural convection. In this paper, we will report on the developments of the CAST3M/ARCTURUS thermal-hydraulics code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases will be detailed, as well as application of the code to benchmark problems such as the GT-MHR or the HTTR. Examples of thermal-hydraulic calculations decay heat removal of fast reactors designs (GCFR) will also be described. (authors)

  3. Analytical validation of the CACECO containment analysis code

    International Nuclear Information System (INIS)

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code

  4. Status of the GAMMA-FR code validation - TES pipe rupture accident of HCCR TBS

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hyung Gon; Lee, Dong Won; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of); Merrill, Brad J. [Idaho National Laboratory, Atomic (United States); Ahn, Mu-Young; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    GAMMA-FR code to code validation is conducted and it shows reasonable agreement, however, near wall effect on the effective thermal conductivity needs to be investigated for better results. The GAMMA-FR code was scheduled for validation during the next two years under UCLA-NFRI collaboration. Through this research, GAMMA-FR will be validated with representative fusion experiments and reference accident cases. The GAMMA-FR (Gas Multicomponent Mixture Transient Analysis for Fusion Reactors) code is an in-house system analysis code to predict the thermal hydraulic and chemical reaction phenomena expected to occur during the thermo-fluid transients in a nuclear fusion system. A safety analysis of the Korea TBS (Test Blanket System) for ITER (International Thermonuclear Experimental Reactor) is underway using this code. This paper describes validation strategy of GAMMA-FR and current status of the validation study with respect to 'TES pipe rupture accident of ITER TBM'.

  5. Status of the GAMMA-FR code validation - TES pipe rupture accident of HCCR TBS

    International Nuclear Information System (INIS)

    GAMMA-FR code to code validation is conducted and it shows reasonable agreement, however, near wall effect on the effective thermal conductivity needs to be investigated for better results. The GAMMA-FR code was scheduled for validation during the next two years under UCLA-NFRI collaboration. Through this research, GAMMA-FR will be validated with representative fusion experiments and reference accident cases. The GAMMA-FR (Gas Multicomponent Mixture Transient Analysis for Fusion Reactors) code is an in-house system analysis code to predict the thermal hydraulic and chemical reaction phenomena expected to occur during the thermo-fluid transients in a nuclear fusion system. A safety analysis of the Korea TBS (Test Blanket System) for ITER (International Thermonuclear Experimental Reactor) is underway using this code. This paper describes validation strategy of GAMMA-FR and current status of the validation study with respect to 'TES pipe rupture accident of ITER TBM'

  6. Knowledge Transfer from Detailed 3-D CFD Codes to System Simulation Tools – CCV Modeling in SI Engine

    Directory of Open Access Journals (Sweden)

    Vítek Oldřich

    2016-06-01

    Full Text Available The paper deals with CCV knowledge transfer from reference data (either experiments or 3-D CFD data into system simulation SW tools (based on 0-D/1-D CFD. It was verified that CCV phenomenon can be modeled by means of combustion model perturbations. The proposed methodology consists of two major steps. First, individual cycle data have to be matched with the 0-D/1-D model, i.e., combustion model parameters are varied to achieve the best possible match of in-cylinder pressure traces. Second, the combustion model parameters (obtained in previous step are statistically evaluated to obtain PDFs and cross-correlations. Then such information is imposed to the 0-D/1-D tool to mimic pressure traces CCV. Good correspondence with the reference data is achieved only if both PDFs and cross-correlations are imposed simultaneously.

  7. An approach to improve the separation of solid-liquid suspensions in inclined plate settlers: CFD simulation and experimental validation.

    Science.gov (United States)

    Salem, A I; Okoth, G; Thöming, J

    2011-05-01

    The most important requirements for achieving effective separation conditions in inclined plate settler (IPS) are its hydraulic performance and the equal distribution of suspensions between settler channels, both of which depend on the inlet configuration. In this study, three different inlet structures were used to explore the effect of feeding a bench scale IPS via a nozzle distributor on its hydraulic performance and separation efficiency. Experimental and Computational Fluid Dynamic (CFD) analyses were carried out to evaluate the hydraulic characteristics of the IPS. Comparing the experimental results with the predicted results by CFD simulation implies that the CFD software can play a useful role in studying the hydraulic performance of the IPS by employing residence time distribution (RTD) curves. The results also show that the use of a nozzle distributor can significantly enhance the hydraulic performance of the IPS, which contributes to the improvement of its separation efficiency. PMID:21546049

  8. Development and validation of high-precision CFD method with volume-tracking algorithm for gas-liquid two-phase flow simulation on unstructured mesh

    International Nuclear Information System (INIS)

    In design studies of Japanese sodium-cooled fast reactors, a compact size reactor vessel is expected to be employed for economical advantages. However, such a design makes coolant velocity higher and may result in occurrence of gas entrainment (GE) phenomena. Since the GE is highly non-linear and too difficult to predict its onset condition by theoretical methods, we are developing a high-precision CFD method to evaluate the GE accurately. The CFD method is formulated on unstructured meshes to establish accurate geometric modeling of complicated reactor systems. As for two-phase flow simulations, a high-precision volume-of-fluid algorithm was employed and newly formulated on unstructured meshes. In the formulation process, a volume-conservative algorithm and a new formulation establishing the mechanical balance between pressure and surface tension were introduced. The developed CFD method was verified by solving well-known driven-cavity and Zalesak's slotted disk rotation problems to show the simulation accuracy. Then, we simulated a rising bubble in liquid under Bhaga et al.'s experimental conditions. As a result, the developed method showed good agreement with the experiment. Finally, the developed method was validated by simulating the GE phenomena in the basic experiment. The developed method succeeded in reproducing the occurrence of the GE under the experimental GE condition. (authors)

  9. Calculations of hydrogen transport for the simulation of a Sbo in the NPP-L V using the code CFD GASFLOW; Calculos de transporte de hidrogeno para la simulacion de un SBO en la CNLV usando el codigo CFD GASFLOW

    Energy Technology Data Exchange (ETDEWEB)

    Gomez T, A. M.; Xolocostli M, V. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Lopez M, R.; Filio L, C.; Mugica R, C. A. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico); Royl, P., E-mail: armando.gomez@inin.gob.mx [Karlsruhe Institute of Technology, Consultor, Hermann-von-Helmholtz-Platz, D-76344 Eggenstein -Leopoldshafen, Karlsruhe (Germany)

    2013-10-15

    The scenario of electric power total loss in the nuclear power plant of Laguna Verde (NPP-L V) has been analyzed using the code MELCOR previously, until reaching fault conditions of the primary container. A mitigation measure to avoid the loss of the primary contention is the realization of a venting toward the secondary contention (reactor building), however this measure bears the potential explosions occurrence risk when the hydrogen accumulated in the primary container with the oxygen of the reactor building atmosphere reacting. In this work a scenario has been supposed that considers the mentioned venting when the pressure of 4.5 kg/cm{sup 2} is reached in the primary container. The information for the hydrogen like an entrance fact is obtained of the MELCOR results and the hydrogen transport in both contentions is analyzed with the code CFD GASFLOW that allows predicting the detailed distribution of the hydrogen volumetric concentration and the possible detonation of flammability conditions in the reactor building. The results show that the venting will produce detonation conditions in the venting level (level 33) and flammability in the level of the recharge floor. The methodology here described constitutes the base of a detailed calculation system of this type of phenomena that can use to make safety evaluations in the NPP-L V on scenarios that include gases transport. (Author)

  10. Prediction and validation of pool fire development in enclosures by means of CFD Models for risk assessment of nuclear power plants (Poolfire) - Report year 2

    Energy Technology Data Exchange (ETDEWEB)

    van Hees, P.; Wahlqvist, J.; Kong, D. [Lund Univ., Lund (Sweden); Hostikka, S.; Sikanen, T. [VTT Technical Research Centre of Finland (Finland); Husted, B. [Haugesund Univ. College, Stord (Norway); Magnusson, T. [Ringhals AB, Vaeroebacka (Sweden); Joerud, F. [European Spallation Source (ESS), Lund (Sweden)

    2013-05-15

    Fires in nuclear power plants can be an important hazard for the overall safety of the facility. One of the typical fire sources is a pool fire. It is therefore important to have good knowledge on the fire behaviour of pool fire and be able to predict the heat release rate by prediction of the mass loss rate. This project envisages developing a pyrolysis model to be used in CFD models. In this report the activities for second year are reported, which is an overview of the experiments conducted, further development and validation of models and cases study to be selected in year 3. (Author)

  11. Prediction and validation of pool fire development in enclosures by means of CFD Models for risk assessment of nuclear power plants (Poolfire) - Report year 2

    International Nuclear Information System (INIS)

    Fires in nuclear power plants can be an important hazard for the overall safety of the facility. One of the typical fire sources is a pool fire. It is therefore important to have good knowledge on the fire behaviour of pool fire and be able to predict the heat release rate by prediction of the mass loss rate. This project envisages developing a pyrolysis model to be used in CFD models. In this report the activities for second year are reported, which is an overview of the experiments conducted, further development and validation of models and cases study to be selected in year 3. (Author)

  12. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    International Nuclear Information System (INIS)

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively

  13. CFD Verification of 5x5 Rod Bundle with Mixing Vane Spacer Grids

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sungkew; Jang, Hyungwook; Lim, Jongseon; Park, Eungjun; Nahm, Keeyil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Results of the CHF test are used for determining the CHF correlation, which is used to evaluate the thermal margin in the reactor core. Computational fluid dynamics (CFD) has been used to save the time and cost for experimental tests, components design and complicated phenomena in all industries including the reactor coolant system. L. D. Smith et al. applied the CFD methodology in a 5x5 rod bundle with the mixing vane spacer grid using the renormalization group (RNG) k-epsilon model. This CFD model agreed reasonably well with the test data. M. E. Conner et al. conducted experiments to validate the CFD methodology for the single-phase flow conditions in PWR fuel assemblies. In this validation case, the CFD code predicted very similar flow field structures as the test data. In this study, a CFD simulation under single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with some mixing vane spacer grids. In this study, a CFD simulation under a single-phase flow condition was conducted for one specific condition in a thermal mixing flow test of 5x5 rod bundle with the mixing vane spacer grids to verify the applicability of the CFD model for predicting the outlet temperature distribution. FLUENT 14.5 Version was used in this CFD analysis. For the successful prediction of the wall bounded turbulent flows, the y+ with 3 prism layers was determined within 5. At this time, k-epsilon standard turbulence model was used. The temperature distribution of CFD for each sub-channel at the outlet region of test bundle showed the difference approximately within 1.1% and 0.2% while comparing to that of test and sub-channel analysis code, respectively.

  14. CFD simulation of thermal discharge behaviour in the Kadra reservoir at the Kaiga atomic power station. Pt. 1. Validation for 2 power plant units in operation

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, P.K.; Goyal, P.; Markandeya, S.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Planning and Coordination Div.; Ghosh, A.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Health Safety and Environment Group

    2011-05-15

    The thermal pollution arising out of discharge of hot water from the power plant condensers into the natural water bodies such as rivers, lakes, reservoirs, oceans etc. has been a serious concern to environmentalists ever since the plants started operating world over. In the past forty to fifty years, the methods of calculations for predicting the velocity and temperature fields in the affected regions of the stagnant/flowing water bodies have undergone a significant improvement. Currently, use of Computational Fluid Dynamics (CFD) codes for performing these calculations is gaining popularity. However, several factors such as the assumed computational domain and its discretisation, the boundary conditions used, representation of hydrodynamic characteristics (laminar/turbulent, buoyant/non-buoyant), etc. have a strong influence on the accuracy of predictions by such a model. A CFD code STAR-CD has been used for analyzing the thermal plume behaviour in the Kadra reservoir at Kaiga Atomic Power Station (KAPS). The predictions from these calculations of two units in operation have been found to be in good agreement with the site data made available from earlier studies. The present paper briefly describes the model developed using STAR-CD and results obtained for the Kadra reservoir at KAPS. (orig.)

  15. CFD simulation of thermal discharge behaviour in the Kadra reservoir at the Kaiga atomic power station. Pt. 1. Validation for 2 power plant units in operation

    International Nuclear Information System (INIS)

    The thermal pollution arising out of discharge of hot water from the power plant condensers into the natural water bodies such as rivers, lakes, reservoirs, oceans etc. has been a serious concern to environmentalists ever since the plants started operating world over. In the past forty to fifty years, the methods of calculations for predicting the velocity and temperature fields in the affected regions of the stagnant/flowing water bodies have undergone a significant improvement. Currently, use of Computational Fluid Dynamics (CFD) codes for performing these calculations is gaining popularity. However, several factors such as the assumed computational domain and its discretisation, the boundary conditions used, representation of hydrodynamic characteristics (laminar/turbulent, buoyant/non-buoyant), etc. have a strong influence on the accuracy of predictions by such a model. A CFD code STAR-CD has been used for analyzing the thermal plume behaviour in the Kadra reservoir at Kaiga Atomic Power Station (KAPS). The predictions from these calculations of two units in operation have been found to be in good agreement with the site data made available from earlier studies. The present paper briefly describes the model developed using STAR-CD and results obtained for the Kadra reservoir at KAPS. (orig.)

  16. Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool

    International Nuclear Information System (INIS)

    Highlights: • No CFD study of LOFA without SCRAM in MTR reactor has been found in the literature. • A chart that provides safety limits during the unprotected LOFA sequences is provided. • The CFD model developed can be adapted for simulating reactivity insertion accident. - Abstract: CFD is expected to feature more frequently in reactor thermal hydraulics. The reason for the increased use of multidimensional CFD methods is not only the increased availability of capable computer systems but also the ongoing drive to improve and reduce uncertainty in our predictions of important phenomena. In this work, a CFD model coupled with the reactor point kinetics equations is developed using the CFD code, Fluent to simulate loss of flow accident (LOFA) without SCRAM in a typical material testing reactor (MTR). The CFD model is used to simulate the core behavior during transient up to the onset of nucleate boiling (ONB) point. PARET code is used not only to validate the CFD model but also to complete simulation during the sub-cooled boiling regime. The focus is on establishing a new CFD approach in the reactor safety analysis and determining the two-phase flow stability boundaries as function of initial reactor conditions. Both ONB and onset of flow instability (OFI) is predicted. Besides a useful chart is provided, which describes the stability region in terms of initial reactor power, core inlet temperature, and power peaking factor

  17. ESE a 2D compressible multiphase flow code developed for MFCI analysis - code validation

    International Nuclear Information System (INIS)

    ESE (Evaluation of Steam Explosions) is a general second order accurate two-dimensional compressible multiphase flow computer code. It has been developed to model the interaction of molten core debris with water during the first premixing stage of a steam explosion. A steam explosion is a physical event, which may occur during a severe reactor accident following core meltdown when the molten fuel comes into contact with the coolant water. Since the exchanges of mass, momentum and energy are regime dependent, different exchange laws have been incorporated in ESE for the major flow regimes. With ESE a number of premixing experiments performed at the Oxford University and at the QUEOS facility at Forschungszentrum Karlsruhe has been simulated. In these premixing experiments different jets of spheres were injected in a water poll. The ESE validation plan was carefully chosen, starting from very simple, well-defined problems, and gradually working up to more complicated ones. The results of ESE simulations, which were compared to experimental data and also to first order accurate calculations, are presented in form graphs. Most of the ESE results agree qualitatively as quantitatively reasonably well with experimental data and in general better than the results obtained with the first order accurate calculation.(author)

  18. Aerosol kinetic code "AERFORM": Model, validation and simulation results

    Science.gov (United States)

    Gainullin, K. G.; Golubev, A. I.; Petrov, A. M.; Piskunov, V. N.

    2016-06-01

    The aerosol kinetic code "AERFORM" is modified to simulate droplet and ice particle formation in mixed clouds. The splitting method is used to calculate condensation and coagulation simultaneously. The method is calibrated with analytic solutions of kinetic equations. Condensation kinetic model is based on cloud particle growth equation, mass and heat balance equations. The coagulation kinetic model includes Brownian, turbulent and precipitation effects. The real values are used for condensation and coagulation growth of water droplets and ice particles. The model and the simulation results for two full-scale cloud experiments are presented. The simulation model and code may be used autonomously or as an element of another code.

  19. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

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

    International Nuclear Information System (INIS)

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

  1. Development of an Auto-Validation Program for MARS Code Assessments

    International Nuclear Information System (INIS)

    MARS (Multi-dimensional Analysis of Reactor Safety) code is a best-estimate thermal hydraulic system analysis code developed at KAERI. It is important for a thermal hydraulic computer code to be assessed against theoretical and experimental data to verify and validate the performance and the integrity of the structure, models and correlations of the code. The code assessment efforts for complex thermal hydraulics code such as MARS code can be tedious, time-consuming and require large amount of human intervention in data transfer to see the results in graphic forms. Code developers produce many versions of a code during development and each version need to be verified for integrity. Thus, for MARS code developers, it is desirable to have an automatic way of carrying out the code assessment calculations. In the present work, an Auto-Validation program that carries out the code assessment efforts has been developed. The program uses the user supplied configuration file (with '.vv' extension) which contain commands to read input file, to execute the user selected MARS program, and to generate result graphs. The program can be useful if a same set of code assessments is repeated with different versions of the code. The program is written with the Delphi program language. The program runs under the Microsoft Windows environment

  2. In-core fuel management code package validation for BWRs

    International Nuclear Information System (INIS)

    The main goal of the present CRP (Coordinated Research Programme) was to develop benchmarks which are appropriate to check and improve the fuel management computer code packages and their procedures. Therefore, benchmark specifications were established which included a set of realistic data for running in-core fuel management codes. Secondly, the results of measurements and/or operating data were also provided to verify and compare with these parameters as calculated by the in-core fuel management codes or code packages. For the BWR it was established that the Mexican Laguna Verde 1 BWR would serve as the model for providing data on the benchmark specifications. It was decided to provide results for the first 2 cycles of Unit 1 of the Laguna Verde reactor. The analyses of the above benchmarks are performed in two stages. In the first stage, the lattice parameters are generated as a function of burnup at different voids and with and without control rod. These lattice parameters form the input for 3-dimensional diffusion theory codes for over-all reactor analysis. The lattice calculations were performed using different methods, such as, Monte Carlo, 2-D integral transport theory methods. Supercell Model and transport-diffusion model with proper correction for burnable absorber. Thus the variety of results should provide adequate information for any institute or organization to develop competence to analyze In-core fuel management codes. 15 refs, figs and tabs

  3. Development and validation of CONV-3D code for calculation of thermal and hydrodynamics of Fast Reactor at the Supercomputer

    International Nuclear Information System (INIS)

    In IBRAE 3D CFD modules (CONV code) for safety analysis of the operated Nuclear Power Plants (NPPs) are developed. 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 iterative Fast Fourier Transformation (FFT) solver for Laplace’s operator as preconditioner 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 Quasi DNS approach is used. The CONV code is fully parallelized and highly effective at the high performance computers such as “Chebyshev”, “Lomonosov” (Moscow State University). 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 software has been applied to the analysis results of test LIVE-L1 (L1 is aimed at investigating the melt pool and crust behaviour during the stages of air circulation at the outer RPV surface with subsequent flooding of the lower head) and joint analyses on transient molten pool thermal hydraulics in the LIVE facility in the framework of ISTC project. Moreover CONV was validated successfully on a series of the experimental tests as: the blind test on simulation of flows in T-junction (OECD/NEA), ERCOFTAC experiment (world database on turbulent flows) natural convection in the closures under extremely high Rayleigh numbers. In all cases the good coincidence of numerical predictions with experimental data was reached, that specifies a possibility of application of the developed approach for a prediction of CFD

  4. CFD Simulation Of Air-Flow Over A „Quarter-Circular” Object Valided By Experimental Measurement

    Directory of Open Access Journals (Sweden)

    Králik Juraj

    2015-12-01

    Full Text Available A Computer-Fluid-Dynamic (CFD simulation of air-flow around quarter-circular object using commercial software ANSYS Fluent was used to study iteration of building to air-flow. Several, well know transient turbulence models were used and results were compared to experimental measurement of this object in Boundary Layer Wind Tunnel (BLWT of Slovak University of Technology (SUT in Bratislava. Main focus of this article is to compare pressure values from CFD in three different elevations, which were obtained from experimental measurement. Polyhedral mesh type was used in the simulation. Best results on the windward face elevations were obtained using LES turbulence model, where the averaged difference was around 7.71 %. On the leeward face elevations it was SAS turbulence model and averaged differences from was 15.91 %. On the circular face it was SAS turbulence model and averaged differences from all elevations was 12.93 %.

  5. Validation of coupled codes in VALCO W P1 using measured WWER data

    International Nuclear Information System (INIS)

    The collection of measured data from transients in WWER type NPPs has been used for the validation of coupled thermal hydraulics / neutron kinetics codes in the previous PHARE project SRRI/95 and continuing now in the Work Package 1 of the VALCO project with new types of transients and new data. Firstly, a collection of five transients was made, then two transients; 'Drop of control rod at nominal power at Bohunice-3' for WWER-440 reactors, were used in code validation. Eight institutes participated with ten calculations for the code validation with five different combinations of coupled codes. Used thermal hydraulic codes were ATHLET, SMABRE and RELAP5 and the neutron kinetic codes DYN3D, HEXTRAN, KIKO3D and BIPR8 (Authors)

  6. Code-to-Code Validation and Application of a Building Dynamic Simulation Tool for the Building Energy Performance Analysis

    OpenAIRE

    Annamaria Buonomano

    2016-01-01

    In this paper details about the results of a code-to-code validation procedure of an in-house developed building simulation model, called DETECt, are reported. The tool was developed for research purposes in order to carry out dynamic building energy performance and parametric analyses by taking into account new building envelope integrated technologies, novel construction materials and innovative energy saving strategies. The reliability and accuracy of DETECt was appropriately tested by mea...

  7. Application of CFD to Safety and Thermal-Hydraulic Analysis of Lead-Cooled Systems

    OpenAIRE

    Jeltsov, Marti

    2011-01-01

    Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety analysis as a tool that enables safety related physical phenomena occurring in the reactor coolant system to be described in more detail and accuracy. Validation is a necessary step in improving predictive capability of a computationa code or coupled computational codes. Validation refers to the assessment of model accuracy incorporating any uncertainties (aleatory and epistemic) that may be of importance....

  8. Validation of computer codes used in the safety analysis of Canadian research reactors

    International Nuclear Information System (INIS)

    AECL has embarked on a validation program for the suite of computer codes that it uses in performing the safety analyses for its research reactors. Current focus is on codes used for the analysis of the two MAPLE reactors under construction at Chalk River but the program will be extended to include additional codes that will be used for the Irradiation Research Facility. The program structure is similar to that used for the validation of codes used in the safety analyses for CANDU power reactors. (author)

  9. Validation of COPERNIC code capabilities with FUMEX III database

    International Nuclear Information System (INIS)

    The development and improvement of fuel performance analysis code highly depends on the experimental research. IAEA has sponsored the FUMEX III (FUel Modeling at Extended Burnup) coordinated research project to improve computer code used for fuel behavior simulation. As one of over thirty international participants, CGNPC (China Guangdong Nuclear Power Group) has been engaged in testing and developing the fuel modeling code COPERNIC against data and cases provided by the IAEA and OECD/NEA. Data from 6 calculation cases have been compared with COPERNIC predictions. Due to different purposes of tests, these cases had different designs including rod refabrication and annular pellet and were under different operation conditions including normal operation and ramp test. The comparison and preliminary analysis between predicted and measured results in such as fuel temperature, cladding strain, cladding corrosion layer thickness, and fission gas release have been conducted. The comparison results demonstrated that the COPERNIC code was applicable to different rod designs under different operation condition with an accurate prediction. (author)

  10. Experimental validation of XRF inversion code for Chandrayaan-1

    CERN Document Server

    Athiray, P S; Tiwari, M K; Narendranath, S; Lodha, G S; Deb, S K; Sreekumar, P; Dash, S K

    2013-01-01

    We have developed an algorithm (x2abundance) to derive the lunar surface chemistry from X-ray fluorescence (XRF) data for the Chandrayaan-1 X-ray Spectrometer (C1XS) experiment. The algorithm converts the observed XRF line fluxes to elemental abundances with uncertainties. We validated the algorithm in the laboratory using high Z elements (20 < Z < 30) published in Athiray et al. (2013). In this paper, we complete the exercise of validation using samples containing low Z elements, which are also analogous to the lunar surface composition (ie., contains major elements between 11 < Z < 30). The paper summarizes results from XRF experiments performed on Lunar simulant (JSC-1A) and anorthosite using a synchrotron beam excitation. We also discuss results from the validation of x2abundance using Monte Carlo simulation (GEANT4 XRF simulation).

  11. Experimental validation of XRF inversion code for Chandrayaan-1

    OpenAIRE

    Athiray, P. S.; M Sudhakar; Tiwari, M K; Narendranath, S.; Lodha, G. S.; Deb, S. K.; Sreekumar, P.; Dash, S. K.

    2013-01-01

    We have developed an algorithm (x2abundance) to derive the lunar surface chemistry from X-ray fluorescence (XRF) data for the Chandrayaan-1 X-ray Spectrometer (C1XS) experiment. The algorithm converts the observed XRF line fluxes to elemental abundances with uncertainties. We validated the algorithm in the laboratory using high Z elements (20 < Z < 30) published in Athiray et al. (2013). In this paper, we complete the exercise of validation using samples containing low Z elements, which are a...

  12. Development, Verification and Validation of Enclosure Radiation Capabilities in the CHarring Ablator Response (CHAR) Code

    Science.gov (United States)

    Salazar, Giovanni; Droba, Justin C.; Oliver, Brandon; Amar, Adam J.

    2016-01-01

    With the recent development of multi-dimensional thermal protection system (TPS) material response codes including the capabilities to account for radiative heating is a requirement. This paper presents the recent efforts to implement such capabilities in the CHarring Ablator Response (CHAR) code developed at NASA's Johnson Space Center. This work also describes the different numerical methods implemented in the code to compute view factors for radiation problems involving multiple surfaces. Furthermore, verification and validation of the code's radiation capabilities are demonstrated by comparing solutions to analytical results, to other codes, and to radiant test data.

  13. Benchmark analyses of sodium convection in the upper plenum of the MONJU reactor vessel - Comparison between plant system analysis code CERES and CFD code -

    International Nuclear Information System (INIS)

    In the CRP of IAEA, the data of the upper plenum geometry of the prototype FBR“MONJU” and the boundary conditions of the plant trip test were provided by JAEA. A plant system analysis code CERES for FBRs was developed by CRIEPI. To verify the CERES code, analyses had been performed for the system test of the MONJU, the results of which showed good agreement with the test. However, the difficulty of accurately reproducing the temperature variation arising from a complex flow in the upper plenum was identified. By using the general-purpose analysis code STAR-CCM+, detailed analysis in the upper plenum was enabled. Based on comparison between analyses of the CERES and STAR-CCM+ codes, parameters that had to be considered to simulate the flow pattern appropriately for plant system analysis codes were discussed. And, the analysis capability of CERES code with appropriate parameter was able to be confirmed. (author)

  14. Prediction and validation of pool fire development in enclosures by means of CFD (Poolfire) Report - Year 1

    Energy Technology Data Exchange (ETDEWEB)

    van Hees, P.; Wahlqvist, J. (Lund Univ., Lund (Sweden)); Hostikka, S.; Sikanen, T. (VTT Technical Research Centre of Finland (Finland)); Husted, B. (Haugesund College, Stord (Norway)); Magnusson, T. (Ringhals AB, Vaeroebacka (Sweden)); Joerud, F. (Oskarshamn Kraftgrupp AB, Oskarshamn (Sweden))

    2012-02-15

    Fires in nuclear power plants can be an important hazard for the overall safety of the facility. One of the typical fire sources is a pool fire. It is therefore important to have good knowledge on the fire behaviour of pool fire and be able to predict the heat release rate by prediction of the mass loss rate. This project envisages developing a pyrolysis model to be used in CFD models. In the this first year report the literature review conducted within the project is reported as well as the first tasks in the evaluation and modelling of the new model. (Author)

  15. Verification and Validation Plan for the Codes LSP and ICARUS (PEGASUS); TOPICAL

    International Nuclear Information System (INIS)

    This report documents the strategies for verification and validation of the codes LSP and ICARUS used for simulating the operation of the neutron tubes used in all modern nuclear weapons. The codes will be used to assist in the design of next generation neutron generators and help resolve manufacturing issues for current and future production of neutron devices. Customers for the software are identified, tube phenomena are identified and ranked, software quality strategies are given, and the validation plan is set forth

  16. Science and code validation program to secure ignition on LMJ

    Science.gov (United States)

    Lefebvre, E.; Boniface, C.; Bonnefille, M.; Casner, A.; Esnault, C.; Galmiche, D.; Gauthier, P.; Girard, F.; Gisbert, R.; Leidinger, J.-P.; Loiseau, P.; Masse, L.; Masson-Laborde, P.-E.; Mignon, P.; Monteil, M.-C.; Seytor, P.; Tassin, V.

    2016-03-01

    The CEA/DAM ICF experimental program is currently conducted on LIL and Omega with the goal of improving our simulation tool, the FCI2 code. In this effort, we focus on typical ICF observables: hohlraum radiation drive history, capsule core shape and neutron emission history, hydrodynamic instability growth. In addition to integrated experiment, specific designs are also helpful to pinpoint a particular phenomenon. In this article, we review our current efforts and status, and our future projects on Omega and LMJ.

  17. MEMOS code validation on JET transient tungsten melting experiments

    Czech Academy of Sciences Publication Activity Database

    Bazylev, B.; Arnoux, G.; Coenen, J.W.; Matthews, G. F.; Mertens, Ph.; Knaup, M.; Jachmich, S.; Clever, M.; Dejarnac, Renaud; Coffey, I.; Corre, Y.; Devaux, S.; Gauthier, E.; Horáček, Jan; Krieger, K.; Marsen, S.; Meigs, A.; Pitts, R.A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.

    Toki City : National Institute for Fusion Science, 2014. P2-098-P2-098. [International Conference on Plasma Surface Interactions 2014/21./. 26.05.2014-30.05.2014, Kanazawa] Institutional support: RVO:61389021 Keywords : Heat loads * misalignment * JET * tungsten * melting * MEMOS * code Subject RIV: BL - Plasma and Gas Discharge Physics http://psi2014.nifs.ac.jp/Files/Files/Abstracts/P2-098_Bazylev_PSI2014.pdf

  18. Status of Verification and Validation of Physics Codes in COSINE Code Package

    International Nuclear Information System (INIS)

    COre and System INtegrated Engine for design and analysis (COSINE), an integrated nuclear engineering code package, is being developed by State Nuclear Power Software Development Center (SNPSDC) in China since 2011. A brief introduction of V and V strategy for the LATC/CORE reactor physics codes in COSINE code package was presented. And some results are shown as above. The preliminary results of V and V shows that the codes could give reasonable results, but still need to be continuously improved. In the next few years, the SNPSDC will build a test data base, including data from the critical experiments and operation plants, in order to continuously carry out the V and V of physics codes in COSINE

  19. Validation of the THIRMAL-1 melt-water interaction code

    Energy Technology Data Exchange (ETDEWEB)

    Chu, C.C.; Sienicki, J.J.; Spencer, B.W. [Argonne National Lab., IL (United States)

    1995-09-01

    The THIRMAL-1 computer code has been used to calculate nonexplosive LWR melt-water interactions both in-vessel and ex-vessel. To support the application of the code and enhance its acceptability, THIRMAL-1 has been compared with available data from two of the ongoing FARO experiments at Ispra and two of the Corium Coolant Mixing (CCM) experiments performed at Argonne. THIRMAL-1 calculations for the FARO Scoping Test and Quenching Test 2 as well as the CCM-5 and -6 experiments were found to be in excellent agreement with the experiment results. This lends confidence to the modeling that has been incorporated in the code describing melt stream breakup due to the growth of both Kelvin-Helmholtz and large wave instabilities, the sizes of droplets formed, multiphase flow and heat transfer in the mixing zone surrounding and below the melt metallic phase. As part of the analysis of the FARO tests, a mechanistic model was developed to calculate the prefragmentation as it may have occurred when melt relocated from the release vessel to the water surface and the model was compared with the relevant data from FARO.

  20. First steps towards a validation of the new burnup and depletion code TNT

    International Nuclear Information System (INIS)

    In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)

  1. Validation of SOCRAT-BN code on the base of reactor experiments

    International Nuclear Information System (INIS)

    SOCRAT-BN code is developed for safety assessment for Liquid Metal Fast Breeder Reactors (LMFBR) with sodium coolant during the design basis accidents and beyond design-basis accidents. Code consists of the several modules, which allows simulating: thermal hydraulic, neutron-physic, thermal-mechanical processes and behavior of fission products in rector vessel in coupled statement. Brief description of SOCRAT-BN code is presented in the paper. Validation on the base of the integral experiments and out-of-pile experiments were carried out to show code workability. Integral tests include experiments provided at BN-600 reactor. Detailed nodalization scheme, consists of primary, secondary and third sides, was performed to simulate BN-600 reactor. To validate SOCRAT-BN code the following BN-600 transients were chosen: natural circulation core cooling at 50% power load, and one loop tripping at about 100% power load. SOCRAT-BN code simulations results showed good agreement with the experimental data. (author)

  2. Validation of MARTHE-REACT coupled surface and groundwater reactive transport code for modeling hydro systems.

    OpenAIRE

    Thiéry, Dominique; Jacquemet, Nicolas; Picot-Colbeaux, Géraldine; Kervévan, Christophe; André, Laurent; Azaroual, Mohamed

    2009-01-01

    This paper presents the validation of the computer code MARTHE-REACT enabling the simulation of reactive transport in hydrosystems. MARTHE-REACT results from coupling the MARTHE code (flow and transport in porous media) with the chemical simulator TOUGHREACT. The resulting coupled model takes advantage of the functionalities already available in each of the two codes. In particular, it is now possible to simulate flow, reactive mass, and energy transfer in both saturated and unsaturated media...

  3. CFD aided investigation of single droplet coalescence

    Institute of Scientific and Technical Information of China (English)

    Felix Gebauer; Mark W Hlawitschka; Hans-Jrg Bart

    2016-01-01

    This article describes the development of a coalescence model using various CFD work packages, and is validated using as toluene water model system. Numerical studies were performed to describe droplet interactions in liq-uid–liquid test systems. Current models use adjustable parameters to describe these phenomena. The research in the past decades led to different correlations to model coalescence and breakage depending on the chemical sys-tem and the apparatus geometry. Especial y the complexity of droplet coalescence requires a detailed investiga-tion of local phenomena during the droplet interaction. Computational fluid dynamics (CFD) studies of single droplet interactions were performed and validated with experimental results to improve the understanding of the local hydrodynamics and film drainage during coalescence. The CFD simulations were performed for the in-teraction of two differently sized droplets at industrial relevant impact velocities. The experimental verification and validation of the numerical results were done with standardized high-speed imaging studies by using a spe-cial test cel with a pendant and a free rising droplet. An experimental based algorithm was implemented in the open source code OpenFOAM to account for the contact time and the dimple formation. The standard European Federation of Chemical Engineering (EFCE) test system toluene/water was used for the numerical studies and the experimental investigations as wel . The results of the CFD simulations are in good accordance with the observed coalescence behavior in the experimental studies. In addition, a detailed description of local phenomena, like film rupture, velocity gradients, pressures and micro-droplet entrainment could be obtained.

  4. Validation of an Euler code for hydraulic turbines

    Science.gov (United States)

    Thibaud, F.; Drotz, A.; Sottas, G.

    1988-12-01

    Validation of a 3-D internal incompressible stationary Euler flow solver was performed. A finite volume discretization scheme with an explicit time integration is used. The influence of the numerical scheme parameters on the solution and on the convergence is extensively studied. The geometry on which the numerical and experimental comparisons are presented is the runner of an hydraulic Francis turbine. The difference between calculated and experimental integral values is less than 0.2 percent.

  5. Theory and Implementation of Nuclear Safety System Codes - Part II: System Code Closure Relations, Validation, and Limitations

    Energy Technology Data Exchange (ETDEWEB)

    Glenn A Roth; Fatih Aydogan

    2014-09-01

    This is Part II of two articles describing the details of thermal-hydraulic sys- tem codes. In this second part of the article series, the system code closure relationships (used to model thermal and mechanical non-equilibrium and the coupling of the phases) for the governing equations are discussed and evaluated. These include several thermal and hydraulic models, such as heat transfer coefficients for various flow regimes, two phase pressure correlations, two phase friction correlations, drag coefficients and interfacial models be- tween the fields. These models are often developed from experimental data. The experiment conditions should be understood to evaluate the efficacy of the closure models. Code verification and validation, including Separate Effects Tests (SETs) and Integral effects tests (IETs) is also assessed. It can be shown from the assessments that the test cases cover a significant section of the system code capabilities, but some of the more advanced reactor designs will push the limits of validation for the codes. Lastly, the limitations of the codes are discussed by considering next generation power plants, such as Small Modular Reactors (SMRs), analyz- ing not only existing nuclear power plants, but also next generation nuclear power plants. The nuclear industry is developing new, innovative reactor designs, such as Small Modular Reactors (SMRs), High-Temperature Gas-cooled Reactors (HTGRs) and others. Sub-types of these reactor designs utilize pebbles, prismatic graphite moderators, helical steam generators, in- novative fuel types, and many other design features that may not be fully analyzed by current system codes. This second part completes the series on the comparison and evaluation of the selected reactor system codes by discussing the closure relations, val- idation and limitations. These two articles indicate areas where the models can be improved to adequately address issues with new reactor design and development.

  6. Validation matrices for the computer codes used for ACR safety analysis

    International Nuclear Information System (INIS)

    This paper discusses the process being used to prepare Validation Matrix Documents (VMs) for validation of the computer codes which will be used for ACR safety analysis. Computer-program validation methodology is organized as a multi-stage process associated with a structured set of documents. The first two stages are implemented with the preparation of a Technical Basis Document and Validation Matrix Documents, which are prepared without reference to specific versions of computer programs. The remaining stages of the process involve the completion of Validation Plans, Validation Exercises and Validation Manuals for specific computer programs. The key features of the ACR design which impact the physical phenomena and the expected ranges of key parameters during postulated accidents are briefly described. An overview is given of the experiments being performed specifically to support the qualification of the safety analysis codes to be used for ACR analysis. Inclusion of these new data sources in the VMs is an important element to ensure that a sound basis is in place for code validation. It is anticipated that the VMs will play a key role in supporting the US NRC's review of the ACR technology base used in the safety analysis code qualification program. The role played by the VMs in linking the fundamental phenomena and the set of design basis events, is also covered in the paper. (author)

  7. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  8. Experimental validation of the thermal-hydraulic code SACATRI

    Energy Technology Data Exchange (ETDEWEB)

    Merroun, O., E-mail: meroun.ossama@gmail.co [LMR/ERSN, Department of Physics, Faculty of Sciences, Abdelmalek Essaadi University, B.P. 2121, Tetouan (Morocco); Al Mers, A. [Department of Energetics, Ecole Nationale Superieure d' Arts et Metiers, Moulay Ismail University, B.P. 4024, Meknes (Morocco); Veloso, M.A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN), Belo Horizonte, MG (Brazil); El Bardouni, T.; El Bakkari, B. [LMR/ERSN, Department of Physics, Faculty of Sciences, Abdelmalek Essaadi University, B.P. 2121, Tetouan (Morocco); Chakir, E. [LRM/EPTN, Department of Physics, Faculty of Sciences, Kenitra (Morocco)

    2009-12-15

    A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.

  9. Experimental validation of the thermal-hydraulic code SACATRI

    International Nuclear Information System (INIS)

    A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.

  10. Validation studies of thermal-hydraulic code for safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.)

  11. Are industry codes and standards a valid cost containment approach

    International Nuclear Information System (INIS)

    The nuclear industry has historically concentrated on safety design features for many years, but recently has been shifting to the reliability of the operating systems and components. The Navy has already gone through this transition and has found that Reliability Centered Maintenance (RCM) is an invaluable tool to improve the reliability of components, systems, ships, and classes of ships. There is a close correlation of Navy ships and equipment to commercial nuclear power plants and equipment. The Navy has a central engineering and configuration management organization (Naval Sea Systems Command) for over 500 ships, where as the over 100 commercial nuclear power plants and 52 nuclear utilities represent a fragmented owner/management structure. This paper suggests that the results of the application of RCM in the Navy can be duplicated to a large degree in the commercial nuclear power industry by the development and utilization of nuclear codes and standards

  12. Evaluation of Computational Fluids Dynamics (CFD) code Open FOAM in the study of the pressurized thermal stress of PWR reactors. Comparison with the commercial code Ansys-CFX

    International Nuclear Information System (INIS)

    In this work is proposed to evaluate the potential of the OpenFOAM code for the simulation of typical fluid flows in reactors PWR, in particular for the study of pressurized thermal stress. Test T1-1 has been simulated , within the OECD ROSA project, with the objective of evaluating the performance of the code OpenFOAM and models of turbulence that has implemented to capture the effect of the thrust forces in the case study.

  13. Phase 1 Validation Testing and Simulation for the WEC-Sim Open Source Code

    Science.gov (United States)

    Ruehl, K.; Michelen, C.; Gunawan, B.; Bosma, B.; Simmons, A.; Lomonaco, P.

    2015-12-01

    WEC-Sim is an open source code to model wave energy converters performance in operational waves, developed by Sandia and NREL and funded by the US DOE. The code is a time-domain modeling tool developed in MATLAB/SIMULINK using the multibody dynamics solver SimMechanics, and solves the WEC's governing equations of motion using the Cummins time-domain impulse response formulation in 6 degrees of freedom. The WEC-Sim code has undergone verification through code-to-code comparisons; however validation of the code has been limited to publicly available experimental data sets. While these data sets provide preliminary code validation, the experimental tests were not explicitly designed for code validation, and as a result are limited in their ability to validate the full functionality of the WEC-Sim code. Therefore, dedicated physical model tests for WEC-Sim validation have been performed. This presentation provides an overview of the WEC-Sim validation experimental wave tank tests performed at the Oregon State University's Directional Wave Basin at Hinsdale Wave Research Laboratory. Phase 1 of experimental testing was focused on device characterization and completed in Fall 2015. Phase 2 is focused on WEC performance and scheduled for Winter 2015/2016. These experimental tests were designed explicitly to validate the performance of WEC-Sim code, and its new feature additions. Upon completion, the WEC-Sim validation data set will be made publicly available to the wave energy community. For the physical model test, a controllable model of a floating wave energy converter has been designed and constructed. The instrumentation includes state-of-the-art devices to measure pressure fields, motions in 6 DOF, multi-axial load cells, torque transducers, position transducers, and encoders. The model also incorporates a fully programmable Power-Take-Off system which can be used to generate or absorb wave energy. Numerical simulations of the experiments using WEC-Sim will be

  14. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  15. Experimental validation of a low-head turbine intake designed by CFD following Fisher and Franke guidelines

    International Nuclear Information System (INIS)

    Model acceptance tests evaluate the response of the turbines at different operating conditions. Model tests are mounted so that the velocity profile at the inlet section is uniform, a condition which is not often met in practice. In fact, divergences might render inaccurate model results, obtaining at prototype scale an efficiency drop, structural vibrations and even component failures, in extreme cases. This concern becomes all the more relevant for low-head turbines, as the intake is closer to the turbine runner. With the aim of best estimating the actual flow conditions at the turbine inlet section as a function of the intake design, Voith designers, Fisher and Franke recommended performing scale model tests of the intake structure and listed a series of requirements that a good intake design should meet. These guidelines have not yet been applied on numerical modeling design but rather on more expensive and time-consuming scale model tests. This work presents the results of a computational fluid dynamics (CFD) design of a low-head turbine intake taking into account an upgraded version of Fisher and Franke recommendations. The optimization process was aimed at obtaining the design that best matches the ideal flow conditions at the inlet section. The physical model was built in a scale of 1:40 and involves the complete turbine intake geometry. Different designs were tested on the basis of the evaluation of their corresponding velocity field distributions at a reference section and the best design was measured with an acoustic Doppler velocimeter (Vectrino). The results show that intake design guidelines are very useful tools that allow hydraulic designers to test their proposals with CFD more quickly, objectively and with enough degree of sensitivity to optimize the intake geometry

  16. Global hydroelastic model for springing and whipping based on a free-surface CFD code (OpenFOAM

    Directory of Open Access Journals (Sweden)

    Seng Sopheak

    2014-12-01

    Full Text Available The theoretical background and a numerical solution procedure for a time domain hydroelastic code are presented in this paper. The code combines a VOF-based free surface flow solver with a flexible body motion solver where the body linear elastic deformation is described by a modal superposition of dry mode shapes expressed in a local floating frame of reference. These mode shapes can be obtained from any finite element code. The floating frame undergoes a pseudo rigid-body motion which allows for a large rigid body translation and rotation and fully preserves the coupling with the local structural deformation. The formulation relies on the ability of the flow solver to provide the total fluid action on the body including e.g. the viscous forces, hydrostatic and hydrodynamic forces, slamming forces and the fluid damping. A numerical simulation of a flexible barge is provided and compared to experiments to show that the VOF-based flow solver has this ability and the code has the potential to predict the global hydroelastic responses accurately.

  17. CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR usando los codigos OpenFOAM y GasFlow

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez P, D. A.

    2014-07-01

    using a limited number of semi-empirical data, and instead, mathematical relationships are used taking into account the various physical phenomena as well the interactions that occur among them, such as heat transfer between the fluid and the solid walls condensation of water vapor on the walls, the turbulent effects in areas of restricted passage, etc. Taking into account these advantages, this study presents a qualitative and quantitative comparison between the CFD codes OpenFOAM and Gas-Flow related to the transport phenomena of Hydrogen and other gases in the primary containment of a BWR reactor. Gas-Flow is a code of commercial license that is well validated, developed in Germany to analyze the transport of gases in nuclear reactor containments. On the other hand, OpenFOAM is an open source CFD code offering several solvers for different phenomena assessments, in this work, the reacting Foam solver is used because it has a strong similarity to the intended application of Hydrogen transport. In this thesis the results obtained using the reacting Foam solver of OpenFOAM for the calculation of transport of Hydrogen are compared with the results of the Gas-Flow code in order to assess if it is feasible to use the open source code OpenFOAM in the case of Hydrogen transport in primary containment of a BWR reactor. Some differences in the qualitative and quantitative results from both codes were found, the differences (with a maximum error rate of 4%) in the quantitative results were found are small and are considered more than acceptable for this type of analysis, moreover, these differences are mainly attributed to the transport models used, mainly because OpenFOAM uses a homogeneous mixture model and Gas-Flow a heterogeneous one. Implementing appropriate solvers in codes like OpenFOAM has the goal to develop own tools that are applicable to the transport of Hydrogen in the primary containment of a BWR reactor and thus, to gain some independence while not relying on

  18. Validation of computer codes used in safety analyses of CANDU power plants

    International Nuclear Information System (INIS)

    Since the 1960s, the CANDU industry has been developing and using scientific computer codes for designing and analysing CANDU power plants. In this endeavour, the industry has been following nuclear quality-assurance practices of the day, including verification and validation of design and analysis methodologies. These practices have resulted in a large body of experience and expertise in the development and application of computer codes and their associated documentation. Major computer codes used in safety analyses of operating plants and those under development have been, and continue to be subjected to rigorous processes of development and application. To provide a systematic framework for the validation work done to date and planned for the future, the industry has decided to adopt the methodology of validation matrices for computer-code validation, similar to that developed by the Nuclear Energy Agency of the Organization for Economic Co-operation and Development and focused on thermalhydraulic phenomena in Light Water Reactors (LWR). To manage the development of validation matrices for CANDU power plants and to engage experts who can work in parallel on several topics, the CANDU task has been divided into six scientific disciplines. Teams of specialists in each discipline are developing the matrices. A review of each matrix will show if there are gaps or insufficient data for validation purposes and will thus help to focus future research and development, if needed. Also, the industry is examining its suite of computer codes, and their specific, additional validation needs, if any, will follow from the work on the validation matrices. The team in System Thermalhydraulics is the furthest advanced, since it had the earliest start and the international precedent on LWRs, and has developed its validation matrix. The other teams are at various stages in this multiphase, multi-year program, and their progress to date is presented. (author)

  19. Validation of the transportation computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND

    International Nuclear Information System (INIS)

    The computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND were used to estimate radiation doses from the transportation of radioactive material in the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement. HIGHWAY and INTERLINE were used to estimate transportation routes for truck and rail shipments, respectively. RADTRAN 4 was used to estimate collective doses from incident-free transportation and the risk (probability x consequence) from transportation accidents. RISKIND was used to estimate incident-free radiation doses for maximally exposed individuals and the consequences from reasonably foreseeable transportation accidents. The purpose of this analysis is to validate the estimates made by these computer codes; critiques of the conceptual models used in RADTRAN 4 are also discussed. Validation is defined as ''the test and evaluation of the completed software to ensure compliance with software requirements.'' In this analysis, validation means that the differences between the estimates generated by these codes and independent observations are small (i.e., within the acceptance criterion established for the validation analysis). In some cases, the independent observations used in the validation were measurements; in other cases, the independent observations used in the validation analysis were generated using hand calculations. The results of the validation analyses performed for HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND show that the differences between the estimates generated using the computer codes and independent observations were small. Based on the acceptance criterion established for the validation analyses, the codes yielded acceptable results; in all cases the estimates met the requirements for successful validation

  20. Validation of SCALE code package on high performance neutron shields

    International Nuclear Information System (INIS)

    The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

  1. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    International Nuclear Information System (INIS)

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated

  2. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kansal, Anuj Kumar, E-mail: akansal@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Maheshwari, Naresh Kumar, E-mail: nmahesh@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Vijayan, Pallippattu Krishnan, E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2015-06-15

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated.

  3. In-vessel core degradation code validation matrix

    International Nuclear Information System (INIS)

    The objective of the current Validation Matrix is to define a basic set of experiments, for which comparison of the measured and calculated parameters forms a basis for establishing the accuracy of test predictions, covering the full range of in-vessel core degradation phenomena expected in light water reactor severe accident transients. The scope of the review covers PWR and BWR designs of Western origin: the coverage of phenomena extends from the initial heat-up through to the introduction of melt into the lower plenum. Concerning fission product behaviour, the effect of core degradation on fission product release is considered. The report provides brief overviews of the main LWR severe accident sequences and of the dominant phenomena involved. The experimental database is summarised. These data are cross-referenced against a condensed set of the phenomena and test condition headings presented earlier, judging the results against a set of selection criteria and identifying key tests of particular value. The main conclusions and recommendations are listed. (K.A.)

  4. Validation of NASA Thermal Ice Protection Computer Codes. Part 1; Program Overview

    Science.gov (United States)

    Miller, Dean; Bond, Thomas; Sheldon, David; Wright, William; Langhals, Tammy; Al-Khalil, Kamel; Broughton, Howard

    1996-01-01

    The Icing Technology Branch at NASA Lewis has been involved in an effort to validate two thermal ice protection codes developed at the NASA Lewis Research Center. LEWICE/Thermal (electrothermal deicing & anti-icing), and ANTICE (hot-gas & electrothermal anti-icing). The Thermal Code Validation effort was designated as a priority during a 1994 'peer review' of the NASA Lewis Icing program, and was implemented as a cooperative effort with industry. During April 1996, the first of a series of experimental validation tests was conducted in the NASA Lewis Icing Research Tunnel(IRT). The purpose of the April 96 test was to validate the electrothermal predictive capabilities of both LEWICE/Thermal, and ANTICE. A heavily instrumented test article was designed and fabricated for this test, with the capability of simulating electrothermal de-icing and anti-icing modes of operation. Thermal measurements were then obtained over a range of test conditions, for comparison with analytical predictions. This paper will present an overview of the test, including a detailed description of: (1) the validation process; (2) test article design; (3) test matrix development; and (4) test procedures. Selected experimental results will be presented for de-icing and anti-icing modes of operation. Finally, the status of the validation effort at this point will be summarized. Detailed comparisons between analytical predictions and experimental results are contained in the following two papers: 'Validation of NASA Thermal Ice Protection Computer Codes: Part 2- The Validation of LEWICE/Thermal' and 'Validation of NASA Thermal Ice Protection Computer Codes: Part 3-The Validation of ANTICE'

  5. Preliminary Validation of the MATRA-LMR Code Using Existing Sodium-Cooled Experimental Data

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sangji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The main objective of the SFR prototype plant is to verify TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. The fuel design limit is highly dependent on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, an accurate temperature calculation in each subassembly is highly important to assure a safe and reliable operation of the reactor systems. The current core thermalhydraulic design is mainly performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code, which has been already validated using the existing sodium-cooled experimental data. In addition to the SLTHEN code, a detailed analysis is performed using the MATRA-LMR (Multichannel Analyzer for Transient and steady-state in Rod Array-Liquid Metal Reactor) code. In this work, the MATRA-LMR code is validated for a single subassembly evaluation using the previous experimental data. The MATRA-LMR code has been validated using existing sodium-cooled experimental data. The results demonstrate that the design code appropriately predicts the temperature distributions compared with the experimental values. Major differences are observed in the experiments with the large pin number due to the radial-wise mixing difference.

  6. Preliminary Validation of the MATRA-LMR Code Using Existing Sodium-Cooled Experimental Data

    International Nuclear Information System (INIS)

    The main objective of the SFR prototype plant is to verify TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. The fuel design limit is highly dependent on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, an accurate temperature calculation in each subassembly is highly important to assure a safe and reliable operation of the reactor systems. The current core thermalhydraulic design is mainly performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code, which has been already validated using the existing sodium-cooled experimental data. In addition to the SLTHEN code, a detailed analysis is performed using the MATRA-LMR (Multichannel Analyzer for Transient and steady-state in Rod Array-Liquid Metal Reactor) code. In this work, the MATRA-LMR code is validated for a single subassembly evaluation using the previous experimental data. The MATRA-LMR code has been validated using existing sodium-cooled experimental data. The results demonstrate that the design code appropriately predicts the temperature distributions compared with the experimental values. Major differences are observed in the experiments with the large pin number due to the radial-wise mixing difference

  7. Verification, validation, and benchmarking report for TRIMHX: A three dimensional hexagonal transient diffusion theory code

    International Nuclear Information System (INIS)

    TRIMHX is a fundamental Reactor Analysis tool in use at the Savannah River Site (SRS) and is an integral part of the Generalized Reactor Analysis Subsystem (GRASS). TRIMHX solves the time dependent multigroup neutron diffusion equation in two and three dimensional hexagonal geometry by standard and coarse mesh finite difference methods. The TRIMHX implementation assumes the solution to this equation can be discretized in space, energy, and time. These are industry accepted approaches which can be found in many nuclear engineering books. This report concerns the verification and validation of TRIMHX, a transient two and three dimensional hex-z diffusion theory code. The validation was performed to determine the accuracy of the code, and the verification was performed to determine if the code was correctly using the correct theory and that all the subroutines function as required. For TRIMHX, the validation requirement was satisfied by comparing the results of the code with experiments and benchmarking the code against other standard or validated code results. The verification requirement for TRIMHX was performed indirectly since it is impossible and not necessary to reverify a large code like TRIMHX line by line. The extensive operations history of TRIMHX in conjunction with the comparisons against many numerical experiments (exact solutions) and other diffusion theory codes is sufficient to establish that the code is functioning as intended and therefore it is verified. This report summarizes four sets of experiments performed in 1974, 1977, and 1988, two DIF3D/TRIMHX comparison problems performed in 1991, a DIF3D/FX2-TH/TRIMHX comparison problem produced for this report, and the comparison of TRIMHX/GRIMHX initial static calculations. The results of these experiments show that TRIMHX was correctly implemented and is ready to submit into SCMS production mode

  8. Modification and validation of ATHLET code for sodium-cooled fast reactor application

    International Nuclear Information System (INIS)

    System analysis code is important for the global simulation of the sodium- cooled fast reactor (SFR) system as well as transient and accident safety analysis. In this paper, the best estimate system code ATHLET for light water reactors, developed by Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS) in Germany, was modified for SFR application. Thermal-dynamic and transport properties as well as heat transfer correlations for sodium were implemented into the ATHLET code. The modified code was then applied to simulate the Phenix reactor in France, and validation of the code was conducted with the Phenix reactor natural convection test. The calculation results were compared with the test data. The results show that the modified ATHLET code has good applicability in simulating SFR systems. (authors)

  9. Correlation Equations of Heat Transfer in Nanofluid Al2O3-Water as Cooling Fluid in a Rectangular Sub Channel Based CFD Code

    Directory of Open Access Journals (Sweden)

    Anwar Ilmar Ramadhan

    2015-03-01

    Full Text Available Safety is a major concern in the design, operation and development of a nuclear reactor. One aspect of nuclear reactor safety factor is thermal-hydraulics aspect. In a PWR-type nuclear power plant has been used lighter fluid coolant is water or H2O. In this research, using nanofluid Al2O3-Water with volume fraction of (1%, (2% and also (3%, used as a cooling fluid in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement. This research was carried out modeling of fuel elements are arranged rectangular, then performed numerical simulations using Computational Fluid Dynamics (CFD code. In order to obtain the characteristic pattern of flow velocity of each fluid, the fluid temperature distribution along the cylinder wall temperature distribution of the fuel element. Then analyzed the heat transfer in a nuclear reactor core with sub channel PWR fuel element rectangular arrangement, including heat transfer coefficient, Nusselt number (Nu, as well as heat transfer correlations. Heat transfer correlation for nanofluid Al2O3-Water (1%, (2% and also (3% proved to core of PWR nuclear reactor fuel element sub channel rectangular arrangement with the Reynolds number (Re is stretched, namely: 404 096

  10. Application perspectives of simulation techniques CFD in nuclear power plants

    International Nuclear Information System (INIS)

    The scenarios simulation in nuclear power plants is usually carried out with system codes that are based on concentrated parameters networks. However situations exist in some components where the flow is predominantly 3-D, as they are the natural circulation, mixed and stratification phenomena. The simulation techniques of computational fluid dynamics (CFD) have the potential to simulate these flows numerically. The use of CFD simulations embraces many branches of the engineering and continues growing, however, in relation to its application with respect to the problems related with the safety in nuclear power plants, has a smaller development, although is accelerating quickly and is expected that in the future they play a more emphasized paper in the analyses. A main obstacle to be able to achieve a general acceptance of the CFD is that the simulations should have very complete validation studies, sometimes not available. In this article a general panorama of the state of the methods application CFD in nuclear power plants is presented and the problem associated to its routine application and acceptance, including the view point of the regulatory authorities. Application examples are revised in those that the CFD offers real benefits and are also presented two illustrative study cases of the application of CFD techniques. The case of a water recipient with a heat source in its interior, similar to spent fuel pool of a nuclear power plant is presented firstly; and later the case of the Boron dilution of a water volume that enters to a nuclear reactor is presented. We can conclude that the CFD technology represents a very important opportunity to improve the phenomena understanding with a strong component 3-D and to contribute in the uncertainty reduction. (Author)

  11. Global hydroelastic model for springing and whipping based on a free-surface CFD code (OpenFOAM)

    DEFF Research Database (Denmark)

    Seng, Sopheak; Jensen, Jørgen Juncher; Malenica, Sime

    2014-01-01

    dry mode shapes expressed in a local floating frame of reference. These mode shapes can be obtained from any finite element code. The floating frame undergoes a pseudo rigid-body motion which allows for a large rigid body translation and rotation and fully preserves the coupling with the local...... structural deformation. The formulation relies on the ability of the flow solver to provide the total fluid action on the body including e.g. the viscous forces, hydrostatic and hydrodynamic forces, slamming forces and the fluid damping. A numerical simulation of a flexible barge is provided and compared to...

  12. Pre-engineering Spaceflight Validation of Environmental Models and the 2005 HZETRN Simulation Code

    Science.gov (United States)

    Nealy, John E.; Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.; Dachev, Ts. P.; Tomov, B. T.; Walker, Steven A.; DeAngelis, Giovanni; Blattnig, Steve R.; Atwell, William

    2006-01-01

    The HZETRN code has been identified by NASA for engineering design in the next phase of space exploration highlighting a return to the Moon in preparation for a Mars mission. In response, a new series of algorithms beginning with 2005 HZETRN, will be issued by correcting some prior limitations and improving control of propagated errors along with established code verification processes. Code validation processes will use new/improved low Earth orbit (LEO) environmental models with a recently improved International Space Station (ISS) shield model to validate computational models and procedures using measured data aboard ISS. These validated models will provide a basis for flight-testing the designs of future space vehicles and systems of the Constellation program in the LEO environment.

  13. Optimised Cockpit Heat Load Analysis using Skin Temperature Predicted by CFD and Validation by Thermal Mapping to Improve the Performance of Fighter Aircraft

    Directory of Open Access Journals (Sweden)

    Paresh Gupta

    2015-03-01

    Full Text Available Designing of optimum environmental control system (ECS plays a major role for increasing performance of fighter aircraft depending upon requirement of engine bleed air for running of ECS. Accurate estimation of cockpit skin temperature for obtaining optimised cockpit heat load helps in estimation of engine bleed air for ECS. Present research evolved a methodology for comparing the theoretically calculated skin temperature with computational fluid dynamics (CFD analysis to obtain optimum skin temperature. Results are validated by flight tests under critical flight conditions using thermal crayons. Based on which the optimized heat load and bleed air requirements has been computed. Uncertainty analysis of skin temperature measurement for thermal crayons have been undertaken. The results indicate that the theoretical skin temperature is -26.70 per cent as that of CFD estimated skin temperature. Optimized average cockpit heat load at critical flight profiles is 0.74 times the theoretical cockpit heat load, leading to reduction of bleed air requirement by 26 per cent as compared to theoretical. Due to this literature survey has pridicted the increase in performance parameters like increase in bleed air pressure by 78 per cent, increase in thrust by 60 per cent, and decrease in specific fuel consumption (SFC by 40 per cent to improve the endurance of aircraft. The research has generated governing equations for variation of cockpit heat loads w.r.t aircraft skin temperatures.Defence Science Journal, Vol. 65, No. 1, January 2015, pp.12-24, DOI:http://dx.doi.org/10.14429/dsj.65.7200

  14. Validation of the ATHLET-SC code by trans-critical transient data

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaojing; Cheng, Xu [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering

    2016-05-15

    For the safety analysis of Supercritical Water-Cooled Reactor (SCWR), one of the challenge tasks is to predict the trans-critical behavior of the reactor system during some accidents. The current safety codes have some shortcomings when the pressure decreases from the supercritical condition to the subcritical state due to the void fraction discontinuity across the critical point. Another challenge is the validation of the system code, which needs the transient experimental data. To overcome the above-mentioned challenges, this paper validates the modified code ATHLET-SC, which is developed based on the pseudo two-phase method. The trans-critical transient data from SWAMUP test facility in Shanghai Jiao Tong University (SJTU) are adopted to compare with the simulation results. The results obtained so far shows that the ATHLET-SC code has good feasibility to the trans-critical simulation of SCWR, and it can be used for transient analysis of SCWR in the future.

  15. First verification and validation steps of mendel release 1.0 cycle code system

    International Nuclear Information System (INIS)

    For each new code system, verification and validation process is a need to prove the efficiency and accuracy of the calculated physical quantities. MENDEL is the new CEA depletion code system, whose first release was done at the end of 2013. It offers iso-capacity with the already well-established DARWIN. MENDEL is the successor of DARWIN, and can be used as a stand-alone code system for reactor cycle studies to compute interest output quantities. MENDEL also provides its depletion solvers to both Monte Carlo TRIPOLI-4® and deterministic APOLLO3® transport code systems. The purpose of this paper is to present the first contributions to MENDEL release 1.0 verification and validation process. This first release has been used with nuclear data coming from both JEFF-3.1.1 and ENDF/B-VII.1 nuclear data evaluations, and its results are compared either with experimental data, either with DARWIN results. (author)

  16. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    1999-09-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotope) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, were considered to be of sufficient quality for depletion code validation.

  17. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses -- Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.

    2000-03-16

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotopes) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data, usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, was considered to be of sufficient quality for depletion code validation.

  18. Validation of the ATHLET-SC code by trans-critical transient data

    International Nuclear Information System (INIS)

    For the safety analysis of Supercritical Water-Cooled Reactor (SCWR), one of the challenge tasks is to predict the trans-critical behavior of the reactor system during some accidents. The current safety codes have some shortcomings when the pressure decreases from the supercritical condition to the subcritical state due to the void fraction discontinuity across the critical point. Another challenge is the validation of the system code, which needs the transient experimental data. To overcome the above-mentioned challenges, this paper validates the modified code ATHLET-SC, which is developed based on the pseudo two-phase method. The trans-critical transient data from SWAMUP test facility in Shanghai Jiao Tong University (SJTU) are adopted to compare with the simulation results. The results obtained so far shows that the ATHLET-SC code has good feasibility to the trans-critical simulation of SCWR, and it can be used for transient analysis of SCWR in the future.

  19. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    International Nuclear Information System (INIS)

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  20. Computational Methods for HSCT-Inlet Controls/CFD Interdisciplinary Research

    Science.gov (United States)

    Cole, Gary L.; Melcher, Kevin J.; Chicatelli, Amy K.; Hartley, Tom T.; Chung, Joongkee

    1994-01-01

    A program aimed at facilitating the use of computational fluid dynamics (CFD) simulations by the controls discipline is presented. The objective is to reduce the development time and cost for propulsion system controls by using CFD simulations to obtain high-fidelity system models for control design and as numerical test beds for control system testing and validation. An interdisciplinary team has been formed to develop analytical and computational tools in three discipline areas: controls, CFD, and computational technology. The controls effort has focused on specifying requirements for an interface between the controls specialist and CFD simulations and a new method for extracting linear, reduced-order control models from CFD simulations. Existing CFD codes are being modified to permit time accurate execution and provide realistic boundary conditions for controls studies. Parallel processing and distributed computing techniques, along with existing system integration software, are being used to reduce CFD execution times and to support the development of an integrated analysis/design system. This paper describes: the initial application for the technology being developed, the high speed civil transport (HSCT) inlet control problem; activities being pursued in each discipline area; and a prototype analysis/design system in place for interactive operation and visualization of a time-accurate HSCT-inlet simulation.

  1. Validation of the Serpent 2 code on TRIGA Mark II benchmark experiments.

    Science.gov (United States)

    Ćalić, Dušan; Žerovnik, Gašper; Trkov, Andrej; Snoj, Luka

    2016-01-01

    The main aim of this paper is the development and validation of a 3D computational model of TRIGA research reactor using Serpent 2 code. The calculated parameters were compared to the experimental results and to calculations performed with the MCNP code. The results show that the calculated normalized reaction rates and flux distribution within the core are in good agreement with MCNP and experiment, while in the reflector the flux distribution differ up to 3% from the measurements. PMID:26516989

  2. Experimental program for real gas flow code validation at NASA Ames Research Center

    Science.gov (United States)

    Deiwert, George S.; Strawa, Anthony W.; Sharma, Surendra P.; Park, Chul

    1989-01-01

    The experimental program for validating real gas hypersonic flow codes at NASA Ames Rsearch Center is described. Ground-based test facilities used include ballistic ranges, shock tubes and shock tunnels, arc jet facilities and heated-air hypersonic wind tunnels. Also included are large-scale computer systems for kinetic theory simulations and benchmark code solutions. Flight tests consist of the Aeroassist Flight Experiment, the Space Shuttle, Project Fire 2, and planetary probes such as Galileo, Pioneer Venus, and PAET.

  3. Validation of ICD-9 Codes for Stable Miscarriage in the Emergency Department

    OpenAIRE

    Quinley, Kelly E.; Falck, Ailsa; Kallan, Michael J.; Datner, Elizabeth M.; Carr, Brendan G.; Schreiber, Courtney A.

    2015-01-01

    Introduction: International Classification of Disease, Ninth Revision (ICD-9) diagnosis codes have not been validated for identifying cases of missed abortion where a pregnancy is no longer viable but the cervical os remains closed. Our goal was to assess whether ICD-9 code “632” for missed abortion has high sensitivity and positive predictive value (PPV) in identifying patients in the emergency department (ED) with cases of stable early pregnancy failure (EPF). Methods:...

  4. Validation of ICD-9 Codes for Stable Miscarriage in the Emergency Department

    OpenAIRE

    Quinley, Kelly E.; Falck, Ailsa; Kallan, Michael J.; Datner, Elizabeth M.; Carr, Brendan G.; Schreiber, Courtney A.

    2015-01-01

    Introduction International Classification of Disease, Ninth Revision (ICD-9) diagnosis codes have not been validated for identifying cases of missed abortion where a pregnancy is no longer viable but the cervical os remains closed. Our goal was to assess whether ICD-9 code “632” for missed abortion has high sensitivity and positive predictive value (PPV) in identifying patients in the emergency department (ED) with cases of stable early pregnancy failure (EPF). Methods We studied females ages...

  5. Validation of capture yield calculations in the Resolved Resonance Energy Range with CONRAD code

    Science.gov (United States)

    Litaize, Olivier; Archier, Pascal; Becker, Bjorn; Schillebeeckx, Peter; Kopecky, Stefan

    2013-03-01

    This paper deals with the validation of the multiple scattering corrections developed in the CONRAD code for the capture yield calculations in the Resolved Resonance energy Range (RRR). In order to calculate the capture yields, analytic and stochastic calculation schemes implemented in CONRAD are described and compared with the analysis code SAMMY/SAMSMC. The results are in excellent agreement for a variety of samples. We concentrate the discussion here on 238U, 197Au and 55Mn.

  6. Validation of capture yield calculations in the Resolved Resonance Energy Range with CONRAD code

    Directory of Open Access Journals (Sweden)

    Schillebeeckx Peter

    2013-03-01

    Full Text Available This paper deals with the validation of the multiple scattering corrections developed in the CONRAD code for the capture yield calculations in the Resolved Resonance energy Range (RRR. In order to calculate the capture yields, analytic and stochastic calculation schemes implemented in CONRAD are described and compared with the analysis code SAMMY/SAMSMC. The results are in excellent agreement for a variety of samples. We concentrate the discussion here on 238U, 197Au and 55Mn.

  7. ANDELA - operating parameters database for validation of the ANDREA code. Technical report

    International Nuclear Information System (INIS)

    Operating data from the Temelin NPP are used as input for validation of the ANDREA neutron physics code and related nuclear data libraries. The auxiliary ANDELA code contains a relational database which stores operating data in the form of condensed histories and makes possible their import directly from the data files supplied by the NPP operator, creation of input files for ANDREA for specific operating cycles, and export of tables of controlled parameters for easy comparison between calculations and observations. (P.A.)

  8. Validation of ICD-9-CM coding algorithm for improved identification of hypoglycemia visits

    OpenAIRE

    Lieberman Rebecca M; Blanc Phillip G; Ginde Adit A; Camargo Carlos A

    2008-01-01

    Abstract Background Accurate identification of hypoglycemia cases by International Classification of Diseases, Ninth Revision, Clinical Modification (ICD-9-CM) codes will help to describe epidemiology, monitor trends, and propose interventions for this important complication in patients with diabetes. Prior hypoglycemia studies utilized incomplete search strategies and may be methodologically flawed. We sought to validate a new ICD-9-CM coding algorithm for accurate identification of hypoglyc...

  9. Validation of the Subchannel Code SUBCHANFLOW Using the NUPEC PWR Tests (PSBT)

    International Nuclear Information System (INIS)

    SUBCHANFLOW is a computer code to analyze thermal-hydraulic phenomena in the core of pressurized water reactors, boiling water reactors, and innovative reactors operated with gas or liquid metal as coolant. As part of the ongoing assessment efforts, the code has been validated by using experimental data from the NUPEC PWR Subchannel and Bundle Tests (PSBT). The database includes single-phase flow bundle outlet temperature distributions, steady state and transient void distributions and critical power measurements. The performed validation work has demonstrated that the two-phase flow empirical knowledge base implemented in SUBCHANFLOW is appropriate to describe key mechanisms of the experimental investigations with acceptable accuracy.

  10. Extensive validation of the code FUROM based on the IFPE database

    International Nuclear Information System (INIS)

    The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. The validation of the code has so far been based on the individual calculation of many relevant experiments. This, however, was a time consuming process that could give rise to errors both at the input and at the comparison stage. A new methodology is implemented to build up a uniform database from the IFPE data and run automated validation tasks depending on the model or phenomenon of interest. The general problems encountered and some results are presented here. (authors)

  11. VULCAN: an Open-Source, Validated Chemical Kinetics Python Code for Exoplanetary Atmospheres

    OpenAIRE

    Tsai, Shang-Min; Lyons, James R.; Grosheintz, Luc; Rimmer, Paul B.; Kitzmann, Daniel; Heng, Kevin

    2016-01-01

    We present an open-source and validated chemical kinetics code for studying hot exoplanetary atmospheres, which we name VULCAN. It is constructed for gaseous chemistry from 500 to 2500 K using a reduced C- H-O chemical network with about 300 reactions. It uses eddy diffusion to mimic atmospheric dynamics and excludes photochemistry. We have provided a full description of the rate coefficients and thermodynamic data used. We validate VULCAN by reproducing chemical equilibrium and by comparing ...

  12. Validation of the TRIAD3 code used for the neutronic simulation of the NRU reactor

    International Nuclear Information System (INIS)

    The neutronic simulation of the NRU research reactor at Chalk River is performed by the TRIAD3 code. TRIAD3 is a three-dimensional code using a modified neutron diffusion theory in two-energy groups. The modification is the use of cell discontinuity factors (cdf) to improve the radial neutron leakage calculation between adjacent cells. This paper describes three validation exercises performed over the past few years. It describes methods of obtaining the flux, power and reactivity measurements from the NRU reactor and presents comparisons between these measurement data and code simulation results. (author)

  13. Validation of the TRIAD3 code used for the neutronic simulation of the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C.; Atfield, M.D. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2009-07-01

    The neutronic simulation of the NRU research reactor at Chalk River is performed by the TRIAD3 code. TRIAD3 is a three-dimensional code using a modified neutron diffusion theory in two-energy groups. The modification is the use of cell discontinuity factors (cdf) to improve the radial neutron leakage calculation between adjacent cells. This paper describes three validation exercises performed over the past few years. It describes methods of obtaining the flux, power and reactivity measurements from the NRU reactor and presents comparisons between these measurement data and code simulation results. (author)

  14. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors

    International Nuclear Information System (INIS)

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d4 code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  15. Development and validation of Monte-Carlo burnup calculation code MCNTRANS

    International Nuclear Information System (INIS)

    A new nuclear fuel burnup calculation code MCNTRANS based on MCNP was introduced in this paper. The neutronics calculation parameter was extracted from the MCNP5 reaction rate tally result, while a graph theory algorithm was implemented to track the burnup chain and the analytic solution of the Bateman equation was given. At the same time, the detailed physical process was considered to improve the accuracy and serviceability of this code, and prediction-correction method was used to allow a large burnup step. The OECD/NEA and JAERI pin cell benchmark problems were used to validate the code MCNTRANS while a reference result was given by other code. It can be concluded that the calculation results of MCNTRANS are generally consistent with the experimental result and that of the other burnup codes, and part of the actinides and fission products calculation result show better accuracy. (authors)

  16. Development and Validation of A Nuclear Fuel Cycle Analysis Tool: A FUTURE Code

    International Nuclear Information System (INIS)

    This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses

  17. Code-to-Code Validation and Application of a Building Dynamic Simulation Tool for the Building Energy Performance Analysis

    Directory of Open Access Journals (Sweden)

    Annamaria Buonomano

    2016-04-01

    Full Text Available In this paper details about the results of a code-to-code validation procedure of an in-house developed building simulation model, called DETECt, are reported. The tool was developed for research purposes in order to carry out dynamic building energy performance and parametric analyses by taking into account new building envelope integrated technologies, novel construction materials and innovative energy saving strategies. The reliability and accuracy of DETECt was appropriately tested by means of the standard BESTEST validation procedure. In the paper, details of this validation process are accurately described. A good agreement between the obtained results and all the reference data of the BESTEST qualification cases is achieved. In particular, the obtained results vs. standard BESTEST output are always within the provided ranges of confidence. In addition, several test cases output obtained by DETECt (e.g., dynamic profiles of indoor air and building surfaces temperature and heat fluxes and spatial trends of temperature across walls are provided.

  18. CFD modeling of the test 25 of the PANDA experiment

    International Nuclear Information System (INIS)

    A large amount of steam and Hydrogen gas is expected to be released within the dry containment of a pressurized water reactor (PWR), after the hypothetical beginning of a severe accident leading to the melting of the core. The accurate modeling of gas distribution in a PWR containment concerns phenomena such as wall condensation, hydrogen accumulation, gas stratification and transport in the different compartments of the containment. The paper presents numerical assessments of CFD solvers NEPTUNE-CFD and Code-Saturne, and is focused on the analysis and the understanding of gas stratification and transport phenomena. NEPTUNE-CFD is dedicated to the simulation of incompressible and compressible multi-component/multi-phase flows, whereas Code-Saturne is dedicated to homogeneous incompressible or low Mach number compressible flows, with only one momentum equation representing the mixture of gases, liquid and particles. NEPTUNE-CFD is mainly used for nuclear engineering, whereas Code-Saturne is used either for nuclear and fossil energy engineering, and for environment (geophysical flows). The NEPTUNE-CFD code 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. Both codes are validated and compared with experimental data corresponding to the test 25 of the PANDA experiment. This test concerns the distribution of mixture of gases (Helium as a simulant of hydrogen and condensing steam) in air over two vertical and cylindrical vessels, interconnected by a horizontal and cylindrical pipe. The overall dimensions of the experiment (Diameter ∼4 m, Height 8 m, Volume of the 2 vessels ∼180 m3) are not yet representative of the true scale of the reactors, but they already provide valuable information when compared to smaller scales (as experience TOSQAN∼7 m3). The computational results with Code-Saturne and NEPTUNE-CFD compare

  19. Code validation for the magnetohydrodynamic flow at high Hartmann Number based on unstructured grid

    International Nuclear Information System (INIS)

    Highlights: • In this paper, a numerical code named MTC-H 2.0 based on unstructured mash was developed and validated. • In the code, a current density conservative scheme is used to ensure the conservative of current, Krylov subspace method and AMG method were used to improve the computing performance. • Both analytical and experimental cases were chosen to validate the code at high Hartmann number. • The code has good computing performance, and numerical results matched well with the analytical and experimental results. -- Abstract: In order to analyze the magnetohydrodynamic (MHD) effect in liquid metal fusion blanket, a parallel and high performance numerical code was developed to study MHD flows at high Hartmann Number based on the unstructured grid. In this code, the induced current and the Lorentz force were calculated with a current density conservative scheme, while the incompressible Navier–Stokes equations with the Lorentz force included as a source term was solved by projection method, a set of method were used to improve the computing performance such as Krylov subspace method and AMG method. To validate this code, three benchmarks of MHD flow at high Hartmann Number were conducted. The first benchmark was the case of Shercliff fully development flow, the second benchmark was the MHD flow in a circular pipe with changing external magnetic field, and the third benchmark was the MHD flow in a pipe with sudden expansion. In these cases the Hartmann Numbers were from 1000 to 6000. The code good computing performance, and numerical results show matched well with the analytical and experimental results

  20. Generic validation of computer codes used in safety analyses of CANDU power plants

    International Nuclear Information System (INIS)

    Since the 1960s, the CANDU industry has been developing and using scientific computer codes, validated according to the quality-assurance practices of the day, for designing and analyzing CANDU power plants. To provide a systematic framework for the validation work done to date and planned for the future, the industry has decided to adopt the methodology of validation matrices, similar to that developed by the Nuclear Energy Agency of the Organization for Economic Co-operation and Development for Light Water Reactors (LWR). Specialists in six scientific disciplines are developing the matrices for CANDU plants, and their progress to date is presented. (author)

  1. CFD Simulation of Liquid Rocket Engine Injectors

    Science.gov (United States)

    Farmer, Richard; Cheng, Gary; Chen, Yen-Sen; Garcia, Roberto (Technical Monitor)

    2001-01-01

    these investigators to be very valuable for code validation because combustion kinetics, turbulence models and atomization models based on low pressure experiments of hydrogen air combustion do not adequately verify analytical or CFD submodels which are necessary to simulate rocket engine combustion. We wish to emphasize that the simulations which we prepared for this meeting are meant to test the accuracy of the approximations used in our general purpose spray combustion models, rather than represent a definitive analysis of each of the experiments which were conducted. Our goal is to accurately predict local temperatures and mixture ratios in rocket engines; hence predicting individual experiments is used only for code validation. To replace the conventional JANNAF standard axisymmetric finite-rate (TDK) computer code 2 for performance prediction with CFD cases, such codes must posses two features. Firstly, they must be as easy to use and of comparable run times for conventional performance predictions. Secondly, they must provide more detailed predictions of the flowfields near the injector face. Specifically, they must accurately predict the convective mixing of injected liquid propellants in terms of the injector element configurations.

  2. The ASBM-SA turbulence closure: Taking advantage of structure-based modeling in current engineering CFD codes

    International Nuclear Information System (INIS)

    Highlights: • We outline a new coupling of the ASBM model with one equation SA closure. • We provide mathematical implementation details and a Fortran module. • The numerical stability and convergence characteristics of ASBM-SA are comparable to that of SA. • The ASBM-SA model provides full Reynolds stress anisotropy information unlike the SA model. • Overall, predictions os ASBM-SA shows moderate accuracy improvements relative to those of SA. - Abstract: Structure-based turbulence models (SBM) carry information about the turbulence structure that is needed for the prediction of complex non-equilibrium flows. SBM have been successfully used to predict a number of canonical flows, yet their adoption rate in engineering practice has been relatively low, mainly because of their departure from standard closure formulations, which hinders easy implementation in existing codes. Here, we demonstrate the coupling between the Algebraic Structure-Based Model (ASBM) and the one-equation Spalart–Allmaras (SA) model, which provides an easy route to bringing structure information in engineering turbulence closures. As the ASBM requires correct predictions of two turbulence scales, which are not taken into account in the SA model, Bradshaw relations and numerical optimizations are used to provide the turbulent kinetic energy and dissipation rate. Attention is paid to the robustness and accuracy of the hybrid model, showing encouraging results for a number of simple test cases. An ASBM module in Fortran-90 is provided along with the present paper in order to facilitate the testing of the model by interested readers

  3. Investigation on effect of equivalence ratio and engine speed on homogeneous charge compression ignition combustion using chemistry based CFD code

    Directory of Open Access Journals (Sweden)

    Ghafouri Jafar

    2014-01-01

    Full Text Available Combustion in a large-bore natural gas fuelled diesel engine operating under Homogeneous Charge Compression Ignition mode at various operating conditions is investigated in the present paper. Computational Fluid Dynamics model with integrated chemistry solver is utilized and methane is used as surrogate of natural gas fuel. Detailed chemical kinetics mechanism is used for simulation of methane combustion. The model results are validated using experimental data by Aceves, et al. (2000, conducted on the single cylinder Volvo TD100 engine operating at Homogeneous Charge Compression Ignition conditions. After verification of model predictions using in-cylinder pressure histories, the effect of varying equivalence ratio and engine speed on combustion parameters of the engine is studied. Results indicate that increasing engine speed provides shorter time for combustion at the same equivalence ratio such that at higher engine speeds, with constant equivalence ratio, combustion misfires. At lower engine speed, ignition delay is shortened and combustion advances. It was observed that increasing the equivalence ratio retards the combustion due to compressive heating effect in one of the test cases at lower initial pressure. Peak pressure magnitude is increased at higher equivalence ratios due to higher energy input.

  4. Intercomparison and validation of computer codes for thermalhydraulic safety analysis of heavy water reactors

    International Nuclear Information System (INIS)

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and co-operative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled: Intercomparison and validation of computer codes for thermalhydraulics safety analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. The RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for this validation project. This report provides a comparison of the results obtained from six participating countries, utilizing four different computer codes. General conclusions are reached and recommendations made

  5. Validity of the coding for herpes simplex encephalitis in the Danish National Patient Registry

    DEFF Research Database (Denmark)

    Jørgensen, Laura Krogh; Dalgaard, Lars Skov; Østergaard, Lars Jørgen;

    2016-01-01

    BACKGROUND: Large health care databases are a valuable source of infectious disease epidemiology if diagnoses are valid. The aim of this study was to investigate the accuracy of the recorded diagnosis coding of herpes simplex encephalitis (HSE) in the Danish National Patient Registry (DNPR). METH...

  6. Decay Heat Code Validation Activities at ORNL: Supporting Expansion of NRC Regulatory Guide 3.54

    International Nuclear Information System (INIS)

    Oak Ridge National Laboratory (ORNL) has a long history of involvement in the development and validation of the ORIGEN series of isotope summation codes and nuclear data libraries, widely recognized and used to predict the decay heat for spent nuclear fuel. In particular, the ORIGEN-S code, the depletion/decay module of the SCALE code system, has been extensively validated using experimental isotopic assay data and decay heat measurements for commercial spent fuel. This work was used in the development of the technical basis for NRC Regulatory Guide 3.54 on spent fuel decay heat. The bulk of the experimental data used to validate spent fuel decay heat predictions are from programs of the 1970s and 1980s and consequently involve older-design fuel assemblies with a relatively low enrichment and burnup. This has led to a situation where the spent fuel now being discharged from operating reactors extends well beyond the regime of the experimental data and area of code applicability based on the data. The absence of validation data for modern fuel designs has potentially serious consequences for decay heat predictions in terms of added safety factors to account for larger uncertainties and lower volumetric transport and storage capacities

  7. Forces and Moments in CFD Analysis

    Czech Academy of Sciences Publication Activity Database

    Říha, Zdeněk; Foldyna, Josef

    Praha : TechSoft Engineering s.r.o., SVS FEM s.r.o., 2010, s. 1-11. ISBN 978-80-254-8388-6. [ ANSYS Konference 2010. Frymburk (CZ), 06.10.2010-09.10.2010] Institutional research plan: CEZ:AV0Z30860518 Keywords : force * CFD analysis * CFD code Subject RIV: JQ - Machines ; Tools

  8. Validation of Numerical Schemes in a Thermal-Hydraulic Analysis Code for a Natural Convection Heat Transfer of a Molten Pool

    International Nuclear Information System (INIS)

    , unsteady turbulence models based on filtered or volume-averaged governing equations have been applied for the turbulent natural convection heat transfer. Tran et al. used large eddy simulation (LES) for the analysis of molten corium coolability. The numerical instability is related to a gravitational force of the molten corium. A staggered grid method on an orthogonal structured grid is used to prohibit a pressure oscillation in the numerical solution. But it is impractical to use the structured grid for a partially filled spherical pool, a cone-type pool or triangular pool. An unstructured grid is an alternative for the nonrectangular pools. In order to remove the checkerboard- like pressure oscillation on the unstructured grid, some special interpolation scheme is required. In order to evaluate in-vessel coolability of the molten corium for a pressurized water reactor (PWR), thermo-hydraulic analysis code LILAC had been developed. LILAC has a capability of multi-layered conjugate heat transfer with melt solidification. A solution domain can be 2-dimensional, axisymmetric, and 3-dimensional. LILAC is based on the unstructured mesh technology to discretized non-rectangular pool geometry. Because of too limited man-power to maintain the code, it becomes more and more difficult to implement new physical and numerical models in the code along with increased complication of the code. Recently, open source CFD code OpenFOAM has been released and applied to many academic and engineering areas. OpenFOAM is based on the very similar numerical schemes to the LILAC code. It has many physical and numerical models for multi-physics analysis. And because it is based on object-oriented programming, it is known that new models can be easily implemented and is very fast with a lower possibility of coding errors. This is a very attractive feature for the development, validation and maintenance of an analysis code. On the contrary to commercial CFD codes, it is possible to modify and add

  9. Validation of the subchannel code CTF against the benchmark data of the OECD/NEA PSBT

    International Nuclear Information System (INIS)

    Nuclear reactor safety analysis has been increased rapidly in the last decades. Due to this, advanced methods for core phenomena predictions are developed, specially related to thermal-hydraulics using subchannel codes. CTF is a version of the subchannel code COBRA-TF improved by the ISIRYM. This code must be validated against full-scale high-quality experimental data, which allow comparing results for several case types. The benchmark of the OECD/NEA PWR Subchannel and Bundle Test (PSBT) has been carried out in order to validate the features of CTF during steady state and transient cases in Light Water Reactors (LWR). The objective of the benchmark is to analyze the void fraction predicted by CTF in 4 different cases for steady state and 4 transient cases. Benchmark cases are performed with 3 different bundle types. (author)

  10. Validation of the HELIOS.HX code for high conversion light water reactor lattice analysis

    International Nuclear Information System (INIS)

    The HELIOS.HX code has been developed for the design study of high conversion light water reactor (HCLWR) lattices. Analysis of the PROTEUS critical experiments at the Swiss Federal Institute for Reactor Research has been carried out as the first step toward validation of the HELIOS.HX code, and indications are that the accuracy may be at a higher or comparable level compared to that of WIMS-D, EPRI-CPM, and SRAC. In addition, comparisons with Monte Carlo calculations have also been performed for an HCLWR fuel assembly benchmark problem, showing that the accuracy is passable in the prediction of important nuclear characteristics, thereby indicating the validity of various approximations involved in the physics methods. These numerical results indicate that the code has basic potential as a tool for HCLWR lattice analysis, but covers only limited HCLWR lattice conditions

  11. Preliminary validation of the MATRA-LMR-FB code for the flow blockage in a subassembly

    International Nuclear Information System (INIS)

    To analyze the flow blockage in a subassembly of a Liquid Metal-cooled Reactor (LMR), the MATRA-LMR-FB code has been developed and validated for the existing experimental data. Compared to the MATRA-LMR code, which had been successfully applied for the core thermal-hydraulic design of KALIMER, the MATRA-LMR-FB code includes some advanced modeling features. Firstly, the Distributed Resistance Model (DRM), which enables a very accurate description of the effects of wire-wrap and blockage in a flow path, is developed for the MATRA-LMR-FB code. Secondly, the hybrid difference method is used to minimize the numerical diffusion especially at the low flow region such as recirculating wakes after blockage. In addition, the code is equipped with various turbulent mixing models to describe the active mixing due to the turbulent motions as accurate as possible. For the validation of the MATRA-LMR-FB code the ORNL THORS test and KOS 169-pin test are analyzed. Based on the analysis results for the temperature data, the accuracy of the code is evaluated quantitatively. The MATRA-LMR-FB code predicts very accurately the exit temperatures measured in the subassembly with wire-wrap. However, the predicted temperatures for the experiment with spacer grid show some deviations from the measured. To enhance the accuracy of the MATRA-LMR-FB for the flow path with grid spacers, it is suggested to improve the models for pressure loss due to spacer grid and the modeling method for blockage itself. The developed MATRA-LMR-FB code is evaluated to be applied to the flow blockage analysis of KALIMER-600 which adopts the wire-wrapped subassemblies

  12. Preliminary validation of the MATRA-LMR-FB code for the flow blockage in a subassembly

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, H. Y.; Ha, K. S.; Kwon, Y. M.; Chang, W. P.; Lee, Y. B.; Heo, S

    2005-01-01

    To analyze the flow blockage in a subassembly of a Liquid Metal-cooled Reactor (LMR), the MATRA-LMR-FB code has been developed and validated for the existing experimental data. Compared to the MATRA-LMR code, which had been successfully applied for the core thermal-hydraulic design of KALIMER, the MATRA-LMR-FB code includes some advanced modeling features. Firstly, the Distributed Resistance Model (DRM), which enables a very accurate description of the effects of wire-wrap and blockage in a flow path, is developed for the MATRA-LMR-FB code. Secondly, the hybrid difference method is used to minimize the numerical diffusion especially at the low flow region such as recirculating wakes after blockage. In addition, the code is equipped with various turbulent mixing models to describe the active mixing due to the turbulent motions as accurate as possible. For the validation of the MATRA-LMR-FB code the ORNL THORS test and KOS 169-pin test are analyzed. Based on the analysis results for the temperature data, the accuracy of the code is evaluated quantitatively. The MATRA-LMR-FB code predicts very accurately the exit temperatures measured in the subassembly with wire-wrap. However, the predicted temperatures for the experiment with spacer grid show some deviations from the measured. To enhance the accuracy of the MATRA-LMR-FB for the flow path with grid spacers, it is suggested to improve the models for pressure loss due to spacer grid and the modeling method for blockage itself. The developed MATRA-LMR-FB code is evaluated to be applied to the flow blockage analysis of KALIMER-600 which adopts the wire-wrapped subassemblies.

  13. On the formulation of species reaction rates in the context of multi-species CFD codes using complex chemistry tabulation techniques

    Energy Technology Data Exchange (ETDEWEB)

    Michel, Jean-Baptiste; Colin, Olivier; Angelberger, Christian [IFP, 1 et 4 Avenue de Bois Preau, F-92500 Rueil Malmaison (France)

    2010-04-15

    In the standard implementation of tabulated combustion models of the FPI or FGM type, the mean species mass fractions are read from look-up tables as functions of a progress variable, mixture fraction and their variances. In multi-species CFD codes however, the mean thermodynamic properties are deduced from the local mean species mass fractions. The unclosed mean source terms appearing in the latter's transport equations must then be given by the chemistry look-up tables. Two possible formulations for this mean source terms are discussed and compared in the present paper. In the reaction rate (RR) formulation, all mean reaction rates are directly read from a look-up table. In the mass fraction (MF) formulation, only the reaction rate for the progress variable is stored, and mean species source terms are constructed to relax the mean mass fractions towards the value stored in the look-up table. After a detailed description of in particular the MF formulation, simple a priori tests of auto-igniting reactors without convection and diffusion are used to illustrate and discuss the differences between the two formulations. Both formulations are then applied to a RANS simulation of the Cabra et al. burner in the context of a PCM-FPI and of an ADF-PCM model. The reported findings confirm the conclusions from the simple tests, highlighting the definitive advantages of the MF formulation. It ensures an accurate reproduction of auto-ignition delays, species evolutions and equilibriums, at the condition that the relaxation parameter is of the order of a characteristic chemical time. Finally, it is shown that the relaxation's effect is only a second order correction. (author)

  14. Further validation and development of the 3-dimensional dynamics code TRAB-3D

    International Nuclear Information System (INIS)

    TRAB-3D, the newest member of VTT's code system for LWR dynamics calculations, is a coupled neutronics-thermal hydraulics code for transient and accident analyses of BWR reactors. The code is largely based on the ID code TRAB and the 3D hexagonal code HEXTRAN which have long been used in safety analyses of Finnish and foreign reactors. In TRAB-3D the two-group neutron diffusion equations are solved in three dimensions in a rectangular fuel assembly geometry by a new method which is similar to the nodal expansion method developed earlier at VTT for hexagonal geometry. The accuracy of the method is shown by comparison with 2D fine-mesh calculations and with 3D calculation for the Olkiluoto reactor with the POLCA-4 code. Capabilities of the code in dynamic analyses is validated with the OECD/NEA LWR benchmark problems and with transient calculations for the Olkiluoto reactor. Further development of the code includes a pin power reconstruction method which makes use of precomputed power distributions within fuel assemblies

  15. CFD aided analysis of a scaled down model of the Brazilian Multipurpose Reactor (RMB) pool

    International Nuclear Information System (INIS)

    Research reactors are commonly built inside deep pools that provide radiological and thermal protection and easy access to its core. Reactors with thermal power in the order of MW usually use an auxiliary thermal-hydraulic circuit at the top of its pool to create a purified hot water layer (HWL). Thermal-hydraulic analysis of the flow configuration in the pool and HWL is paramount to insure radiological protection. A useful tool for these analyses is the application of CFD (Computational Fluid Dynamics). To obtain satisfactory results using CFD it is necessary the verification and validation of the CFD numerical model. Verification is divided in code and solution verifications. In the first one establishes the correctness of the CFD code implementation and in the former estimates the numerical accuracy of a particular calculation. Validation is performed through comparison of numerical and experimental results. This paper presents a dimensional analysis of the RMB (Brazilian Multipurpose Reactor) pool to determine a scaled down experimental installation able to aid in the HWL numerical investigation. Two CFD models were created one with the same dimensions and boundary conditions of the reactor prototype and the other with 1/10 proportion size and boundary conditions set to achieve the same inertial and buoyant forces proportions represented by Froude Number between the two models. Results comparing the HWL thickness show consistence between the prototype and the scaled down model behavior. (author)

  16. CFD aided analysis of a scaled down model of the Brazilian Multipurpose Reactor (RMB) pool

    Energy Technology Data Exchange (ETDEWEB)

    Schweizer, Fernando L.A.; Lima, Claubia P.B.; Costa, Antonella L.; Veloso, Maria A.F., E-mail: ando.schweizer@gmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: antonella@nuclear.ufmg.br, E-mail: mdora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Santos, Andre A.C.; Costa, Antonio C.L., E-mail: aacs@cdtn.br, E-mail: aclc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN/-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    Research reactors are commonly built inside deep pools that provide radiological and thermal protection and easy access to its core. Reactors with thermal power in the order of MW usually use an auxiliary thermal-hydraulic circuit at the top of its pool to create a purified hot water layer (HWL). Thermal-hydraulic analysis of the flow configuration in the pool and HWL is paramount to insure radiological protection. A useful tool for these analyses is the application of CFD (Computational Fluid Dynamics). To obtain satisfactory results using CFD it is necessary the verification and validation of the CFD numerical model. Verification is divided in code and solution verifications. In the first one establishes the correctness of the CFD code implementation and in the former estimates the numerical accuracy of a particular calculation. Validation is performed through comparison of numerical and experimental results. This paper presents a dimensional analysis of the RMB (Brazilian Multipurpose Reactor) pool to determine a scaled down experimental installation able to aid in the HWL numerical investigation. Two CFD models were created one with the same dimensions and boundary conditions of the reactor prototype and the other with 1/10 proportion size and boundary conditions set to achieve the same inertial and buoyant forces proportions represented by Froude Number between the two models. Results comparing the HWL thickness show consistence between the prototype and the scaled down model behavior. (author)

  17. Validation of thermal hydraulic computer codes for advanced light water reactor

    International Nuclear Information System (INIS)

    The Czech Republic operates 4 WWER-440 units, two WWER-1000 units are being finalised (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppressure system are modelled with RALOC and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems. An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. The paper provides a concise information on these activities of the NRI and its Thermal-hydraulics Department. A detailed example of the system code validation and the consequent utilisation of the results for a real NPP purposes is included. (author)

  18. Test and validation of the iterative code for the neutrons spectrometry and dosimetry: NSDUAZ

    International Nuclear Information System (INIS)

    In this work was realized the test and validation of an iterative code for neutronic spectrometry known as Neutron Spectrometry and Dosimetry of the Universidad Autonoma de Zacatecas (NSDUAZ). This code was designed in a user graph interface, friendly and intuitive in the environment programming of LabVIEW using the iterative algorithm known as SPUNIT. The main characteristics of the program are: the automatic selection of the initial spectrum starting from the neutrons spectra catalog compiled by the International Atomic Energy Agency, the possibility to generate a report in HTML format that shows in graph and numeric way the neutrons flowing and calculates the ambient dose equivalent with base to this. To prove the designed code, the count rates of a spectrometer system of Bonner spheres were used with a detector of 6LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: 252Cf and 239PuBe were used to validate the code, the obtained results were compared against those obtained using the BUNKIUT code. We find that the reconstructed spectra present an error that is inside the limit reported in the literature that oscillates around 15%. Therefore, it was concluded that the designed code presents similar results to those techniques used at the present time. (Author)

  19. Two-phase CFD modeling of flow causing the heater vibration

    International Nuclear Information System (INIS)

    Vibrations of heater rods were observed in a heated annulus with water flow under boiling conditions. In order to find out the cause of such vibrations, CFD model of this annulus has been prepared in CFD code STAR-CCM+. Two-phase flow in the annulus was described using a two-fluid model with number of sub-models to describe the mass, momentum and energy transfer between phases. The model was validated using experimental data from reference. The validated model was used to perform a steady state calculation of flow parameters under different conditions. Results of CFD simulations were compared to experimentally detected vibration offset. It was found out that vibration increase caused by heating the channel is connected with the vibration offset. The results and their extension to nuclear safety were discussed. (author)

  20. Validation of the KARATE code system by using the reevaluated ZR-6 measurements

    International Nuclear Information System (INIS)

    In the frame of upgrading the KARATE-440 code system the application of ENDF/B-VI nuclear data had been decided in all stages of the calculation. Recently the multi group libraries used by the MULTICELL lattice calculation code and the few group libraries used by the COREMICRO 2D fine-mesh diffusion code have been prepared. The validation of these libraries has been performed using measurements on zero power critical facilities. The standard data base used in that work was extended by some temperature dependent parameters of the slightly enriched uranium piles of ZR-6. In the whole work the experimental data of ZR-6 reevaluated by the RFIT code was used (Authors)

  1. Validation of sonic boom propagation codes using SR-71 flight test data

    Science.gov (United States)

    Ivanteyeva, Lyudmila G.; Kovalenko, Victor V.; Pavlyukov, Evgeny V.; Teperin, Leonid L.; Rackl, Robert G.

    2002-01-01

    The results of two sonic boom propagation codes, ZEPHYRUS (NASA) and BOOM (TsAGI), are compared with SR-71 flight test data from 1995. Options available in the computational codes are described briefly. Special processing methods are described which were applied to the experimental data. A method to transform experimental data at close ranges to the supersonic aircraft into initial data required by the codes was developed; it is applicable at any flight regime. Data are compared in near-, mid-, and far fields. The far-field observation aircraft recorded both direct and reflected waves. Comparison of computed and measured results shows good agreement with peak pressure, duration, and wave shape for direct waves, thus validating the computational codes.

  2. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  3. Validation of the modified ATHLET code with the natural convection test of the PHENIX reactor

    International Nuclear Information System (INIS)

    Highlights: • Modification of system code ATHLET for Sodium-cooled Fast Reactors application. • Development of a properties package as well as a heat transfer package for sodium. • Validation of the modified code with the PHENIX reactor ultimate natural convection test. - Abstract: This paper presents the modification, validation and application of the system code ATHLET for Sodium-cooled Fast Reactors. The ATHLET code is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) in Germany for Light Water Reactors application. The code structure is modified that it can be easily extended to different fluids. For the present analysis of SFR, a sodium property package as well as heat transfer correlations for sodium are implemented into the code. To evaluate the feasibility of the modified code, the PHENIX reactor, a SFR operated by French Alternative Energies and Atomic Energy Commission (CEA) from 1973 to 2009, is modeled. Two different modeling approaches of the hot and cold plenum of the PHENIX reactor are adopted, i.e. the one-dimensional representation with standard pipes, and the pseudo-three-dimensional representation with parallel channels connected via cross connections. Both models are used for simulation of the natural convection ultimate test scenario of the PHENIX reactor, and the results are compared with the measured data. Improvement approach for the pseudo-three-dimensional modeling method is proposed and realized. The results reveal advantage as well as limitation of the current models, and good applicability of the modified ATHLET code to sodium-cooled reactor systems

  4. Methods for Computationally Efficient Structured CFD Simulations of Complex Turbomachinery Flows

    Science.gov (United States)

    Herrick, Gregory P.; Chen, Jen-Ping

    2012-01-01

    This research presents more efficient computational methods by which to perform multi-block structured Computational Fluid Dynamics (CFD) simulations of turbomachinery, thus facilitating higher-fidelity solutions of complicated geometries and their associated flows. This computational framework offers flexibility in allocating resources to balance process count and wall-clock computation time, while facilitating research interests of simulating axial compressor stall inception with more complete gridding of the flow passages and rotor tip clearance regions than is typically practiced with structured codes. The paradigm presented herein facilitates CFD simulation of previously impractical geometries and flows. These methods are validated and demonstrate improved computational efficiency when applied to complicated geometries and flows.

  5. Experience from the HELHEX code package validation at the Kozloduy NPP

    International Nuclear Information System (INIS)

    A new code package HELHEX (HELios-HEX3dp) for steady-state neutron physics calculations of WWER-1000 reactors has been developed at the University of Sofia. The authors of the code are Prof. Petko Petkov, Prof. Ivaylo Christoskov and Mr. Plamen Petkov. HELHEX consists of a 3D two-group nodal diffusion code HEX3DA and a 3D two group pin-by-pin diffusion code HEX3DP. The nodal and pin-by-pin cross-section libraries are generated using the Helios-1.5 lattice code. The albedo coefficients for the radial and axial boundaries are calculated with the Helios-Mariko system. A visualization module HEX3VI is also included. This paper contains a brief account of the HELHEX code package validation work performed at the Kozloduy NPP. The HELHEX calculated neutron-physics characteristics have been compared with relevant measured data and with results produced by the KASKAD code package. For all calculated data the applicable acceptance criteria are satisfied. This result allows the introduction of the new software and XS-libraries for regular use at the Kozloduy NPP. (authors)

  6. Verification and Validation: High Charge and Energy (HZE) Transport Codes and Future Development

    Science.gov (United States)

    Wilson, John W.; Tripathi, Ram K.; Mertens, Christopher J.; Blattnig, Steve R.; Clowdsley, Martha S.; Cucinotta, Francis A.; Tweed, John; Heinbockel, John H.; Walker, Steven A.; Nealy, John E.

    2005-01-01

    In the present paper, we give the formalism for further developing a fully three-dimensional HZETRN code using marching procedures but also development of a new Green's function code is discussed. The final Green's function code is capable of not only validation in the space environment but also in ground based laboratories with directed beams of ions of specific energy and characterized with detailed diagnostic particle spectrometer devices. Special emphasis is given to verification of the computational procedures and validation of the resultant computational model using laboratory and spaceflight measurements. Due to historical requirements, two parallel development paths for computational model implementation using marching procedures and Green s function techniques are followed. A new version of the HZETRN code capable of simulating HZE ions with either laboratory or space boundary conditions is under development. Validation of computational models at this time is particularly important for President Bush s Initiative to develop infrastructure for human exploration with first target demonstration of the Crew Exploration Vehicle (CEV) in low Earth orbit in 2008.

  7. Database of Temelin NPP Operational States and Its Use for Neutron Codes Validation

    International Nuclear Information System (INIS)

    Analogous to NPP Dukovany is made for NPP Temelin database of operational states. The database ETEBase is needed for the validation of various reactor computing codes, which will be developed during NPP Temelin life cycle and used for WWER-1000 core analyses. To obtain licenses in Czech Republic for new neutron codes programs it is needed to publish technical report about validation and evaluated precision of the computer codes. Benchmark data sets are processed from operational measurements data on Temelin WWER-1000 reactors. The input data from the NPP are verified; errors and inaccuracies are filtered out. Required data are chosen and processed, and then data are transferred to a form suitable for input data for neutron codes and for validation. Main data contained in benchmark dataset: effective time, boron concentration, thermal power, position of working group control clusters, inlet coolant temperature and flow rate of coolant water. Additional 3D-data are stored only for chosen time points (approx. 40 per cycle) - axial and radial power distribution in full and 60-degree core symmetry. Also datasets contain core description and list of outages during the cycle. At present, ETEBase contains processed data from these unit/cycles: 1-1, 1-2, 1-3 (partial of data), 2-1, 2-2 (Author)

  8. Application of Simple CFD Models in Smoke Ventilation Design

    DEFF Research Database (Denmark)

    Brohus, Henrik; Nielsen, Peter Vilhelm; la Cour-Harbo, Hans;

    2004-01-01

    The paper examines the possibilities of using simple CFD models in practical smoke ventilation design. The aim is to assess if it is possible with a reasonable accuracy to predict the behaviour of smoke transport in case of a fire. A CFD code mainly applicable for “ordinary” ventilation design is...... used for the examination. The CFD model is compared with benchmark tests and results from a special application fire simulation CFD code. Apart from benchmark tests two practical applications are examined in shape of modelling a fire in a theatre and a double façade, respectively. The simple CFD model...

  9. Development of PIRT and assessment matrix for verification and validation of sodium fire analysis codes

    International Nuclear Information System (INIS)

    Thermodynamic consequence in liquid sodium leak and fire accident is one of the important issues to be evaluated when considering the safety aspect of fast reactor plant building. The authors are therefore initiating systematic verification and validation (V and V) activity to assure and demonstrate the reliability of numerical simulation tool for sodium fire analysis. The V and V activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The events to be evaluated are hypothetical sodium spray, pool, or combined fire accidents followed by thermodynamic behaviors postulated in a plant building. The present paper describes that the 'Phenomena Identification and Ranking Table (PIRT)' is developed at first for clarifying the important validation points in the sodium fire analysis codes, and that an 'Assessment Matrix' is proposed which summarizes both separate effect tests and integral effect tests for validating the computational models or whole code for important phenomena. Furthermore, the paper shows a practical validation with a separate effect test in which the spray droplet combustion model of SPHINCS and AQUA-SF predicts the burned amount of a falling sodium droplet with the error mostly less than 30%. (author)

  10. Development of Validation System for Subchannel Analysis Codes under Steady-State PWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, Kyung Won; Kwon, Hyouk [KAERI, Daejeon (Korea, Republic of)

    2010-10-15

    Subchannel analysis code plays an essential role for the thermal hydraulic design of PWR cores. It calculates subchannel-wise local properties under single- and two-phase flow conditions which are used as input parameters for evaluating design parameters such as minimum DNBR and maximum fuel temperature. A validation system for subchannel analysis codes for steady-state conditions is provided in this study through investigation of validation status of existing subchannel codes, review of licensing guideline for thermal-hydraulic codes, and establishment of thermal hydraulic data base for test bundles. Thermal hydraulic test data has been procured for the following experiments: CNEN 4x4 flow distribution test, CU 4x4 flow and enthalpy distribution test, PNL 7x7 flow blockage test, GE 3x3 two-phase flow and quality distribution test, CE 15x15 inlet jetting test, WH 14x14 inlet blockage test, PNL 2x6 low flow test, ISPRA 4x4 two-phase flow and quality distribution test, FRIGG 36-rod void distribution test, Zion-1 plant FA exit temperature distribution test. Sampling analysis for each test has been conducted by employing KAERI inhouse subchannel analysis code MATRA. The state-of-the-art for subchannel thermal hydraulic analysis models such as void fraction correlations, crossflow and turbulent mixing models, and heat transfer correlations was also investigated

  11. Validation of a Subchannel Analysis Code MATRA Version 1.0

    International Nuclear Information System (INIS)

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

  12. Validation of a Subchannel Analysis Code MATRA Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  13. High Fidelity CFD Analysis and Validation of Rotorcraft Gearbox Aerodynamics Under Operational and Oil-Out Conditions

    Science.gov (United States)

    Kunz, Robert F.

    2014-01-01

    This document represents the evolving formal documentation of the NPHASE-PSU computer code. Version 3.15 is being delivered along with the software to NASA in 2013.Significant upgrades to the NPHASE-PSU have been made since the first delivery of draft documentation to DARPA and USNRC in 2006. These include a much lighter, faster and memory efficient face based front end, support for arbitrary polyhedra in front end, flow-solver and back-end, a generalized homogeneous multiphase capability, and several two-fluid modelling and algorithmic elements. Specific capability installed for the NASA Gearbox Windage Aerodynamics NRA are included in this version: Hybrid Immersed Overset Boundary Method (HOIBM) [Noack et. al (2009)] Periodic boundary conditions for multiple frames of reference, Fully generalized immersed boundary method, Fully generalized conjugate heat transfer, Droplet deposition, bouncing, splashing models, and, Film transport and breakup.

  14. Experimental Investigation of Coolant Mixing in WWER and PWR Reactor Fuel Bundles by Laser Optical Techniques for CFD Validation

    International Nuclear Information System (INIS)

    Non intrusive laser optical measurements have been carried out to investigate the coolant mixing in a model of the head part of a fuel assembly of a WWER reactor. The goal of this research was to investigate the coolant flow around the point based in-core thermocouple; and also provide experimental database as a validation tool for computational fluid dynamics calculations. The experiments have been carried out on a full size scale model of the head part of WWER-440/213 fuel assembly. In this paper first the previous results of the research project is summarised, when full field velocity vectors and temperature were obtained by particle image velocimetry and planar laser induced fluorescence, respectively. Then, preliminary results of the investigation of the influence of the flow in the central tube will be reported by presenting velocity measurement results. In order to have well measurable effect, extreme flow rates have been set in the central tube by applying an inner tube with controlled flow rates. Despite the extreme conditions, the influence of the central tube to the velocity field proved to be significant. Further measurement will be done for the investigation of the effect of the gaps at the spacer fixings by displacing the inner tube vertically, and also the temperature distribution will also be determined at similar geometries by laser induced fluorescence. The aim of the measurements was to establish an experimental database, as well as the validation of computational fluid dynamics calculations. (Authors)

  15. Toward a CFD-grade database addressing LWR containment phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Paladino, Domenico, E-mail: domenico.paladino@psi.ch [Laboratory for Thermal-Hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland); Andreani, Michele; Zboray, Robert; Dreier, Joerg [Laboratory for Thermal-Hydraulics, Nuclear Energy and Safety Department, Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer The SETH-2 PANDA tests have supplied data with CFD-grade on plumes and jets at large-scale. Black-Right-Pointing-Pointer The PANDA tests have contributed to the understanding of phenomena with high safety relevance for LWRs. Black-Right-Pointing-Pointer The analytical activities related increased confidence in the use of various computational tools for safety analysis. - Abstract: The large-scale, multi-compartment PANDA facility (located at PSI in Switzerland) is one of the state-of-the-art facilities which is continuously upgraded to progressively match the requirements of CFD-grade experiments. Within the OECD/SETH projects, the PANDA facility has been used for the creation of an experimental database on basic containment phenomena e.g. gas mixing, transport, stratification, condensation. In the PANDA tests, these phenomena are driven by large scale plumes or jets. In the paper is presented a selection of the SETH PANDA experimental results. Examples of analytical activities performed at PSI using the GOTHIC, CFX-4 and CFX-5 codes will be used to illustrate how the spatial and temporal resolutions of the measurement grid in PANDA tests are adequate for CFD code (and advanced containment codes) assessment and validation purposes.

  16. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    International Nuclear Information System (INIS)

    This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

  17. Heat removal (wetting, heat transfer, T/H, secondary circuit, code validation etc.)

    Energy Technology Data Exchange (ETDEWEB)

    Dury, T.; Siman-Tov, M.

    1996-06-01

    This working group provided a comprehensive list of feasibility and uncertainty issues. Most of the issues seem to fall into the `needed but can be worked out` category. They feel these can be worked out as the project develops. A few issues can be considered critical or feasibility issues (that must be proven to be feasible). Those include: (1) Thermal shock and its mitigation (>1 MW); how to inject the He bubbles (if used) - back pressure into He lines - mercury traces in He lines; how to maintain proper bubble distribution and size (static and dynamic; if used); vibrations and fatigue (dynamic); possibility of cavitation from thermal shock. (2) Wetting and/or non-wetting of mercury on containment walls with or without gases and its effect on heat transfer (and materials). (3) Prediction capabilities in the CFD code; bubbles behavior in mercury (if used) - cross stream turbulence (ESS only) - wetting/non-wetting effects. (4) Cooling of beam `windows`; concentration of local heat deposition at center, especially if beam is of parabolic profile.

  18. Revised Burnup Code System SWAT: Description and Validation Using Postirradiation Examination Data

    International Nuclear Information System (INIS)

    The burnup code system Step-Wise Burnup Analysis Code System (SWAT) is revised for use in a burnup credit analysis. An important feature of the revised SWAT is that its functions are achieved by calling validated neutronics codes without any changes to the original codes. This feature is realized with a system function of the operating system, which allows the revised SWAT to be independent of the development status of each code.A package of the revised SWAT contains the latest libraries based on JENDL-3.2 and the second version of the JNDC FP library. These libraries allow us to analyze burnup problems, such as an analysis of postirradiation examination (PIE), using the latest evaluated data of not only cross sections but also fission yield and decay constants.Another function of the revised SWAT is a library generator for the ORIGEN2 code, which is one of the most reliable burnup codes. ORIGEN2 users can obtain almost the same results with the revised SWAT using the library prepared by this function.The validation of the revised SWAT is conducted by calculation of the Organization for Economic Cooperation and Development/Nuclear Energy Agency burnup credit criticality safety benchmark Phase I-B and analyses of PIE data for spent fuel from Takahama Unit 3. The analysis of PIE data shows that the revised SWAT can predict the isotopic composition of main uranium and plutonium with a deviation of 5% from experimental results taken from UO2 fuels of 17 x 17 fuel assemblies. Many results of fission products including samarium are within a deviation of 10%. This means that the revised SWAT has high reliability to predict the isotopic composition for pressurized water reactor spent fuel

  19. Development and validation of the ENIGMA code for MOX fuel performance modelling

    International Nuclear Information System (INIS)

    The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

  20. Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly

    International Nuclear Information System (INIS)

    The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k∞) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.