WorldWideScience

Sample records for cfd code validation

  1. 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

  2. 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)

  3. 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.

  4. 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)

  5. 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.

  6. 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

  7. 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.

  8. 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

    with sheared inflow are used to compare the vortex code performance with CFD and lidar measurements. Laminar CFD computations are used to evaluate the performance of the viscous models. Consistent results between the vortex code and CFD tool are obtained up to three diameters downstream. The modelling......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...... of 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....

  9. 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.

  10. 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

  11. 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)

  12. 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)

  13. 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.

  14. 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.

  15. 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)

  16. 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

  17. 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 ...

  18. 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)

  19. 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...

  20. 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.

  1. 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.

  2. 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)

  3. 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...

  4. 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

  5. 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

  6. 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...

  7. 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

  8. 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

  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. 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)

  11. CFD code comparison for 2D airfoil flows

    DEFF Research Database (Denmark)

    Sørensen, Niels N.; Méndez, B.; Muñoz, A.;

    2016-01-01

    The current paper presents the effort, in the EU AVATAR project, to establish the necessary requirements to obtain consistent lift over drag ratios among seven CFD codes. The flow around a 2D airfoil case is studied, for both transitional and fully turbulent conditions at Reynolds numbers of 3 × ...

  12. 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

  13. 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

  14. Methodology for computational fluid dynamics code verification/validation

    Energy Technology Data Exchange (ETDEWEB)

    Oberkampf, W.L.; Blottner, F.G.; Aeschliman, D.P.

    1995-07-01

    The issues of verification, calibration, and validation of computational fluid dynamics (CFD) codes has been receiving increasing levels of attention in the research literature and in engineering technology. Both CFD researchers and users of CFD codes are asking more critical and detailed questions concerning the accuracy, range of applicability, reliability and robustness of CFD codes and their predictions. This is a welcomed trend because it demonstrates that CFD is maturing from a research tool to the world of impacting engineering hardware and system design. In this environment, the broad issue of code quality assurance becomes paramount. However, the philosophy and methodology of building confidence in CFD code predictions has proven to be more difficult than many expected. A wide variety of physical modeling errors and discretization errors are discussed. Here, discretization errors refer to all errors caused by conversion of the original partial differential equations to algebraic equations, and their solution. Boundary conditions for both the partial differential equations and the discretized equations will be discussed. Contrasts are drawn between the assumptions and actual use of numerical method consistency and stability. Comments are also made concerning the existence and uniqueness of solutions for both the partial differential equations and the discrete equations. Various techniques are suggested for the detection and estimation of errors caused by physical modeling and discretization of the partial differential equations.

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. Validation of CFD-models for non-stoichiometric oxycoal combustion

    Energy Technology Data Exchange (ETDEWEB)

    Bohn, Jan-Peter; Goanta, Adrian; Baumgartner, Andreas; Blume, Maximilian; Spliethoff, Hartmut [Technische Univ. Muenchen, Garching (Germany). Inst. of Energy Systems

    2013-07-01

    To compensate the drawback of high flue gas recirculation rates specific for oxyfuel processes, a new concept based on staged combustion, called controlled staging with non-stoichiometric burners (CSNB) was investigated. A combination of over- and sub-stoichiometric burners avoids inadmissible high flame temperatures even with oxygen concentrations up to 40 vol.-% in the oxidant. The non-stoichiometric burners are arranged in such a way that the overall stoichiometry at the combustion chamber outlet is slightly over-stoichiometric similar to conventional combustion processes, so that full burn out is secured. This concept aims at a more efficient oxyfuel process due to the decreasing effort in the recirculation loop and a new designed, more cost effective, steam generator. For further process optimization of the CSNB concept and the adjacent steam generator layout are validated CFD simulations urgently required. This paper shows first steps in validation of the CSNB combustion concept against state of the art CFD codes. The CFD code was optimized for oxyfuel combustion including a new char combustion and gas radiation model. Experimental investigations and CFD modelling are showing good agreement in concerns of temperature, CO and CO{sub 2} profiles. However the prediction of the oxygen concentration differs significantly between experiment and simulation.

  1. 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

  2. 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.

  3. 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

  4. 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)

  5. 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.

  6. 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.

  7. 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...

  8. 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.

  9. 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

  10. 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.)

  11. 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)

  12. Teaching CFD as a Black Box: A Validation and Verification Approach

    Science.gov (United States)

    Hertzberg, Jean

    2012-11-01

    There are a number of good reasons for NOT teaching computational fluid dynamics to undergraduates: a reluctance to make room in an already-compressed curriculum, the sophistication of the computational techniques and mathematics involved, the cost of licensing a professional quality code, and above all, the danger that a shallow understanding of CFD will lead to blithely accepted incorrect results. Nevertheless, as today's students enter the workplace they are routinely expected to be able to use CFD and other high level software packages. Industry's response to the necessity of minimally trained engineers using such software is a series of tests prior to accepting the results: verification and validation (V&V), or more specifically independent software verification and validation (ISVV). The verification question asks ``is the software producing correct answers, given the inputs?'' while the validation question asks ``is this the right set of inputs, are the right physics being addressed?'' A recent attempt to implement a V&V approach to CFD in the required undergraduate curriculum at the University of Colorado will be described.

  13. A 3D-CFD code for accurate prediction of fluid flows and fluid forces in seals

    Science.gov (United States)

    Athavale, M. M.; Przekwas, A. J.; Hendricks, R. C.

    1994-01-01

    Current and future turbomachinery requires advanced seal configurations to control leakage, inhibit mixing of incompatible fluids and to control the rotodynamic response. In recognition of a deficiency in the existing predictive methodology for seals, a seven year effort was established in 1990 by NASA's Office of Aeronautics Exploration and Technology, under the Earth-to-Orbit Propulsion program, to develop validated Computational Fluid Dynamics (CFD) concepts, codes and analyses for seals. The effort will provide NASA and the U.S. Aerospace Industry with advanced CFD scientific codes and industrial codes for analyzing and designing turbomachinery seals. An advanced 3D CFD cylindrical seal code has been developed, incorporating state-of-the-art computational methodology for flow analysis in straight, tapered and stepped seals. Relevant computational features of the code include: stationary/rotating coordinates, cylindrical and general Body Fitted Coordinates (BFC) systems, high order differencing schemes, colocated variable arrangement, advanced turbulence models, incompressible/compressible flows, and moving grids. This paper presents the current status of code development, code demonstration for predicting rotordynamic coefficients, numerical parametric study of entrance loss coefficients for generic annular seals, and plans for code extensions to labyrinth, damping, and other seal configurations.

  14. 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)

  15. 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)

  16. 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)

  17. 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...

  18. Validation and evaluation of the advanced aeronautical CFD system SAUNA: A method developer's view

    Science.gov (United States)

    Shaw, J. A.; Peace, A. J.; Georgala, J. M.; Childs, P. N.

    1993-09-01

    This paper is concerned with a detailed validation and evaluation of the SAUNA CFD system for complex aircraft configurations. The methodology of the complete system is described in brief, including its unique use of differing grid generation strategies (structured, unstructured or both) depending on the geometric complexity of the configuration. A wide range of configurations and flow conditions are chosen in the validation and evaluation exercise to demonstrate the scope of SAUNA. A detailed description of the results from the method is preceded by a discussion on the philosophy behind the strategy followed in the exercise, in terms of equality assessment and the differing roles of the code developer and the code user. It is considered that SAUNA has grown into a highly usable tool for the aircraft designer, in combining flexibility and accuracy in an efficient manner.

  19. 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)

  20. 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)

  1. 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)

  2. Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

    Energy Technology Data Exchange (ETDEWEB)

    Delchini, Marc-Olivier [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Popov, Emilian L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-04-29

    This report presents the state of advancement of a Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the computational fluid dynamics (CFD) code Nek5000 using the Dakota package for flows encountered in the nuclear engineering industry. Nek5000 is a high order spectral element CFD code developed at Argonne National Laboratory for high resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds averaged Navier-Stokes (URANS) simulations.

  3. Validation of a loss of vacuum accident (LOVA) Computational Fluid Dynamics (CFD) model

    Energy Technology Data Exchange (ETDEWEB)

    Bellecci, C.; Gaudio, P. [EURATOM-Faculty of Engineering, University of Rome ' Tor Vergata' Via del Politecnico 1, 00133 Rome (Italy); Lupelli, I., E-mail: ivan.lupelli@uniroma2.it [EURATOM-Faculty of Engineering, University of Rome ' Tor Vergata' Via del Politecnico 1, 00133 Rome (Italy); Malizia, A. [EURATOM-Faculty of Engineering, University of Rome ' Tor Vergata' Via del Politecnico 1, 00133 Rome (Italy); Porfiri, M.T. [ENEA Nuclear Fusion Tecnologies, Via Enrico Fermi, 45 I-00044 Frascati (Italy); Quaranta, R.; Richetta, M. [EURATOM-Faculty of Engineering, University of Rome ' Tor Vergata' Via del Politecnico 1, 00133 Rome (Italy)

    2011-10-15

    Intense thermal loads in fusion devices occur during plasma disruptions, Edge Localized Modes (ELM) and Vertical Displacement Events (VDE). They will result in macroscopic erosion of the plasma facing materials and consequent accumulation of activated dust into the ITER Vacuum Vessel (VV). A recognized safety issue for future fusion reactors fueled with deuterium and tritium is the generation of sizeable quantities of dust. In case of LOVA, air inlet occurs due to the pressure difference between the atmospheric condition and the internal condition. It causes mobilization of the dust that can exit the VV threatening public safety because it may contain tritium, may be radioactive from activation products, and may be chemically reactive and/or toxic (Sharpe et al.; Sharpe and Humrickhouse). Several experiments have been conducted with STARDUST facility in order to reproduce a low pressurization rate (300 Pa/s) LOVA event in ITER due to a small air leakage for two different positions of the leak, at the equatorial port level and at the divertor port level, in order to evaluate the velocity magnitude in case of a LOVA that is strictly connected with dust mobilization phenomena. A two-dimensional (2D) modelling of STARDUST, made with the CFD commercial code FLUENT, has been carried out. The results of these simulations were compared against the experimental data for CFD code validation. For validation purposes, the CFD simulation data were extracted at the same locations as the experimental data were collected. In this paper, the authors present and discuss the computer-simulation data and compare them with data collected during the laboratory studies at the University of Rome 'Tor Vergata' Quantum Electronics and Plasmas lab.

  4. 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...

  5. 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.

  6. Evaluation of heat transfer surfaces for compact recuperator using a CFD code

    Science.gov (United States)

    Ashok Babu, T. P.; Talekala, Mohammad Shekoor

    2009-04-01

    Exhaust recovery recuperator is mandatory in order to realize a thermal efficiency of 30% or higher for micro turbines. In this work an attempt is made to select the cross corrugated heat transfer surface with minimum core volume of a recuperator matrix using a CFD code. Analysis is carried out for selected cross corrugated heat transfer surface configurations. The relation between the minimum core volume from design calculation and average skin friction coefficient from CFD analysis has been established.

  7. 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.

  8. 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.)

  9. 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)

  10. 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. ‎

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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.

  20. 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.

  1. 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 ...

  2. 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 Schmidt;

    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....... Although it is not necessary to maintain the same format of reporting the CFD modeling as presented in this paper, the authors would suggest including all the information related to the selection of turbulence models, difference schemes, convergence criteria, boundary conditions, geometry simplification......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...

  3. 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

  4. 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.

  5. 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 ...

  6. 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

  7. 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

  8. VOLVOF: An update of the CFD code, SOLA-VOF

    Energy Technology Data Exchange (ETDEWEB)

    Park, J.E.

    1999-12-14

    The SOLA-VOF code developed by the T-3 (Theoretical Physics, Fluids) group at the Los Alamos National Laboratory (LANL) has been extensively modified at the Oak Ridge National Laboratory (ORNL). The modified and improved version has been dubbed ''VOLVOF,'' to acknowledge the state of Tennessee, the Volunteer state and home of ORNL. Modifications include generalization of boundary conditions, additional flexibility in setting up problems, addition of a problem interruption and restart capability, segregation of graphics functions to allow utilization of modern commercial graphics programs, and addition of time and date stamps to output files. Also, the pressure iteration has been restructured to exploit the much greater system memory available on modern workstations and personal computers. A solution monitoring capability has been added to utilize the multi-tasking capability of modern computer operating systems. These changes are documented in the following report. NAMELIST input variables are defined, and input files and the resulting output are given for two test problems. Modification and documentation of a working technical computer program is almost never complete. This is certainly true for the present effort. However, the impending retirement of the writer dictates that the current configuration and capability be reported.

  9. 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.

  10. 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)

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  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 simula-tions 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, imagesof these flowfields were constructed from numerical results based on the principle of the optical flowvisualization, and compared directly with experimental interferograms. Because both experimental and numerical results axe of identical physical representation, the agreement between them can be evaluatedeffectively 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. 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.

  17. 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

  18. 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.

  19. 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....

  20. 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...

  1. 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)

  2. 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)

  3. 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

    of full-scale experiments in a room ventilated by the mixing principle have been performed for validation of the models. The experimental results include measurements of temperature as well as measurements of velocity and turbulence by Laser Doppler Anemometry (LDA). A simple model of the complex inlet...

  4. 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...

  5. 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

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. Experience with Aero- and Fluid-Dynamic Testing for Engineering and CFD Validation

    Science.gov (United States)

    Ross, James C.

    2016-01-01

    Ever since computations have been used to simulate aerodynamics the need to ensure that the computations adequately represent real life has followed. Many experiments have been performed specifically for validation and as computational methods have improved, so have the validation experiments. Validation is also a moving target because computational methods improve requiring validation for the new aspect of flow physics that the computations aim to capture. Concurrently, new measurement techniques are being developed that can help capture more detailed flow features pressure sensitive paint (PSP) and particle image velocimetry (PIV) come to mind. This paper will present various wind-tunnel tests the author has been involved with and how they were used for validation of various kinds of CFD. A particular focus is the application of advanced measurement techniques to flow fields (and geometries) that had proven to be difficult to predict computationally. Many of these difficult flow problems arose from engineering and development problems that needed to be solved for a particular vehicle or research program. In some cases the experiments required to solve the engineering problems were refined to provide valuable CFD validation data in addition to the primary engineering data. All of these experiments have provided physical insight and validation data for a wide range of aerodynamic and acoustic phenomena for vehicles ranging from tractor-trailers to crewed spacecraft.

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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.

  16. 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.

  17. Modeling HCCI using CFD and Detailed Chemistry with Experimental Validation and a Focus on CO Emissions

    Energy Technology Data Exchange (ETDEWEB)

    Hessel, R; Foster, D; Aceves, S; Flowers, D; Pitz, B; Dec, J; Sjoberg, M; Babajimopoulos, A

    2007-04-23

    Multi-zone CFD simulations with detailed kinetics were used to model engine experiments performed on a diesel engine that was converted for single cylinder, HCCI operation, here using iso-octane as the fuel. The modeling goals were to validate the method (multi-zone combustion modeling) and the reaction mechanism (LLNL 857 species iso-octane), both of which performed very well. The purpose of this paper is to document the validation findings and to set the ground work for further analysis of the results by first looking at CO emissions characteristics with varying equivalence ratio.

  18. 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

  19. 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

  20. 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.

  1. Prediction of flow in mix-proof valve by use of CFD - Validation by LDA

    DEFF Research Database (Denmark)

    Jensen, Bo Boye Busk; Friis, Alan

    2004-01-01

    was done on a spherical shaped mix-proof valve (MPV). Flow were predicted by Computational Fluid Dynamics (CFD) and validated by data obtained from experiments using laser sheet visualization and laser Doppler anemometry. Correction of the measured velocities and probe location was required as refraction...... of laser beams through the curved surfaces of the valve house changes both intersection angle and intersection location. CFD simulations were performed by use of both the standard wall function and the two-layer model of Norris and Reynolds for near-wall description. The importance of resolving the near......-wall region is shown. Fully 3D flow patterns were identified and valuable information was obtained for further investigations concerning prediction of cleanability in the MPV based on knowledge of the hydrodynamics herein....

  2. Overview of NASA Multi-Dimensional Stirling Convertor Code Development and Validation Effort

    Science.gov (United States)

    Tew, Roy C.; Cairelli, James E.; Ibrahim, Mounir B.; Simon, Terrence W.; Gedeon, David

    2003-01-01

    A NASA grant has been awarded to Cleveland State University (CSU) to develop a multi-dimensional (multi-D) Stirling computer code with the goals of improving loss predictions and identifying component areas for improvements. The University of Minnesota (UMN) and Gedeon Associates are teamed with CSU. Development of test rigs at UMN and CSU and validation of the code against test data are part of the effort. The one-dimensional (1-D) Stirling codes used for design and performance prediction do not rigorously model regions of the working space where abrupt changes in flow area occur (such as manifolds and other transitions between components). Certain hardware experiences have demonstrated large performance gains by varying manifolds and heat exchanger designs to improve flow distributions in the heat exchangers. 1-D codes were not able to predict these performance gains. An accurate multi-D code should improve understanding of the effects of area changes along the main flow axis, sensitivity of performance to slight changes in internal geometry, and, in general, the understanding of various internal thermodynamic losses. The commercial CFD-ACE code has been chosen for development of the multi-D code. This 2-D/3-D code has highly developed pre- and post-processors, and moving boundary capability. Preliminary attempts at validation of CFD-ACE models of MIT gas spring and ``two space'' test rigs were encouraging. Also, CSU's simulations of the UMN oscillating-flow rig compare well with flow visualization results from UMN. A complementary Department of Energy (DOE) Regenerator Research effort is aiding in development of regenerator matrix models that will be used in the multi-D Stirling code. This paper reports on the progress and challenges of this multi-D code development effort.

  3. 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.

  4. Construction of TH code development and validation environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyungjun; Kim, Hee-Kyung; Bae, Kyoo-Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, each component of code development and validation system, i.e. IVS and Mercurial will be introduced and Redmine, the integrated platform of IVS and Mercurial, will be explained later. Integrated TH code validation system, IVS and code development and management environment are constructed. The code validation could be achieved by a comparison of results with corresponding experiments. The development of thermal-hydraulic (TH) system code for nuclear reactor requires much time and effort, also for its validation and verification(V and V). In previous, TASS/SMR-S code (hereafter TASS) for SMART is developed by KAERI through V and V process. On the way of code development, the version control of source code has great importance. Also, during the V and V process, the way to reduce repeated labor- and time-consuming work of running the code before releasing new version of TH code, is required. Therefore, the integrated platform for TH code development and validation environment is constructed. Finally, Redmine, the project management and issue tracking system, is selected as platform, Mercurial (hg) for source version control and IVS (Integrated Validation System) for TASS is constructed as a prototype for automated V and V. IVS is useful before release a new code version. The code developer can validate code result easily using IVS. Even during code development, IVS could be used for validation of code modification. Using Redmine and Mercurial, users and developers can use IVS result more effectively.

  5. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    OpenAIRE

    Kim, Hyoung Tae; Chang, Se-Myong; Shin, Jong-Hyeon; Kim, Yong Gwon

    2016-01-01

    The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchm...

  6. 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

  7. PIV Measurements of the CEV Hot Abort Motor Plume for CFD Validation

    Science.gov (United States)

    Wernet, Mark; Wolter, John D.; Locke, Randy; Wroblewski, Adam; Childs, Robert; Nelson, Andrea

    2010-01-01

    NASA s next manned launch platform for missions to the moon and Mars are the Orion and Ares systems. Many critical aspects of the launch system performance are being verified using computational fluid dynamics (CFD) predictions. The Orion Launch Abort Vehicle (LAV) consists of a tower mounted tractor rocket tasked with carrying the Crew Module (CM) safely away from the launch vehicle in the event of a catastrophic failure during the vehicle s ascent. Some of the predictions involving the launch abort system flow fields produced conflicting results, which required further investigation through ground test experiments. Ground tests were performed to acquire data from a hot supersonic jet in cross-flow for the purpose of validating CFD turbulence modeling relevant to the Orion Launch Abort Vehicle (LAV). Both 2-component axial plane Particle Image Velocimetry (PIV) and 3-component cross-stream Stereo Particle Image Velocimetry (SPIV) measurements were obtained on a model of an Abort Motor (AM). Actual flight conditions could not be simulated on the ground, so the highest temperature and pressure conditions that could be safely used in the test facility (nozzle pressure ratio 28.5 and a nozzle temperature ratio of 3) were used for the validation tests. These conditions are significantly different from those of the flight vehicle, but were sufficiently high enough to begin addressing turbulence modeling issues that predicated the need for the validation tests.

  8. Loss of vacuum accident (LOVA): Comparison of computational fluid dynamics (CFD) flow velocities against experimental data for the model validation

    Energy Technology Data Exchange (ETDEWEB)

    Bellecci, C.; Gaudio, P.; Lupelli, I. [Faculty of Engineering, University of Rome ' Tor Vergata' , Via del Politecnico 1, 00133 Rome (Italy); Malizia, A., E-mail: malizia@ing.uniroma2.it [Faculty of Engineering, University of Rome ' Tor Vergata' , Via del Politecnico 1, 00133 Rome (Italy); Porfiri, M.T. [ENEA Nuclear Fusion Technologies, Via Enrico Fermi 45 I, 00044, Frascati (Italy); Quaranta, R.; Richetta, M. [Faculty of Engineering, University of Rome ' Tor Vergata' , Via del Politecnico 1, 00133 Rome (Italy)

    2011-06-15

    A recognized safety issue for future fusion reactors fueled with deuterium and tritium is the generation of sizeable quantities of dust. Several mechanisms resulting from material response to plasma bombardment in normal and off-normal conditions are responsible for generating dust of micron and sub-micron length scales inside the VV (Vacuum Vessel) of experimental fusion facilities. The loss of coolant accidents (LOCA), loss of coolant flow accidents (LOFA) and loss of vacuum accidents (LOVA) are types of accidents, expected in experimental fusion reactors like ITER, that may jeopardize components and plasma vessel integrity and cause dust mobilization risky for workers and public. The air velocity is the driven parameter for dust resuspension and its characterization, in the very first phase of the accidents, is critical for the dust release. To study the air velocity trend a small facility, Small Tank for Aerosol Removal and Dust (STARDUST), was set up at the University of Rome 'Tor Vergata', in collaboration with ENEA Frascati laboratories. It simulates a low pressurization rate (300 Pa/s) LOVA event in ITER due to a small air inlet from two different positions of the leak: at the equatorial port level and at the divertor port level. The velocity magnitude in STARDUST was investigated in order to map the velocity field by means of a punctual capacitive transducer placed inside STARDUST without obstacles. FLUENT was used to simulate the flow behavior for the same LOVA scenarios used during the experimental tests. The results of these simulations were compared against the experimental data for CFD code validation. For validation purposes, the CFD simulation data were extracted at the same locations as the experimental data were collected for the first four seconds, because at the beginning of the experiments the maximum velocity values (that could cause the almost complete dust mobilization) have been measured. In this paper the authors present and

  9. Verification and Validation of Kinetic Codes

    Science.gov (United States)

    Christlieb, Andrew

    2014-10-01

    We review the last three workshops held on Validation and Verification of Kinetic Codes. The goal of the workshops was to highlight the need to develop benchmark test problems beyond traditional test problems such as Landau damping and the two-stream instability. These test problems provide a limited understanding how a code might perform and mask key issues in more complicated situations. Developing these test problems highlights the strengths and weaknesses of both mesh- and particle-based codes. One outcome is that designing test problems that clearly deliver a path forward for developing improved methods is complicated by the need to create a completely self-consistent model. For example, two test cases proposed by the authors as simple test cases turn out to be ill defined. The first case is the modeling of sheath formation in a 1D 1V collisionless plasma. We found that losses to the wall lead to discontinuous distribution functions, a challenge for high order mesh-based solvers. The semi-infinite case was problematic because the far field boundary condition poses difficulty in computing on a finite domain. Our second case was flow of a collisionless electron beam in a pipe. Here, numerical diffusion is a key problem we are testing; however, two-stream instability at the beam edges introduces other issues in terms of finding convergent solutions. For mesh-based codes, before particle trapping takes place, mesh-based methods find themselves outside of the asymptotic regime. Another conclusion we draw from this exercise is that including collisional models in benchmark test problems for mesh-based plasma simulation tools is an important step in providing robust test problems for mesh-based kinetic solvers. In collaboration with Yaman Guclu, David Seal, and John Verboncoeur, Michigan State University.

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

    Directory of Open Access Journals (Sweden)

    Fabio Moretti

    2008-11-01

    Full Text Available 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, and a stratified steam-water flow. Some standard and new implemented models were tested: modified turbulent k-ε models with turbulence production induced by interfacial friction, models for the drag coefficient, and interfacial heat transfer models. Quite good agreement with experimental data was achieved with best performing models for both test cases, even if a further improvement in phase change modelling would be suitable for nuclear technology applications.

  11. A study on the dependency between turbulent models and mesh configurations of CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Jungjin; Heo, Yujin; Jerng, Dong-Wook [CAU, Seoul (Korea, Republic of)

    2015-10-15

    This paper focuses on the analysis of the behavior of hydrogen mixing and hydrogen stratification, using the GOTHIC code and the CFD code. Specifically, we examined the mesh sensitivity and how the turbulence model affects hydrogen stratification or hydrogen mixing, depending on the mesh configuration. In this work, sensitivity analyses for the meshes and the turbulence models were conducted for missing and stratification phenomena. During severe accidents in a nuclear power plants, the generation of hydrogen may occur and this will complicate the atmospheric condition of the containment by causing stratification of air, steam, and hydrogen. This could significantly impact containment integrity analyses, as hydrogen could be accumulated in local region. From this need arises the importance of research about stratification of gases in the containment. Two computation fluid dynamics code, i.e. GOTHIC and STAR-CCM+ were adopted and the computational results were benchmarked against the experimental data from PANDA facility. The main findings observed through the present work can be summarized as follows: 1) In the case of the GOTHIC code, it was observed that the aspect ratio of the mesh was found more important than the mesh size. Also, if the number of the mesh is over 3,000, the effects of the turbulence models were marginal. 2) For STAR-CCM+, the tendency is quite different from the GOTHIC code. That is, the effects of the turbulence models were small for fewer number of the mesh, however, as the number of mesh increases, the effects of the turbulence models becomes significant. Another observation is that away from the injection orifice, the role of the turbulence models tended to be important due to the nature of mixing process and inducted jet stream.

  12. 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.

  13. 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)

  14. 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.

  15. Investigation of the validity of BEM for simulation of wind turbines in complex load cases and comparison with experiment and CFD

    Science.gov (United States)

    Rahimi, H.; Dose, B.; Stoevesandt, B.; Peinke, J.

    2016-09-01

    The aim of this work is to investigate the validity of simulation codes based on the Blade Element Momentum (BEM) theory for three important design load conditions. This paper includes the cases of yawed inflow, rotor tower interaction for downwind turbines and the standstill case. Computational Fluid Dynamics (CFD) and experimental data (when available) are used for the evaluation of the obtained results. For the yawed inflow, the results indicate that significant deviations between BEM and experiments & CFD can be observed. This discrepancy is caused by unsteady phenomena such as the advancing & retreating blade effect and the skewed wake effect. In the case of the rotor and tower interaction of the downwind turbine, the results show that the BEM based code overpredicts the sectional forces in terms of the normal and tangential forces by 20%. In the case of standstill, the evaluation of the results based on tip deflections shows clear differences in the output of both numerical approaches. While the flapwise deflections show a reasonable agreement, the CFD-based coupled solver predicts much larger edgewise vibrations.

  16. 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.

  17. 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)

  18. 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)

  19. 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.

  20. 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.

  1. 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

  2. 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

  3. 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.

  4. 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)

  5. CFD Recombiner Modelling and Validation on the H2-Par and Kali-H2 Experiments

    Directory of Open Access Journals (Sweden)

    Stéphane Mimouni

    2011-01-01

    Full Text Available A large amount of Hydrogen gas is expected to be released within the dry containment of a pressurized water reactor (PWR, shortly after the hypothetical beginning of a severe accident leading to the melting of the core. According to local gas concentrations, the gaseous mixture of hydrogen, air and steam can reach the flammability limit, threatening the containment integrity. In order to prevent mechanical loads resulting from a possible conflagration of the gas mixture, French and German reactor containments are equipped with passive autocatalytic recombiners (PARs which preventively oxidize hydrogen for concentrations lower than that of the flammability limit. The objective of the paper is to present numerical assessments of the recombiner models implemented in CFD solvers NEPTUNE_CFD and Code_Saturne. Under the EDF/EPRI agreement, CEA has been committed to perform 42 tests of PARs. The experimental program named KALI-H2, consists checking the performance and behaviour of PAR. Unrealistic values for the gas temperature are calculated if the conjugate heat transfer and the wall steam condensation are not taken into account. The combined effects of these models give a good agreement between computational results and experimental data.

  6. 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.

  7. 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)

  8. 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

  9. 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...

  10. 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

  11. 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

  12. 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.......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...

  13. 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

  14. 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

  15. 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....

  16. 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.

  17. 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)

  18. 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

  19. 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)

  20. CFD analysis of turbopump volutes

    Science.gov (United States)

    Ascoli, Edward P.; Chan, Daniel C.; Darian, Armen; Hsu, Wayne W.; Tran, Ken

    1993-07-01

    An effort is underway to develop a procedure for the regular use of CFD analysis in the design of turbopump volutes. Airflow data to be taken at NASA Marshall will be used to validate the CFD code and overall procedure. Initial focus has been on preprocessing (geometry creation, translation, and grid generation). Volute geometries have been acquired electronically and imported into the CATIA CAD system and RAGGS (Rockwell Automated Grid Generation System) via the IGES standard. An initial grid topology has been identified and grids have been constructed for turbine inlet and discharge volutes. For CFD analysis of volutes to be used regularly, a procedure must be defined to meet engineering design needs in a timely manner. Thus, a compromise must be established between making geometric approximations, the selection of grid topologies, and possible CFD code enhancements. While the initial grid developed approximated the volute tongue with a zero thickness, final computations should more accurately account for the geometry in this region. Additionally, grid topologies will be explored to minimize skewness and high aspect ratio cells that can affect solution accuracy and slow code convergence. Finally, as appropriate, code modifications will be made to allow for new grid topologies in an effort to expedite the overall CFD analysis process.

  1. 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。

  2. 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)

  3. 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

  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. 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)

  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. 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

  8. 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

  9. 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.

  10. 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.

  11. Criticality benchmarks validation of the Monte Carlo code TRIPOLI-2

    Energy Technology Data Exchange (ETDEWEB)

    Maubert, L. (Commissariat a l' Energie Atomique, Inst. de Protection et de Surete Nucleaire, Service d' Etudes de Criticite, 92 - Fontenay-aux-Roses (France)); Nouri, A. (Commissariat a l' Energie Atomique, Inst. de Protection et de Surete Nucleaire, Service d' Etudes de Criticite, 92 - Fontenay-aux-Roses (France)); Vergnaud, T. (Commissariat a l' Energie Atomique, Direction des Reacteurs Nucleaires, Service d' Etudes des Reacteurs et de Mathematique Appliquees, 91 - Gif-sur-Yvette (France))

    1993-04-01

    The three-dimensional energy pointwise Monte-Carlo code TRIPOLI-2 includes metallic spheres of uranium and plutonium, nitrate plutonium solutions, square and triangular pitch assemblies of uranium oxide. Results show good agreements between experiments and calculations, and avoid a part of the code and its ENDF-B4 library validation. (orig./DG)

  12. The Facial Expression Coding System (FACES): Development, Validation, and Utility

    Science.gov (United States)

    Kring, Ann M.; Sloan, Denise M.

    2007-01-01

    This article presents information on the development and validation of the Facial Expression Coding System (FACES; A. M. Kring & D. Sloan, 1991). Grounded in a dimensional model of emotion, FACES provides information on the valence (positive, negative) of facial expressive behavior. In 5 studies, reliability and validity data from 13 diverse…

  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. 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.

  15. 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

  16. 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)

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

    Directory of Open Access Journals (Sweden)

    Xiaotie Zhang

    Full Text Available 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.

  18. 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

  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. 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

  1. 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)

  2. 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

  3. 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.

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Caron-Charles, M.; Blumenfeld, L. [CEA Saclay, 91 - Gif sur Yvette (France)

    2001-07-01

    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)

  13. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2016-01-01

    Full Text Available The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor, has been modeled in multidimension for the computation based on CFD (computational fluid dynamics technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors.

  14. 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.

  15. 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.

  16. 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.

  17. 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...

  18. 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)

  19. 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)

  20. 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...

  1. 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

  2. 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

  3. 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.

  4. Eu-NORSEWInD - Assessment of Viability of Open Source CFD Code for the Wind Industry

    DEFF Research Database (Denmark)

    Stickland, Matt; Scanlon, Tom; Fabre, Sylvie;

    2009-01-01

    . The cost of the extra licences can become the limit on the final number of nodes employed. Whilst there are significant benefits to be found when using a commercial code which has a user friendly interface and has undergone significant verification testing the financial advantages of using an open source...... However; the extent to which the flow field above the various mounting platforms would be distorted was unknown. Therefore, part of the fundamental research incorporated into the NORSEWInD project was a computational and experimental investigation into the possible scale and extent of the interference...... between the results of simulations created by the commercial code FLUENT and the open source code OpenFOAM. An assessment of the ease with which the open source code can be used is also included....

  5. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Farah, A.; Haines, P.; Harvel, G.; Pioro, I., E-mail: amjad.farah@yahoo.com, E-mail: patrickjhaines@gmail.com, E-mail: glenn.harvel@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science,Oshawa, Ontario (Canada)

    2012-07-01

    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)

  7. 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)

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. The Initial Atmospheric Transport (IAT) Code: Description and Validation

    Energy Technology Data Exchange (ETDEWEB)

    Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-10-01

    The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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, ...

  3. 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.

  4. 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.)

  5. Benchmark Solutions for Computational Aeroacoustics (CAA) Code Validation

    Science.gov (United States)

    Scott, James R.

    2004-01-01

    NASA has conducted a series of Computational Aeroacoustics (CAA) Workshops on Benchmark Problems to develop a set of realistic CAA problems that can be used for code validation. In the Third (1999) and Fourth (2003) Workshops, the single airfoil gust response problem, with real geometry effects, was included as one of the benchmark problems. Respondents were asked to calculate the airfoil RMS pressure and far-field acoustic intensity for different airfoil geometries and a wide range of gust frequencies. This paper presents the validated that have been obtained to the benchmark problem, and in addition, compares them with classical flat plate results. It is seen that airfoil geometry has a strong effect on the airfoil unsteady pressure, and a significant effect on the far-field acoustic intensity. Those parts of the benchmark problem that have not yet been adequately solved are identified and presented as a challenge to the CAA research community.

  6. CFD and Ventilation Research

    DEFF Research Database (Denmark)

    Li, Y.; Nielsen, Peter V.

    2011-01-01

    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 theoretical......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...

  7. 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.

  8. 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.

  9. 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...

  10. 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

  11. 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

  12. 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.

  13. Issues in computational fluid dynamics code verification and validation

    Energy Technology Data Exchange (ETDEWEB)

    Oberkampf, W.L.; Blottner, F.G.

    1997-09-01

    A broad range of mathematical modeling errors of fluid flow physics and numerical approximation errors are addressed in computational fluid dynamics (CFD). It is strongly believed that if CFD is to have a major impact on the design of engineering hardware and flight systems, the level of confidence in complex simulations must substantially improve. To better understand the present limitations of CFD simulations, a wide variety of physical modeling, discretization, and solution errors are identified and discussed. Here, discretization and solution errors refer to all errors caused by conversion of the original partial differential, or integral, conservation equations representing the physical process, to algebraic equations and their solution on a computer. The impact of boundary conditions on the solution of the partial differential equations and their discrete representation will also be discussed. Throughout the article, clear distinctions are made between the analytical mathematical models of fluid dynamics and the numerical models. Lax`s Equivalence Theorem and its frailties in practical CFD solutions are pointed out. Distinctions are also made between the existence and uniqueness of solutions to the partial differential equations as opposed to the discrete equations. Two techniques are briefly discussed for the detection and quantification of certain types of discretization and grid resolution errors.

  14. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

    Pla, P.; Reventos, F.; Pretel, C. [Catalonia Technical Univ., Barcelona (Spain); Pla, P.; Giannotti, W.; D' Auria, F. [Pisa Univ. (Italy); Annunziato, A. [Joint Research Centre of the European Commission, Ispra (Italy); Sol, I. [Associacio Nuclear Asco-Vandellos 2, L' Hospitalet de l' Infant (Tarragona)(Spain)

    2007-07-01

    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.

  16. 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.

  17. 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...

  18. 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

  19. VERIFICATION OF SMOM AND QMOM POPULATION BALANCE MODELING IN CFD CODE USING ANALYTICAL SOLUTIONS FOR BATCH PARTICULATE PROCESSES

    Institute of Scientific and Technical Information of China (English)

    Bin Wan; Terry A.Ring

    2006-01-01

    For many processes of industrial significance, due to the strong coupling between particle interactions and fluid dynamics, the population balance must be solved as part of a computational fluid dynamics (CFD) simulation. In this work, a CFD based population balance model is tested using a batch crystallization reactor. In this CFD model, the population balance is solved by the standard method of moments (SMOM) and the quadrature method of moments (QMOM). The results of these simulations are compared to analytical solutions for the population balance in a batch tank where 1) nucleation, 2) growth, 3) aggregation, and 4) breakage are taking place separately. The results of these comparisons show that the first 6 moments of the population balance are accurately predicted for nucleation, growth, aggregation and breakage at all times.

  20. 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

  1. 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.

  2. 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

  3. ClinicalCodes: an online clinical codes repository to improve the validity and reproducibility of research using electronic medical records.

    Directory of Open Access Journals (Sweden)

    David A Springate

    Full Text Available Lists of clinical codes are the foundation for research undertaken using electronic medical records (EMRs. If clinical code lists are not available, reviewers are unable to determine the validity of research, full study replication is impossible, researchers are unable to make effective comparisons between studies, and the construction of new code lists is subject to much duplication of effort. Despite this, the publication of clinical codes is rarely if ever a requirement for obtaining grants, validating protocols, or publishing research. In a representative sample of 450 EMR primary research articles indexed on PubMed, we found that only 19 (5.1% were accompanied by a full set of published clinical codes and 32 (8.6% stated that code lists were available on request. To help address these problems, we have built an online repository where researchers using EMRs can upload and download lists of clinical codes. The repository will enable clinical researchers to better validate EMR studies, build on previous code lists and compare disease definitions across studies. It will also assist health informaticians in replicating database studies, tracking changes in disease definitions or clinical coding practice through time and sharing clinical code information across platforms and data sources as research objects.

  4. 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...

  5. 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)

  6. 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

  7. Verification and Validation of MERCURY: A Modern, Monte Carlo Particle Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Procassini, R J; Cullen, D E; Greenman, G M; Hagmann, C A

    2004-12-09

    Verification and Validation (V&V) is a critical phase in the development cycle of any scientific code. The aim of the V&V process is to determine whether or not the code fulfills and complies with the requirements that were defined prior to the start of the development process. While code V&V can take many forms, this paper concentrates on validation of the results obtained from a modern code against those produced by a validated, legacy code. In particular, the neutron transport capabilities of the modern Monte Carlo code MERCURY are validated against those in the legacy Monte Carlo code TART. The results from each code are compared for a series of basic transport and criticality calculations which are designed to check a variety of code modules. These include the definition of the problem geometry, particle tracking, collisional kinematics, sampling of secondary particle distributions, and nuclear data. The metrics that form the basis for comparison of the codes include both integral quantities and particle spectra. The use of integral results, such as eigenvalues obtained from criticality calculations, is shown to be necessary, but not sufficient, for a comprehensive validation of the code. This process has uncovered problems in both the transport code and the nuclear data processing codes which have since been rectified.

  8. 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)

  9. 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'.

  10. 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'

  11. 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

  12. 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.

  13. 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)

  14. 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

  15. 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.

  16. 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)

  17. 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)

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. 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.)

  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. 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....

  6. 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).

  7. 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)

  8. 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

  9. 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.

  10. 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...

  11. 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.

  12. Validation of daylighting model in CODYRUN building simulation code

    OpenAIRE

    Boyer, Harry; Boyer, H.; Guichard, Stéphane; Guichard, S; Jean, Aurélien; Jean, A.; Libelle, Teddy; Libelle, T; Bigot, Dimitri; Miranville, F; Miranville, Frédéric; Bojić, M.

    2015-01-01

    International audience CODYRUN is a multi-zone software integrating thermal building simulation, airflow, and pollutant transfer. A first question thus arose as to the integration of indoor lighting conditions into the simulation, leading to a new model calculating natural and artificial lighting. The results of this new daylighting module were then compared with results of other simulation codes and experimental cases both in artificial and natural environments. Excellent agreements were ...

  13. 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.

  14. Automated facial coding: validation of basic emotions and FACS AUs in FaceReader

    NARCIS (Netherlands)

    P. Lewinski; T.M. den Uyl; C. Butler

    2014-01-01

    In this study, we validated automated facial coding (AFC) software—FaceReader (Noldus, 2014)—on 2 publicly available and objective datasets of human expressions of basic emotions. We present the matching scores (accuracy) for recognition of facial expressions and the Facial Action Coding System (FAC

  15. 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.

  16. 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)

  17. 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.

  18. 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%.

  19. 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%.

  20. 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.

  1. 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.

  2. 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.)

  3. 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

  4. 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.

  5. 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)

  6. 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

  7. 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

  8. Benchmark and partial validation testing of the FLASH computer code, Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Martian, P.; Smith, C.S.

    1993-09-01

    This document presents methods and results of benchmark testing (i.e., code-to-code comparisons) and partial validation testing (i.e., tests which compare field data to the computer generated solutions) of the FLASH computer code, Version 3.0, which were conducted to determine if the code is ready for performance assessment studies of the Radioactive Waste Management Complex. Three test problems are presented that were designed to check computational efficiency, accuracy of the numerical algorithms, and the capability of the code to simulate diverse hydrological conditions. These test problems were designed to specifically test the code`s ability to simulate, (a) seasonal infiltration in response to meteorological conditions, (b) changing watertable elevations due to a transient areal source of water, (i.e., influx from spreading basins), and (c) infiltration into fractured basalt as a result of seasonal water in drainage ditches. The FLASH simulations generally compared well with the benchmark codes, indicating good stability and acceptable computational efficiency while simulating a wide range of conditions. The code appears operational for modeling both unsaturated and saturated flow in fractured, heterogeneous porous media. However, the code failed to converge when a unsaturated to saturated transition occurred. Consequently, the code should not be used when this condition occurs or is expected to occur, i.e. when perched water is present or when infiltration rates exceed the saturated conductivity of the soil.

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. 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.

  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. 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.

  16. 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.

  17. 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.

  18. 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...

  19. 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

  20. 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)

  1. 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)

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

    Directory of Open Access Journals (Sweden)

    Uwe Imke

    2012-01-01

    Full Text Available 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.

  3. Automated face analysis by feature point tracking has high concurrent validity with manual FACS coding.

    Science.gov (United States)

    Cohn, J F; Zlochower, A J; Lien, J; Kanade, T

    1999-01-01

    The face is a rich source of information about human behavior. Available methods for coding facial displays, however, are human-observer dependent, labor intensive, and difficult to standardize. To enable rigorous and efficient quantitative measurement of facial displays, we have developed an automated method of facial display analysis. In this report, we compare the results with this automated system with those of manual FACS (Facial Action Coding System, Ekman & Friesen, 1978a) coding. One hundred university students were videotaped while performing a series of facial displays. The image sequences were coded from videotape by certified FACS coders. Fifteen action units and action unit combinations that occurred a minimum of 25 times were selected for automated analysis. Facial features were automatically tracked in digitized image sequences using a hierarchical algorithm for estimating optical flow. The measurements were normalized for variation in position, orientation, and scale. The image sequences were randomly divided into a training set and a cross-validation set, and discriminant function analyses were conducted on the feature point measurements. In the training set, average agreement with manual FACS coding was 92% or higher for action units in the brow, eye, and mouth regions. In the cross-validation set, average agreement was 91%, 88%, and 81% for action units in the brow, eye, and mouth regions, respectively. Automated face analysis by feature point tracking demonstrated high concurrent validity with manual FACS coding.

  4. Some Examples of the Application and Validation of the NUFT Subsurface Flow and Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Nitao, J J

    2001-08-01

    This report was written as partial fulfillment of a subcontract from DOD/DOE Strategic Environmental Research and Development Program (SERDP) as part of a project directed by the U.S. Army Engineer Research and Development Center, Waterways Experiment Station (WES), Vicksburg, Mississippi. The report documents examples of field validation of the Non-isothermal Unsaturated-saturated Flow and Transport model (NUFT) code for environmental remediation, with emphasis on soil vapor extraction, and describes some of the modifications needed to integrate the code into the DOD Groundwater Modeling System (GMS, 2000). Note that this report highlights only a subset of the full capabilities of the NUFT code.

  5. 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)

  6. 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.

  7. 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

  8. 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 ...

  9. 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

  10. The Crucial Role of Error Correlation for Uncertainty Modeling of CFD-Based Aerodynamics Increments

    Science.gov (United States)

    Hemsch, Michael J.; Walker, Eric L.

    2011-01-01

    The Ares I ascent aerodynamics database for Design Cycle 3 (DAC-3) was built from wind-tunnel test results and CFD solutions. The wind tunnel results were used to build the baseline response surfaces for wind-tunnel Reynolds numbers at power-off conditions. The CFD solutions were used to build increments to account for Reynolds number effects. We calculate the validation errors for the primary CFD code results at wind tunnel Reynolds number power-off conditions and would like to be able to use those errors to predict the validation errors for the CFD increments. However, the validation errors are large compared to the increments. We suggest a way forward that is consistent with common practice in wind tunnel testing which is to assume that systematic errors in the measurement process and/or the environment will subtract out when increments are calculated, thus making increments more reliable with smaller uncertainty than absolute values of the aerodynamic coefficients. A similar practice has arisen for the use of CFD to generate aerodynamic database increments. The basis of this practice is the assumption of strong correlation of the systematic errors inherent in each of the results used to generate an increment. The assumption of strong correlation is the inferential link between the observed validation uncertainties at wind-tunnel Reynolds numbers and the uncertainties to be predicted for flight. In this paper, we suggest a way to estimate the correlation coefficient and demonstrate the approach using code-to-code differences that were obtained for quality control purposes during the Ares I CFD campaign. Finally, since we can expect the increments to be relatively small compared to the baseline response surface and to be typically of the order of the baseline uncertainty, we find that it is necessary to be able to show that the correlation coefficients are close to unity to avoid overinflating the overall database uncertainty with the addition of the increments.

  11. 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)

  12. Neonatal Facial Coding System for Assessing Postoperative Pain in Infants: Item Reduction is Valid and Feasible

    NARCIS (Netherlands)

    Peters, J.W.B.; Koot, H.M.; Grunau, R.E.; Boer, J. de; Druenen, M.J. van; Tibboel, D.; Duivenvoorden, H.J.

    2003-01-01

    Objective: The objectives of this study were to: (1) evaluate the validity of the Neonatal Facial Coding System (NFCS) for assessment of postoperative pain and (2) explore whether the number of NFCS facial actions could be reduced for assessing postoperative pain. Design: Prospective, observational

  13. 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.

  14. 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

  15. 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......). METHODS: The DNPR was used to identify all hospitalized patients, aged ≥15 years, with a first-time diagnosis of HSE according to the International Classification of Diseases, tenth revision (ICD-10), from 2004 to 2014. To validate the coding of HSE, we collected data from the Danish Microbiology Database......, from departments of clinical microbiology, and from patient medical records. Cases were classified as confirmed, probable, or no evidence of HSE. We estimated the positive predictive value (PPV) of the HSE diagnosis coding stratified by diagnosis type, study period, and department type. Furthermore, we...

  16. 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)

  17. 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)

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

    Directory of Open Access Journals (Sweden)

    Jørgensen LK

    2016-05-01

    Full Text Available Laura Krogh Jørgensen,1 Lars Skov Dalgaard,1 Lars Jørgen Østergaard,1 Nanna Skaarup Andersen,2 Mette Nørgaard,3 Trine Hyrup Mogensen1 1Department of Infectious Diseases, Aarhus University Hospital, Aarhus, 2Department of Clinical Microbiology, Odense University Hospital, Odense, 3Department of Clinical Epidemiology, Aarhus University Hospital, Aarhus, Denmark 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. Methods: The DNPR was used to identify all hospitalized patients, aged ≥15 years, with a first-time diagnosis of HSE according to the International Classification of Diseases, tenth revision (ICD-10, from 2004 to 2014. To validate the coding of HSE, we collected data from the Danish Microbiology Database, from departments of clinical microbiology, and from patient medical records. Cases were classified as confirmed, probable, or no evidence of HSE. We estimated the positive predictive value (PPV of the HSE diagnosis coding stratified by diagnosis type, study period, and department type. Furthermore, we estimated the proportion of HSE cases coded with nonspecific ICD-10 codes of viral encephalitis and also the sensitivity of the HSE diagnosis coding. Results: We were able to validate 398 (94.3% of the 422 HSE diagnoses identified via the DNPR. Hereof, 202 (50.8% were classified as confirmed cases and 29 (7.3% as probable cases providing an overall PPV of 58.0% (95% confidence interval [CI]: 53.0–62.9. For “Encephalitis due to herpes simplex virus” (ICD-10 code B00.4, the PPV was 56.6% (95% CI: 51.1–62.0. Similarly, the PPV for “Meningoencephalitis due to herpes simplex virus” (ICD-10 code B00.4A was 56.8% (95% CI: 39.5–72.9. “Herpes viral encephalitis” (ICD-10 code G05.1E had a PPV

  19. 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.

  20. 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.

  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. 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)

  3. 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)

  4. 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)

  5. 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.

  6. 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)

  7. Update on the development and validation of MERCURY: a modern, Monte Carlo particle transport code

    Energy Technology Data Exchange (ETDEWEB)

    Procassini, R.; Taylor, J.; McKinley, S.; Greenman, G. [Dermott Cullen, Matthew O' Brien, Bret Beck and Christian Hagmann, Lawrence Livermore National Lab., Livermore, CA (United States)

    2005-07-01

    An update on the development and validation of the MERCURY Monte Carlo particle transport code is presented. MERCURY is a modern, parallel, general-purpose Monte Carlo code being developed at the Lawrence Livermore National Laboratory. During the past year, several major algorithm enhancements have been completed. These include the addition of particle trackers for 3-dimensional combinatorial geometry (CG), 1-dimensional radial meshes, 2-dimensional quadrilateral unstructured meshes, as well as a feature known as templates for defining recursive, repeated structures in CG. New physics capabilities include an elastic-scattering neutron thermalization model for free gas and bound, S({alpha}, {beta}) molecular scattering, as well as support for continuous energy cross sections. Each of these new physics features has been validated through code-to-code comparisons with another Monte Carlo transport code. Several important computer science features have been developed, including an extensible input-parameter parser based upon the XML data description language, and a dynamic load-balance methodology for efficient parallel calculations. This paper discusses the recent work in each of these areas, and describes a plan for future extensions that are required to meet the needs of our ever expanding user base. (authors)

  8. Update on the Development and Validation of MERCURY: A Modern, Monte Carlo Particle Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Procassini, R J; Taylor, J M; McKinley, M S; Greenman, G M; Cullen, D E; O' Brien, M J; Beck, B R; Hagmann, C A

    2005-06-06

    An update on the development and validation of the MERCURY Monte Carlo particle transport code is presented. MERCURY is a modern, parallel, general-purpose Monte Carlo code being developed at the Lawrence Livermore National Laboratory. During the past year, several major algorithm enhancements have been completed. These include the addition of particle trackers for 3-D combinatorial geometry (CG), 1-D radial meshes, 2-D quadrilateral unstructured meshes, as well as a feature known as templates for defining recursive, repeated structures in CG. New physics capabilities include an elastic-scattering neutron thermalization model, support for continuous energy cross sections and S ({alpha}, {beta}) molecular bound scattering. Each of these new physics features has been validated through code-to-code comparisons with another Monte Carlo transport code. Several important computer science features have been developed, including an extensible input-parameter parser based upon the XML data description language, and a dynamic load-balance methodology for efficient parallel calculations. This paper discusses the recent work in each of these areas, and describes a plan for future extensions that are required to meet the needs of our ever expanding user base.

  9. 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

  10. 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.

  11. 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.

  12. A method for detecting code security vulnerability based on variables tracking with validated-tree

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    SQL injection poses a major threat to the application level security of the database and there is no systematic solution to these attacks.Different from traditional run time security strategies such as IDS and fire wall,this paper focuses on the solution at the outset;it presents a method to find vulnerabilities by analyzing the source codes.The concept of validated tree is developed to track variables referenced by database operations in scripts.By checking whether these variables are influenced by outside inputs,the database operations are proved to be secure or not.This method has advantages of high accuracy and efficiency as well as low costs,and it is universal to any type of web application platforms.It is implemented by the SOftware code vulnerabilities of SQL injection detector(CVSID).The validity and efficiency are demonstrated with an example.

  13. Extensible Validation Framework for DSLs using MontiCore on the Example of Coding Guidelines

    OpenAIRE

    Berger, Christian; Rumpe, Bernhard; Völkel, Steven

    2014-01-01

    Unit test environments are today's state of the art for many programming languages to keep the software's quality above a certain level. However, the software's syntactic quality necessary for the developers themselves is not covered by the aforementioned frameworks. This paper presents a tool realized using the DSL framework MontiCore for automatically validating easily extensible coding guidelines for any domain specific language or even general purpose languages like C++ and its applicatio...

  14. IAEA programme to support development and validation of advanced design and safety analysis codes

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has been organized many international collaboration programs to support the development and validation of design and safety analysis computer codes for nuclear power plants. These programs are normally implemented with a frame of Coordinated Research Project (CRP) or International Collaborative Standard Problem (ICSP). This paper introduces CRPs and ICSPs currently being organized or recently completed by IAEA for this purpose. (author)

  15. Validation of GEANT4 Monte Carlo Simulation Code for 6 MV Varian Linac Photon Beam

    International Nuclear Information System (INIS)

    The head of a clinical linear accelerator based on the manufacturer detailed information is simulated by using GEANT4. Percentage Depth Dose (PDD) and flatness symmetry (lateral dose profiles) in water phantom were evaluated. Comparisons between experimental and simulated data were carried out for two field sizes; 5 × 5, and 10 ×10 cm2. The obtained results indicated that GEANT4 code is a promising and validated Monte Carlo program for using in radiotherapy applications

  16. Validation of the THIRST steam generator thermalhydraulic code against the CLOTAIRE phase II experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Pietralik, J.M.; Campagna, A.O.; Frisina, V.C

    1999-04-01

    Steam generator thermalhydraulic codes are frequently used to calculate both global and local parameters inside a stern generator. The global parameters include heat transfer output, recirculation ratio, outlet temperatures, and pressure drops for operating and abnormal conditions. The local parameters are used in further analyses of flow-induced vibration, fretting wear, sludge deposition, and flow-accelerated corrosion. For these purposes, detailed, 3-dimensional 2-phase flow and heat transfer parameters are needed. To make the predictions more accurate and reliable, the codes need to be validated in geometries representative of real conditions. One such study is an international co-operative experimental program called CLOTAIRE, which is based in France. The CANDU Owners Group(COG) participated in the first two phases of the program. The results of the validation of Phase 1 were presented at the 1994 Steam Generator and Heat Exchanger Conference, and the results of the validation of Phase II are the subject of this report. THIRST is a thermalhydraulic, finite-volume code used to predict flow and heat transfer in steam generators. The local results of CLOTAIRE Phase II were used to validate the code. The results consist of the measurements of void fraction and axial gas-phase velocity in the U-bend region. The measurements were done using bi-optical probes. A comparison of global results indicates that the THIRST predictions, with the Chisholm void fraction model, are within 2% to 3% of the experimental results. Using THIRST with the homogeneous void fraction model, the global results were less accurate but still gave very good predictions; the greatest error was 10% for the separator pressure drop. Comparisons of the local predictions for void fraction and axial gas-phase velocity show good agreement. The Chisholm void fraction model generally gives better agreement with the experimental data, whereas the homogeneous model tends to overpredict the void fraction

  17. Validation of the THIRST steam generator thermalhydraulic code against the CLOTAIRE phase II experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Pietralik, J.M.; Campagna, A.O.; Frisina, V.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1998-07-01

    Steam generator thermalhydraulic codes are used frequently to calculate both global and local parameters inside the steam generator. The former include heat transfer output, recirculation ratio, outlet temperatures, and pressure drops for operating and abnormal conditions. The latter are used in further analyses of flow-induced vibration, fretting wear, sludge deposition, and flow accelerated corrosion. For these purposes, detailed, three-dimensional two-phase flow and heat transfer parameters are needed. To make the predictions more accurate and reliable, the codes need to be validated in geometries representative of real conditions. One such study is an international cooperative experimental program called CLOTAIRE based in France. COG participated in the first two phases of the program; the results of the validation of Phase 1 were presented at the 1994 Steam Generator and Heat Exchanger Conference, and the results of the validation of Phase II are the subject of this paper. THIRST is a thermalhydraulic, finite volume code to predict the flow and heat transfer in steam generators. The local results of CLOTAIRE Phase II have been used to validate the code. These consist of the measurements of void fraction and axial gas-phase velocity in the U-bend region. The measurements were done using bi-optical probes. A comparison of global results indicates that the THIRST predictions, with the Chisholm void fraction model, are within 2 to 3% of the experimental results. Using THIRST with the homogeneous void fraction model, the global results were less accurate but still well predicted with the greatest error of 10% for the separator pressure drop. Comparisons of the local predictions for void fraction and axial gas-phase show good agreement. The Chisholm void fraction model generally gives better agreement with the experimental data while the homogeneous model tends to overpredict the void fraction and underpredict the gas velocity. (author)

  18. AEEW comments on the NNC/CEGB LOCA code validation report RX 440-A

    International Nuclear Information System (INIS)

    Comments are made on the NNC/CEGB report PWR/RX 440-A, Review of Validation for the ECCS Evaluation Model Codes, by K.T. Routledge et al, 1982. This set out to review methods and models used in the LOCA safety case for Sizewell B. These methods are embodied in the Evaluation Model Computer codes SATAN-VI, WREFLOOD, WFLASH, LOCTA-IV and COCO. The main application of these codes is the determination of peak clad temperature and overall containment pressure. The comments represent the views of a group which has been involved for a number of years in the development and application of Best-Estimate methods for LOCA analysis. It is the judgement of this group that, overall, the EM methods can be used to make an acceptable safety case, but there are a number of points of detail still to be resolved. (U.K.)

  19. Simulating fuel assemblies with low resolution CFD approaches

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, F., E-mail: roelofs@nrg.eu [NRG, Petten (Netherlands); Gopala, V.R. [NRG, Petten (Netherlands); Chandra, L. [IIT Rajasthan (India); Viellieber, M.; Class, A. [KIT, Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Justification for development of low resolution mesh approaches. Black-Right-Pointing-Pointer Mathematical background of the approaches. Black-Right-Pointing-Pointer Meshing considerations for different approaches are presented. Black-Right-Pointing-Pointer Examples of applications are provided. - Abstract: In addition to the traditional fuel assembly simulations 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 are 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 are difficult to 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 (small flows in the direction perpendicular to the main flow) is not feasible. Using this approach, the consequences of flow blockages for detection possibilities and cladding temperatures can be discussed. The goal of the CGCFD approach with SGM is that it 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. Within the CGCFD 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 current paper discusses and presents both, the CGCFD and the LRGR approaches.

  20. A Methodology to Validate 3-D Arbitrary Lagrangian Eulerian Codes with Applications to Alegra

    Energy Technology Data Exchange (ETDEWEB)

    Chhabildas, L.C.; Duggins, B.D.; Konrad, C.H.; Mosher, D.A.; Perry, J.S.; Reinhart, W.D.; Summers, R.M.; Trucano, T.G.

    1998-11-04

    In this study we provided an experimental test bed for validating features of the Arbitrary Lagrangian Eulerian Grid for Research Applications (ALEGRA) code over a broad range of strain rates with overlapping diagnostics that encompass the multiple responses. A unique feature of the 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 (107/s) and those characteristics of structural response (102/s). Most previous code validation experimental &udies, however, have been restricted to simulating or investigating a single strain-rate regime. What is new and different in this investigation is that we have performed well-controlled and 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. The current experiments span strain rate regimes of over 107/s to less than 102/s in a single experiment. The input conditions were extremely well defined. Velocity interferometers were used to record the high' strain-rate response, while low strain rate data were collected using strain gauges. Although the current tests were conducted at a nominal velocity of - 1.5 km/s, it is the test methodology that is being emphasized herein. Results of a three-dimensional experiment are also presented.

  1. CFD Simulation of Polydispersed Bubbly Two-Phase Flow around an Obstacle

    Directory of Open Access Journals (Sweden)

    E. Krepper

    2009-01-01

    Full Text Available This paper concerns the model of a polydispersed bubble population in the frame of an ensemble averaged two-phase flow formulation. The ability of the moment density approach to represent bubble population size distribution within a multi-dimensional CFD code based on the two-fluid model is studied. Two different methods describing the polydispersion are presented: (i a moment density method, developed at IRSN, to model the bubble size distribution function and (ii a population balance method considering several different velocity fields of the gaseous phase. The first method is implemented in the Neptune_CFD code, whereas the second method is implemented in the CFD code ANSYS/CFX. Both methods consider coalescence and breakup phenomena and momentum interphase transfers related to drag and lift forces. Air-water bubbly flows in a vertical pipe with obstacle of the TOPFLOW experiments series performed at FZD are then used as simulations test cases. The numerical results, obtained with Neptune_CFD and with ANSYS/CFX, allow attesting the validity of the approaches. Perspectives concerning the improvement of the models, their validation, as well as the extension of their applicability range are discussed.

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

    Directory of Open Access Journals (Sweden)

    Thomas Höhne

    2009-01-01

    Full Text Available 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 generator through the steam valves at low reactor power and with all main coolant pumps in operation. CFD calculations have been performed using a numerical grid model of 4.7 million tetrahedral elements. The Best Practice Guidelines in using CFD in nuclear reactor safety applications has been used. Different advanced turbulence models were utilized in the numerical simulation. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet, which corresponds to the measured values. The maximum local values of the relative temperature rise in the calculation are in the same range of the experiment. Due to this result, it is now possible to improve the mixing models which are usually used in system codes.

  3. Active-passive measurements and CFD based modelling for indoor radon dispersion study

    International Nuclear Information System (INIS)

    Computational fluid dynamics (CFD) play a significant role in indoor pollutant dispersion study. Radon is an indoor pollutant which is radioactive and inert gas in nature. The concentration level and spatial distribution of radon may be affected by the dwelling's ventilation conditions. Present work focus at the study of indoor radon gas distribution via measurement and CFD modeling in naturally ventilated living room. The need of the study is the prediction of activity level and to study the effect of natural ventilation on indoor radon. Two measurement techniques (Passive measurement using pin-hole dosimeters and active measurement using continuous radon monitor (SRM)) were used for the validation purpose of CFD results. The CFD simulation results were compared with the measurement results at 15 points, 3 XY planes at different heights along with the volumetric average concentration. The simulation results found to be comparable with the measurement results. The future scope of these CFD codes is to study the effect of varying inflow rate of air on the radon concentration level and dispersion pattern. - Highlights: • The distribution of radon gas in indoor environment was simulated using CFD modelling. • The distribution of radon was found to be more homogenous in open room condition. • The radon concentration level in open room was low as compare to closed room due to enhanced ventilation rate. • Simulation results are in agreement with active and passive measurements results

  4. Systematic review of validated case definitions for diabetes in ICD-9-coded and ICD-10-coded data in adult populations

    Science.gov (United States)

    Khokhar, Bushra; Jette, Nathalie; Metcalfe, Amy; Cunningham, Ceara Tess; Kaplan, Gilaad G; Butalia, Sonia; Rabi, Doreen

    2016-01-01

    Objectives With steady increases in ‘big data’ and data analytics over the past two decades, administrative health databases have become more accessible and are now used regularly for diabetes surveillance. The objective of this study is to systematically review validated International Classification of Diseases (ICD)-based case definitions for diabetes in the adult population. Setting, participants and outcome measures Electronic databases, MEDLINE and Embase, were searched for validation studies where an administrative case definition (using ICD codes) for diabetes in adults was validated against a reference and statistical measures of the performance reported. Results The search yielded 2895 abstracts, and of the 193 potentially relevant studies, 16 met criteria. Diabetes definition for adults varied by data source, including physician claims (sensitivity ranged from 26.9% to 97%, specificity ranged from 94.3% to 99.4%, positive predictive value (PPV) ranged from 71.4% to 96.2%, negative predictive value (NPV) ranged from 95% to 99.6% and κ ranged from 0.8 to 0.9), hospital discharge data (sensitivity ranged from 59.1% to 92.6%, specificity ranged from 95.5% to 99%, PPV ranged from 62.5% to 96%, NPV ranged from 90.8% to 99% and κ ranged from 0.6 to 0.9) and a combination of both (sensitivity ranged from 57% to 95.6%, specificity ranged from 88% to 98.5%, PPV ranged from 54% to 80%, NPV ranged from 98% to 99.6% and κ ranged from 0.7 to 0.8). Conclusions Overall, administrative health databases are useful for undertaking diabetes surveillance, but an awareness of the variation in performance being affected by case definition is essential. The performance characteristics of these case definitions depend on the variations in the definition of primary diagnosis in ICD-coded discharge data and/or the methodology adopted by the healthcare facility to extract information from patient records. PMID:27496226

  5. Clinicalcodes: an online clinical codes repository to improve the validity and reproducibility of research using electronic medical records

    OpenAIRE

    Springate, DA; Kontopantelis, E; Ashcroft, DM; Olier, I.; Parisi, R; Chamapiwa, E; Reeves, D

    2014-01-01

    Lists of clinical codes are the foundation for research undertaken using electronic medical records (EMRs). If clinical code lists are not available, reviewers are unable to determine the validity of research, full study replication is impossible, researchers are unable to make effective comparisons between studies, and the construction of new code lists is subject to much duplication of effort. Despite this, the publication of clinical codes is rarely if ever a requirement for obtaining gran...

  6. Development of the Verification and Validation Matrix for Safety Analysis Code SPACE

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yo Han; Ha, Sang Jun; Yang, Chang Keun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Korea Electric Power Research Institute (KEPRI) has been developed the safety analysis code, called as SPACE (Safety and Performance Analysis CodE for Nuclear Power Plant), for typical pressurized water reactors (PWR). Current safety analysis codes were conducted from foreign vendors, such as Westinghouse Electric Corp., ABB Combustion Engineering Inc., Kraftwerk Union, etc. Considering the conservatism and inflexibility of the foreign code systems, it is difficult to expand the application areas and analysis scopes. To overcome the mentioned problems KEPRI has launched the project to develop the native safety analysis code with Korea Power Engineering Co.(KOPEC), Korea Atomic Energy Research Inst.(KAERI), Korea Nuclear Fuel(KNF), and Korea Hydro and Nuclear Power Co.(KHNP) under the funding of Ministry of Knowledge Economy (MKE). As a result of the project, the demo-version of SPACE has been released in July 2009. As an advance preparation of the next step, KEPRI and colleagues have developed the verification and validation (V and V) matrix for SPACE. To develop the matrix, the preceding studies and experiments were reviewed. After mature consideration, the V and V matrix has been developed and the experiment plans were designed for the next step to compensate the lack of data.

  7. CFD modeling and experimental validation of heat and mass transfer in wood poles subjected to high temperatures: a conjugate approach

    Science.gov (United States)

    Younsi, R.; Kocaefe, D.; Poncsak, S.; Kocaefe, Y.; Gastonguay, L.

    2008-03-01

    In this article, a coupling method is presented in the case of high thermal treatment of a wood pole and a three-dimensional numerical simulation is proposed. The conservation equations for the wood sample are obtained using diffusion equation with variables diffusion coefficients and the incompressible Reynolds averaged Navier Stokes equations have been solved for the flow field. The connection between the two problems is achieved by expressing the continuity of the state variables and their respective fluxes through the interface. Turbulence closure is obtained by the use of the standard k ɛ model with the usual wall function treatment. The model equations are solved numerically by the commercial package ANSYS-CFX10. The wood pole was subjected to high temperature treatment under different operating conditions. The model validation is carried out via a comparison between the predicted values with those obtained experimentally. The comparison of the numerical and experimental results shows good agreement, implying that the proposed numerical algorithm can be used as a useful tool in designing high-temperature wood treatment processes. A parametric study was also carried out to determine the effects of several parameters such as initial moisture content, wood aspect ratio and final gas temperature on temperature and moisture content distributions within the samples during heat treatment.

  8. Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-04-07

    The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

  9. Radiant Energy Measurements from a Scaled Jet Engine Axisymmetric Exhaust Nozzle for a Baseline Code Validation Case

    Science.gov (United States)

    Baumeister, Joseph F.

    1994-01-01

    A non-flowing, electrically heated test rig was developed to verify computer codes that calculate radiant energy propagation from nozzle geometries that represent aircraft propulsion nozzle systems. Since there are a variety of analysis tools used to evaluate thermal radiation propagation from partially enclosed nozzle surfaces, an experimental benchmark test case was developed for code comparison. This paper briefly describes the nozzle test rig and the developed analytical nozzle geometry used to compare the experimental and predicted thermal radiation results. A major objective of this effort was to make available the experimental results and the analytical model in a format to facilitate conversion to existing computer code formats. For code validation purposes this nozzle geometry represents one validation case for one set of analysis conditions. Since each computer code has advantages and disadvantages based on scope, requirements, and desired accuracy, the usefulness of this single nozzle baseline validation case can be limited for some code comparisons.

  10. Verification & Validation Toolkit to Assess Codes: Is it Theory Limitation, Numerical Method Inadequacy, Bug in the Code or a Serious Flaw?

    Science.gov (United States)

    Bombardelli, F. A.; Zamani, K.

    2014-12-01

    We introduce and discuss an open-source, user friendly, numerical post-processing piece of software to assess reliability of the modeling results of environmental fluid mechanics' codes. Verification and Validation, Uncertainty Quantification (VAVUQ) is a toolkit developed in Matlab© for general V&V proposes. In this work, The VAVUQ implementation of V&V techniques and user interfaces would be discussed. VAVUQ is able to read Excel, Matlab, ASCII, and binary files and it produces a log of the results in txt format. Next, each capability of the code is discussed through an example: The first example is the code verification of a sediment transport code, developed with the Finite Volume Method, with MES. Second example is a solution verification of a code for groundwater flow, developed with the Boundary Element Method, via MES. Third example is a solution verification of a mixed order, Compact Difference Method code of heat transfer via MMS. Fourth example is a solution verification of a 2-D, Finite Difference Method code of floodplain analysis via Complete Richardson Extrapolation. In turn, application of VAVUQ in quantitative model skill assessment studies (validation) of environmental codes is given through two examples: validation of a two-phase flow computational modeling of air entrainment in a free surface flow versus lab measurements and heat transfer modeling in the earth surface versus field measurement. At the end, we discuss practical considerations and common pitfalls in interpretation of V&V results.

  11. Application of a CFD based containment model to different large-scale hydrogen distribution experiments

    International Nuclear Information System (INIS)

    Highlights: • A CFD based model developed in ANSYS-FLUENT for simulating the distribution of hydrogen in the containment of a nuclear power plant during a severe accident is validated against four large-scale experiments. • The successive formation and mixing of a stratified gas-layer in experiments performed in the THAI and PANDA facilities are predicted well by the CFD model. • The pressure evolution and related condensation rate during different mixed convection flow conditions in the TOSQAN facility are predicted well by the CFD model. • The results give confidence in the general applicability of the CFD model and model settings. - Abstract: In the event of core degradation during a severe accident in water-cooled nuclear power plants (NPPs), large amounts of hydrogen are generated that may be released into the reactor containment. As the hydrogen mixes with the air in the containment, it can form a flammable mixture. Upon ignition it can damage relevant safety systems and put the integrity of the containment at risk. Despite the installation of mitigation measures, it has been recognized that the temporary existence of combustible or explosive gas clouds cannot be fully excluded during certain postulated accident scenarios. The distribution of hydrogen in the containment and mitigation of the risk are, therefore, important safety issues for NPPs. Complementary to lumped parameter code modelling, Computational Fluid Dynamics (CFD) modelling is needed for the detailed assessment of the hydrogen risk in the containment and for the optimal design of hydrogen mitigation systems in order to reduce this risk as far as possible. The CFD model applied by NRG makes use of the well-developed basic features of the commercial CFD package ANSYS-FLUENT. This general purpose CFD package is complemented with specific user-defined sub-models required to capture the relevant thermal-hydraulic phenomena in the containment during a severe accident as well as the effect of

  12. HYDRA-II: A hydrothermal analysis computer code: Volume 3, Verification/validation assessments

    International Nuclear Information System (INIS)

    HYDRA-II is a hydrothermal computer code capable of three-dimensional analysis of coupled conduction, convection, and thermal radiation problems. This code is especially appropriate for simulating the steady-state performance of spent fuel storage systems. The code has been evaluated for this application for the US Department of Energy's Commercial Spent Fuel Management Program. HYDRA-II provides a finite difference solution in cartesian coordinates to the equations governing the conservation of mass, momentum, and energy. A cylindrical coordinate system may also be used to enclose the cartesian coordinate system. This exterior coordinate system is useful for modeling cylindrical cask bodies. The difference equations for conservation of momentum are enhanced by the incorporation of directional porosities and permeabilities that aid in modeling solid structures whose dimensions may be smaller than the computational mesh. The equation for conservation of energy permits modeling of orthotropic physical properties and film resistances. Several automated procedures are available to model radiation transfer within enclosures and from fuel rod to fuel rod. The documentation of HYDRA-II is presented in three separate volumes. Volume I - Equations and Numerics describes the basic differential equations, illustrates how the difference equations are formulated, and gives the solution procedures employed. Volume II - User's Manual contains code flow charts, discusses the code structure, provides detailed instructions for preparing an input file, and illustrates the operation of the code by means of a model problem. This volume, Volume III - Verification/Validation Assessments, provides a comparison between the analytical solution and the numerical simulation for problems with a known solution. This volume also documents comparisons between the results of simulations of single- and multiassembly storage systems and actual experimental data. 11 refs., 55 figs., 13 tabs

  13. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    thermal-hydraulic feedback effects. Thus, in VALCO work package 3 (WP 3) stand-alone three-dimensional neutron-kinetic codes have been validated. Measurements carried out in an original-size VVER-1000 mock-up (V-1000 facility, Kurchatov Institute Moscow) were used for the validation of the codes DYN3D, HEXTRAN, KIKO3D and BIPR-8, which are chiefly designed for VVER safety calculations. The significant neutron flux tilt measured in the V-1000 core, which is caused only by radial-reflector asymmetries, was successfully modelled. A good agreement between calculated and measured steady-state powers has been achieved, for relative assembly powers and inner-assembly pin power distributions. Calculated effective multiplication factors exceed unity in all cases. (orig.)

  14. MELCOR and SCDAP/RELAP5 code validation by simulation of TMI-2

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Chris [400 Central Drive, School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Liao, Yehong; Vierow, Karen

    2005-07-01

    Full text of publication follows: A comparison of the two severe accident codes, MELCOR and SCDAP/RELAP5, within the scope of thermal-hydraulic and core degradation models, has been previously performed by the authors for a hypothetical station blackout severe accident of a typical 4-loop PWR. This paper describes a validation of the codes. In the simulation of the TMI-2 severe accident, the data recorded during the accident and inferred from post-accident phenomena were used to investigate the soundness and compare the capabilities of MELCOR and SCDAP/RELAP5. With versatile control functions, best estimate thermal-hydraulic component models and detailed core models, MELCOR and SCDAP/RELAP5 produced similar predictions of the progression of TMI-2 accident, in simulating the actions of the plant control room operators, the system thermal-hydraulic response, the fuel damages, the core degradation and relocation. Input models and assumptions were modified to be as consistent yet true to the actual plant as possible. Due to different development approaches and some unique models, some minor discrepancies were observed between the predictions of MELCOR and SCDAP/RELAP5, which are within the uncertainties of the code numerical computation and the physics models. Some significant discrepancies in a few key areas were resolved either with a sophisticated method comparing with phenomena instead of raw code output information, or with a model modification based on actual plant data and experiment results. (authors)

  15. Development and validation of ALEPH2 Monte Carlo burn-up code

    Energy Technology Data Exchange (ETDEWEB)

    Van Den Eynde, G.; Stankovskiy, A.; Fiorito, L.; Broustaut, M. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2013-07-01

    The ALEPH2 Monte Carlo depletion code has two principal features that make it a flexible and powerful tool for reactor analysis. First of all, its comprehensive nuclear data library ensures the consistency between steady-state Monte Carlo and deterministic depletion modules. It covers neutron and proton induced reactions, neutron and proton fission product yields, spontaneous fission product yields, radioactive decay data and total recoverable energies per fission. Secondly, ALEPH2 uses an advanced numerical solver for the first order ordinary differential equations describing the isotope balances, namely a Radau IIA implicit Runge-Kutta method. The versatility of the code allows using it for time behavior simulation of various systems ranging from single pin model to full-scale reactor model. The code is extensively used for the neutronics design of the MYRRHA research fast spectrum facility which will operate in both critical and sub-critical modes. The code has been validated on the decay heat data from JOYO experimental fast reactor. (authors)

  16. Validation of a Computer Code for Use in the Mechanical Design of Spallation Neutron Targets

    CERN Document Server

    Montanez, P A

    2000-01-01

    The present work concentrates on comparing a numerical code and a closed-form analytic solution for determining transient stress waves generated by an impinging, high-intensity proton pulse onto a perfectly elastic solid cylindrical target. The comparison of the two methods serves both to benchmark the physics and numerical methods of the codes, and to verify them against analytic expressions that can be established for calculating the response of the target for simple cases of loading and geometry. Additionally, the comparison elucidated the effects of approximations used in the computation of the analytic results. Two load cases have been investigated: (1) an instantaneously uniform thermal loading along the central core, and (2) a ramped and uniform thermal load applied along the central core. In addition, the influence of the approximations applied to the accurate analytic forms has been elucidated. By validating these analytical results, the closed-form solution may be confidently used to "bound" the sol...

  17. CFD simulation of shear-induced aggregation and breakage in turbulent Taylor-Couette flow.

    Science.gov (United States)

    Wang, Liguang; Vigil, R Dennis; Fox, Rodney O

    2005-05-01

    An experimental and computational investigation of the effects of local fluid shear rate on the aggregation and breakage of approximately 10 microm latex spheres suspended in an aqueous solution undergoing turbulent Taylor-Couette flow was carried out. First, computational fluid dynamics (CFD) simulations were performed and the flow field predictions were validated with data from particle image velocimetry experiments. Subsequently, the quadrature method of moments (QMOM) was implemented into the CFD code to obtain predictions for mean particle size that account for the effects of local shear rate on the aggregation and breakage. These predictions were then compared with experimental data for latex sphere aggregates (using an in situ optical imaging method). Excellent agreement between the CFD-QMOM and experimental results was observed for two Reynolds numbers in the turbulent-flow regime. PMID:15797411

  18. A Complex-Geometry Validation Experiment for Advanced Neutron Transport Codes

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Anthony W. LaPorta; Joseph W. Nielsen; James Parry; Mark D. DeHart; Samuel E. Bays; William F. Skerjanc

    2013-11-01

    The Idaho National Laboratory (INL) has initiated a focused effort to upgrade legacy computational reactor physics software tools and protocols used for support of core fuel management and experiment management in the Advanced Test Reactor (ATR) and its companion critical facility (ATRC) at the INL.. This will be accomplished through the introduction of modern high-fidelity computational software and protocols, with appropriate new Verification and Validation (V&V) protocols, over the next 12-18 months. Stochastic and deterministic transport theory based reactor physics codes and nuclear data packages that support this effort include MCNP5[1], SCALE/KENO6[2], HELIOS[3], SCALE/NEWT[2], and ATTILA[4]. Furthermore, a capability for sensitivity analysis and uncertainty quantification based on the TSUNAMI[5] system has also been implemented. Finally, we are also evaluating the Serpent[6] and MC21[7] codes, as additional verification tools in the near term as well as for possible applications to full three-dimensional Monte Carlo based fuel management modeling in the longer term. On the experimental side, several new benchmark-quality code validation measurements based on neutron activation spectrometry have been conducted using the ATRC. Results for the first four experiments, focused on neutron spectrum measurements within the Northwest Large In-Pile Tube (NW LIPT) and in the core fuel elements surrounding the NW LIPT and the diametrically opposite Southeast IPT have been reported [8,9]. A fifth, very recent, experiment focused on detailed measurements of the element-to-element core power distribution is summarized here and examples of the use of the measured data for validation of corresponding MCNP5, HELIOS, NEWT, and Serpent computational models using modern least-square adjustment methods are provided.

  19. Validation of lattice code 'EXCEL' with TIC experiments on uniform and regularly perturbed lattices

    International Nuclear Information System (INIS)

    Temporary International Collective (TIC) was established in 1972 by an agreement among seven countries, namely, Bulgaria, Czechoslovakia, Germany, Hungary, Poland, Romania and Union of Soviet Socialist Republics. The main objective of TIC was to provide the experimental data for the reactor physics analysis of water cooled and water moderated power reactors (WWER). Extensive experimental work for different core configurations was carried out by TIC countries to investigate the physics behaviour of WWER lattices and the results were published in TIC volumes. Two VVER-1000 MWe reactors are currently in an advanced stage of construction and due for commissioning in Kudankulam, Tamil Nadu, India. Indigenous development of in-core fuel management computer codes for the analysis of hexagonal lattice cores is also in an advanced stage to address various design, operation and safety issues of VVER type cores. The validation of the above TIC lattice experiments will help in the identification of deficiencies in reactor physics design computational codes and the associated nuclear data libraries. In this paper, TIC experiments on uniform and regularly perturbed lattices have been analyzed as part of the validation of indigenous computer codes, EXCEL, TRIHEX-FA and HEXPIN developed at Light Water Reactors Physics Section, B.A.R.C. Neutron-nuclear multi-group cross-section libraries in WIMS/D format in 69/172 energy groups have been released by IAEA at the conclusion of WIMS library update project (WLUP). In the present study we have used libraries based on ENDF/B-6, ENDF/B-7, JEFF3.1 and JENDL3.2 evaluated nuclear datasets. The results of the theoretical analyses bring out the performance of the code system and various cross-section libraries.

  20. Validation of computer code TRAFIC used for estimation of charcoal heatup in containment ventilation systems

    International Nuclear Information System (INIS)

    Full text of publication follows: Standard Indian PHWRs are provided with a Primary Containment Filtration and Pump-Back System (PCFPB) incorporating charcoal filters in the ventilation circuit to remove radioactive iodine that may be released from reactor core into the containment during LOCA+ECCS failure which is a Design Basis Accident for containment of radioactive release. This system is provided with two identical air circulation loops, each having 2 full capacity fans (1 operating and 1 standby) for a bank of four combined charcoal and High Efficiency Particulate Activity (HEPA) filters, in addition to other filters. While the filtration circuit is designed to operate under forced flow conditions, it is of interest to understand the performance of the charcoal filters, in the event of failure of the fans after operating for some time, i.e., when radio-iodine inventory is at its peak value. It is of interest to check whether the buoyancy driven natural circulation occurring in the filtration circuit is sufficient enough to keep the temperature in the charcoal under safe limits. A computer code TRAFIC (Transient Analysis of Filters in Containment) was developed using conservative one dimensional model to analyze the system. Suitable parametric studies were carried out to understand the problem and to identify the safety of existing system. TRAFIC Code has two important components. The first one estimates the heat generation in charcoal filter based on 'Source Term'; while the other one performs thermal-hydraulic computations. In an attempt validate the Code, experimental studies have been carried out. For this purpose, an experimental set up comprising of scaled down model of filtration circuit with heating coils embedded in charcoal for simulating the heating effect due to radio iodine has been constructed. The present work of validation consists of utilizing the results obtained from experiments conducted for different heat loads, elevations and adsorbent

  1. Integrating CFD and building simulation

    DEFF Research Database (Denmark)

    Bartak, M.; Beausoleil-Morrison, I.; Clarke, J.A.;

    2002-01-01

    To provide practitioners with the means to tackle problems related to poor indoor environments, building simulation and computational 3uid dynamics can usefully be integrated within a single computational framework. This paper describes the outcomes from a research project sponsored by the European...... Commission, which furthered the CFD modelling aspects of the ESP-r system. The paper summarises the form of the CFD model, describes the method used to integrate the thermal and 3ow domains and reports the outcome from an empirical validation exercise. © 2002 Published by Elsevier Science Ltd....

  2. Development of an adjoint sensitivity method for site characterization, uncertainty analysis, and code calibration/validation

    Energy Technology Data Exchange (ETDEWEB)

    Lu, A.H.

    1991-09-01

    The adjoint method is applied to groundwater flow-mass transport coupled equations in variably saturated media. The sensitivity coefficients derived by this method can be calculated by a single execution for each performance measure regardless of the number of parameters in question. The method provides an efficient and effective way to rank the importance of the parameters, so that data collection can be guided in support of site characterization programs. The developed code will facilitate the sensitivity/uncertainty analysis in both model prediction and model calibration/validation. 13 refs., 1 tab.

  3. Validation of a pre-coded food record for infants and young children

    OpenAIRE

    Gondolf, Ulla Holmboe; Tetens, Inge; Hills, Andrew; Michaelsen, Kim Fleischer; Trolle, Ellen

    2011-01-01

    Abstract Background/Objectives: To assess the validity of a 7-day pre-coded food record (PFR) method in 9-month-old infants against metabolisable energy intake (MEDLW) measured by doubly labelled water (DLW); additionally to compare PFR with a 7-day weighed food record (WFR) in 9-month-old infants and 36-month-old children. Subjects/Methods: The study population consisted of 36 infants (age: 9.03?0.2 months) and 36 young children (age: 36.1?0.3 months) enrolled in a cross-over ...

  4. Validation of simulation codes for future systems: motivations, approach, and the role of nuclear data

    International Nuclear Information System (INIS)

    The validation of advanced simulation tools will still play a very significant role in several areas of reactor system analysis. This is the case of reactor physics and neutronics, where nuclear data uncertainties still play a crucial role for many core and fuel cycle parameters. The present paper gives a summary of validation motivations, objectives and approach. A validation effort is in particular necessary in the frame of advanced (e.g. Generation-IV or GNEP) reactors and associated fuel cycles assessment and design. Validation of simulation codes is complementary to the 'verification' process. In fact, 'verification' addresses the question 'are we solving the equations correctly' while validation addresses the question 'are we solving the correct equations with the correct parameters'. Verification implies comparisons with 'reference' equation solutions or with analytical solutions, when they exist. Most of what is called 'numerical validation' falls in this category. Validation strategies differ according to the relative weight of the methods and of the parameters that enter into the simulation tools. Most validation is based on experiments, and the field of neutronics where a 'robust' physics description model exists and which is function of 'input' parameters not fully known, will be the focus of this paper. In fact, in the case of reactor core, shielding and fuel cycle physics the model (theory) is well established (the Boltzmann and Bateman equations) and the parameters are the nuclear cross-sections, decay data etc. Two types of validation approaches can and have been used: (a) Mock-up experiments ('global' validation): need for a very close experimental simulation of a reference configuration. Bias factors cannot be extrapolated beyond reference configuration; (b) Use of 'clean', 'representative' integral experiments ('bias factor and adjustment' method). Allows to define bias factors, uncertainties and can be used for a wide range of applications. It

  5. PIV validation of blood-heart valve leaflet interaction modelling.

    Science.gov (United States)

    Kaminsky, R; Dumont, K; Weber, H; Schroll, M; Verdonck, P

    2007-07-01

    The aim of this study was to validate the 2D computational fluid dynamics (CFD) results of a moving heart valve based on a fluid-structure interaction (FSI) algorithm with experimental measurements. Firstly, a pulsatile laminar flow through a monoleaflet valve model with a stiff leaflet was visualized by means of Particle Image Velocimetry (PIV). The inflow data sets were applied to a CFD simulation including blood-leaflet interaction. The measurement section with a fixed leaflet was enclosed into a standard mock loop in series with a Harvard Apparatus Pulsatile Blood Pump, a compliance chamber and a reservoir. Standard 2D PIV measurements were made at a frequency of 60 bpm. Average velocity magnitude results of 36 phase-locked measurements were evaluated at every 10 degrees of the pump cycle. For the CFD flow simulation, a commercially available package from Fluent Inc. was used in combination with inhouse developed FSI code based on the Arbitrary Lagrangian-Eulerian (ALE) method. Then the CFD code was applied to the leaflet to quantify the shear stress on it. Generally, the CFD results are in agreement with the PIV evaluated data in major flow regions, thereby validating the FSI simulation of a monoleaflet valve with a flexible leaflet. The applicability of the new CFD code for quantifying the shear stress on a flexible leaflet is thus demonstrated.

  6. Modellierung und CFD Simulation von viskoelastischen Ein- und Mehrphasenströmungen

    OpenAIRE

    Habla, Florian

    2015-01-01

    In this work a new model for describing viscoelastic two-phase flows is developed and implemented in an open-source CFD software together with a non-isothermal model and a method for modeling the temperature rise in single-screw extruders. Underlying numerical algorithms are improved in terms of stability, accuracy and efficiency and the developed code is thoroughly validated and verified with existing analytical solutions, other numerical predictions and experimental measurements. In dies...

  7. Application perspectives of simulation techniques CFD in nuclear power plants; Perspectivas de aplicacion de tecnicas de modelado CFD en plantas nucleoelectricas

    Energy Technology Data Exchange (ETDEWEB)

    Galindo G, I. F., E-mail: igalindo@iie.org.mx [Instituto de Investigaciones Electricas, Reforma No. 113, Col. Palmira, 62490 Cuernavaca, Morelos (Mexico)

    2013-10-15

    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)

  8. Validation analysis of pool fire experiment (Run-F7) using SPHINCS code

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, Akira [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Tajima, Yuji

    1998-04-01

    SPHINCS (Sodium Fire Phenomenology IN multi-Cell System) code has been developed for the safety analysis of sodium fire accident in a Fast Breeder Reactor. The main features of the SPHINCS code with respect to the sodium pool fire phenomena are multi-dimensional modeling of the thermal behavior in sodium pool and steel liner, modeling of the extension of sodium pool area based on the sodium mass conservation, and equilibrium model for the chemical reaction of pool fire on the flame sheet at the surface of sodium pool during. Therefore, the SPHINCS code is capable of temperature evaluation of the steel liner in detail during the small and/or medium scale sodium leakage accidents. In this study, Run-F7 experiment in which the sodium leakage rate is 11.8 kg/hour has been analyzed. In the experiment the diameter of the sodium pool is approximately 60 cm and the maximum steel liner temperature was 616 degree C. The analytical results tell us the agreement between the SPHINCS analysis and the experiment is excellent with respect to the time history and spatial distribution of the liner temperature, sodium pool extension behavior, as well as atmosphere gas temperature. It is concluded that the pool fire modeling of the SPHINCS code has been validated for this experiment. The SPHINCS code is currently applicable to the sodium pool fire phenomena and the temperature evaluation of the steel liner. The experiment series are continued to check some parameters, i.e., sodium leakage rate and the height of sodium leakage. Thus, the author will analyze the subsequent experiments to check the influence of the parameters and applies SPHINCS to the sodium fire consequence analysis of fast reactor. (author)

  9. 3-D CFD Simulation and Validation of Oxygen-Rich Hydrocarbon Combustion in a Gas-Centered Swirl Coaxial Injector using a Flamelet-Based Approach

    Science.gov (United States)

    Richardson, Brian; Kenny, Jeremy

    2015-01-01

    Injector design is a critical part of the development of a rocket Thrust Chamber Assembly (TCA). Proper detailed injector design can maximize propulsion efficiency while minimizing the potential for failures in the combustion chamber. Traditional design and analysis methods for hydrocarbon-fuel injector elements are based heavily on empirical data and models developed from heritage hardware tests. Using this limited set of data produces challenges when trying to design a new propulsion system where the operating conditions may greatly differ from heritage applications. Time-accurate, Three-Dimensional (3-D) Computational Fluid Dynamics (CFD) modeling of combusting flows inside of injectors has long been a goal of the fluid analysis group at Marshall Space Flight Center (MSFC) and the larger CFD modeling community. CFD simulation can provide insight into the design and function of an injector that cannot be obtained easily through testing or empirical comparisons to existing hardware. However, the traditional finite-rate chemistry modeling approach utilized to simulate combusting flows for complex fuels, such as Rocket Propellant-2 (RP-2), is prohibitively expensive and time consuming even with a large amount of computational resources. MSFC has been working, in partnership with Streamline Numerics, Inc., to develop a computationally efficient, flamelet-based approach for modeling complex combusting flow applications. In this work, a flamelet modeling approach is used to simulate time-accurate, 3-D, combusting flow inside a single Gas Centered Swirl Coaxial (GCSC) injector using the flow solver, Loci-STREAM. CFD simulations were performed for several different injector geometries. Results of the CFD analysis helped guide the design of the injector from an initial concept to a tested prototype. The results of the CFD analysis are compared to data gathered from several hot-fire, single element injector tests performed in the Air Force Research Lab EC-1 test facility

  10. CFD activities in support of thermal-hydraulic modeling of SFR fuel bundles

    International Nuclear Information System (INIS)

    Extensive testing and validation work is being performed to assess and validate Computational Fluid Dynamics (CFD) applicability to the simulation of SFR fuel assemblies. The demonstrated robustness of the method allows extending the CFD analysis to distorted fuel configurations, which will inevitably occur during extended fuel operation. The subchannel code COBRA-IV-I-MIT is adopted to evaluate the range of applicability of lumped parameter methods. Comparisons of mixing simulations show some intrinsic limitation in the subchannel methods, but allow confirming its overall applicability to nominal and mildly deformed assembly configurations. For significantly deformed geometries CFD is the recommend approach and is applied in this work. Deformed geometries considered include duct swelling, rod swelling, rod bowing, rod twisting, and various combinations of the simple deformations. While not derived from the realistic analysis of the in-core fuel behavior, the distorted geometries have been designed to embrace all conceptual worst case scenarios. The work focuses on the evaluation of the influence of the deformation on the fuel behavior, rather than on the actual fuel performance. Such approach is driven by the objective of deriving general understanding, and evaluating the applicability of subchannel analysis codes to long life fuel design, possibly in combination with distorted-channel factors derived from the CFD analyses. (author)

  11. CFD Analysis of an Aerofoil

    Directory of Open Access Journals (Sweden)

    Karna S. Patel

    2014-03-01

    Full Text Available In this report we have obtained the drag and lift forces using CFD which can also be determined through experiments using wind tunnel testing. In experimental setup, the design model has to be placed in the test section. This process is quite laborious & (surely cost more than CFD techniques cost for the same. Thus we have gone through analytical method then it can be validated by experimental testing. The analysis of the two dimensional subsonic flow over a NACA 0012 airfoil at various angles of attack and operating at a Reynolds number of 3×E+06 is presented. The CFD simulation results show close agreement with those of the experiments, thus suggesting a reliable alternative to experimental method in determining drag and lift.

  12. New weighted sum of gray gases model applicable to Computational Fluid Dynamics (CFD) modeling of oxy-fuel combustion

    DEFF Research Database (Denmark)

    Yin, Chungen; Johansen, Lars Christian Riis; Rosendahl, Lasse;

    2010-01-01

    Radiation is the principal mode of heat transfer in furnaces. Models for gaseous radiative properties have been well established for air combustion. However, there is uncertainty regarding their applicability to oxy-fuel conditions. In this paper, a new and complete set of weighted sum of gray......), and the calculated results are validated in very details against data in literature. Then the validated code is used to generate emissivity databases for representative air-firing and oxy-firing conditions, for each of which a refined WSGGM with new parameters is derived. The practical way to implement the model...... into CFD simulations of combustion systems is given. Finally, as a demonstration, the new model is implemented into CFD modeling of two furnaces of very different beam lengths, respectively. The CFD results are compared with those based on the widely used WSGGM in literature, from which some useful...

  13. Validation of the XLACS code related to contribution of resolved and unresolved resonances and background cross sections

    International Nuclear Information System (INIS)

    The procedures for calculating contributions of resolved and unresolved resonances and background cross sections, in XLACS code, were revised. Constant weighting function and zero Kelvin temperature were considered. Discrepancies found were corrected and now the validated XLACS code generates results that are correct and in accordance with its originally established procedures. (author)

  14. The BOUT Project; Validation and Benchmark of BOUT Code and Experimental Diagnostic Tools for Fusion Boundary Turbulence

    Institute of Scientific and Technical Information of China (English)

    徐学桥

    2001-01-01

    A boundary plasma turbulence code BOUT is presented. The preliminary encour aging results have been obtained when comparing with probe measurements for a typical Ohmic discharge in HT-7 tokamak. The validation and benchmark of BOUT code and experimental diagnostic tools for fusion boundary plasma turbulence is proposed.

  15. Decay heat experiment and validation of calculation code systems for fusion reactor

    International Nuclear Information System (INIS)

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)

  16. Decay heat experiment and validation of calculation code systems for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

    1999-10-01

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

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

    International Nuclear Information System (INIS)

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

  18. Experiments to validate computer codes used in the safety assessment of concrete containments

    International Nuclear Information System (INIS)

    The safety analysis for the hazardous plants with reinforced and prestressed concrete containments includes the assessment of the containment performance under severe accident loading. Such assessment is normally based on the prediction using computer codes, supported by the measured evidence of small scale experiment. A program of small scale experiments is in progress at AEE Winfrith. The first series included five tests on simple concrete frame specimens for providing the basic response data under static loading. The second series included the test on reinforced concrete slab specimens having the geometry representing steel-lined containment walls. The experiment, the properties of the materials used and the measurement are reported. The selection of the measured results is presented. The validation of the finite element computer codes, ABAQUS and DYNA 3D, by collating with the measured results is in progress. Two aspects of the safety analysis for concrete containments under severe accident loading are the need for computer codes to predict accurately the structural response at re-entrant corners and the integrity of liners. (K.I.)

  19. Validity of the Child Facial Coding System for the Assessment of Acute Pain in Children With Cerebral Palsy.

    Science.gov (United States)

    Hadden, Kellie L; LeFort, Sandra; O'Brien, Michelle; Coyte, Peter C; Guerriere, Denise N

    2016-04-01

    The purpose of the current study was to examine the concurrent and discriminant validity of the Child Facial Coding System for children with cerebral palsy. Eighty-five children (mean = 8.35 years, SD = 4.72 years) were videotaped during a passive joint stretch with their physiotherapist and during 3 time segments: baseline, passive joint stretch, and recovery. Children's pain responses were rated from videotape using the Numerical Rating Scale and Child Facial Coding System. Results indicated that Child Facial Coding System scores during the passive joint stretch significantly correlated with Numerical Rating Scale scores (r = .72, P valid method of identifying pain in children with cerebral palsy.

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

    CERN Document Server

    Tsai, Shang-Min; 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 its output versus the disequilibrium-chemistry calculations of Moses et al. and Rimmer & Helling. It reproduces the models of HD 189733b and HD 209458b by Moses et al., which employ a network with nearly 1600 reactions. Further validation of VULCAN is made by examining the theoretical trends produced when the temperature-pressure profile and carbon-to-oxygen ratio are varied. Assisted by a sensitivity test designed to identify the key reactions responsible for producing a specific molecule, we revisit the quenching ap...

  1. CFD Turbulence Study of PWR Spacer-Grids in a Rod Bundle

    Directory of Open Access Journals (Sweden)

    C. Peña-Monferrer

    2014-01-01

    the flow dynamics and heat transfer phenomena along the fuel rods. This work presents the analysis of the turbulence effects of a split-type and swirl-type spacer-grid geometries on single phase in a PWR (pressurized water reactor rod bundle. Various computational fluid dynamics (CFD calculations have been performed and the results validated with the experiments of the OECD/NEA-KAERI rod bundle CFD blind benchmark exercise on turbulent mixing in a rod bundle with spacers at the MATiS-H facility. Simulation of turbulent phenomena downstream of the spacer-grid presents high complexity issues; a wide range of length scales are present in the domain increasing the difficulty of defining in detail the transient nature of turbulent flow with ordinary turbulence models. This paper contains a complete description of the procedure to obtain a validated CFD model for the simulation of the spacer-grids. Calculations were performed with the commercial code ANSYS CFX using large eddy simulation (LES turbulence model and the CFD modeling procedure validated by comparison with measurements to determine their suitability in the prediction of the turbulence phenomena.

  2. Adaption, validation and application of advanced codes with 3-dimensional neutron kinetics for accident analysis calculations - STC with Bulgaria

    International Nuclear Information System (INIS)

    In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.)

  3. CFD study of isothermal water flow in rod bundle with split-type spacer grid

    Science.gov (United States)

    Batta, A.; Class, A. G.

    2014-06-01

    The design of rod bundles in nuclear application nowadays is assessed by CFD (computational fluid dynamics). The accuracy of CFD models need validation. Within the OECD/NEA benchmark MATiS-H (Measurement and Analysis of Turbulent Mixing in Sub-channels - Horizontal) a single-phase water flow in a 5x5 rod bundle is studied. In the benchmark, two types of spacer grids are tested, the swirl type and the split type, where the current study focuses on the split type spacer grid. Comparison of CFD results obtained at Karlsruhe Institut of Technology (KIT) with experimental results of KAERI (Korea Atomic Energy Research Institute) are presented. In the benchmark velocities components along selected lines downstream of the spacer grid are measured and compared to CFD results. The CFD code STAR CCM+ with the Realized k-ɛ model is used. Comparisons with experimental results show quantitative and qualitative agreement for the averaged values of velocity components. Comparisons of results to other benchmark partners using different modeling show that the selected mesh size and models for the analysis of the current case gives relatively accurate results. However, the used turbulent model (Realized k-ɛ does not capture the turbulent intensity correctly. Computation shows that the flow has very high mixing due to the spacer grid, which does not decay within the measurements domain (z/ DH =0-10 downstream of spacer grid). The same conclusion can be drawn from experimental data.

  4. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

    International Nuclear Information System (INIS)

    This report presents the results of the COBRA-SFS (Spent Fuel Storage) computer code validation effort. COBRA-SFS, while refined and specialized for spent fuel storage system analyses, is a lumped-volume thermal-hydraulic analysis computer code that predicts temperature and velocity distributions in a wide variety of systems. Through comparisons of code predictions with spent fuel storage system test data, the code's mathematical, physical, and mechanistic models are assessed, and empirical relations defined. The six test cases used to validate the code and code models include single-assembly and multiassembly storage systems under a variety of fill media and system orientations and include unconsolidated and consolidated spent fuel. In its entirety, the test matrix investigates the contributions of convection, conduction, and radiation heat transfer in spent fuel storage systems. To demonstrate the code's performance for a wide variety of storage systems and conditions, comparisons of code predictions with data are made for 14 runs from the experimental data base. The cases selected exercise the important code models and code logic pathways and are representative of the types of simulations required for spent fuel storage system design and licensing safety analyses. For each test, a test description, a summary of the COBRA-SFS computational model, assumptions, and correlations employed are presented. For the cases selected, axial and radial temperature profile comparisons of code predictions with test data are provided, and conclusions drawn concerning the code models and the ability to predict the data and data trends. Comparisons of code predictions with test data demonstrate the ability of COBRA-SFS to successfully predict temperature distributions in unconsolidated or consolidated single and multiassembly spent fuel storage systems

  5. Validation of a pre-coded food record for infants and young children

    DEFF Research Database (Denmark)

    Gondolf, Ulla Holmboe; Tetens, Inge; Hills, A. P.;

    2012-01-01

    Background/Objectives:To assess the validity of a 7-day pre-coded food record (PFR) method in 9-month-old infants against metabolizable energy intake (ME(DLW)) measured by doubly labeled water (DLW); additionally to compare PFR with a 7-day weighed food record (WFR) in 9-month-old infants and 36......, crossing over to the alternative method in week 2. Total energy expenditure (TEE) and ME(DLW) were obtained in the 9-month-old infants using the DLW technique for 7 days while recording with PFR.Results:For the 9-month-old group, PFR showed a mean bias of +726âkJ/day, equivalent to 24%, (P...

  6. ASTEC V2 severe accident integral code: Fission product modelling and validation

    Energy Technology Data Exchange (ETDEWEB)

    Cantrel, L., E-mail: laurent.cantrel@irsn.fr; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-06-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix.

  7. Status and Strategy of the GAMMA-FR code Validation for ITER TBM and Fusion Reactor System in Korea

    International Nuclear Information System (INIS)

    Korea has developed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) and Helium Cooled Ceramic Reflector (HCCR) TBM to be tested in the ITER. The main purpose for developing the TBM is to develop the design technology for the DEMO and fusion reactor, which should be proved experimentally in the ITER. Therefore, we developed a design scheme and codes including the safety analysis capability for obtaining the license for testing in the ITER. The GAMMA-FR code is a domestic 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) is underway using this code, and not MELCOR, which is a representative code for ITER. Therefore, validation of GAMMA-FR is one of the most primary interests, and validation using MELCOR V and V list has top priority. The GAMMA-FR code was scheduled for validation during the next three years under UCLA-NFRI collaboration. Through this research, GAMMA-FR will be validated with representative fusion experiments and reference accident cases

  8. Validation of plant dynamics analysis code Super-COPD by MONJU startup tests

    International Nuclear Information System (INIS)

    The plant dynamics analysis code COPD was validated by the experimental fast reactor JOYO, the 50MWt class steam generator (SG) experiments and the in-pile experiments for MONJU components. The flow network models and the control systems of the MONJU plant in the COPD code, even those of the main components, were divided into simple calculation modules. These modular programs have been developed as the Super-COPD code and made it possible to construct the flow networks, the control systems and main components of arbitral liquid metal fast reactors by connecting these modules. However, it is necessary to validate the constructed flow networks in each component, because the flow paths and the heat transport routes change by the plant operational conditions and affect the whole plant dynamics. In this study, flow network models of the upper plenum of MONJU reactor vessel (RV) and those of the primary inlet plenum of the intermediate heat exchanger (IHX) were modified correctly by using the startup test conditions. Then the whole system from RV to turbines was calculated by Super-COPD and the plant dynamics were validated by the measured data. The evaluated cases were the plant trip transient from 40% rated operational condition (Case 1), the natural circulation in the primary loops (Case 2) and that in the secondary loops (Case 3), which were both caused by the released heat from the primary pump operation. The calculated results of Case 1 and the measured data are shown. In this transient, the primary and secondary main pump operations were switched from the flow coast down to the pony motor operations and the air cooler (AC) operations in the secondary loops were switched from the SG operations. The RV outlet sodium temperatures agreed well with each other until 3600 s. The calculated inlet sodium temperature were approximately 20 deg. C higher larger than the experiments from approximately 400 s to 1300 s. The outlet temperature of IHX in secondary heat transport

  9. Correlation of Puma airloads: Evaluation of CFD prediction methods

    Science.gov (United States)

    Strawn, Roger C.; Desopper, Andre; Miller, Judith; Jones, Alan

    1989-01-01

    A cooperative program was undertaken by research organizations in England, France, Australia and the U.S. to study the capabilities of computational fluid dynamics codes (CFD) to predict the aerodynamic loading on helicopter rotor blades. The program goal is to compare predictions with experimental data for flight tests of a research Puma helicopter with rectangular and swept tip blades. Two topics are studied. First, computed results from three CFD codes are compared for flight test cases where all three codes use the same partial inflow-angle boundary conditions. Second, one of the CFD codes (FPR) is iteratively coupled with the CAMRAD/JA helicopter performance code. These results are compared with experimental data and with an uncoupled CAMRAD/JA solution. The influence of flow field unsteadiness is found to play an important role in the blade aerodynamics. Alternate boundary conditions are suggested in order to properly model this unsteadiness in the CFD codes.

  10. Correlation of Puma airfoils - Evaluation of CFD prediction methods

    Science.gov (United States)

    Strawn, Roger C.; Desopper, Andre; Miller, Judith; Jones, Alan

    1989-01-01

    A cooperative program was undertaken by research organizations in England, France, Australia and the U.S. to study the capabilities of computational fluid dynamics codes (CFD) to predict the aerodynamic loading on helicopter rotor blades. The program goal is to compare predictions with experimental data for flight tests of a research Puma helicopter with rectangular and swept tip blades. Two topics are studied. First, computed results from three CFD codes are compared for flight test cases where all three codes use the same partial inflow-angle boundary conditions. Second, one of the CFD codes (FPR) is iteratively coupled with the CAMRAD/JA heilcopter performance code. These results are compared with experimental data and with an uncoupled CAMRAD/JA solution. The influence of flow field unsteadiness is found to play an important role in the blade aerodynamics. Alternate boundary conditions are suggested in order to properly model this unsteadiness in the CFD codes.

  11. Computer code validation study of PWR core design system, CASMO-3/MASTER-{alpha}

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Kim, M. H. [Kyounghee Univ., Taejon (Korea, Republic of); Woo, S. W. [KINS, Taejon (Korea, Republic of)

    1999-05-01

    In this paper, the feasibility of CASMO-3/MASTER-{alpha} nuclear design system was investigated for commercial PWR core. Validation calculation was performed as follows. Firstly, the accuracy of cross section generation from table set using linear feedback model was estimated. Secondly, the results of CASMO-3/MASTER-{alpha} was compared with CASMO-3/NESTLE 5.02 for a few benchmark problems. Microscopic cross sections computed from table set were almost the same with those from CASMO-3. There were small differences between calculated results of two code systems. Thirdly, the repetition of CASMO-3/MASTER-{alpha} calculation for Younggwang Unit-3, Cycle-1 core was done and their results were compared with nuclear design report(NDR) and uncertainty analysis results of KAERI. It was found that uncertainty analysis results were reliable enough because results were agreed each other. It was concluded that the use of nuclear design system CASMO-3/MASTER-{alpha} was validated for commercial PWR core.

  12. Fast-Track Design Efforts Using CFD: Bonneville Second Powerhouse

    Energy Technology Data Exchange (ETDEWEB)

    Rakowski, Cynthia L.; Ebner, Laurie L.; Richmond, Marshall C.

    2007-10-10

    A set of three-dimensional, computational fluid dynamics (CFD) models were developed and used for the Bonneville Project tailrace to study the impact of a proposed outfall structure on the tailrace hydraulics; these structures were designed to improve the survival of downstream migrant (juvenile) salmon. Flows were simulated by solving the Reynolds-Averaged Navier-Stokes equations together with a two-equation k-epsilon turbulences model in a commercial CFD code. The numerical model was validated using field-measured velocity data. The model results identified undesirable combinations of outfall location and operational scenarios and helped to identify the location in which the outfall structure was built. The numerical model provided a relatively low-cost tool to rapidly simulate and visualize the flow field for multiple proposed outfall locations for a large number of operational scenarios. The visualizations of the results from the CFD model provided insights to hydraulic engineers and fisheries biologists working on the design and placement of the outfall structure.

  13. Investigation on Improved Correlation of CFD and EFD for Supercritical Airfoil

    Directory of Open Access Journals (Sweden)

    Xin Xu

    2014-02-01

    Full Text Available It is necessary to improve the correlation between CFD and EFD through the correction of EFD results and validation of CFD method, thus investigating the aerodynamic characteristics of supercritical airfoil perfectly. In this study, NASA SC (2 -0714 airfoil is numerically simulated and compared with NASA corrected experimental results to validate the CFD method. The Barnwell-Sewell method is applied to correct sidewall effects for experimental results of typical supercritical airfoil CH obtained in NF-6 wind tunnel. It is shown that there was large disparity between CFD and uncorrected EFD results, while CFD and EFD compared well after correction. The CFD method is validated and the Barnwell-Sewell method is feasible for sidewall effects correction. The correlation of EFD and CFD improved after the correction of EFD results and validation of CFD method.

  14. Asynchronous Parallelization of a CFD Solver

    Directory of Open Access Journals (Sweden)

    Daniel S. Abdi

    2015-01-01

    Full Text Available A Navier-Stokes equations solver is parallelized to run on a cluster of computers using the domain decomposition method. Two approaches of communication and computation are investigated, namely, synchronous and asynchronous methods. Asynchronous communication between subdomains is not commonly used in CFD codes; however, it has a potential to alleviate scaling bottlenecks incurred due to processors having to wait for each other at designated synchronization points. A common way to avoid this idle time is to overlap asynchronous communication with computation. For this to work, however, there must be something useful and independent a processor can do while waiting for messages to arrive. We investigate an alternative approach of computation, namely, conducting asynchronous iterations to improve local subdomain solution while communication is in progress. An in-house CFD code is parallelized using message passing interface (MPI, and scalability tests are conducted that suggest asynchronous iterations are a viable way of parallelizing CFD code.

  15. Validity of ICD-9-CM Coding for Identifying Incident Methicillin-Resistant Staphylococcus aureus (MRSA) Infections: Is MRSA Infection Coded as a Chronic Disease?

    Science.gov (United States)

    Schweizer, Marin L.; Eber, Michael R.; Laxminarayan, Ramanan; Furuno, Jon P.; Popovich, Kyle J.; Hota, Bala; Rubin, Michael A.; Perencevich, Eli N.

    2013-01-01

    BACKGROUND AND OBJECTIVE Investigators and medical decision makers frequently rely on administrative databases to assess methicillin-resistant Staphylococcus aureus (MRSA) infection rates and outcomes. The validity of this approach remains unclear. We sought to assess the validity of the International Classification of Diseases, 9th Revision, Clinical Modification (ICD-9-CM) code for infection with drug-resistant microorganisms (V09) for identifying culture-proven MRSA infection. DESIGN Retrospective cohort study. METHODS All adults admitted to 3 geographically distinct hospitals between January 1, 2001, and December 31, 2007, were assessed for presence of incident MRSA infection, defined as an MRSA-positive clinical culture obtained during the index hospitalization, and presence of the V09 ICD-9-CM code. The k statistic was calculated to measure the agreement between presence of MRSA infection and assignment of the V09 code. Sensitivities, specificities, positive predictive values, and negative predictive values were calculated. RESULTS There were 466,819 patients discharged during the study period. Of the 4,506 discharged patients (1.0%) who had the V09 code assigned, 31% had an incident MRSA infection, 20% had prior history of MRSA colonization or infection but did not have an incident MRSA infection, and 49% had no record of MRSA infection during the index hospitalization or the previous hospitalization. The V09 code identified MRSA infection with a sensitivity of 24% (range, 21%–34%) and positive predictive value of 31% (range, 22%–53%). The agreement between assignment of the V09 code and presence of MRSA infection had a κ coefficient of 0.26 (95% confidence interval, 0.25–0.27). CONCLUSIONS In its current state, the ICD-9-CM code V09 is not an accurate predictor of MRSA infection and should not be used to measure rates of MRSA infection. PMID:21460469

  16. CFD Analysis of Centrifugal Pump: A Review

    Directory of Open Access Journals (Sweden)

    Narayan P. Jaiswal

    2014-05-01

    Full Text Available The main objective of this work is to understand role of the computational fluid dynamics (CFD technique in analyzing and predicting the performance of centrifugal pump. Computational Fluid Dynamics (CFD is the present day state-of-art technique for fluid flow analysis. The critical review of CFD analysis of CFD analysis of centrifugal pump along with future scope for further improvement is presented in this paper. Different solver like ANSYS-CFX, FLUENT etc can be used for simulations. Shear stress transport model has been found appropriate as turbulence model. Study of pressure contours, velocity contours, flow streamlines etc can be studied by CFD techniques. Unsteady Reynolds Averaged Navier Stokes (URANS equations are solved by solver to get flow simulation results inside centrifugal pump. CFD results has to be validated with testing results or with performance characteristics curves. Performance prediction at design and off-design conditions, parametric study, cavitation analysis, diffuser pump analysis, performance of pump running in turbine mode etc. are possible with CFD simulation techniques.

  17. Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

    International Nuclear Information System (INIS)

    The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors keff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors keff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors keff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

  18. A Mode Propagation Database Suitable for Code Validation Utilizing the NASA Glenn Advanced Noise Control Fan and Artificial Sources

    Science.gov (United States)

    Sutliff, Daniel L.

    2014-01-01

    The NASA Glenn Research Center's Advanced Noise Control Fan (ANCF) was developed in the early 1990s to provide a convenient test bed to measure and understand fan-generated acoustics, duct propagation, and radiation to the farfield. A series of tests were performed primarily for the use of code validation and tool validation. Rotating Rake mode measurements were acquired for parametric sets of: (i) mode blockage, (ii) liner insertion loss, (iii) short ducts, and (iv) mode reflection.

  19. CFD computations of the second round of MEXICO rotor measurements

    DEFF Research Database (Denmark)

    Sørensen, Niels N.; Zahle, Frederik; Boorsma, K.;

    2016-01-01

    A comparison, between selected wind tunnel data from the NEW MEXICO measuring campaign and CFD computations are shown. The present work, documents that a state of the art CFD code, including a laminar turbulent transition model, can provide good agreement with experimental data. Good agreement...

  20. Modelling of Air Flow trough a Slatted Floor by CFD

    DEFF Research Database (Denmark)

    Svidt, Kjeld; Bjerg, Bjarne; Morsing, Svend;

    In this paper two different CFD-approaches are investigated to model the airflow through a slatted floor. Experiments are carried out in a full-scale test room. The computer simulations are carried out with the CFD-code FLOVENT, which solves the time-averaged Navier-Stokes equations by use of the...

  1. Containment code comparison exercise on experiment ThAI TH7

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, K.; Rastogi, A. K. [Becker Technologies GmbH, Eschborn (Germany); Braun, T.; Drath, T. [Ruhr-Universitaet Bochum, Bochum (Germany); Lyubar, A. [TU Muenchen, Garching (Germany); Schwarz, S. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH, Garching (Germany)

    2003-07-01

    The paper summarises the results of a code comparison exercise based on the containment thermal- hydraulic experiment TH7 in the ThAI test facility. Phenomena addressed in the experiment are atmospheric thermal stratification and mixing, steam injection in form of a free jet and against an impingement plate, heat transfer and steam condensation at walls, condensate collection in pools, fog formation in the atmosphere, complex geometric configuration. The test was simulated by three types of codes: the advanced lumped-parameter containment code COCOSYS, the GOTHIC code (developed for nuclear reactor applications) operating under CFD model option, and three industrial CFD codes STAR-CD, CFX and FLUENT for general simulation purposes. The comparison of blind predictions with test data indicates that a necessary requirement for realistic simulation is the availability of a reliable model for steam condensation on walls. While such model was available in the nuclear codes, it was missing in the industrial CFD codes, and efforts to implement such model via user coding were only partly successful under the given time restrictions of the exercise. Phenomena of second order importance like fog formation and transport or condensate runoff are simulated to various degrees of detail and realism, or completely omitted. For the simulation of the thermal stratification, no advantage of CFD over lumped parameter models was found. Experimental code validation data on flow velocity distributions in the foggy atmosphere are lacking due to limitations of the optical measurement systems.

  2. Containment code comparison exercise on experiment ThAI TH7

    International Nuclear Information System (INIS)

    The paper summarises the results of a code comparison exercise based on the containment thermal- hydraulic experiment TH7 in the ThAI test facility. Phenomena addressed in the experiment are atmospheric thermal stratification and mixing, steam injection in form of a free jet and against an impingement plate, heat transfer and steam condensation at walls, condensate collection in pools, fog formation in the atmosphere, complex geometric configuration. The test was simulated by three types of codes: the advanced lumped-parameter containment code COCOSYS, the GOTHIC code (developed for nuclear reactor applications) operating under CFD model option, and three industrial CFD codes STAR-CD, CFX and FLUENT for general simulation purposes. The comparison of blind predictions with test data indicates that a necessary requirement for realistic simulation is the availability of a reliable model for steam condensation on walls. While such model was available in the nuclear codes, it was missing in the industrial CFD codes, and efforts to implement such model via user coding were only partly successful under the given time restrictions of the exercise. Phenomena of second order importance like fog formation and transport or condensate runoff are simulated to various degrees of detail and realism, or completely omitted. For the simulation of the thermal stratification, no advantage of CFD over lumped parameter models was found. Experimental code validation data on flow velocity distributions in the foggy atmosphere are lacking due to limitations of the optical measurement systems

  3. A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

    Directory of Open Access Journals (Sweden)

    Jorge Pérez Mañes

    2014-01-01

    Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

  4. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  5. A CFD/CSD Interaction Methodology for Aircraft Wings

    Science.gov (United States)

    Bhardwaj, Manoj K.

    1997-01-01

    With advanced subsonic transports and military aircraft operating in the transonic regime, it is becoming important to determine the effects of the coupling between aerodynamic loads and elastic forces. Since aeroelastic effects can contribute significantly to the design of these aircraft, there is a strong need in the aerospace industry to predict these aero-structure interactions computationally. To perform static aeroelastic analysis in the transonic regime, high fidelity computational fluid dynamics (CFD) analysis tools must be used in conjunction with high fidelity computational structural fluid dynamics (CSD) analysis tools due to the nonlinear behavior of the aerodynamics in the transonic regime. There is also a need to be able to use a wide variety of CFD and CSD tools to predict these aeroelastic effects in the transonic regime. Because source codes are not always available, it is necessary to couple the CFD and CSD codes without alteration of the source codes. In this study, an aeroelastic coupling procedure is developed which will perform static aeroelastic analysis using any CFD and CSD code with little code integration. The aeroelastic coupling procedure is demonstrated on an F/A-18 Stabilator using NASTD (an in-house McDonnell Douglas CFD code) and NASTRAN. In addition, the Aeroelastic Research Wing (ARW-2) is used for demonstration of the aeroelastic coupling procedure by using ENSAERO (NASA Ames Research Center CFD code) and a finite element wing-box code (developed as part of this research).

  6. A generic data translation scheme for the coupling of high-fidelity fusion neutronics and CFD calculations

    International Nuclear Information System (INIS)

    Highlights: • A data translation scheme has been developed for coupling Monte Carlo neutronics and CFD simulations. • It contains a generic data translation kernel, and interfaces for the MCNP, CFX and Fluent code. • A blanket test case model was investigated for validation and verification purposes. • Results of the so-called Inversion Check are very close to MCNP calculated results. - Abstract: The design of fusion device components is achieved through iterative coupled neutronics and thermal hydraulics analyses. A translation scheme has been developed for transferring the nuclear heating data from Monte Carlo (MC) neutronic calculations to CFD simulations. It contains a generic data translation kernel which supports the high-fidelity data mapping of MC meshes on CFD meshes, and provides interfaces for processing the nuclear response data on the meshes for CFD codes. This translation scheme has been implemented in the open-source pre- and post-processing platform SALOME to extend its capabilities on data manipulations and visualizations. For verification purposes, a blanket test case based on the Helium Cooled Pebble Bed Test Blanket Module was investigated. The processing of the heating distribution data was validated through a so-called Inversion Check comparing the inverted heating field with the original MC tally distribution. The results of the verification have been discussed in detail, and the reliability of the data translation scheme is concluded

  7. Validation of the TRACR3D code for soil water flow under saturated/unsaturated conditions in three experiments

    International Nuclear Information System (INIS)

    Validation of the TRACR3D code in a one-dimensional form was obtained for flow of soil water in three experiments. In the first experiment, a pulse of water entered a crushed-tuff soil and initially moved under conditions of saturated flow, quickly followed by unsaturated flow. In the second experiment, steady-state unsaturated flow took place. In the final experiment, two slugs of water entered crushed tuff under field conditions. In all three experiments, experimentally measured data for volumetric water content agreed, within experimental errors, with the volumetric water content predicted by the code simulations. The experiments and simulations indicated the need for accurate knowledge of boundary and initial conditions, amount and duration of moisture input, and relevant material properties as input into the computer code. During the validation experiments, limitations on monitoring of water movement in waste burial sites were also noted. 5 references, 34 figures, 9 tables

  8. CSNI Integral test facility validation matrix for the assessment of thermal-hydraulic codes for LWR LOCA and transients

    International Nuclear Information System (INIS)

    This report deals with an internationally agreed integral test facility (ITF) matrix for the validation of best estimate thermal-hydraulic computer codes. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a life of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of such a matrix is an attempt to collect together in a systematic way the best sets of openly available test data for code validation, assessment and improvement, including quantitative assessment of uncertainties in the modelling of phenomena by the codes. In addition to this objective, it is an attempt to record information which has been generated around the world over the last 20 years so that it is more accessible to present and future workers in that field than would otherwise be the case

  9. Gasificaton Transport: A Multiphase CFD Approach & Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Dimitri Gidaspow; Veeraya Jiradilok; Mayank Kashyap; Benjapon Chalermsinsuwan

    2009-02-14

    The objective of this project was to develop predictive theories for the dispersion and mass transfer coefficients and to measure them in the turbulent fluidization regime, using existing facilities. A second objective was to use our multiphase CFD tools to suggest optimized gasifier designs consistent with aims of Future Gen. We have shown that the kinetic theory based CFD codes correctly compute: (1) Dispersion coefficients; and (2) Mass transfer coefficients. Hence, the kinetic theory based CFD codes can be used for fluidized bed reactor design without any such inputs. We have also suggested a new energy efficient method of gasifying coal and producing electricity using a molten carbonate fuel cell. The principal product of this new scheme is carbon dioxide which can be converted into useful products such as marble, as is done very slowly in nature. We believe this scheme is a lot better than the canceled FutureGen, since the carbon dioxide is safely sequestered.

  10. Turbomachinery Heat Transfer and Loss Modeling for 3D Navier-Stokes Codes

    Science.gov (United States)

    DeWitt, Kenneth; Ameri, Ali

    2005-01-01

    This report's contents focus on making use of NASA Glenn on-site computational facilities,to develop, validate, and apply models for use in advanced 3D Navier-Stokes Computational Fluid Dynamics (CFD) codes to enhance the capability to compute heat transfer and losses in turbomachiney.

  11. Experimental Measurements and CFD Simulations

    Directory of Open Access Journals (Sweden)

    Arijit A. Ganguli

    2012-01-01

    Full Text Available Bubble dynamics of a single condensing vapor bubble in a subcooled pool boiling system with a centrally heated cylindrical tank has been studied in the Rayleigh number range 7.9×1012CFD investigation has been performed using Volume of Fluid (VOF method. The heat source has been modeled using simple heat balance. The rise behavior of condensing bubbles (change in size during rise and path tracking was studied and the CFD model was validated both quantitatively and qualitatively.

  12. Source strength and dispersion of CO2 releases from high-pressure pipelines: CFD model using real gas equation of state

    International Nuclear Information System (INIS)

    Highlights: • Validated CFD models for decompression and dispersion of CO2 releases from pipelines. • Incorporation of real gas EOS into CFD code for source strength estimation. • Demonstration of better performance of SST k–ω turbulence model for jet flow. • Demonstration of better performance of real gas EOS compared to ideal gas EOS. • Demonstration of superiority of CFD models over a commercial risk assessment package. - Abstract: Transportation of CO2 in high-pressure pipelines forms a crucial link in the ever-increasing application of Carbon Capture and Storage (CCS) technologies. An unplanned release of CO2 from a pipeline presents a risk to human and animal populations and the environment. Therefore it is very important to develop a deeper understanding of the atmospheric dispersion of CO2 before the deployment of CO2 pipelines, to allow the appropriate safety precautions to be taken. This paper presents a two-stage Computational Fluid Dynamics (CFD) study developed (1) to estimate the source strength, and (2) to simulate the subsequent dispersion of CO2 in the atmosphere, using the source strength estimated in stage (1). The Peng–Robinson (PR) EOS was incorporated into the CFD code. This enabled accurate modelling of the CO2 jet to achieve more precise source strength estimates. The two-stage simulation approach also resulted in a reduction in the overall computing time. The CFD models were validated against experimental results from the British Petroleum (BP) CO2 dispersion trials, and also against results produced by the risk management package Phast. Compared with the measurements, the CFD simulation results showed good agreement in both source strength and dispersion profile predictions. Furthermore, the effect of release direction on the dispersion was studied. The presented research provides a viable method for the assessment of risks associated with CCS

  13. CFD analysis on heat transfer in low Prandtl number fluid flows

    International Nuclear Information System (INIS)

    Use of Computational Fluid Dynamics (CFD) code is helpful for designing liquid metal cooled nuclear reactor systems. Before using any CFD code proper evaluation of the code is essential for simulation of heat transfer in liquid metal flow. In this paper, a review of the literature on the correlations for liquid metal heat transfer is carried out and a comparison with experimental results is performed. CFD analysis is carried out using PHOENICS-3.6 code on heat transfer in molten Lead Bismuth Eutectic (LBE) flowing through tube. Turbulent flow analyses are carried out for the evaluation of the CFD code. The CFD results are compared with the available correlations. Assessment of various turbulence models and correlations for turbulent Prandtl number in the tube geometry are carried out. From the analysis it is found that, the CFD prediction can be improved with modified turbulent Prandtl number in the turbulence models. (author)

  14. SLURRY FLOW MODELLING BY CFD

    Directory of Open Access Journals (Sweden)

    K.C. Ghanta

    2010-12-01

    Full Text Available An attempt has been made in the present study to develop a generalized slurry flow model using CFD and utilize the model to predict concentration profile. The purpose of the CFD model is to gain better insight into the solid liquid slur¬ry flow in pipelines. Initially a three-dimensional model problem was developed to understand the influence of the particle drag coefficient on the solid concen¬tration profile. The preliminary simulations highlighted the need for correct mo¬delling of the inter phase drag force. The various drag correlations available in the literature were incorporated into a two-fluid model (Euler-Euler along with the standard k- turbulence model with mixture properties to simulate the tur¬bulent solid-liquid flow in a pipeline. The computational model was mapped on to a commercial CFD solver FLUENT6.2 (of Fluent Inc., USA. To push the en¬velope of applicability of the simulation, recent data from Kaushal (2005 (with solid concentration up to 50% was selected to validate the three dimensional simulations. The experimental data consisted of water-glass bead slurry at 125 and 440-micron particle with different flow velocity (from 1 to 5 m/s and overall concentration up to 10 to 50% by volume. The predicted pressure drop and concentration profile were validated by experimental data and showed excel-lent agreement. Interesting findings came out from the parametric study of ve-locity and concentration profiles. The computational model and results discus¬sed in this work would be useful for extending the applications of CFD models for simulating large slurry pipelines.

  15. Experimental Space Shuttle Orbiter Studies to Acquire Data for Code and Flight Heating Model Validation

    Science.gov (United States)

    Wadhams, T. P.; Holden, M. S.; MacLean, M. G.; Campbell, Charles

    2010-01-01

    thin-film resolution in both the span and chord direction in the area of peak heating. Additional objectives of this first study included: obtaining natural or tripped turbulent wing leading edge heating levels, assessing the effectiveness of protuberances and cavities placed at specified locations on the orbiter over a range of Mach numbers and Reynolds numbers to evaluate and compare to existing engineering and computational tools, obtaining cavity floor heating to aid in the verification of cavity heating correlations, acquiring control surface deflection heating data on both the main body flap and elevons, and obtain high speed schlieren videos of the interaction of the orbiter nose bow shock with the wing leading edge. To support these objectives, the stainless steel 1.8% scale orbiter model in addition to the sensors on the wing leading edge was instrumented down the windward centerline, over the wing acreage on the port side, and painted with temperature sensitive paint on the starboard side wing acreage. In all, the stainless steel 1.8% scale Orbiter model was instrumented with over three-hundred highly sensitive thin-film heating sensors, two-hundred of which were located in the wing leading edge shock interaction region. Further experimental studies will also be performed following the successful acquisition of flight data during the Orbiter Entry Boundary Layer Flight Experiment and HYTHIRM on STS-119 at specific data points simulating flight conditions and geometries. Additional instrumentation and a protuberance matching the layout present during the STS-119 boundary layer transition flight experiment were added with testing performed at Mach number and Reynolds number conditions simulating conditions experienced in flight. In addition to the experimental studies, CUBRC also performed a large amount of CFD analysis to confirm and validate not only the tunnel freestream conditions, but also 3D flows over the orbiter acreage, wing leading edge, and

  16. CFD simulation of IPR-R1 Triga subchannels fluid flow

    International Nuclear Information System (INIS)

    Computational fluid dynamics (CFD) codes have been extensively used in engineering problems, with increasing use in nuclear engineering. One of these computer codes is OpenFOAM. It is freely distributed with source code and offers a great flexibility in simulating particular conditions like those found in many problems in nuclear reactor analysis. The aim of this work is to simulate fluid flow and heat flux in three different configurations of subchannels of IPR-R1 TRIGA reactor using OpenFOAM. The data will be then validated against real experimental data obtained during the operation of the reactor at 100kW. This validation process is fundamental to allow the use of the software and associated model to simulate reactor's operation at different conditions, namely different power e fluid flow velocities. (author)

  17. CFD simulation of IPR-R1 Triga subchannels fluid flow

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Vitor V.; Santos, A.; Mesquita, Amir Z.; Silva, P.S. da, E-mail: vitors@cdtn.br, E-mail: aacs@cdtn.br, E-mail: amir@cdtn.br, E-mail: psblsg@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN - MG), Belo Horizonte, MG (Brazil); Pereira, C., E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    Computational fluid dynamics (CFD) codes have been extensively used in engineering problems, with increasing use in nuclear engineering. One of these computer codes is OpenFOAM. It is freely distributed with source code and offers a great flexibility in simulating particular conditions like those found in many problems in nuclear reactor analysis. The aim of this work is to simulate fluid flow and heat flux in three different configurations of subchannels of IPR-R1 TRIGA reactor using OpenFOAM. The data will be then validated against real experimental data obtained during the operation of the reactor at 100kW. This validation process is fundamental to allow the use of the software and associated model to simulate reactor's operation at different conditions, namely different power e fluid flow velocities. (author)

  18. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  19. Applications of CFD in Hydraulics and River Engineering

    Science.gov (United States)

    Nguyen, Van Thinh; Nestmann, Franz

    2004-02-01

    In this paper, various applications and developments of CFD technology in hydraulics and river engineering are presented. Numerical studies of three-dimensional turbulent flow fields in open channels and rivers are carried out by CFD packages such as the finite element code FIDAP and finite volume code COMET. Meshing procedures are implemented by GAMBIT or CFD-GEOM. To calculate the position of the free surface two methods are applied, free surface tracking and volume-of-fluid, and some comparisons of these methods are discussed.

  20. Simulation of plasma turbulence in scrape-off layer conditions: the GBS code, simulation results and code validation

    Science.gov (United States)

    Ricci, P.; Halpern, F. D.; Jolliet, S.; Loizu, J.; Mosetto, A.; Fasoli, A.; Furno, I.; Theiler, C.

    2012-12-01

    Based on the drift-reduced Braginskii equations, the Global Braginskii Solver, GBS, is able to model the scrape-off layer (SOL) plasma turbulence in terms of the interplay between the plasma outflow from the tokamak core, the turbulent transport, and the losses at the vessel. Model equations, the GBS numerical algorithm, and GBS simulation results are described. GBS has been first developed to model turbulence in basic plasma physics devices, such as linear and simple magnetized toroidal devices, which contain some of the main elements of SOL turbulence in a simplified setting. In this paper we summarize the findings obtained from the simulation carried out in these configurations and we report the first simulations of SOL turbulence. We also discuss the validation project that has been carried out together with the GBS development.

  1. Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety Applications

    International Nuclear Information System (INIS)

    In May 2002, an 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' was held at Aix-en-Provence, France. One of three recommended actions was the formation of this writing group to report on the need for guidelines for use of CFD in single phase Nuclear Reactor Safety (NRS) applications. CSNI approved this writing group at the end of 2002, and work began in March 2003. A final report was submitted to GAMA in September 2004, summarizing existing Best Practice Guidelines (BPG) for CFD, and recommending creation of a BPG document for Nuclear Reactor Safety (NRS) applications. The present document is intended to provide an internally complete set of guidelines for a range of single phase applications of CFD to NRS problems. However, it is not meant to be comprehensive; it is recognized that for any specific application a higher level of specificity is possible on questions of nodalization, model selection, and validation. This document should provide direct guidance on the key considerations in known single phase applications, and general directions for resolving remaining details. The intent is that it will serve as a template for further application specific (e.g. PTS, induced break) BPG documents that will provide much more detailed information and examples. The document begins with a summary of NRS related CFD analysis in countries represented by the authors. Chapter 3 deals with definition of the problem and its solution approach. This includes isolation of the portion of the NRS problem most in need of CFD, and use of a classic thermal hydraulic (TH) safety code to provide boundary conditions for the CFD based upon less detailed simulation of the balance of plant. Chapter 4 provides guidance in choosing between various options, and also discusses use of a transient calculation with tightly coupled CFD and TH codes. Chapter 5 discusses selection of physical

  2. Using CFD as a Rocket Injector Design Tool: Recent Progress at Marshall Space Flight Center

    Science.gov (United States)

    Tucker, Kevin; West, Jeff; Williams, Robert; Lin, Jeff; Canabal, Francisco; Rocker, marvin; Robles, Bryan; Garcia, Robert; Chenoweth, James

    2005-01-01

    New programs are forcing American propulsion system designers into unfamiliar territory. For instance, industry s answer to the cost and reliability goals set out by the Next Generation Launch Technology Program are engine concepts based on the Oxygen- Rich Staged Combustion Cycle. Historical injector design tools are not well suited for this new task. The empirical correlations do not apply directly to the injector concepts associated with the ORSC cycle. These legacy tools focus primarily on performance with environment evaluation a secondary objective. Additionally, the environmental capability of these tools is usually one-dimensional while the actual environments are at least two- and often three-dimensional. CFD has the potential to calculate performance and multi-dimensional environments but its use in the injector design process has been retarded by long solution turnaround times and insufficient demonstrated accuracy. This paper has documented the parallel paths of program support and technology development currently employed at Marshall Space Flight Center in an effort to move CFD to the forefront of injector design. MSFC has established a long-term goal for use of CFD for combustion devices design. The work on injector design is the heart of that vision and the Combustion Devices CFD Simulation Capability Roadmap that focuses the vision. The SRL concept, combining solution fidelity, robustness and accuracy, has been established as a quantitative gauge of current and desired capability. Three examples of current injector analysis for program support have been presented and discussed. These examples are used to establish the current capability at MSFC for these problems. Shortcomings identified from this experience are being used as inputs to the Roadmap process. The SRL evaluation identified lack of demonstrated solution accuracy as a major issue. Accordingly, the MSFC view of code validation and current MSFC-funded validation efforts were discussed in

  3. Experimental benchmark of non-local-thermodynamic-equilibrium plasma atomic physics codes; Validation experimentale des codes de physique atomique des plasmas hors equilibre thermodynamique local

    Energy Technology Data Exchange (ETDEWEB)

    Nagels-Silvert, V

    2004-09-15

    The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)

  4. Validation of an electroseismic and seismoelectric modeling code, for layered earth models, by the explicit homogeneous space solutions

    OpenAIRE

    N. Grobbe; Slob, E.C.

    2013-01-01

    We have developed an analytically based, energy fluxnormalized numerical modeling code (ESSEMOD), capable of modeling the wave propagation of all existing ElectroSeismic and SeismoElectric source-receiver combinations in horizontally layered configurations. We compare the results of several of these modeled source-receiver combinations in a homogeneous medium with explicitly derived homogeneous space Green’s function solutions, in order to be able to validate the results of ESSEMOD both in ar...

  5. Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

  6. RELAP5 code validation using a medium-size break LOCA experiment at the PMK-2 test facility

    International Nuclear Information System (INIS)

    For the analyses of loss of coolant accidents (LOCA) the thermohydraulic computer code capabilities for eastern-type reactors like VVER-440 must be validated by pre- and post test calculations of suitable experiments. Such experiments are performed on PMK-2 integral-type test facility in KFKI Atomic Energy Research Institute, Budapest, which is a volume-scaled model of the primary and secondary system of the Paks Nuclear Power Plant. One of these experiments is the pressuriser surge line break which correspond to a 22% leak. The most important phenomena of the experiment are the behavior of hot leg loop seal and the core dry-out with refill-reflood. Posttest calculations were performed by use of the code version RELAP5/mod.3.2. The results of the calculation and experiment are compared. The code properly simulate the analyzed transient.(author)

  7. Validation of the BISON 3D Fuel Performance Code: Temperature Comparisons for Concentrically and Eccentrically Located Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Hales; D. M. Perez; R. L. Williamson; S. R. Novascone; B. W. Spencer

    2013-03-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behaviour and is used to analyse either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods. Halden IFA experiments constitute a large percentage of the current BISON validation base. The validation emphasis here is centreline temperatures at the beginning of fuel life, with comparisons made to seven rods from the IFA-431 and 432 assemblies. The principal focus is IFA-431 Rod 4, which included concentric and eccentrically located fuel pellets. This experiment provides an opportunity to explore 3D thermomechanical behaviour and assess the 3D simulation capabilities of BISON. Analysis results agree with experimental results showing lower fuel centreline temperatures for eccentric fuel with the peak temperature shifted from the centreline. The comparison confirms with modern 3D analysis tools that the measured temperature difference between concentric and eccentric pellets is not an artefact and provides a quantitative explanation for the difference.

  8. Validity of the recorded codes of gonadotropin-releasing hormone agonist treatment and orchiectomies in the Danish National Patient Registry

    Directory of Open Access Journals (Sweden)

    Jespersen CG

    2012-06-01

    Full Text Available Christina Gade Jespersen,1,2 Michael Borre,1 Mette Nørgaard21Department of Urology, Aarhus University Hospital, Aarhus, Denmark; 2Department of Clinical Epidemiology, Institute of Clinical Medicine, Aarhus University Hospital, Aarhus, DenmarkPurpose: Large-scale observational studies based on existing medical databases may have an important role in studies of long-term effects of different treatments in prostate cancer patients if the coding of the treatment is valid. We therefore estimated the positive predictive value (PPV and negative predictive value (NPV of hospital codes for gonadotropin-releasing hormone (GnRH agonist treatment and orchiectomies in the Danish National Patient Registry (DNPR.Patients and methods: From Danish prostate cancer patients we selected 100 patients who were registered as users of GnRH agonists, 100 patients who were registered as nonusers of GnRH agonists, 50 patients who were registered as bilateral orchidectomized, and 50 patients who were not registered as orchidectomized in the DNPR between January 1, 2002 and December 31, 2008. From the patients' medical files we recorded codes for GnRH agonist treatment and orchiectomies, including dates of treatment from date of first prostate cancer diagnosis and onward.Results: The PPV of GnRH agonist treatment coding in the DNPR was 93% (95% confidence interval [CI]: 86.1–97.1, and the NPV was 94% (95% CI: 87.4–97.8. Both the PPV and NPV of orchiectomy coding in the DNPR were 100% (97.5% CI: 92.9–100.Conclusion: We measured the validity of codes for GnRH agonist treatment and orchiectomies in the DNPR among prostate cancer patients and found high PPV and NPV. Thus, the DNPR remains a valuable tool for clinical epidemiological studies of GnRH agonist treatment and orchiectomies in the treatment of prostate cancer.Keywords: prostate cancer, orchiectomy, positive predictive value, negative predictive value

  9. SIGACE Code for Generating High-Temperature ACE Files; Validation and Benchmarking

    Science.gov (United States)

    Sharma, Amit R.; Ganesan, S.; Trkov, A.

    2005-05-01

    A code named SIGACE has been developed as a tool for MCNP users within the scope of a research contract awarded by the Nuclear Data Section of the International Atomic Energy Agency (IAEA) (Ref: 302-F4-IND-11566 B5-IND-29641). A new recipe has been evolved for generating high-temperature ACE files for use with the MCNP code. Under this scheme the low-temperature ACE file is first converted to an ENDF formatted file using the ACELST code and then Doppler broadened, essentially limited to the data in the resolved resonance region, to any desired higher temperature using SIGMA1. The SIGACE code then generates a high-temperature ACE file for use with the MCNP code. A thinning routine has also been introduced in the SIGACE code for reducing the size of the ACE files. The SIGACE code and the recipe for generating ACE files at higher temperatures has been applied to the SEFOR fast reactor benchmark problem (sodium-cooled fast reactor benchmark described in ENDF-202/BNL-19302, 1974 document). The calculated Doppler coefficient is in good agreement with the experimental value. A similar calculation using ACE files generated directly with the NJOY system also agrees with our SIGACE computed results. The SIGACE code and the recipe is further applied to study the numerical benchmark configuration of selected idealized PWR pin cell configurations with five different fuel enrichments as reported by Mosteller and Eisenhart. The SIGACE code that has been tested with several FENDL/MC files will be available, free of cost, upon request, from the Nuclear Data Section of the IAEA.

  10. Numerical CFD Comparison of Lillgrund Employing RANS

    DEFF Research Database (Denmark)

    Simisiroglou, N.; Breton, S.-P.; Crasto, G.;

    2014-01-01

    The following article will validate the results obtained using the actuator disc method in the state of the art numerical Computational Fluid Dynamic (CFD) tool WindSim using on-site measurements from the offshore wind farm Lillgrund. WindSim solves the mass, momentum and energy conservation...

  11. Subjective Rights that may be Acquisitively Prescribed in the System of the Valid Civil Code

    OpenAIRE

    Peptan, Rodica

    2010-01-01

    From the text of art. 1837, art. 1844 and art. 1895 Civil Code, we get the idea that the acquisitive prescription is applied to the property right. Other regulations (1846 Civil Code , art. 623 and 624 Civil Code ), but, I complete that, by supporting the conclusion of the judicial doctrine and practice, namely not only the property right, but also the other main real rights – usufruct, use, occupancy, servitude and superficies – may be gained by means of the short or long term acquisitive pr...

  12. Validation of the Monteburns code for criticality calculation of TRIGA reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)

    2002-07-01

    Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)

  13. SediFoam: A general-purpose, open-source CFD-DEM solver for particle-laden flow with emphasis on sediment transport

    Science.gov (United States)

    Sun, Rui; Xiao, Heng

    2016-04-01

    With the growth of available computational resource, CFD-DEM (computational fluid dynamics-discrete element method) becomes an increasingly promising and feasible approach for the study of sediment transport. Several existing CFD-DEM solvers are applied in chemical engineering and mining industry. However, a robust CFD-DEM solver for the simulation of sediment transport is still desirable. In this work, the development of a three-dimensional, massively parallel, and open-source CFD-DEM solver SediFoam is detailed. This solver is built based on open-source solvers OpenFOAM and LAMMPS. OpenFOAM is a CFD toolbox that can perform three-dimensional fluid flow simulations on unstructured meshes; LAMMPS is a massively parallel DEM solver for molecular dynamics. Several validation tests of SediFoam are performed using cases of a wide range of complexities. The results obtained in the present simulations are consistent with those in the literature, which demonstrates the capability of SediFoam for sediment transport applications. In addition to the validation test, the parallel efficiency of SediFoam is studied to test the performance of the code for large-scale and complex simulations. The parallel efficiency tests show that the scalability of SediFoam is satisfactory in the simulations using up to O(107) particles.

  14. Validity of the International Classification of Diseases 10th revision code for hospitalisation with hyponatraemia in elderly patients

    Science.gov (United States)

    Gandhi, Sonja; Shariff, Salimah Z; Fleet, Jamie L; Weir, Matthew A; Jain, Arsh K; Garg, Amit X

    2012-01-01

    Objective To evaluate the validity of the International Classification of Diseases, 10th Revision (ICD-10) diagnosis code for hyponatraemia (E87.1) in two settings: at presentation to the emergency department and at hospital admission. Design Population-based retrospective validation study. Setting Twelve hospitals in Southwestern Ontario, Canada, from 2003 to 2010. Participants Patients aged 66 years and older with serum sodium laboratory measurements at presentation to the emergency department (n=64 581) and at hospital admission (n=64 499). Main outcome measures Sensitivity, specificity, positive predictive value and negative predictive value comparing various ICD-10 diagnostic coding algorithms for hyponatraemia to serum sodium laboratory measurements (reference standard). Median serum sodium values comparing patients who were code positive and code negative for hyponatraemia. Results The sensitivity of hyponatraemia (defined by a serum sodium ≤132 mmol/l) for the best-performing ICD-10 coding algorithm was 7.5% at presentation to the emergency department (95% CI 7.0% to 8.2%) and 10.6% at hospital admission (95% CI 9.9% to 11.2%). Both specificities were greater than 99%. In the two settings, the positive predictive values were 96.4% (95% CI 94.6% to 97.6%) and 82.3% (95% CI 80.0% to 84.4%), while the negative predictive values were 89.2% (95% CI 89.0% to 89.5%) and 87.1% (95% CI 86.8% to 87.4%). In patients who were code positive for hyponatraemia, the median (IQR) serum sodium measurements were 123 (119–126) mmol/l and 125 (120–130) mmol/l in the two settings. In code negative patients, the measurements were 138 (136–140) mmol/l and 137 (135–139) mmol/l. Conclusions The ICD-10 diagnostic code for hyponatraemia differentiates between two groups of patients with distinct serum sodium measurements at both presentation to the emergency department and at hospital admission. However, these codes underestimate the true incidence of hyponatraemia

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

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

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

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of prediting CHF at these local conditions, makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries

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

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting critical heat flux (CHF) at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is the only tool available to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries. 28 refs., 12 figs

  18. Safety analysis of MNSR reactor during reactivity insertion accident using the validated code PARET

    International Nuclear Information System (INIS)

    In the framework of the IAEA CRP project (J7.10.10) on 'Safety significance of postulated initiating events for various types of research reactors and assessment of analytical tools' the Syrian team contributed in the assessment of computational codes related to the safety analysis of research reactors. During the project implementation the codes PARET and MERSAT have been tested, modified and verified regarding specific phenomena related to safety analysis of research reactors. In the framework of this contribution the code PARET has been applied to model the core of the Syrian MNSR reactor. The code analysis includes the simulation of steady state operation and a group of selected reactivity insertion accident (RIA) including the design basis accidents dealing with the insertion of total available excess reactivity

  19. CFD Analysis for Natural Convection Induced by Steam Condensation in the THAI Facility

    International Nuclear Information System (INIS)

    A series of tests were performed to investigate this design feature of the PWR in Germany. One of the tests was to investigate the dissolution of a steam-air stratification by natural convection in the THAI (Thermal hydraulics, Hydrogen, Aerosol, and Iodine) facility of 9.2 m height and 3.2 m diameter. In addition, the test results are used as validation data for development of numerical models in the lumped parameter codes and the computational fluid dynamics (CFD) codes for simulating the multiphase flow field in the containment. In this study, STAR-CCM+ 9.04 was used to evaluate its models for simulating the dissolution of a steam-air stratification induced by the natural convection in the THAI facility. Through the comparison of the simulated results with the test results performed in the THAI facility, we found that STAR-CCM+ 9.04 with the fluid film model simulating the steam condensation predicted the steam concentration, the gas temperature, and the vessel wall temperature with an error range of about ±20%. In order to decrease the discrepancy between the CFD and test results, a detailed analysis on the fluid film model and the conjugate heat transfer through the vessel wall should be performed. Furthermore, the total calculation time should be extended to about 2000 s for better comparison between the CFD results and test data

  20. Overview on CSNI Separate Effects Test Facility Matrices for Validation of Best Estimate Thermal-Hydraulic Computer Codes

    International Nuclear Information System (INIS)

    An internationally agreed separate effects test (Set) Validation Matrix for thermal-hydraulic system codes has been established by a sub-group of the Task Group on Thermal Hydraulic System Behaviour as requested by the OECD/Nea Committee on Safety of Nuclear Installations (CSNI) Principal Working Group No. 2 on Coolant System Behaviour. The construction of such a Matrix is an attempt to collect together in a systematic way the best sets of openly available test data for code validation, assessment and improvement and also for quantitative code assessment with respect to quantification of uncertainties to the modelling of individual phenomena by the codes. The methodology that has been developed during the process of establishing CSNI-Set validation matrix was an important outcome of the work on Set matrix. In the paper, the methodology developed will be discussed in detail, together with the examples from the Set matrix. In addition, all the choices, which have been made from the 187 identified facilities covering the 67 phenomena, will be investigated together with some discussions on the data-base. Facilities and phenomena have been cross-referenced in a separate effects test cross reference matrix, while the selected separate effects tests themselves are listed against the thermal-hydraulic phenomena for which they can provide validation data. As a preliminary to the classification of facilities and test data, it was necessary to identify a sufficiently complete list of relevant phenomena for LOCA and non-LOCA transient applications of PWRs and BWRs. The majority of these phenomena are also relevant to Advanced Water Cooled Reactors. To this end, 67 phenomena were identified for inclusion in the Set matrix and, in all; about 2094 tests are included in the Set matrix. The Set matrix, as it stands, is representative of the major part of the experimental work, which has been carried out in the LWR-safety thermal hydraulics field, covering a large number of

  1. CFD modelling of subcooled flow boiling for nuclear engineering applications

    International Nuclear Information System (INIS)

    In this work a general-purpose CFD code CFX-5 was used for simulations of subcooled flow boiling. The subcooled boiling model, available in a custom version of CFX-5, uses a special treatment of the wall boiling boundary, which assures the grid invariant solution. The simulation results have been validated against the published experimental data [1] of high-pressure flow boiling in a vertical pipe covering a wide range of conditions (relevant to the pressurized water reactor). In general, a good agreement with the experimental data has been achieved. To adequately predict the lateral distribution of two-phase flow parameters, the modelling of two-phase flow turbulence and non-drag forces under wall boiling conditions have been also investigated in the paper. (author)

  2. Use of an Accurate DNS Particulate Flow Method to Supply and Validate Boundary Conditions for the MFIX Code

    Energy Technology Data Exchange (ETDEWEB)

    Zhi-Gang Feng

    2012-05-31

    The simulation of particulate flows for industrial applications often requires the use of two-fluid models, where the solid particles are considered as a separate continuous phase. One of the underlining uncertainties in the use of the two-fluid models in multiphase computations comes from the boundary condition of the solid phase. Typically, the gas or liquid fluid boundary condition at a solid wall is the so called no-slip condition, which has been widely accepted to be valid for single-phase fluid dynamics provided that the Knudsen number is low. However, the boundary condition for the solid phase is not well understood. The no-slip condition at a solid boundary is not a valid assumption for the solid phase. Instead, several researchers advocate a slip condition as a more appropriate boundary condition. However, the question on the selection of an exact slip length or a slip velocity coefficient is still unanswered. Experimental or numerical simulation data are needed in order to determinate the slip boundary condition that is applicable to a two-fluid model. The goal of this project is to improve the performance and accuracy of the boundary conditions used in two-fluid models such as the MFIX code, which is frequently used in multiphase flow simulations. The specific objectives of the project are to use first principles embedded in a validated Direct Numerical Simulation particulate flow numerical program, which uses the Immersed Boundary method (DNS-IB) and the Direct Forcing scheme in order to establish, modify and validate needed energy and momentum boundary conditions for the MFIX code. To achieve these objectives, we have developed a highly efficient DNS code and conducted numerical simulations to investigate the particle-wall and particle-particle interactions in particulate flows. Most of our research findings have been reported in major conferences and archived journals, which are listed in Section 7 of this report. In this report, we will present a

  3. Low Reynolds turbulence model CFD simulation for complex electronic system: an industrial point of view

    Science.gov (United States)

    Giannuzzi, M.

    2014-07-01

    In electronic systems the presence of bluff bodies, sharp corners and bends are the cause of flow separation and large recirculation bubbles. Since the recirculation vortices develop they encapsulate the heat from an electronic component becoming one of the major contributors of malfunction. Going in depth in this, some numerical simulations of conjugate heat transfer for a heat wall-mounted cube have been performed using the commercial CFD code scSTREAM V11 by Software Cradle Co, Ltd. It is well known that the reliability of CFD analysis depends heavily on the turbulent model employed together with the wall functions implemented. The three low- Reynolds k - epsilon turbulent models developed by Abe-Nagano-Kondoh have been validated against experimental data consisting mainly of velocity profiles and surface temperature distributions provided in literature. The performed validation shows a satisfactory agreement between the measured and simulated data. The turbulent model chosen is then used for the CFD simulation of a complex electronic system.

  4. Low Reynolds turbulence model CFD simulation for complex electronic system: an industrial point of view

    International Nuclear Information System (INIS)

    In electronic systems the presence of bluff bodies, sharp corners and bends are the cause of flow separation and large recirculation bubbles. Since the recirculation vortices develop they encapsulate the heat from an electronic component becoming one of the major contributors of malfunction. Going in depth in this, some numerical simulations of conjugate heat transfer for a heat wall-mounted cube have been performed using the commercial CFD code scSTREAM V11 by Software Cradle Co, Ltd. It is well known that the reliability of CFD analysis depends heavily on the turbulent model employed together with the wall functions implemented. The three low- Reynolds k – ε turbulent models developed by Abe-Nagano-Kondoh have been validated against experimental data consisting mainly of velocity profiles and surface temperature distributions provided in literature. The performed validation shows a satisfactory agreement between the measured and simulated data. The turbulent model chosen is then used for the CFD simulation of a complex electronic system.

  5. Synthesis of the turbulent mixing in a rod bundle with vaned spacer grids based on the OECD-KAERI CFD benchmark exercise

    International Nuclear Information System (INIS)

    Highlights: • OECD/KAERI international CFD benchmark exercise was operated by KAERI. • The purpose is to validate relevant CFD codes based on the MATiS-H experiments. • Blind calculation results were synthesized in terms of mean velocity and RMS. • Quality of control volume rather than the number of it was emphasized. • Major findings were followed OECD/NEA CSNI report. - Abstract: The second international CFD benchmark exercise on turbulent mixing in a rod bundle has been launched by OECD/NEA, to validate relevant CFD (Computational Fluid Dynamics) codes and develop problem-specific Best Practice Guidelines (BPG) based on the KAERI (Korea Atomic Energy Research Institute) MATiS-H experiments on the turbulent mixing in a 5 × 5 rod array having two different types of vaned spacer grids: split and swirl types. For this 2nd international benchmark exercise (IBE-2), the MATiS-H testing provided a unique set of experimental data such as axial and lateral velocity components, turbulent intensity, and vorticity information. Blind CFD calculation results were submitted by twenty-five (25) participants to KAERI, who is the host organization of the IBE-2, and then analyzed and synthesized by comparing them with the MATiS-H data. Based on the synthesis of the results from both the experiments and blind CFD calculations for the IBE-2, and also by comparing with the IBE-1 benchmark exercise on the mixing in a T-junction, useful information for simulating this kind of complicated physical problem in a rod bundle was obtained. And some additional Best Practice Guidelines (BPG) are newly proposed. A summary of the synthesis results obtained in the IBE-2 is presented in this paper

  6. The stability analysis using two fluids (SAT) code for boiling flow systems: Volume 4, Experiments and model validation

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.P.; Dykhuizen, R.C.; Su, M.G.; Jain, P.

    1988-12-01

    This report presents analyses of dynamic instability and frequency response characteristics of boiling flow systems based on an unequal velocity, unequal temperature two-fluid model of such flow. The dynamic instability analyses in the time domain are incorporated into three options of a computer code SAT (steady state, or equilibrium point analyses; linear stability analysis; and nonlinear analysis). The frequency response analysis is incorporated into a fourth option FREQ. Results from dynamic instability experiments carried out in a Refrigerant-113 boiling flow rig are also reported as are comparison of these with linear stability analysis predictions. Descriptions of the model, the computational techniques, the computer codes, the experiments and model validation are divided into the following volumes: Volume 1, theoretical model and computational formulation; Volume 2, coding description; Volume 3, user's manual; and Volume 4, experiments and model validation. Instability experiments run in our Refrigerant-113 boiling flow facility are described in this document. Results from these experiments are compared with predictions of the theoretical model. Instability experiment data from two other facilities and frequency response results from one are compared with theoretical model predictions also. 19 refs., 41 figs.

  7. Validation of plant dynamics analysis code Super-COPD by MONJU startup tests

    International Nuclear Information System (INIS)

    The advanced flow network models of the RV upper plenum and the IHX inlet plenum of MONJU were explained and validated by the previous SSTs. The whole plant dynamics of MONJU were also predicted using the validated flow networks. The natural circulation experiments both in the PHTS and the SHTS were conducted applying the previous SST conditions. The whole plant dynamics model with the advanced IHX model was also validated by these test results. Through these validations, we concluded that (1) the present plant dynamics model of Super-COPD could simulate the whole plant dynamics in good accuracy, which was applicable to the next SSTs, and (2) a proper multi-dimensional thermal-hydraulic analysis would be required to predict the detailed behavior of future FBR plants by Super-COPD, especially in the different sizes and conditions from MONJU. (author)

  8. TRACE code validation for BWR spray cooling injection based on GOTA facility experiments

    Energy Technology Data Exchange (ETDEWEB)

    Racca, S. [San Piero a Grado Nuclear Research Group (GRNSPG), Pisa (Italy); Kozlowski, T. [Royal Inst. of Tech., Stockholm (Sweden)

    2011-07-01

    Best estimate codes have been used in the past thirty years for the design, licensing and safety of NPP. Nevertheless, large efforts are necessary for the qualification and the assessment of such codes. The aim of this work is to study the main phenomena involved in the emergency spray cooling injection in a Swedish designed BWR. For this purpose, data from the Swedish separate effect test facility GOTA have been simulated using TRACE version 5.0 Patch 2. Furthermore, uncertainty calculations have been performed with the propagation of input errors method and the identification of the input parameters that mostly influence the peak cladding temperature has been performed. (author)

  9. CFD analysis of jet mixing in low NOx flametube combustors

    Science.gov (United States)

    Talpallikar, M. V.; Smith, C. E.; Lai, M. C.; Holdeman, J. D.

    1991-01-01

    The Rich-burn/Quick-mix/Lean-burn (RQL) combustor was identified as a potential gas turbine combustor concept to reduce NO(x) emissions in High Speed Civil Transport (HSCT) aircraft. To demonstrate reduced NO(x) levels, cylindrical flametube versions of RQL combustors are being tested at NASA Lewis Research Center. A critical technology needed for the RQL combustor is a method of quickly mixing by-pass combustion air with rich-burn gases. Jet mixing in a cylindrical quick-mix section was numerically analyzed. The quick-mix configuration was five inches in diameter and employed twelve radial-inflow slots. The numerical analyses were performed with an advanced, validated 3-D Computational Fluid Dynamics (CFD) code named REFLEQS. Parametric variation of jet-to-mainstream momentum flux ratio (J) and slot aspect ratio was investigated. Both non-reacting and reacting analyses were performed. Results showed mixing and NO(x) emissions to be highly sensitive to J and slot aspect ratio. Lowest NO(x) emissions occurred when the dilution jet penetrated to approximately mid-radius. The viability of using 3-D CFD analyses for optimizing jet mixing was demonstrated.

  10. Overview of the system alone and system/CFD coupled calculations of the PHENIX Natural Circulation Test within the THINS project

    International Nuclear Information System (INIS)

    Highlights: • The PHENIX natural convection test performed during the end of life tests program. • The calculation with system codes and theirs limits. • The calculation with coupling CFD and system code, which allows better prediction. • The tasks of code validation have been done in the frame of the THINS project. - Abstract: The PHENIX sodium cooled fast reactor started operation in 1973 and was shut down in 2009. Before decommissioning, an ultimate test program was designed and performed to provide valuable data for the development of future sodium cooled fast reactors, as the so-called Astrid prototype in France. Among these ultimate tests, a thermal-hydraulic Natural Convection Test (NCT) was set-up in June 2009. Starting from a reduced power state of 120 MWt, the NCT consists of a loss of the heat sink combined with a reactor scram and a primary pumps trip leading to stabilized natural circulation in the primary sodium system. The thermal-hydraulics innovative system project (THINS project), sponsored by the European Community in the frame of the 7th FP has selected this transient for validation of both stand-alone system code simulations and coupled simulations using system and CFD codes. Participants from three organizations (CEA, IRSN and KIT) have addressed this transient using different system codes (CATHARE, DYN2B and ATHLET) and CFD codes (TRIO-U and OPEN FOAM). The present paper depicts the different modeling approaches, methodologies and compares the numerical results with the available experimental data. Finally, the main lessons learned from the work performed within the THINS project on the PHENIX NCT with respect to code development and validation are summarized

  11. Overview of the system alone and system/CFD coupled calculations of the PHENIX Natural Circulation Test within the THINS project

    Energy Technology Data Exchange (ETDEWEB)

    Pialla, David, E-mail: david.pialla@cea.fr [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 17 rue des martyrs, 38054 Grenoble Cedex 9 (France); Tenchine, Denis [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 17 rue des martyrs, 38054 Grenoble Cedex 9 (France); Li, Simon [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 91191 Gif-sur-Yvette Cedex (France); Gauthe, Paul; Vasile, Alfredo [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DER/SESI, 13108 Saint Paul Lez Durance Cedex (France); Baviere, Roland; Tauveron, Nicolas; Perdu, Fabien [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), DEN/DM2S/STMF, 17 rue des martyrs, 38054 Grenoble Cedex 9 (France); Maas, Ludovic; Cocheme, François [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN/SEMIA/BAST, B.P. 17, 92262 Fontenay-aux-Roses Cedex (France); Huber, Klaus; Cheng, Xu [Karlsruhe Institute of Technology (KIT), Institute of Fusion and Reactor Technology (IFRT), Kaiserstraße 12, Building 07.08, 76131 Karlsruhe (Germany)

    2015-08-15

    Highlights: • The PHENIX natural convection test performed during the end of life tests program. • The calculation with system codes and theirs limits. • The calculation with coupling CFD and system code, which allows better prediction. • The tasks of code validation have been done in the frame of the THINS project. - Abstract: The PHENIX sodium cooled fast reactor started operation in 1973 and was shut down in 2009. Before decommissioning, an ultimate test program was designed and performed to provide valuable data for the development of future sodium cooled fast reactors, as the so-called Astrid prototype in France. Among these ultimate tests, a thermal-hydraulic Natural Convection Test (NCT) was set-up in June 2009. Starting from a reduced power state of 120 MWt, the NCT consists of a loss of the heat sink combined with a reactor scram and a primary pumps trip leading to stabilized natural circulation in the primary sodium system. The thermal-hydraulics innovative system project (THINS project), sponsored by the European Community in the frame of the 7th FP has selected this transient for validation of both stand-alone system code simulations and coupled simulations using system and CFD codes. Participants from three organizations (CEA, IRSN and KIT) have addressed this transient using different system codes (CATHARE, DYN2B and ATHLET) and CFD codes (TRIO-U and OPEN FOAM). The present paper depicts the different modeling approaches, methodologies and compares the numerical results with the available experimental data. Finally, the main lessons learned from the work performed within the THINS project on the PHENIX NCT with respect to code development and validation are summarized.

  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; Evaluacion del codigo de Dinamica de Fluidos Computacional (CFD) Open FOAM en el estudio del estres termico presurizado de los reactores PWR. Comparacion con el codigo comercial Ansys-CFX

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, M.; Barrachina, T.; Miro, R.; Verdu Martin, G.; Chiva, S.

    2012-07-01

    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. Implementation of a Transition Model in a NASA Code and Validation Using Heat Transfer Data on a Turbine Blade

    Science.gov (United States)

    Ameri, Ali A.

    2012-01-01

    The purpose of this report is to summarize and document the work done to enable a NASA CFD code to model laminar-turbulent transition process on an isolated turbine blade. The ultimate purpose of the present work is to down-select a transition model that would allow the flow simulation of a variable speed power turbine to be accurately performed. The flow modeling in its final form will account for the blade row interactions and their effects on transition which would lead to accurate accounting for losses. The present work only concerns itself with steady flows of variable inlet turbulence. The low Reynolds number k- model of Wilcox and a modified version of the same model will be used for modeling of transition on experimentally measured blade pressure and heat transfer. It will be shown that the k- model and its modified variant fail to simulate the transition with any degree of accuracy. A case is thus made for the adoption of more accurate transition models. Three-equation models based on the work of Mayle on Laminar Kinetic Energy were explored. The three-equation model of Walters and Leylek was thought to be in a relatively mature state of development and was implemented in the Glenn-HT code. Two-dimensional heat transfer predictions of flat plate flow and two-dimensional and three-dimensional heat transfer predictions on a turbine blade were performed and reported herein. Surface heat transfer rate serves as sensitive indicator of transition. With the newly implemented model, it was shown that the simulation of transition process is much improved over the baseline k- model for the single Reynolds number and pressure ratio attempted; while agreement with heat transfer data became more satisfactory. Armed with the new transition model, total-pressure losses of computed three-dimensional flow of E3 tip section cascade were compared to the experimental data for a range of incidence angles. The results obtained, form a partial loss bucket for the chosen blade

  14. Validity of diagnostic codes and laboratory measurements to identify patients with idiopathic acute liver injury in a hospital database

    DEFF Research Database (Denmark)

    Udo, Renate; Maitland-van der Zee, Anke H; Egberts, Toine C G;

    2016-01-01

    of liver enzyme values (ALT > 2× upper limit of normal (ULN); AST > 1ULN + AP > 1ULN + bilirubin > 1ULN; ALT > 3ULN; ALT > 3ULN + bilirubin > 2ULN; ALT > 10ULN) and (II) algorithms based on solely liver enzyme values (ALT > 3ULN + bilirubin > 2ULN; ALT > 10ULN). Hospital medical records were reviewed......PURPOSE: The development and validation of algorithms to identify cases of idiopathic acute liver injury (ALI) are essential to facilitate epidemiologic studies on drug-induced liver injury. The aim of this study is to determine the ability of diagnostic codes and laboratory measurements...... 32% (13/41) to 48% (43/90) with the highest PPV found with ALT > 2ULN. The PPV for (II) algorithms with liver test abnormalities was maximally 26% (150/571). CONCLUSIONS: The algorithm based on ICD-9-CM codes indicative of ALI combined with abnormal liver-related laboratory tests is the most...

  15. Qualification of CFD-models for multiphase flows

    Energy Technology Data Exchange (ETDEWEB)

    Lucas, Dirk [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany)

    2016-05-15

    While Computational Fluid Dynamics (CFD) is already an accepted industrial tool for single phase flows it is not yet mature for two-phase flows. For this reason the qualification of CFD for reactor safety relevant applications which involve multiphase flows is a present topic of research. At the CFD division of Helmholtz-Zentrum Dresden-Rossendorf (HZDR) hereby beside an application-oriented model development and validation also more generic investigations are done. Thus, the baseline model strategy aims on the consolidation of the CFD-modelling for multiphase to enable reliable predictions for well-defined flow pattern in future. In addition the recently developed GENTOP-concept broadens the range of applicability of CFD. Different flow morphologies including transitions between them can be considered in frame of this concept.

  16. Qualification of CFD-models for multiphase flows

    International Nuclear Information System (INIS)

    While Computational Fluid Dynamics (CFD) is already an accepted industrial tool for single phase flows it is not yet mature for two-phase flows. For this reason the qualification of CFD for reactor safety relevant applications which involve multiphase flows is a present topic of research. At the CFD division of Helmholtz-Zentrum Dresden-Rossendorf (HZDR) hereby beside an application-oriented model development and validation also more generic investigations are done. Thus, the baseline model strategy aims on the consolidation of the CFD-modelling for multiphase to enable reliable predictions for well-defined flow pattern in future. In addition the recently developed GENTOP-concept broadens the range of applicability of CFD. Different flow morphologies including transitions between them can be considered in frame of this concept.

  17. Investigation on Improved Correlation of CFD and EFD for Supercritical Airfoil

    OpenAIRE

    Xin Xu; Da-wei Liu; De-hua Chen; Zhi Wei; Yuan-jing Wang

    2014-01-01

    It is necessary to improve the correlation between CFD and EFD through the correction of EFD results and validation of CFD method, thus investigating the aerodynamic characteristics of supercritical airfoil perfectly. In this study, NASA SC (2) -0714 airfoil is numerically simulated and compared with NASA corrected experimental results to validate the CFD method. The Barnwell-Sewell method is applied to correct sidewall effects for experimental results of typical supercritical airfoil CH obta...

  18. CFD modeling of a boiler's tubes rupture

    International Nuclear Information System (INIS)

    This paper reports the results of a study on the reason for tubes damage in the superheater Platen section of the 320 MW Bisotoun power plant, Iran. The boiler has three types of superheater tubes and the damage occurs in a series of elbows belongs to the long tubes. A three-dimensional modeling was performed using an in-house computational fluid dynamics (CFD) code in order to explore the reason. The code has ability of simultaneous solving of the continuity, the Reynolds-Averaged Navier-Stokes (RANS) equations and employing the turbulence, combustion and radiation models. The whole boiler including; walls, burners, air channels, three types of tubes, etc., was modeled in the real scale. The boiler was meshed into almost 2,000,000 tetrahedral control volumes and the standard k-ε turbulence model and the Rosseland radiation model were used in the model. The theoretical results showed that the inlet 18.9 MPa saturated steam becomes superheated inside the tubes and exit at a pressure of 17.8 MPa. The predicted results showed that the temperature of the steam and tube's wall in the long tubes is higher than the short and medium size tubes. In addition, the predicted steam mass flow rate in the long tube was lower than other ones. Therefore, it was concluded that the main reason for the rupture in the long tubes elbow is changing of the tube's metal microstructure due to working in a temperature higher than the design temperature. In addition, the structural fatigue tension makes the last elbow of the long tube more ready for rupture in comparison with the other places. The concluded result was validated by observations from the photomicrograph of the tube's metal samples taken from the damaged and undamaged sections

  19. 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...... uses a standard k-ε turbulence model. Simulations comprise both steady-state and dynamic approaches. Several boundary conditions are tested. Finally, the paper discusses the prospects of simple CFD models in smoke ventilation design including the inherent limitations....

  20. Relative validity of the pre-coded food diary used in the Danish National Survey of Diet and Physical Activity

    DEFF Research Database (Denmark)

    Knudsen, Vibeke Kildegaard; Gille, Maj-Britt; Nielsen, Trine Holmgaard;

    2011-01-01

    weighed food record. Intakes of foods and drinks were estimated, and nutrient intakes were calculated. Means and medians of intake were compared, and crossclassification of individuals according to intake was performed. To assess agreement between the two methods, Pearson and Spearman’s correlation...... coefficients and weighted kappa coefficients were calculated. Setting: Validation study of the pre-coded food diary against a 4 d weighed food record. Subjects: Seventy-two volunteer, healthy free-living adults (thirty-five males, thirty-seven females). Results: Intakes of cereals and vegetables were higher...

  1. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    International Nuclear Information System (INIS)

    of the flow and energy transport as applied to nuclear reactor safety. However, it is expected that these practices and procedures will require updating from time to time as research and development affect them or replace them with better procedures. The practices and procedures are categorized into five groups. These are: (1) Code Verification; (2) Code and Calculation Documentation; (3) Reduction of Numerical Error; (4) Quantification of Numerical Uncertainty (Calculation Verification); and (5) Calculation Validation. These five categories have been identified from procedures currently required of CFD simulations such as those required for publication of a paper in the ASME Journal of Fluids Engineering and from the literature such as Roache [1998]. Code verification refers to the demonstration that the equations of fluid and energy transport have been correctly coded in the CFD code. Code and calculation documentation simply means that the equations and their discretizations, etc., and boundary and initial conditions used to pose the fluid flow problem are fully described in available documentation. Reduction of numerical error refers to practices and procedures to lower numerical errors to negligible or very low levels as is reasonably possible (such as avoiding use of first-order discretizations). The quantification of numerical uncertainty is also known as calculation verification. This means that estimates are made of numerical error to allow the characterization of the numerical results with a certain confidence level. Numerical error in this case does not include error due to models such as turbulence models. Calculation validation is the process of comparing simulation results to experimental data to demonstrate level of agreement. Validation does include the effects of modeling errors as well as numerical and experimental errors. A key issue in the validation process of numerical results is the existence of appropriate experimental data to use for

  2. Validation of the HZETRN code for laboratory exposures with 1A GeV iron ions in several targets.

    Science.gov (United States)

    Walker, S A; Tweed, J; Wilson, J W; Cucinotta, F A; Tripathi, R K; Blattnig, S; Zeitlin, C; Heilbronn, L; Miller, J

    2005-01-01

    A new version of the HZETRN code capable of validation with HZE ions in either the laboratory or the space environment is under development. The computational model consists of the lowest order asymptotic approximation followed by a Neumann series expansion with non-perturbative corrections. The physical description includes energy loss with straggling, nuclear attenuation, nuclear fragmentation with energy dispersion and downshift. Measurements to test the model were performed at the Alternating Gradient Synchrotron and the NASA Space Radiation Laboratory at Brookhaven National Laboratory with iron ions. Surviving beam particles and produced fragments were measured with solid-state detectors. Beam analysis software has been written to relate the computational results to the measured energy loss spectra of the incident ions for rapid validation of modeled target transmission functions.

  3. Flow induced vibration forces on a fuel rod by LES CFD analysis

    International Nuclear Information System (INIS)

    The purpose of the present study is to evaluate the feasibility of use of CFD Large Eddy Simulation (LES) modeling techniques in CD-adapco CFD code STAR-CCM+ to calculate the instantaneous stress tensor on the fuel rod wall and then utilize these data for mechanical calculations. Transient hydraulic forces on the fuel rod resulting from the CFD model are linked to the Westinghouse VITRAN code to predict fuel rod vibration response. The coupled CFD/mechanical solution has provided a reasonable prediction of fuel rod vibration and a more accurate representation of all the important physics and excitation forces. (author)

  4. TRIPOLI-4{sup ®} Monte Carlo code ITER A-lite neutronic model validation

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, Jean-Charles, E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Cayla, Pierre-Yves; Fausser, Clement [MILLENNIUM, 16 Av du Québec Silic 628, F-91945 Villebon sur Yvette (France); Damian, Frederic; Lee, Yi-Kang; Puma, Antonella Li; Trama, Jean-Christophe [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France)

    2014-10-15

    3D Monte Carlo transport codes are extensively used in neutronic analysis, especially in radiation protection and shielding analyses for fission and fusion reactors. TRIPOLI-4{sup ®} is a Monte Carlo code developed by CEA. The aim of this paper is to show its capability to model a large-scale fusion reactor with complex neutron source and geometry. A benchmark between MCNP5 and TRIPOLI-4{sup ®}, on the ITER A-lite model was carried out; neutron flux, nuclear heating in the blankets and tritium production rate in the European TBMs were evaluated and compared. The methodology to build the TRIPOLI-4{sup ®} A-lite model is based on MCAM and the MCNP A-lite model. Simplified TBMs, from KIT, were integrated in the equatorial-port. A good agreement between MCNP and TRIPOLI-4{sup ®} is shown; discrepancies are mainly included in the statistical error.

  5. Simulation of the ICE P1 test for a validation of COCOSYS and ASTEC codes

    International Nuclear Information System (INIS)

    Paper presents simulation of the upgraded ICE facility test P1 performed using COCOSYS and ASTEC lumped parameter codes. The experimental ICE facility was constructed at the Japan Atomic Energy Research Institute (JAERI) to obtain experimental data on the ingress-of-coolant event which could occur in the ITER. COCOSYS and ASTEC are computer codes used to simulate accident scenarios in the light water reactors. However, they’ve been also used for fusion safety applications and ASTEC is being extended to better address main accident sequences of fusion installations. It is shown in the paper that pressure and temperature evolutions of the main volumes were predicted with reasonable accuracy, however discrepancies in the suppression and drain volumes suggest underestimated vaporization rate

  6. Prediction of steam condensation in the presence of noncondensable gases using a CFD-based approach

    Energy Technology Data Exchange (ETDEWEB)

    Dehbi, A., E-mail: abdel.dehbi@psi.ch [Laboratory for Thermal-Hydraulics, Paul Scherrer Institut, Villigen 5232 (Switzerland); Janasz, F., E-mail: filip.janasz@psi.ch [Laboratory for Thermal-Hydraulics, Paul Scherrer Institut, Villigen 5232 (Switzerland); Bell, B., E-mail: brian.bell@ansys.com [ANSYS Inc., Lebanon, NH 03766 (United States)

    2013-05-15

    Highlights: ► A model of condensation with noncondensable gases is integrated in the Fluent code. ► Condensation is modeled as sink terms in the conservation equations. ► A best-estimate parameter is proposed for heat transfer enhancement due to suction. ► Validation is conducted for a wide range of flow conditions and geometries. ► Predictions are in good agreement with experimental correlations. -- Abstract: We integrate in the ANSYS CFD code Fluent a model for wall condensation from a vapor–noncondensable gas mixture. The condensation phenomenon is modeled from first principles as sink terms for the mass, momentum, species and energy conservation equations. The condensation rate is obtained by requiring the condensate–gas interface to be impermeable to the noncondensable gas. The model assumes in addition that the thermal resistance of the liquid film is negligible, and hence the predictions are only valid for relatively large mass fractions of the noncondensable gas (above 0.1). When the condensation rates are high, a best-estimate suction correction factor is proposed for CFD codes that impose the no-slip boundary conditions at the wall surfaces. In such a way, the enhancement in the heat transfer due to suction is accounted for. We first simulate condensation in laminar and turbulent forced flows along a cold flat plate. More challenging simulations are subsequently conducted for the case where vapor is introduced into closed vessels containing a noncondensable gas and in which stand condensing surfaces held at constant cold temperature. The flow transient is computed until steady conditions are reached, at which point the condensation flow rate equals the injected steam flow rate. Overall, the predicted heat transfer rates are in good agreement with available analytical solutions as well as experimental correlations. CFD Best Practice Guidelines are followed to a large extent. In particular, a hierarchy of grids is used to ensure mesh

  7. Methods of fluid properties for compressible refrigerant CFD analysis

    OpenAIRE

    Branch, Scott

    2014-01-01

    There are numerous ways of defining fluids properties for computational fluid dynamic (CFD) simulations. This paper examines three methods of defining fluid properties that are available in a commercial CFD code. Simulations were carried out for an R410a scroll compressor used in air conditioning applications and the effects are illustrated through the compression process. The first method used is the real gas property data from the National Institute of Standards and Technology (NIST). While...

  8. A computational design system for rapid CFD analysis

    Science.gov (United States)

    Ascoli, E. P.; Barson, S. L.; Decroix, M. E.; Sindir, Munir M.

    1992-01-01

    A computation design system (CDS) is described in which these tools are integrated in a modular fashion. This CDS ties together four key areas of computational analysis: description of geometry; grid generation; computational codes; and postprocessing. Integration of improved computational fluid dynamics (CFD) analysis tools through integration with the CDS has made a significant positive impact in the use of CFD for engineering design problems. Complex geometries are now analyzed on a frequent basis and with greater ease.

  9. Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels

    International Nuclear Information System (INIS)

    Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)

  10. Validation of Framework Code Approach to a Life Prediction System for Fiber Reinforced Composites

    Science.gov (United States)

    Gravett, Phillip

    1997-01-01

    The grant was conducted by the MMC Life Prediction Cooperative, an industry/government collaborative team, Ohio Aerospace Institute (OAI) acted as the prime contractor on behalf of the Cooperative for this grant effort. See Figure I for the organization and responsibilities of team members. The technical effort was conducted during the period August 7, 1995 to June 30, 1996 in cooperation with Erwin Zaretsky, the LERC Program Monitor. Phil Gravett of Pratt & Whitney was the principal technical investigator. Table I documents all meeting-related coordination memos during this period. The effort under this grant was closely coordinated with an existing USAF sponsored program focused on putting into practice a life prediction system for turbine engine components made of metal matrix composites (MMC). The overall architecture of the NMC life prediction system was defined in the USAF sponsored program (prior to this grant). The efforts of this grant were focussed on implementing and tailoring of the life prediction system, the framework code within it and the damage modules within it to meet the specific requirements of the Cooperative. T'he tailoring of the life prediction system provides the basis for pervasive and continued use of this capability by the industry/government cooperative. The outputs of this grant are: 1. Definition of the framework code to analysis modules interfaces, 2. Definition of the interface between the materials database and the finite element model, and 3. Definition of the integration of the framework code into an FEM design tool.

  11. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  12. Calibration and Validation of the Dynamic Wake Meandering Model for Implementation in an Aeroelastic Code

    DEFF Research Database (Denmark)

    Aagaard Madsen, Helge; Larsen, Gunner Chr.; Larsen, Torben J.;

    2010-01-01

    in an aeroelastic model. Calibration and validation of the different parts of the model is carried out by comparisons with actuator disk and actuator line (ACL) computations as well as with inflow measurements on a full-scale 2 MW turbine. It is shown that the load generating part of the increased turbulence...

  13. Turbulence Modeling Verification and Validation

    Science.gov (United States)

    Rumsey, Christopher L.

    2014-01-01

    steps in the process. Verification insures that the CFD code is solving the equations as intended (no errors in the implementation). This is typically done either through the method of manufactured solutions (MMS) or through careful step-by-step comparisons with other verified codes. After the verification step is concluded, validation is performed to document the ability of the turbulence model to represent different types of flow physics. Validation can involve a large number of test case comparisons with experiments, theory, or DNS. Organized workshops have proved to be valuable resources for the turbulence modeling community in its pursuit of turbulence modeling verification and validation. Workshop contributors using different CFD codes run the same cases, often according to strict guidelines, and compare results. Through these comparisons, it is often possible to (1) identify codes that have likely implementation errors, and (2) gain insight into the capabilities and shortcomings of different turbulence models to predict the flow physics associated with particular types of flows. These are valuable lessons because they help bring consistency to CFD codes by encouraging the correction of faulty programming and facilitating the adoption of better models. They also sometimes point to specific areas needed for improvement in the models. In this paper, several recent workshops are summarized primarily from the point of view of turbulence modeling verification and validation. Furthermore, the NASA Langley Turbulence Modeling Resource website is described. The purpose of this site is to provide a central location where RANS turbulence models are documented, and test cases, grids, and data are provided. The goal of this paper is to provide an abbreviated survey of turbulence modeling verification and validation efforts, summarize some of the outcomes, and give some ideas for future endeavors in this area.

  14. CFD analysis of pump consortium impeller

    Science.gov (United States)

    Cheng, Gary C.; Chen, Y. S.; Williams, R. W.

    1992-01-01

    Current design of high performance turbopumps for rocket engines requires effective and robust analytical tools to provide design impact in a productive manner. The main goal of this study is to develop a robust and effective computational fluid dynamics (CFD) pump model for general turbopump design and analysis applications. A Navier-Stokes flow solver, FDNS, embedded with the extended k-epsilon turbulence model and with appropriate moving interface boundary conditions, is developed to analyze turbulent flows in the turbomachinery devices. The FDNS code was benchmarked with its numerical predictions of the pump consortium inducer, and provides satisfactory results. In the present study, a CFD analysis of the pump consortium impeller will be conducted with the application of the FDNS code. The pump consortium impeller, with partial blades, is the new design concept of the advanced rocket engine.

  15. Reproducible and replicable CFD: it's harder than you think

    CERN Document Server

    Mesnard, Olivier

    2016-01-01

    Completing a full replication study of our previously published findings on bluff-body aerodynamics was harder than we thought. Despite the fact that we have good reproducible-research practices, sharing our code and data openly. Here's what we learned from three years, four CFD codes and hundreds of runs.

  16. A CFD/CSD interaction methodology for aircraft wings

    Energy Technology Data Exchange (ETDEWEB)

    Bhardwaj, M.K.; Kapania, R.K. [Virginia Polytechnic Inst. and State Univ., Blacksburg, VA (United States); Reichenbach, E. [Boeing Co., St. Louis, MO (United States); Guruswamy, G.P. [NASA, Moffett Field, CA (United States). Ames Research Center

    1998-01-01

    With advanced subsonic transports and military aircraft operating in the transonic regime, it is becoming important to determine the effects of the coupling between aerodynamic loads and elastic forces. Since aeroelastic effects can significantly impact the design of these aircraft, there is a strong need in the aerospace industry to predict these interactions computationally. Such an analysis in the transonic regime requires high fidelity computational fluid dynamics (CFD) analysis tools, due to the nonlinear behavior of the aerodynamics in the transonic regime and also high fidelity computational structural dynamics (CSD) analysis tools. Also, there is a need to be able to use a wide variety of CFD and CSD methods to predict aeroelastic effects. Since source codes are not always available, it is necessary to couple the CFD and CSD codes without alteration of the source codes. In this study, an aeroelastic coupling procedure is developed to determine the static aeroelastic response of aircraft wings using any CFD and CSD code with little code integration. The aeroelastic coupling procedure is demonstrated on an F/A-18 Stabilator using NASTD (an in-house McDonnell Douglas CFD code) and NASTRAN. In addition, the Aeroelastic Research Wing (ARW-2) is used for demonstration of the aeroelastic coupling procedure by using ENSAERO (NASA Ames Research Center CFD code) and a finite element wing-box code. The results obtained from the present study are compared with those available from an experimental study conducted at NASA Langley Research Center and a study conducted at NASA Ames Research Center using ENSAERO and modal superposition. The results compare well with experimental data.

  17. Idaho National Laboratory Experimental Program to Measure the Flow Phenomena in a Scaled Model of a Prismatic Gas-Cooled Reactor Lower Plenum for Validation of CFD Codes

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

    2008-09-01

    The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal developing, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet velocity profiles is also presented.

  18. CFD design analysis of ventilated disc brakes

    OpenAIRE

    Pulugundla, Gautam

    2008-01-01

    This thesis reports the numerical investigation of the automotive ventilated disc brake rotor. Disc brakes operate on the principle of friction by converting kinetic energy into heat energy. The main objective of a disc brake rotor is to store this heat energy and dissipate it as soon as possible. This work is carried out in a area where there is very limited understanding. Commercial CFD code FLUENT was used for carrying out the simulations with the rotor rotating in still ...

  19. Investigation of a two-phase nozzle flow and validation of several computer codes by the experimental data

    International Nuclear Information System (INIS)

    Stationary experiments with a convergent nozzle are performed in order to validate advanced two-phase computer codes, which find application in the blowdown-phase of a loss-of-coolant accident (LOCA). The steam/water flow presents a broad variety of initial conditions: The pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of subcritical as well as critical experiments with different flow pattern is investigated. Additional air/water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiments. The layout of the nozzle and the applied measurement technique allow for a separate testing of physical models and the determination of empirical model parameters, respectively: In the four codes DUESE, DRIX-20, RELAP4/MOD6 and STRUYA the models - if they exist - for slip between the phases, thermodynamic non-equilibrium, pipe friction and critical mass flow rate are validated and criticised in comparison with the experimental data, and the corresponding model parameters are determined. The parameters essentially are a function of the void fraction. (orig.)

  20. Validation of fast reactor thermomechanical and thermohydraulic codes. Final report of a co-ordinated research project. 1996-1999

    International Nuclear Information System (INIS)

    This report is a summary of the work performed under a co-ordinated research project (CRP) entitled Harmonization and Validation of Fast Reactor Thermomechanical and Thermo-Hydraulic Codes and Relations using Experimental Data. The project was organized by the IAEA on the recommendation of the IAEA's Technical Working Group on Fast Reactors (TWGFR) and carried out from 1996 to 1999. In certain conditions, temperature fluctuations in the coolant close to a structure caused by thermal striping can lead to thermomechanical damage to structures. Institutes from a number of Member States have an interest in improving engineering tools and prediction techniques concerning the characterization of the thermal striping effects, in which numerical models have a major role. Therefore, the IAEA through its advanced reactor technology development programme supports the activities of Member States in this area. Design analyses applied to thermal striping phenomena need to be firmly established, and the CRP provided a valuable tool in assessing their reliability. Eleven institutes from France, India, Italy, Japan, the Republic of Korea, the Russian Federation and the United Kingdom co-operated in this CRP. This report documents the CRP activities, provides the main results and recommendations and includes the work carried out by the research groups at the participating institutes within the CRP on harmonization and validation of fast reactor thermomechanical and thermohydraulic codes and relations

  1. Thermal hydraulic simulations, error estimation and parameter sensitivity studies in Drekar::CFD

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Thomas Michael; Shadid, John N; Pawlowski, Roger P; Cyr, Eric C; Wildey, Timothy Michael

    2014-01-01

    This report describes work directed towards completion of the Thermal Hydraulics Methods (THM) CFD Level 3 Milestone THM.CFD.P7.05 for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Nuclear Hub effort. The focus of this milestone was to demonstrate the thermal hydraulics and adjoint based error estimation and parameter sensitivity capabilities in the CFD code called Drekar::CFD. This milestone builds upon the capabilities demonstrated in three earlier milestones; THM.CFD.P4.02 [12], completed March, 31, 2012, THM.CFD.P5.01 [15] completed June 30, 2012 and THM.CFD.P5.01 [11] completed on October 31, 2012.

  2. A study on the flow characteristics of a direct drive turbine for energy conversion generation by experiment and CFD

    Science.gov (United States)

    Cho, Y. J.; Zullah, M. A.; Faizal, M.; Choi, Y. D.; Lee, Y. H.

    2012-11-01

    A variety of technologies has been proposed to capture the energy from waves. Some of the more promising designs are undergoing demonstration testing at commercial scales. Due to the complexity of most offshore wave energy devices and their motion response in different sea states, physical tank tests are common practice for WEC design. Full scale tests are also necessary, but are expensive and only considered once the design has been optimized. Computational Fluid Dynamics (CFD) is now recognized as an important complement to traditional physical testing techniques in offshore engineering. Once properly calibrated and validated to the problem, CFD offers a high density of test data and results in a reasonable timescale to assist with design changes and improvements to the device. The purpose of this study is to investigate the performance of a newly developed direct drive hydro turbine (DDT), which will be built in a caisson for extraction of wave energy. Experiments and CFD analysis are conducted to clarify the turbine performance and internal flow characteristics. The results show that commercial CFD code can be applied successfully to the simulation of the wave motion in the water tank. The performance of the turbine for wave energy converter is studied continuously for a ongoing project.

  3. Validating and Verifying a New Thermal-Hydraulic Analysis Tool

    International Nuclear Information System (INIS)

    The Idaho National Engineering and Environmental Laboratory (INEEL) has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3DC/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 RELAP5-3DC/ATHENA. Both steady-state and transient calculations can be performed, using many working fluids and point to three-dimensional neutronics. A general description of the techniques used to couple the codes is given. The validation and verification (V and V) matrix is outlined. V and V is presently ongoing. (authors)

  4. Validation of eureka-2/rr code for analysis of pulsing parameters of triga mark ii research reactor in bangladesh

    International Nuclear Information System (INIS)

    Some parametric studies on pulsing mode for fresh core of TRIGA Mark II research reactor in AERE, Savar, have been carried out with coupled thermal-hydraulics code EUREKA-2/RR in association with neutronics code SRAC. At the beginning, role of some important parameters in pulsing like delayed neutron fraction (beta eff) and reactivity insertion have been studied keeping prompt neutron life time (lp) fixed at 33.4 micro-sec. After a series of experiments, we found that the pulsing peak that is consistent with the Safety Analysis Report (SAR) is for the delayed neutron fraction (beta eff) of 0.007 and reactivity insertion of 2. Study has determined the pulsing peak of the fresh core for this particular condition to be 857.86 MW which is 852 MW according to SAR. Experiment also shows the pulsing peak increases with the increase of reactivity insertion whereas decreases with increase of delayed neutron fraction. With the utilization of the particular values of these parameters, pulsing parameters like prompt energy released, reactor period, pulse width at half maxima, alongwith safety parameters including peak power and clad maximum temperature, have been analyzed. The clad maximum temperature for fresh core is simulated to be 144.54 MW, which is much less than the SAR Value, ensuring the validity of codes and the safety of pulsing in that particular condition. (author)

  5. Preliminary result of validation study in SAS-SFR (SAS4A) code in simulated top and undercooled overpower conditions

    International Nuclear Information System (INIS)

    SAS-SFR(SAS4A) is presently the most advanced computer code for simulating the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors (SFR). Intensive model improvement and validation effort have been performed for SAS-SFR(SAS4A) utilizing the experimental data from the CABRI programs. A PIRT (Phenomena Identification and Ranking Table) for CDA of MOX-fueled fast reactors has been developed under an international collaboration framework between JAEA and European organizations (CEA and KIT). Important phenomena were identified for respective typical CDA categories. Based on this PIRT, a systematic and comprehensive validation test matrix was defined and validation work has been started. The first systematic and comprehensive validation work for SAS-SFR(SAS4A) was conducted focusing on the fuel failure position under Transient Overpower (TOP) and transient undercooled overpower conditions. The fuel failure position is an important parameter, which dominates the analytical results and has a meaningful effect on the release energy of the initiating phase of a reactor calculation. The typical CABRI experiments, AI3 and BI4 were selected for this validation. In the calculations of the selected CABRI experiments, the fuel failure positions under TOP and transient undercooled overpower were reproduced, using the failure prediction model of SAS-SFR(SAS4A). The mechanistic behavior of these results agrees with the experimental interpretation on the failure location in these experiments, in which pressurized molten fuel cavity could cause cladding failure at the most weakened position due to its elevated temperature. It was confirmed that the failure prediction model of SAS-SFR(SAS4A) could reproduce the typical CABRI experiment results. (author)

  6. Best Practice Guidelines for the Use of CFD in Nuclear Reactor Safety Applications - Revision

    International Nuclear Information System (INIS)

    This document is intended to provide an internally complete set of guidelines for a range of single phase applications of CFD to NRS problems. However, it is not meant to be comprehensive. We recognize that for any specific application a higher level of specificity is possible on questions of nodalization, model selection, and validation. This document should provide direct guidance on the key considerations in known single phase applications, and general directions for resolving remaining details. After review of other Best Practice Guidelines, and discussion with many CFD practitioners and developers, we have assembled guidance covering a fully verified and validated NRS analysis. The document begins with a summary of NRS related CFD analysis in countries represented by the authors, to give a feeling for the existing range of experience. Some key terminology in the field is defined in the field. These definitions are not meant simply for novices, but also provide experienced users with an understanding of how some terms (e.g. verification and validation) are used within this document. Chapter 3 deals with definition of the problem and its solution approach. This includes isolation of the portion of the NRS problem most in need of CFD, and use of a classic thermal-hydraulic (TH) safety code to provide boundary conditions for the CFD based upon less detailed simulation of the balance of plant. The chapter discusses the Phenomena Identification and Ranking Table (PIRT) process, which identifies phenomena critical to the problem, provides a basis for selection of an appropriate simulation tool, and establishes the foundation for the validation process needed for confidence in final results. The chapter also discusses theory and modelling needs associated with a number of special phenomena important to NRS but not commonly modelled in the CFD community. Chapter 5 discusses selection of physical models available as user options. As is appropriate for single phase CFD

  7. RELAP5 code study of ROSA/LSTF validation tests for PWR safety system using SG secondary-side depressurization

    International Nuclear Information System (INIS)

    RELAP5 code post-test analyses were performed on two ROSA/large scale test facility (LSTF) validation tests for PWR safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using steam generator (SG) secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs a little after a safety injection signal. Auxiliary feedwater injection was done into the secondary-side of both SGs thereafter. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by boil-off. Core collapsed liquid level recovered after the initiation of accumulator (ACC) coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection (LPI) system. In the 4-in. break test, on the other hand, no core uncovery and heatup happened due to the coolant injection from the ACC and LPI systems. Adjustment of break discharge coefficient for two-phase discharge flow predicted the break flow rate reasonably well. The code predicted well the overall trend of the major thermal-hydraulic response observed in the two LSTF tests. The code, however, overpredicted the peak cladding temperature (PCT) because of underprediction of the core collapsed liquid level due to inadequate prediction of the ACC flow rate in the 8-in. break case. Sensitivity analyses with the RELAP5 code indicated that a time delay for the SG depressurization start and break discharge coefficient for two-phase discharge flow affect the PCT significantly in the 8-in. break case. (author)

  8. Verification and validation of LMFBR static core mechanics codes. Pt. 2

    International Nuclear Information System (INIS)

    Although core static mechanics had received sporadic attention at international gatherings in the 1970s (e.g. the SMIRT series), the first major coordinated international review was organized as an IWGFR Specialists' Meeting which was held in the United Kingdom at Wilmslow, Cheshire in October 1984. The problems of structural analysis raised by core static mechanics were novel and difficult. They involved analysing the non-linear interactive behaviour of hundreds of wrappers separated by gaps which might be open or closed. Because this problem had not responded to conventional analysis, each country had set about solving this ''discontinuum'' problem by specialist coding. The current document presents proceedings of the two Research Coordination Meetings held in Vienna (March 1987) and in Oarai, Japan (May 1989). The proceeding of the first meeting contains 11 presentations and the second one has 14 presentations. A separate abstract was prepared for each of these 25 papers. Refs, figs and tabs

  9. CFD Technology for Rotorcraft Gearbox Windage Aerodynamics Simulation

    Science.gov (United States)

    Handschuh, Robert; Hill, Matthew; Kunz, Robert; Long, Lyle; Morris, Philip; Noack, Ralph

    2009-01-01

    A computational fluid dynamics (CFD) method is adapted, validated and applied to spinning gear systems with emphasis on predicting windage losses. Several spur gears and a disc are studied. The CFD simulations return good agreement with measured windage power loss. Turbulence modeling choices, the relative importance of viscous and pressure torques with gear speed and the physics of the complex 3-D unsteady flow field in the vicinity of the gear teeth are studied.

  10. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of

  11. A CFD analysis of blade row interactions within a high-speed axial compressor

    Science.gov (United States)

    Richman, Michael Scott

    Aircraft engine design provides many technical and financial hurdles. In an effort to streamline the design process, save money, and improve reliability and performance, many manufacturers are relying on computational fluid dynamic simulations. An overarching goal of the design process for military aircraft engines is to reduce size and weight while maintaining (or improving) reliability. Designers often turn to the compression system to accomplish this goal. As pressure ratios increase and the number of compression stages decrease, many problems arise, for example stability and high cycle fatigue (HCF) become significant as individual stage loading is increased. CFD simulations have recently been employed to assist in the understanding of the aeroelastic problems. For accurate multistage blade row HCF prediction, it is imperative that advanced three-dimensional blade row unsteady aerodynamic interaction codes be validated with appropriate benchmark data. This research addresses this required validation process for TURBO, an advanced three-dimensional multi-blade row turbomachinery CFD code. The solution/prediction accuracy is characterized, identifying key flow field parameters driving the inlet guide vane (IGV) and stator response to the rotor generated forcing functions. The result is a quantified evaluation of the ability of TURBO to predict not only the fundamental flow field characteristics but the three dimensional blade loading.

  12. CFD Analysis for H2 Flame Propagation during Spray Operation in the THAI Facility

    International Nuclear Information System (INIS)

    In this paper discussion is focused on the three tests HD-30, HD-31, and HD-32.1 conducted in the THAI test facility to investigate the influence of water spray operation on hydrogen deflagration behavior. Test results have been extensively used by the project partners for further development and validation of computational codes within the frame of the OECD THAI-2 project. In order to quantify the influence of spray operation on hydrogen deflagration behavior, test results are compared with the reference tests conducted with same initial thermal-hydraulic conditions but without spray in the frame of OECD-THAI project. KAERI performed the CFD calculation of the tests HD-30, HD-31, and HD-32..1 conducted in the THAI facility to observe the influence of spray operation on hydrogen combustion for further development and validation of computational codes within the frame of the OECD THAI-2 project. We accurately simulated the measured peak pressure in the tests with an error range of about 5%. However, we cannot accurately simulate the disturbance of the chemical reaction of H2-Air mixture owing to the spray water. Thus, the calculated gas temperature during the hydrogen combustion was overestimated In order to reduce this discrepancy between the CFD results and test data, we have to modify the correlation of the turbulent flame speed in the BVM for simulating the hydrogen deflagration during the spray operation

  13. Massive separation around bluff bodies: comparisons among different cfd solvers and turbulence models

    Science.gov (United States)

    Armenio, Vincenzo; Fakhari, Ahmad; Petronio, Andrea; Padovan, Roberta; Pittaluga, Chiara; Caprino, Giovanni

    2015-11-01

    Massive flow separation is ubiquitous in industrial applications, ruling drag and hydrodynamic noise. In spite of considerable efforts, its numerical prediction still represents a challenge for CFD models in use in engineering. Aside commercial software, over the latter years the opensource software OpenFOAMR (OF) has emerged as a valid tool for prediction of complex industrial flows. In the present work, we simulate two flows representative of a class of situations occurring in industrial problems: the flow around sphere and that around a wall-mounted square cylinder at Re = 10000 . We compare the performance two different tools, namely OF and ANSYS CFX 15.0 (CFX) using different unstructured grids and turbulence models. The grids have been generated using SNAPPYHEXMESH and ANSYS ICEM CFD 15.0 with different near wall resolutions. The codes have been run in a RANS mode using k - ɛ model (OF) and SST - k - ω (CFX) with and without wall-layer models. OF has been also used in LES, WMLES and DES mode. Regarding the sphere, RANS models were not able to catch separation, while good prediction of separation and distribution of stresses over the surface were obtained using LES, WMLES and DES. Results for the second test case are currently under analysis. Financial support from COSMO ``cfd open source per opera mortta'' PAR FSC 2007-2013, Friuli Venezia Giulia.

  14. CFD Simulation and Optimisation of a Low Energy Ventilation and Cooling System

    Directory of Open Access Journals (Sweden)

    John Kaiser Calautit

    2015-04-01

    Full Text Available Mechanical Heating Ventilation and Air-Conditioning (HVAC systems account for 60% of the total energy consumption of buildings. As a sector, buildings contributes about 40% of the total global energy demand. By using passive technology coupled with natural ventilation from wind towers, significant amounts of energy can be saved, reducing the emissions of greenhouse gases. In this study, the development of Computational Fluid Dynamics (CFD analysis in aiding the development of wind towers was explored. Initial concepts of simple wind tower mechanics to detailed design of wind towers which integrate modifications specifically to improve the efficiency of wind towers were detailed. From this, using CFD analysis, heat transfer devices were integrated into a wind tower to provide cooling for incoming air, thus negating the reliance on mechanical HVAC systems. A commercial CFD code Fluent was used in this study to simulate the airflow inside the wind tower model with the heat transfer devices. Scaled wind tunnel testing was used to validate the computational model. The airflow supply velocity was measured and compared with the numerical results and good correlation was observed. Additionally, the spacing between the heat transfer devices was varied to optimise the performance. The technology presented here is subject to a patent application (PCT/GB2014/052263.

  15. CFD simulation of aggregation and breakage processes in laminar Taylor-Couette flow.

    Science.gov (United States)

    Wang, L; Marchisio, D L; Vigil, R D; Fox, R O

    2005-02-15

    An experimental and computational investigation of the effects of local fluid shear rate on the aggregation and breakage of approximately 10 microm latex spheres suspended in an aqueous solution undergoing laminar Taylor-Couette flow was carried out according to the following program. First, computational fluid dynamics (CFD) simulations were performed and the flow field predictions were validated with data from particle image velocimetry experiments. Subsequently, the quadrature method of moments (QMOM) was implemented into the CFD code to obtain predictions for mean particle size that account for the effects of local shear rate on the aggregation and breakage. These predictions were then compared with experimental data for latex sphere aggregates (using an in situ optical imaging method) and with predictions using spatial average shear rates. The mean particle size evolution predicted by CFD and QMOM using appropriate kinetic expressions that incorporate information concerning the particle morphology (fractal dimension) and the local fluid viscous effects on aggregation collision efficiency match well with the experimental data. PMID:15589543

  16. Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

    Science.gov (United States)

    Ramamoorthy, Karthikeyan

    The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant

  17. Validation of the solar heating and cooling high speed performance (HISPER) computer code

    Science.gov (United States)

    Wallace, D. B.

    1980-10-01

    Developed to give a quick and accurate predictions HISPER, a simplification of the TRNSYS program, achieves its computational speed by not simulating detailed system operations or performing detailed load computations. In order to validate the HISPER computer for air systems the simulation was compared to the actual performance of an operational test site. Solar insolation, ambient temperature, water usage rate, and water main temperatures from the data tapes for an office building in Huntsville, Alabama were used as input. The HISPER program was found to predict the heating loads and solar fraction of the loads with errors of less than ten percent. Good correlation was found on both a seasonal basis and a monthly basis. Several parameters (such as infiltration rate and the outside ambient temperature above which heating is not required) were found to require careful selection for accurate simulation.

  18. Validation of the solar heating and cooling high speed performance (HISPER) computer code

    Science.gov (United States)

    Wallace, D. B.

    1980-01-01

    Developed to give a quick and accurate predictions HISPER, a simplification of the TRNSYS program, achieves its computational speed by not simulating detailed system operations or performing detailed load computations. In order to validate the HISPER computer for air systems the simulation was compared to the actual performance of an operational test site. Solar insolation, ambient temperature, water usage rate, and water main temperatures from the data tapes for an office building in Huntsville, Alabama were used as input. The HISPER program was found to predict the heating loads and solar fraction of the loads with errors of less than ten percent. Good correlation was found on both a seasonal basis and a monthly basis. Several parameters (such as infiltration rate and the outside ambient temperature above which heating is not required) were found to require careful selection for accurate simulation.

  19. A Global Approach to the Physics Validation of Simulation Codes for Future Nuclear Systems

    Energy Technology Data Exchange (ETDEWEB)

    Giuseppe Palmiotti; Massimo Salvatores; Gerardo Aliberti; Hikarui Hiruta; R. McKnight; P. Oblozinsky; W. S. Yang

    2008-09-01

    This paper presents a global approach to the validation of the parameters that enter into the neutronics simulation tools for advanced fast reactors with the objective to reduce the uncertainties associated to crucial design parameters. This global approach makes use of sensitivity/uncertainty methods; statistical data adjustments; integral experiment selection, analysis and “representativity” quantification with respect to a reference system; scientifically based cross section covariance data and appropriate methods for their use in multigroup calculations. This global approach has been applied to the uncertainty reduction on the criticality of the Advanced Burner Reactor, (both metal and oxide core versions) presently investigated in the frame of the GNEP initiative. The results obtained are very encouraging and allow to indicate some possible improvements of the ENDF/B-VII data file.

  20. MiniPanda - a small-scale containment test facility with novel instrumentation is used for code validation for an air ingress scenario

    International Nuclear Information System (INIS)

    “MiniPanda”, a small-scale containment test facility built at ETH Zurich, was equipped with novel field measurement techniques. The capabilities of the facility were demonstrated in a first test series on an air ingress scenario. The ingress of air into a helium cooled reactor is considered to be one of the most severe accidents for the GenIV-type helium cooled reactor. The air once arrived inside the reactor can cause the oxidation of the graphite structures. The ingress of air into a helium environment was investigated experimentally and analytically using the commercial CFX-13 and StarCCM+ 5.06 CFD codes. The experimental volume consists of two cylindrical vessels that are filled separately with air or helium. The experiment is initiated by removing the blockage from the pipe connecting the two vessels. The experimental and analytical results of the consequent buoyancy-driven air ingress are compared against each other. (author)