WorldWideScience

Sample records for validate aeroelastic codes

  1. The aeroelastic code FLEXLAST

    Energy Technology Data Exchange (ETDEWEB)

    Visser, B. [Stork Product Eng., Amsterdam (Netherlands)

    1996-09-01

    To support the discussion on aeroelastic codes, a description of the code FLEXLAST was given and experiences within benchmarks and measurement programmes were summarized. The code FLEXLAST has been developed since 1982 at Stork Product Engineering (SPE). Since 1992 FLEXLAST has been used by Dutch industries for wind turbine and rotor design. Based on the comparison with measurements, it can be concluded that the main shortcomings of wind turbine modelling lie in the field of aerodynamics, wind field and wake modelling. (au)

  2. 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....... Finally, added turbulence characteristics are compared with correlation results from literature. ©2010 American Society of Mechanical Engineers...

  3. Centrifugal Compressor Aeroelastic Analysis Code

    Science.gov (United States)

    Keith, Theo G., Jr.; Srivastava, Rakesh

    2002-01-01

    Centrifugal compressors are very widely used in the turbomachine industry where low mass flow rates are required. Gas turbine engines for tanks, rotorcraft and small jets rely extensively on centrifugal compressors for rugged and compact design. These compressors experience problems related with unsteadiness of flowfields, such as stall flutter, separation at the trailing edge over diffuser guide vanes, tip vortex unsteadiness, etc., leading to rotating stall and surge. Considerable interest exists in small gas turbine engine manufacturers to understand and eventually eliminate the problems related to centrifugal compressors. The geometric complexity of centrifugal compressor blades and the twisting of the blade passages makes the linear methods inapplicable. Advanced computational fluid dynamics (CFD) methods are needed for accurate unsteady aerodynamic and aeroelastic analysis of centrifugal compressors. Most of the current day industrial turbomachines and small aircraft engines are designed with a centrifugal compressor. With such a large customer base and NASA Glenn Research Center being, the lead center for turbomachines, it is important that adequate emphasis be placed on this area as well. Currently, this activity is not supported under any project at NASA Glenn.

  4. Aeroelastic code development activities in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Wright, A.D. [National Renewable Energy Lab., Golden, Colorado (United States)

    1996-09-01

    Designing wind turbines to be fatigue resistant and to have long lifetimes at minimal cost is a major goal of the federal wind program and the wind industry in the United States. To achieve this goal, we must be able to predict critical loads for a wide variety of different wind turbines operating under extreme conditions. The codes used for wind turbine dynamic analysis must be able to analyze a wide range of different wind turbine configurations as well as rapidly predict the loads due to turbulent wind inflow with a minimal set of degrees of freedom. Code development activities in the US have taken a two-pronged approach in order to satisfy both of these criteria: (1) development of a multi-purpose code which can be used to analyze a wide variety of wind turbine configurations without having to develop new equations of motion with each configuration change, and (2) development of specialized codes with minimal sets of specific degrees of freedom for analysis of two- and three-bladed horizontal axis wind turbines and calculation of machine loads due to turbulent inflow. In the first method we have adapted a commercial multi-body dynamics simulation package for wind turbine analysis. In the second approach we are developing specialized codes with limited degrees of freedom, usually specified in the modal domain. This paper will summarize progress to date in the development, validation, and application of these codes. (au) 13 refs.

  5. Implementation of the Actuator Cylinder Flow Model in the HAWC2 code for Aeroelastic Simulations on Vertical Axis Wind Turbines

    DEFF Research Database (Denmark)

    Aagaard Madsen, Helge; Larsen, Torben J.; Schmidt Paulsen, Uwe

    2013-01-01

    The paper presents the implementation of the Actuator Cylinder (AC) flow model in the HAWC2 aeroelastic code originally developed for simulation of Horizontal Axis Wind Turbine (HAWT) aeroelasticity. This is done within the DeepWind project where the main objective is to explore the competitiveness...

  6. GAROS, an aeroelastic code for coupled fixed-rotating structures

    Energy Technology Data Exchange (ETDEWEB)

    Rees, M. [Aerodyn Energiestyseme GmbH, Rendsburg (Germany); Vollan, A. [Pilatus Flugzeugwerke, Stans (Switzerland)

    1996-09-01

    The GAROS (General Analysis of Rotating Structures) program system has been specially designed to calculate aeroelastic stability and dynamic response of horizontal axis wind energy converters. Nevertheless it is also suitable for the dynamic analysis of helicopter rotors and has been used in the analysis of car bodies taking account of rotating wheels. GAROS was developed over the last 17 years. In the following the mechanical and the aerodynamic model will be discussed in detail. A short overview of the solution methods for the equation of motion in time and frequency domain will ge given. After this one example for the FEM model of the rotor and tower will be discussed. (EG)

  7. The aeroelastic code HawC - model and comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Thirstrup Petersen, J. [Risoe National Lab., The Test Station for Wind Turbines, Roskilde (Denmark)

    1996-09-01

    A general aeroelastic finite element model for simulation of the dynamic response of horizontal axis wind turbines is presented. The model has been developed with the aim to establish an effective research tool, which can support the general investigation of wind turbine dynamics and research in specific areas of wind turbine modelling. The model concentrates on the correct representation of the inertia forces in a form, which makes it possible to recognize and isolate effects originating from specific degrees of freedom. The turbine structure is divided into substructures, and nonlinear kinematic terms are retained in the equations of motion. Moderate geometric nonlinearities are allowed for. Gravity and a full wind field including 3-dimensional 3-component turbulence are included in the loading. Simulation results for a typical three bladed, stall regulated wind turbine are presented and compared with measurements. (au)

  8. OC3—Benchmark Exercise of Aero-elastic Offshore Wind Turbine Codes

    Science.gov (United States)

    Passon, P.; Kühn, M.; Butterfield, S.; Jonkman, J.; Camp, T.; Larsen, T. J.

    2007-07-01

    This paper introduces the work content and status of the first international investigation and verification of aero-elastic codes for offshore wind turbines as performed by the "Offshore Code Comparison Collaboration"(OC3) within the "IEA Wind Annex XXIII - Subtask 2". An overview is given on the state-of-the-art of the concerned offshore wind turbine simulation codes. Exemplary results of benchmark simulations from the first phase of the project are presented and discussed while subsequent phases are introduced. Furthermore, the paper discusses areas where differences between the codes have been identified and the sources of those differences, such as the differing theories implemented into the individual codes. Finally, further research and code development needs are presented based on the latest findings from the current state of the project.

  9. Development of an aeroelastic code based on three-dimensional viscous–inviscid method for wind turbine computations

    DEFF Research Database (Denmark)

    Sessarego, Matias; Ramos García, Néstor; Sørensen, Jens Nørkær

    2017-01-01

    Aerodynamic and structural dynamic performance analysis of modern wind turbines are routinely estimated in the wind energy field using computational tools known as aeroelastic codes. Most aeroelastic codes use the blade element momentum (BEM) technique to model the rotor aerodynamics and a modal......, multi-body or the finite-element approach to model the turbine structural dynamics. The present work describes the development of a novel aeroelastic code that combines a three-dimensional viscous–inviscid interactive method, method for interactive rotor aerodynamic simulations (MIRAS...... Code Comparison Collaboration Project. Simulation tests consist of steady wind inflow conditions with different combinations of yaw error, wind shear, tower shadow and turbine-elastic modeling. Turbulent inflow created by using a Mann box is also considered. MIRAS-FLEX results, such as blade tip...

  10. Validation of thermalhydraulic codes

    International Nuclear Information System (INIS)

    Wilkie, D.

    1992-01-01

    Thermalhydraulic codes require to be validated against experimental data collected over a wide range of situations if they are to be relied upon. A good example is provided by the nuclear industry where codes are used for safety studies and for determining operating conditions. Errors in the codes could lead to financial penalties, to the incorrect estimation of the consequences of accidents and even to the accidents themselves. Comparison between prediction and experiment is often described qualitatively or in approximate terms, e.g. ''agreement is within 10%''. A quantitative method is preferable, especially when several competing codes are available. The codes can then be ranked in order of merit. Such a method is described. (Author)

  11. Containment Code Validation Matrix

    International Nuclear Information System (INIS)

    Chin, Yu-Shan; Mathew, P.M.; Glowa, Glenn; Dickson, Ray; Liang, Zhe; Leitch, Brian; Barber, Duncan; Vasic, Aleks; Bentaib, Ahmed; Journeau, Christophe; Malet, Jeanne; Studer, Etienne; Meynet, Nicolas; Piluso, Pascal; Gelain, Thomas; Michielsen, Nathalie; Peillon, Samuel; Porcheron, Emmanuel; Albiol, Thierry; Clement, Bernard; Sonnenkalb, Martin; Klein-Hessling, Walter; Arndt, Siegfried; Weber, Gunter; Yanez, Jorge; Kotchourko, Alexei; Kuznetsov, Mike; Sangiorgi, Marco; Fontanet, Joan; Herranz, Luis; Garcia De La Rua, Carmen; Santiago, Aleza Enciso; Andreani, Michele; Paladino, Domenico; Dreier, Joerg; Lee, Richard; Amri, Abdallah

    2014-01-01

    The Committee on the Safety of Nuclear Installations (CSNI) formed the CCVM (Containment Code Validation Matrix) task group in 2002. The objective of this group was to define a basic set of available experiments for code validation, covering the range of containment (ex-vessel) phenomena expected in the course of light and heavy water reactor design basis accidents and beyond design basis accidents/severe accidents. It was to consider phenomena relevant to pressurised heavy water reactor (PHWR), pressurised water reactor (PWR) and boiling water reactor (BWR) designs of Western origin as well as of Eastern European VVER types. This work would complement the two existing CSNI validation matrices for thermal hydraulic code validation (NEA/CSNI/R(1993)14) and In-vessel core degradation (NEA/CSNI/R(2001)21). The report initially provides a brief overview of the main features of a PWR, BWR, CANDU and VVER reactors. It also provides an overview of the ex-vessel corium retention (core catcher). It then provides a general overview of the accident progression for light water and heavy water reactors. The main focus is to capture most of the phenomena and safety systems employed in these reactor types and to highlight the differences. This CCVM contains a description of 127 phenomena, broken down into 6 categories: - Containment Thermal-hydraulics Phenomena; - Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena; - Aerosol and Fission Product Behaviour Phenomena; - Iodine Chemistry Phenomena; - Core Melt Distribution and Behaviour in Containment Phenomena; - Systems Phenomena. A synopsis is provided for each phenomenon, including a description, references for further information, significance for DBA and SA/BDBA and a list of experiments that may be used for code validation. The report identified 213 experiments, broken down into the same six categories (as done for the phenomena). An experiment synopsis is provided for each test. Along with a test description

  12. Flight code validation simulator

    Science.gov (United States)

    Sims, Brent A.

    1996-05-01

    An End-To-End Simulation capability for software development and validation of missile flight software on the actual embedded computer has been developed utilizing a 486 PC, i860 DSP coprocessor, embedded flight computer and custom dual port memory interface hardware. This system allows real-time interrupt driven embedded flight software development and checkout. The flight software runs in a Sandia Digital Airborne Computer and reads and writes actual hardware sensor locations in which Inertial Measurement Unit data resides. The simulator provides six degree of freedom real-time dynamic simulation, accurate real-time discrete sensor data and acts on commands and discretes from the flight computer. This system was utilized in the development and validation of the successful premier flight of the Digital Miniature Attitude Reference System in January of 1995 at the White Sands Missile Range on a two stage attitude controlled sounding rocket.

  13. Implementation of a Forth-Order Aeroelastic Coupling into a Viscous-Inviscid Flow Solver with Experimental Validation (for One Degree of Freedom)

    Science.gov (United States)

    Bartholomay, Sirko; Ramos-García, Néstor; Mikkelsen, Robert Flemming; Technical University of Denmark (DTU)-WInd Energy Team

    2014-11-01

    The viscous-inviscid flow solver Q3UIC for 2D aerodynamics has recently been developed at the Technical University of Denmark. The Q3UIC solver takes viscous and unsteady effects into account by coupling an unsteady inviscid panel method with the integral boundary layer equations by means of a strong coupling between the viscous and inviscid parts, and in this respect differs from other classic panel codes e.g. Xfoil. In the current work a Runge-Kutta-Nyström scheme was employed to couple inertial, elastic and aerodynamical forces and moments calculated by Q3UIC for a two-dimensional blade section in the time-domain. Numerical simulations are validated by a three step experimental verification process carried out in the low-turbulence wind tunnel at DTU. First, a comparison against steady experiments for a NACA 64418 profile and a flexible trailing edge flap is presented for different fixed flap angles, and second, the measured aerodynamic characteristics considering prescribed motion of the airfoil with a moving flap are compared to the Q3UIC predictions. Finally, an aeroelastic experiment for one degree of freedom-airfoil pitching- is used to evaluate the accuracy of aeroelastic coupling.

  14. Nonlinear aeroelastic modelling for wind turbine blades based on blade element momentum theory and geometrically exact beam theory

    International Nuclear Information System (INIS)

    Wang, Lin; Liu, Xiongwei; Renevier, Nathalie; Stables, Matthew; Hall, George M.

    2014-01-01

    Due to the increasing size and flexibility of large wind turbine blades, accurate and reliable aeroelastic modelling is playing an important role for the design of large wind turbines. Most existing aeroelastic models are linear models based on assumption of small blade deflections. This assumption is not valid anymore for very flexible blade design because such blades often experience large deflections. In this paper, a novel nonlinear aeroelastic model for large wind turbine blades has been developed by combining BEM (blade element momentum) theory and mixed-form formulation of GEBT (geometrically exact beam theory). The nonlinear aeroelastic model takes account of large blade deflections and thus greatly improves the accuracy of aeroelastic analysis of wind turbine blades. The nonlinear aeroelastic model is implemented in COMSOL Multiphysics and validated with a series of benchmark calculation tests. The results show that good agreement is achieved when compared with experimental data, and its capability of handling large deflections is demonstrated. Finally the nonlinear aeroelastic model is applied to aeroelastic modelling of the parked WindPACT 1.5 MW baseline wind turbine, and reduced flapwise deflection from the nonlinear aeroelastic model is observed compared to the linear aeroelastic code FAST (Fatigue, Aerodynamics, Structures, and Turbulence). - Highlights: • A novel nonlinear aeroelastic model for wind turbine blades is developed. • The model takes account of large blade deflections and geometric nonlinearities. • The model is reliable and efficient for aeroelastic modelling of wind turbine blades. • The accuracy of the model is verified by a series of benchmark calculation tests. • The model provides more realistic aeroelastic modelling than FAST (Fatigue, Aerodynamics, Structures, and Turbulence)

  15. Verification of aero-elastic offshore wind turbine design codes under IEA Wind Task XXIII

    DEFF Research Database (Denmark)

    Vorpahl, Fabian; Strobel, Michael; Jonkman, Jason M.

    2014-01-01

    with the incident waves, sea current, hydrodynamics and foundation dynamics of the support structure. A large set of time series simulation results such as turbine operational characteristics, external conditions, and load and displacement outputs was compared and interpreted. Load cases were defined and run...... to differences in the model fidelity, aerodynamic implementation, hydrodynamic load discretization and numerical difficulties within the codes. The comparisons resulted in a more thorough understanding of the modeling techniques and better knowledge of when various approximations are not valid.More importantly...... is to summarize the lessons learned and present results that code developers can compare to. The set of benchmark load cases defined and simulated during the course of this project—the raw data for this paper—is available to the offshore wind turbine simulation community and is already being used for testing...

  16. Development and validation of sodium fire codes

    International Nuclear Information System (INIS)

    Morii, Tadashi; Himeno Yoshiaki; Miyake, Osamu

    1989-01-01

    Development, verification, and validation of the spray fire code, SPRAY-3M, the pool fire codes, SOFIRE-M2 and SPM, the aerosol behavior code, ABC-INTG, and the simultaneous spray and pool fires code, ASSCOPS, are presented. In addition, the state-of-the-art of development of the multi-dimensional natural convection code, SOLFAS, for the analysis of heat-mass transfer during a fire, is presented. (author)

  17. Validation of the reactor dynamics code HEXTRAN

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1994-05-01

    HEXTRAN is a new three-dimensional, hexagonal reactor dynamics code developed in the Technical Research Centre of Finland (VTT) for VVER type reactors. This report describes the validation work of HEXTRAN. The work has been made with the financing of the Finnish Centre for Radiation and Nuclear Safety (STUK). HEXTRAN is particularly intended for calculation of such accidents, in which radially asymmetric phenomena are included and both good neutron dynamics and two-phase thermal hydraulics are important. HEXTRAN is based on already validated codes. The models of these codes have been shown to function correctly also within the HEXTRAN code. The main new model of HEXTRAN, the spatial neutron kinetics model has been successfully validated against LR-0 test reactor and Loviisa plant measurements. Connected with SMABRE, HEXTRAN can be reliably used for calculation of transients including effects of the whole cooling system of VVERs. Further validation plans are also introduced in the report. (orig.). (23 refs., 16 figs., 2 tabs.)

  18. Cable SGEMP Code Validation Study

    Energy Technology Data Exchange (ETDEWEB)

    Ballard, William Parker [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Center for CA Weapons Systems Engineering

    2013-05-01

    This report compared data taken on the Modular Bremsstrahlung Simulator using copper jacketed (cujac) cables with calculations using the RHSD-RA Cable SGEMP analysis tool. The tool relies on CEPXS/ONBFP to perform radiation transport in a series of 1D slices through the cable, and then uses a Green function technique to evaluate the expected current drive on the center conductor. The data were obtained in 2003 as part of a Cabana verification and validation experiment using 1-D geometries, but were not evaluated until now. The agreement between data and model is not adequate unless gaps between the dielectric and outer conductor (ground) are assumed, and these gaps are large compared with what is believed to be in the actual cable.

  19. New aeroelastic studies for a morphing wing

    Directory of Open Access Journals (Sweden)

    Ruxandra Mihaela BOTEZ*

    2012-06-01

    Full Text Available For this study, the upper surface of a rectangular finite aspect ratio wing, with a laminar airfoil cross-section, was made of a carbon-Kevlar composite material flexible skin. This flexible skin was morphed by use of Shape Memory Alloy actuators for 35 test cases characterized by combinations of Mach numbers, Reynolds numbers and angles of attack. The Mach numbers varied from 0.2 to 0.3 and the angles of attack ranged between -1° and 2°. The optimized airfoils were determined by use of the CFD XFoil code. The purpose of this aeroelastic study was to determine the flutter conditions to be avoided during wind tunnel tests. These studies show that aeroelastic instabilities for the morphing configurations considered appeared at Mach number 0.55, which was higher than the wind tunnel Mach number limit speed of 0.3. The wind tunnel tests could thus be performed safely in the 6’×9’ wind tunnel at the Institute for Aerospace Research at the National Research Council Canada (IAR/NRC, where the new aeroelastic studies, applied on morphing wings, were validated.

  20. Validations and applications of the FEAST code

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Z.; Tayal, M.; Lau, J.H.; Evinou, D. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Jun, J.S. [Korea Atomic Energy Research Inst. (Korea, Republic of)

    1999-07-01

    The FEAST (Finite Element Analysis for STresses) code is part of a suite of computer codes that are used to assess the structural integrity of CANDu fuel elements and bundles. A detailed validation of the FEAST code was recently performed. The FEAST calculations are in good agreement with a variety of analytical solutions (18 cases) for stresses, strains and displacements. This consistency shows that the FEAST code correctly incorporates the fundamentals of stress analysis. Further, the calculations of the FEAST code match the variations in axial and hoop strain profiles, measured by strain gauges near the sheath-endcap weld during an out-reactor compression test. The code calculations are also consistent with photoelastic measurements in simulated endcaps. (author)

  1. Validations and applications of the FEAST code

    International Nuclear Information System (INIS)

    Xu, Z.; Tayal, M.; Lau, J.H.; Evinou, D.; Jun, J.S.

    1999-01-01

    The FEAST (Finite Element Analysis for STresses) code is part of a suite of computer codes that are used to assess the structural integrity of CANDu fuel elements and bundles. A detailed validation of the FEAST code was recently performed. The FEAST calculations are in good agreement with a variety of analytical solutions (18 cases) for stresses, strains and displacements. This consistency shows that the FEAST code correctly incorporates the fundamentals of stress analysis. Further, the calculations of the FEAST code match the variations in axial and hoop strain profiles, measured by strain gauges near the sheath-endcap weld during an out-reactor compression test. The code calculations are also consistent with photoelastic measurements in simulated endcaps. (author)

  2. Validation of comprehensive space radiation transport code

    International Nuclear Information System (INIS)

    Shinn, J.L.; Simonsen, L.C.; Cucinotta, F.A.

    1998-01-01

    The HZETRN code has been developed over the past decade to evaluate the local radiation fields within sensitive materials on spacecraft in the space environment. Most of the more important nuclear and atomic processes are now modeled and evaluation within a complex spacecraft geometry with differing material components, including transition effects across boundaries of dissimilar materials, are included. The atomic/nuclear database and transport procedures have received limited validation in laboratory testing with high energy ion beams. The codes have been applied in design of the SAGE-III instrument resulting in material changes to control injurious neutron production, in the study of the Space Shuttle single event upsets, and in validation with space measurements (particle telescopes, tissue equivalent proportional counters, CR-39) on Shuttle and Mir. The present paper reviews the code development and presents recent results in laboratory and space flight validation

  3. 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. ...

  4. Validation of the reactor dynamics code TRAB

    International Nuclear Information System (INIS)

    Raety, H.; Kyrki-Rajamaeki, R.; Rajamaeki, M.

    1991-05-01

    The one-dimensional reactor dynamics code TRAB (Transient Analysis code for BWRs) developed at VTT was originally designed for BWR analyses, but it can in its present version be used for various modelling purposes. The core model of TRAB can be used separately for LWR calculations. For PWR modelling the core model of TRAB has been coupled to circuit model SMABRE to form the SMATRA code. The versatile modelling capabilities of TRAB have been utilized also in analyses of e.g. the heating reactor SECURE and the RBMK-type reactor (Chernobyl). The report summarizes the extensive validation of TRAB. TRAB has been validated with benchmark problems, comparative calculations against independent analyses, analyses of start-up experiments of nuclear power plants and real plant transients. Comparative RBMES type reactor calculations have been made against Soviet simulations and the initial power excursion of the Chernobyl reactor accident has also been calculated with TRAB

  5. Experimental validation of the HARMONIE code

    International Nuclear Information System (INIS)

    Bernard, A.; Dorsselaere, J.P. van

    1984-01-01

    An experimental program of deformation, in air, of different groups of subassemblies (7 to 41 subassemblies), was performed on a scale 1 mock-up in the SPX1 geometry, in order to achieve a first experimental validation of the code HARMONIE. The agreement between tests and calculations was suitable, qualitatively for all the groups and quantitatively for regular groups of 19 subassemblies at most. The differences come mainly from friction between pads, and secondly from the foot gaps. (author)

  6. Measurement of reactivity coefficients for code validation

    International Nuclear Information System (INIS)

    Nuding, Matthias; Loetsch, Thomas

    2005-01-01

    In the year 2003 measurements in the cold reactor state have been performed at the NPP KKI 2 in order to validate the codes that are used for reactor core calculations and especially for the proof of the shutdown margin that is produced by calculations only. For full power states code verification is quite easy because the calculations can be compared with different measured values, e.g. with the activation values determined by the aeroball system. For cold reactor states, however the data base is smaller, especially for reactor cores that are quite 'inhomogeneous' and have rather high Pu-fiss-and 235 U-contents. At the same time the cold reactor state is important regarding the shutdown margin. For these reasons the measurements mentioned above have been performed in order to check the accuracy of the codes that are used by the operator and by our organization for many years. Basically, boron concentrations and control rod worths for different configurations have been measured. The results of the calculation show a very good agreement with the measured values. Therefore, it can be stated that the operator's as well as our code system is suitable for routine use, e.g. during licensing procedures (Authors)

  7. A Nonlinear Modal Aeroelastic Solver for FUN3D

    Science.gov (United States)

    Goldman, Benjamin D.; Bartels, Robert E.; Biedron, Robert T.; Scott, Robert C.

    2016-01-01

    A nonlinear structural solver has been implemented internally within the NASA FUN3D computational fluid dynamics code, allowing for some new aeroelastic capabilities. Using a modal representation of the structure, a set of differential or differential-algebraic equations are derived for general thin structures with geometric nonlinearities. ODEPACK and LAPACK routines are linked with FUN3D, and the nonlinear equations are solved at each CFD time step. The existing predictor-corrector method is retained, whereby the structural solution is updated after mesh deformation. The nonlinear solver is validated using a test case for a flexible aeroshell at transonic, supersonic, and hypersonic flow conditions. Agreement with linear theory is seen for the static aeroelastic solutions at relatively low dynamic pressures, but structural nonlinearities limit deformation amplitudes at high dynamic pressures. No flutter was found at any of the tested trajectory points, though LCO may be possible in the transonic regime.

  8. KNOW-BLADE task-4 report. Navier-Stokes aeroelasticity

    Energy Technology Data Exchange (ETDEWEB)

    Politis, E.S.; Nikolaou, I.G.; Chaviaropoulos, P.K.; Bertagnolio, F.; Soerensen, N.N.; Johansen, J.

    2005-01-01

    The problem of the aeroelastic stability of wind turbine blades is addressed in this report by advancing the aerodynamic modelling in the beam element type codes from the engineering-type empirical models to unsteady, 2D or 3D, Navier-Stokes solvers. In this project, structural models for the full wind turbine blade have been combined with 2D and 3D unsteady Navier-Stokes solvers. The relative disadvantage of the quasi-3D approach (where the elastic solver is coupled with a 2D Navier-Stokes solver) is its inability to model induced flow. The lack of a validation test case did not allow for quantitative comparisons with experimental data to be carried out; instead the results of the advanced aeroelastic tools are qualitatively cross-compared. All investigated methods predicted qualitatively similar results. They all resulted in positive aerodynamic damping values for the flap mode, in a decrease in damping with the increase of wind speeds and in a minimum value for the damping for wind speed around 15{approx}m/s. The eigenvalue analyses resulted in steeper distributions for this mode. The agreement in aerodynamic damping decrease with the increase of wind speed is also observed in the distributions for the lead-lag mode. In perspective, the uncoupled, linear method results in higher values of aerodynamic damping compared to the 3D aeroelastic tool. The quasi-3D tool results in lower aerodynamic damping values in the higher wind speeds and in lower damping values in the lower wind speed regime. Apart from the computations for the full blade, 2D computations for the so-called 'typical section' have been carried out. The 2D aeroelastic tools resulted in similar aerodynamic damping values. Qualitative agreement was better for the lead-lag mode. The presence of roughness tapes has a small, rather negligible impact on aeroelastic stability as depicted by the results of both aeroelastic tools. On the other hand, in conformity to the inability of the adopted

  9. Harmonic Balance Computations of Fan Aeroelastic Stability

    Science.gov (United States)

    Bakhle, Milind A.; Reddy, T. S. R.

    2010-01-01

    A harmonic balance (HB) aeroelastic analysis, which has been recently developed, was used to determine the aeroelastic stability (flutter) characteristics of an experimental fan. To assess the numerical accuracy of this HB aeroelastic analysis, a time-domain aeroelastic analysis was also used to determine the aeroelastic stability characteristics of the same fan. Both of these three-dimensional analysis codes model the unsteady flowfield due to blade vibrations using the Reynolds-averaged Navier-Stokes (RANS) equations. In the HB analysis, the unsteady flow equations are converted to a HB form and solved using a pseudo-time marching method. In the time-domain analysis, the unsteady flow equations are solved using an implicit time-marching approach. Steady and unsteady computations for two vibration modes were carried out at two rotational speeds: 100 percent (design) and 70 percent (part-speed). The steady and unsteady results obtained from the two analysis methods compare well, thus verifying the recently developed HB aeroelastic analysis. Based on the results, the experimental fan was found to have no aeroelastic instability (flutter) at the conditions examined in this study.

  10. X-HALE: The Development of a Research Platform for the Validation of Nonlinear Aeroelastic Codes

    Science.gov (United States)

    2011-03-01

    general, whenever the number of plies or the laminate direction is specifically modified or selected for a composite aircraft, the aircraft’s design is...cm 35 X-HALE’s wings are composed primarily of Hexcel E-Glass 120/F155 prepreg fabric and Rohacell Foam. Hexcel E-Glass 120/F155 prepreg

  11. Nuclear data to support computer code validation

    International Nuclear Information System (INIS)

    Fisher, S.E.; Broadhead, B.L.; DeHart, M.D.; Primm, R.T. III

    1997-04-01

    The rate of plutonium disposition will be a key parameter in determining the degree of success of the Fissile Materials Disposition Program. Estimates of the disposition rate are dependent on neutronics calculations. To ensure that these calculations are accurate, the codes and data should be validated against applicable experimental measurements. Further, before mixed-oxide (MOX) fuel can be fabricated and loaded into a reactor, the fuel vendors, fabricators, fuel transporters, reactor owners and operators, regulatory authorities, and the Department of Energy (DOE) must accept the validity of design calculations. This report presents sources of neutronics measurements that have potential application for validating reactor physics (predicting the power distribution in the reactor core), predicting the spent fuel isotopic content, predicting the decay heat generation rate, certifying criticality safety of fuel cycle facilities, and ensuring adequate radiation protection at the fuel cycle facilities and the reactor. The U.S. in-reactor experience with MOX fuel is first presented, followed by information related to other aspects of the MOX fuel performance information that is valuable to this program, but the data base remains largely proprietary. Thus, this information is not reported here. It is expected that the selected consortium will make the necessary arrangements to procure or have access to the requisite information

  12. Transonic aeroelastic numerical simulation in aeronautical engineering

    International Nuclear Information System (INIS)

    Yang, G.

    2005-01-01

    An LU-SGS (lower-upper symmetric Gauss-Seidel) subiteration scheme is constructed for time-marching of the fluid equations. The HLLEW (Harten-Lax-van Leer-Einfeldt-Wada) scheme is used for the spatial discretization. The same subiteration formulation is applied directly to the structural equations of motion in generalized coordinates. Through subiteration between the fluid and structural equations, a fully implicit aeroelastic solver is obtained for the numerical simulation of fluid/structure interaction. To improve the ability for application to complex configurations, a multiblock grid is used for the flow field calculation and Transfinite Interpolation (TFI) is employed for the adaptive moving grid deformation. The infinite plate spline (IPS) and the principal of virtual work are utilized for the data transformation between the fluid and structure. The developed code was first validated through the comparison of experimental and computational results for the AGARD 445.6 standard aeroelastic wing. Then the flutter character of a tail wing with control surface was analyzed. Finally, flutter boundaries of a complex aircraft configuration were predicted. (author)

  13. Validation and verification plan for safety and PRA codes

    International Nuclear Information System (INIS)

    Ades, M.J.; Crowe, R.D.; Toffer, H.

    1991-04-01

    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

  14. Sensitivity Analysis and Error Control for Computational Aeroelasticity, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of this proposal is the development of a next-generation computational aeroelasticity code, suitable for real-world complex geometries, and...

  15. A fuel performance code TRUST VIc and its validation

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, M; Kogai, T [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan)

    1997-08-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs.

  16. A fuel performance code TRUST VIc and its validation

    International Nuclear Information System (INIS)

    Ishida, M.; Kogai, T.

    1997-01-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs

  17. Gap Conductance model Validation in the TASS/SMR-S code using MARS code

    International Nuclear Information System (INIS)

    Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae

    2010-01-01

    Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case

  18. Development and validation of sodium fire analysis code ASSCOPS

    International Nuclear Information System (INIS)

    Ohno, Shuji

    2001-01-01

    A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)

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

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Neu, K.; Dorsselaere, J.P. Van

    2005-01-01

    The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes. The basis version V0 of ASTEC (Accident Source Term Evaluation Code)-commonly developed and basically validated by GRS and IRSN-was made available in late 2000 for the EVITA partners on their individual platforms. Users' training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A 'hot line' assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:- for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases; - for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models. Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users' tools. As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme

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

    International Nuclear Information System (INIS)

    Sills, H.E.

    1995-01-01

    Several of the OECD/CSNI members have adapted a proposed methodology for code validation and uncertainty assessment. Although the validation process adapted by members has a high degree of commonality, the uncertainty assessment processes selected are more variable, ranaing from subjective to formal. This paper describes the validation and uncertainty assessment process, the sources of uncertainty, methods of reducing uncertainty, and methods of assessing uncertainty.Examples are presented from the Ontario Hydro application of the validation methodology and uncertainty assessment to the system thermal hydraulics discipline and the TUF (1) system thermal hydraulics code. (author)

  1. Computer code validation by high temperature chemistry

    International Nuclear Information System (INIS)

    Alexander, C.A.; Ogden, J.S.

    1988-01-01

    At least five of the computer codes utilized in analysis of severe fuel damage-type events are directly dependent upon or can be verified by high temperature chemistry. These codes are ORIGEN, CORSOR, CORCON, VICTORIA, and VANESA. With the exemption of CORCON and VANESA, it is necessary that verification experiments be performed on real irradiated fuel. For ORIGEN, the familiar knudsen effusion cell is the best choice and a small piece of known mass and known burn-up is selected and volatilized completely into the mass spectrometer. The mass spectrometer is used in the integral mode to integrate the entire signal from preselected radionuclides, and from this integrated signal the total mass of the respective nuclides can be determined. For CORSOR and VICTORIA, experiments with flowing high pressure hydrogen/steam must flow over the irradiated fuel and then enter the mass spectrometer. For these experiments, a high pressure-high temperature molecular beam inlet must be employed. Finally, in support of VANESA-CORCON, the very highest temperature and molten fuels must be contained and analyzed. Results from all types of experiments will be discussed and their applicability to present and future code development will also be covered

  2. Verification and validation of XSDRNPM code for tank waste calculations

    International Nuclear Information System (INIS)

    ROGERS, C.A.

    1999-01-01

    This validation study demonstrates that the XSDRNPM computer code accurately calculates the infinite neutron multiplication for water-moderated systems of low enriched uranium, plutonium, and iron. Calculations are made on a 200 MHz Brvo MS 5200M personal

  3. 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

  4. Improvements, verifications and validations of the BOW code

    International Nuclear Information System (INIS)

    Yu, S.D.; Tayal, M.; Singh, P.N.

    1995-01-01

    The BOW code calculates the lateral deflections of a fuel element consisting of sheath and pellets, due to temperature gradients, hydraulic drag and gravity. the fuel element is subjected to restraint from endplates, neighboring fuel elements and the pressure tube. Many new features have been added to the BOW code since its original release in 1985. This paper outlines the major improvements made to the code and verification/validation results. (author)

  5. Validation of Dose Calculation Codes for Clearance

    International Nuclear Information System (INIS)

    Menon, S.; Wirendal, B.; Bjerler, J.; Studsvik; Teunckens, L.

    2003-01-01

    Various international and national bodies such as the International Atomic Energy Agency, the European Commission, the US Nuclear Regulatory Commission have put forward proposals or guidance documents to regulate the ''clearance'' from regulatory control of very low level radioactive material, in order to allow its recycling as a material management practice. All these proposals are based on predicted scenarios for subsequent utilization of the released materials. The calculation models used in these scenarios tend to utilize conservative data regarding exposure times and dose uptake as well as other assumptions as a safeguard against uncertainties. None of these models has ever been validated by comparison with the actual real life practice of recycling. An international project was organized in order to validate some of the assumptions made in these calculation models, and, thereby, better assess the radiological consequences of recycling on a practical large scale

  6. Numerical Simulations of the Aeroelastic Behavior of Large Horizontal-Axis Wind Turbines: The Drivetrain Case

    DEFF Research Database (Denmark)

    Gebhardt, Cristian; Veluri, Badrinath; Preidikman, Sergio

    2010-01-01

    In this work an aeroelastic model that describes the interaction between aerodynamics and drivetrain dynamics of a large horizontal–axis wind turbine is presented. Traditional designs for wind turbines are based on the output of specific aeroelastic simulation codes. The output of these codes giv...

  7. Preliminary Computational Analysis of the (HIRENASD) Configuration in Preparation for the Aeroelastic Prediction Workshop

    Science.gov (United States)

    Chwalowski, Pawel; Florance, Jennifer P.; Heeg, Jennifer; Wieseman, Carol D.; Perry, Boyd P.

    2011-01-01

    This paper presents preliminary computational aeroelastic analysis results generated in preparation for the first Aeroelastic Prediction Workshop (AePW). These results were produced using FUN3D software developed at NASA Langley and are compared against the experimental data generated during the HIgh REynolds Number Aero- Structural Dynamics (HIRENASD) Project. The HIRENASD wind-tunnel model was tested in the European Transonic Windtunnel in 2006 by Aachen University0s Department of Mechanics with funding from the German Research Foundation. The computational effort discussed here was performed (1) to obtain a preliminary assessment of the ability of the FUN3D code to accurately compute physical quantities experimentally measured on the HIRENASD model and (2) to translate the lessons learned from the FUN3D analysis of HIRENASD into a set of initial guidelines for the first AePW, which includes test cases for the HIRENASD model and its experimental data set. This paper compares the computational and experimental results obtained at Mach 0.8 for a Reynolds number of 7 million based on chord, corresponding to the HIRENASD test conditions No. 132 and No. 159. Aerodynamic loads and static aeroelastic displacements are compared at two levels of the grid resolution. Harmonic perturbation numerical results are compared with the experimental data using the magnitude and phase relationship between pressure coefficients and displacement. A dynamic aeroelastic numerical calculation is presented at one wind-tunnel condition in the form of the time history of the generalized displacements. Additional FUN3D validation results are also presented for the AGARD 445.6 wing data set. This wing was tested in the Transonic Dynamics Tunnel and is commonly used in the preliminary benchmarking of computational aeroelastic software.

  8. Validity of vascular trauma codes at major trauma centres.

    Science.gov (United States)

    Altoijry, Abdulmajeed; Al-Omran, Mohammed; Lindsay, Thomas F; Johnston, K Wayne; Melo, Magda; Mamdani, Muhammad

    2013-12-01

    The use of administrative databases in vascular injury research has been increasing, but the validity of the diagnosis codes used in this research is uncertain. We assessed the positive predictive value (PPV) of International Classification of Diseases, tenth revision (ICD-10), vascular injury codes in administrative claims data in Ontario. We conducted a retrospective validation study using the Canadian Institute for Health Information Discharge Abstract Database, an administrative database that records all hospital admissions in Canada. We evaluated 380 randomly selected hospital discharge abstracts from the 2 main trauma centres in Toronto, Ont., St.Michael's Hospital and Sunnybrook Health Sciences Centre, between Apr. 1, 2002, and Mar. 31, 2010. We then compared these records with the corresponding patients' hospital charts to assess the level of agreement for procedure coding. We calculated the PPV and sensitivity to estimate the validity of vascular injury diagnosis coding. The overall PPV for vascular injury coding was estimated to be 95% (95% confidence interval [CI] 92.3-96.8). The PPV among code groups for neck, thorax, abdomen, upper extremity and lower extremity injuries ranged from 90.8 (95% CI 82.2-95.5) to 97.4 (95% CI 91.0-99.3), whereas sensitivity ranged from 90% (95% CI 81.5-94.8) to 98.7% (95% CI 92.9-99.8). Administrative claims hospital discharge data based on ICD-10 diagnosis codes have a high level of validity when identifying cases of vascular injury. Observational Study Level III.

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

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

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

  10. Results from the First Validation Phase of CAP code

    International Nuclear Information System (INIS)

    Choo, Yeon Joon; Hong, Soon Joon; Hwang, Su Hyun; Kim, Min Ki; Lee, Byung Chul; Ha, Sang Jun; Choi, Hoon

    2010-01-01

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

  11. WSRC approach to validation of criticality safety computer codes

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. Contributions to the validation of the ASTEC V1 code

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei; Turcu, Ilie

    2004-01-01

    In the frame of PHEBEN2 project (Validation of the severe accidents codes for applications to nuclear power plants, based on the PHEBUS FP experiments), a project developed within the EU research Frame Program 5 (FP5), the INR-Pitesti's team has received the task of determining the ASTEC code sensitivity. The PHEBEN2 project has been initiated in 1998 and gathered 13 partners from 6 EU member states. To the project 4 partners from 3 candidate states (Hungary, Bulgaria and Romania) joined later. The works were contracted with the European Commission (under FIKS-CT1999-00009 contract) that supports financially the research effort up to about 50%. According to the contract provisions, INR's team participated in developing the Working Package 1 (WP1) which refers to validation of the integral computation codes that use the PHOEBUS experimental data and the Working Package 3 (WP3) referring to the evaluation of the codes to be applied in nuclear power plants for risk evaluation, nuclear safety margin evaluation and determination/evaluation of the measures to be adopted in case of severe accident. The present work continues the efforts to validate preliminarily the ASTEC code. Focused are the the stand-alone sensitivity analyses applied to two most important modules of the code, namely DIVA and SOPHAEROS

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

    International Nuclear Information System (INIS)

    Nowakowski, Pedro Mariano

    2004-01-01

    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

  14. Verification and Validation of the Tritium Transport Code TMAP7

    International Nuclear Information System (INIS)

    Longhurst, Glen R.; Ambrosek, James

    2005-01-01

    The TMAP code has been upgraded to version 7, which includes radioactive decay along with many features implemented in prior versions. Pursuant to acceptance and release for distribution, the code was exercised in a variety of problem types to demonstrate that it provides results in agreement with theoretical results for cases where those are available. It has also been used to model certain experimental results. In this paper, the capabilities of the TMAP7 code are demonstrated by presenting some of the results from the verification and validation process

  15. Validating the BISON fuel performance code to integral LWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

    2016-05-15

    Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

  16. Viscous and Aeroelastic Effects on Wind Turbine Blades. The VISCEL Project. Part II: Aeroelastic Stability Investigations

    Science.gov (United States)

    Chaviaropoulos, P. K.; Soerensen, N. N.; Hansen, M. O. L.; Nikolaou, I. G.; Aggelis, K. A.; Johansen, J.; Gaunaa, Mac; Hambraus, T.; Frhr. von Geyr, Heiko; Hirsch, Ch.; Shun, Kang; Voutsinas, S. G.; Tzabiras, G.; Perivolaris, Y.; Dyrmose, S. Z.

    2003-10-01

    The recent introduction of ever larger wind turbines poses new challenges with regard to understanding the mechanisms of unsteady flow-structure interaction. An important aspect of the problem is the aeroelastic stability of the wind turbine blades, especially in the case of combined flap/lead-lag vibrations in the stall regime. Given the limited experimental information available in this field, the use of CFD techniques and state-of-the-art viscous flow solvers provides an invaluable alternative towards the identification of the underlying physics and the development and validation of sound engineering-type aeroelastic models. Navier-Stokes-based aeroelastic stability analysis of individual blade sections subjected to combined pitch/flap or flap/lead-lag motion has been attempted by the present consortium in the framework of the concluded VISCEL JOR3-CT98-0208 Joule III project.

  17. Research in aeroelasticity EFP-2007-II

    Energy Technology Data Exchange (ETDEWEB)

    Buhl, T. (ed.)

    2009-06-15

    This report contains results from the EFP-2007-II project 'Program for Research in Applied Aeroelasticity'. The main results can be summed up into the following bullets: 1) 2D CFD was used to investigate tower shadow effects on both upwind and downwind turbines, and was used to validate the tower shadow models implemented in the aeroelastic code HAWC2. 2) Using a streamlined tower reduces the tower shadow by 50% compared to a cylindrical tower. Similar reductions can be achieved using a four legged lattice tower. 3) The application of laminar/turbulent transition in CFD computations for airfoils is demonstrated. For attached flow over thin airfoils (18%) 2D computations provide good results while a combination of Detached Eddy Simulation and laminar/ turbulent transition modeling improve the results in stalled conditions for a thick airfoil. 4) The unsteady flow in the nacelle region of a wind turbine is dominated by large flow gradients caused by unsteady shedding of vortices from the root sections of the blades. 5) The averaged nacelle wind speed compares well to the freestream wind speed, whereas the nacelle flow angle is highly sensitive to vertical positioning and tilt in the inflow. 6) The trailing edge noise model, TNO, was implemented and validated. The results showed that the noise was not predicted accurately, but the model captured the trends and can be used in airfoil design. The model was implemented in the airfoil design tool AIRFOILOPT and existing airfoils can be adjusted to maintain the aerodynamic characteristics, but with reduced noise in the order of up to 3dB in total sound power level and up to 1dB with A-weighting. 7) 2D CFD simulations are performed to verify their capability in predicting multi element airfoil configurations. The present computations show good agreement with measured performance from wind tunnel experiments. 8) The stochastic fluctuations of the aerodynamic forces on blades in deep-stall have an insignificant

  18. Lawrence Livermore National Laboratory Probabilistic Seismic Hazard Codes Validation

    International Nuclear Information System (INIS)

    Savy, J B

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Caron-Charles, M.; Blumenfeld, L.

    2001-01-01

    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)

  20. Experimental validation of the containment codes ASTARTE and SEURBNUK

    International Nuclear Information System (INIS)

    Kendall, K.C.; Arnold, L.A.; Broadhouse, B.J.; Jones, A.; Yerkess, A.; Benuzzi, A.

    1979-10-01

    The fast reactor containment codes ASTARTE and SEURBNUK are being validated against data from the COVA series of small scale experiments being performed jointly by the UKAEA and JRC Ispra. The experimental programme is nearly complete, and data are given. (U.K.)

  1. Validation of Magnetic Reconstruction Codes for Real-Time Applications

    International Nuclear Information System (INIS)

    Mazon, D.; Murari, A.; Boulbe, C.; Faugeras, B.; Blum, J.; Svensson, J.; Quilichini, T.; Gelfusa, M.

    2010-01-01

    The real-time reconstruction of the plasma magnetic equilibrium in a tokamak is a key point to access high-performance regimes. Indeed, the shape of the plasma current density profile is a direct output of the reconstruction and has a leading effect for reaching a steady-state high-performance regime of operation. The challenge is thus to develop real-time methods and algorithms that reconstruct the magnetic equilibrium from the perspective of using these outputs for feedback control purposes. In this paper the validation of the JET real-time equilibrium reconstruction codes using both a Bayesian approach and a full equilibrium solver named Equinox will be detailed, the comparison being performed with the off-line equilibrium code EFIT (equilibrium fitting) or the real-time boundary reconstruction code XLOC (X-point local expansion). In this way a significant database, a methodology, and a strategy for the validation are presented. The validation of the results has been performed using a validated database of 130 JET discharges with a large variety of magnetic configurations. Internal measurements like polarimetry and motional Stark effect have been also used for the Equinox validation including some magnetohydrodynamic signatures for the assessment of the reconstructed safety profile and current density. (authors)

  2. Generator dynamics in aeroelastic analysis and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, T.J.; Hansen, M.H.; Iov, F.

    2003-05-01

    This report contains a description of a dynamic model for a doubly-fed induction generator implemented in the aeroelastic code HAWC. The model has physical input parameters (resistance, reactance etc.) and input variables (stator and rotor voltage and rotor speed). The model can be used to simulate the generator torque as well as the rotor and stator currents, active and reactive power. A perturbation method has been used to reduce the original generator model equations to a set of equations which can be solved with the same time steps as a typical aeroelastic code. The method is used to separate the fast transients of the model from the slow variations and deduce a reduced order expression for the slow part. Dynamic effects of the first order terms in the model as well as the influence on drive train eigenfrequencies and damping has been investigated. Load response during time simulation of wind turbine response have been compared to simulations with a linear static generator model originally implemented i HAWC. A 2 MW turbine has been modelled in the aeroelastic code HAWC. When using the new dynamic generator model there is an interesting coupling between the generator dynamics and a global turbine vibration mode at 4.5 Hz, which only occurs when a dynamic formulation of the generator equations is applied. This frequency can especially be seen in the electrical power of the generator and the rotational speed of the generator, but also as torque variations in the drive train. (au)

  3. KNOW-BLADE task-4 report: Navier-Stokes aeroelasticity

    DEFF Research Database (Denmark)

    Politis, E.S.; Nikolaou, I.G.; Chaviaropoulos, P.K.

    2004-01-01

    wind turbine blade have been combined with 2D and 3D unsteady Navier-Stokes solvers. The relative disadvantage of the quasi-3D approach (where the elastic solver is coupled with a 2D Navier-Stokes solver) isits inability to model induced flow. The lack of a validation test case did not allow...... the computations for the full blade, 2D computations for the so-called “typical section” have been carried out. The 2D aeroelastic tools resulted in similar aerodynamic damping values. Qualitative agreement was better for the lead-lagmode. The presence of roughness tapes has a small, rather negligible impact...... on aeroelastic stability as depicted by the results of both aeroelastic tools. On the other hand, in conformity to the inability of the adopted computational model to successfullypredict the corresponding test cases under work package 2 of the project, the aeroelastic tools are not capable to predict the correct...

  4. System verification and validation report for the TMAD code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

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

  5. Validation of OPERA3D PCMI Analysis Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)

    2013-10-15

    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.

  6. Verification and Validation of Heat Transfer Model of AGREE Code

    Energy Technology Data Exchange (ETDEWEB)

    Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)

    2013-05-15

    The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.

  7. Validation of containment thermal hydraulic computer codes for VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

    2005-07-01

    Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

  8. The WINCON programme - validation of fast reactor primary containment codes

    International Nuclear Information System (INIS)

    Sidoli, J.E.A.; Kendall, K.C.

    1988-01-01

    In the United Kingdom safety studies for the Commercial Demonstration Fast Reactor (CDFR) include an assessment of the capability of the primary containment in providing an adequate containment for defence against the hazards resulting from a hypothetical Whole Core Accident (WCA). The assessment is based on calculational estimates using computer codes supported by measured evidence from small-scale experiments. The hydrodynamic containment code SEURBNUK-EURDYN is capable of representing a prescribed energy release, the sodium coolant and cover gas, and the main containment and safety related internal structures. Containment loadings estimated using SEURBNUK-EURDYN are used in the structural dynamic code EURDYN-03 for the prediction of the containment response. The experiments serve two purposes, they demonstrate the response of the CDFR containment to accident loadings and provide data for the validation of the codes. This paper summarises the recently completed WINfrith CONtainment (WINCON) experiments that studied the response of specific features of current CDFR design options to WCA loadings. The codes have been applied to some of the experiments and a satisfactory prediction of the global response of the model containment is obtained. This provides confidence in the use of the codes in reactor assessments. (author)

  9. Development and preliminary validation of flux map processing code MAPLE

    International Nuclear Information System (INIS)

    Li Wenhuai; Zhang Xiangju; Dang Zhen; Chen Ming'an; Lu Haoliang; Li Jinggang; Wu Yuanbao

    2013-01-01

    The self-reliant flux map processing code MAPLE was developed by China General Nuclear Power Corporation (CGN). Weight coefficient method (WCM), polynomial expand method (PEM) and thin plane spline (TPS) method were applied to fit the deviation between measured and predicted detector signal results for two-dimensional radial plane, to interpolate or extrapolate the non-instrumented location deviation. Comparison of results in the test cases shows that the TPS method can better capture the information of curved fitting lines than the other methods. The measured flux map data of the Lingao Nuclear Power Plant were processed using MAPLE as validation test cases, combined with SMART code. Validation results show that the calculation results of MAPLE are reasonable and satisfied. (authors)

  10. Validation of the TAC/BLOOST code (Contract research)

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

    2005-06-01

    Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30% (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test. (author)

  11. Chaotic Patterns in Aeroelastic Signals

    Directory of Open Access Journals (Sweden)

    F. D. Marques

    2009-01-01

    patterns. With the reconstructed state spaces, qualitative analyses may be done, and the attractors evolutions with parametric variation are presented. Overall results reveal complex system dynamics associated with highly separated flow effects together with nonlinear coupling between aeroelastic modes. Bifurcations to the nonlinear aeroelastic system are observed for two investigations, that is, considering oscillations-induced aeroelastic evolutions with varying freestream speed, and aeroelastic evolutions at constant freestream speed and varying oscillations. Finally, Lyapunov exponent calculation is proceeded in order to infer on chaotic behavior. Poincaré mappings also suggest bifurcations and chaos, reinforced by the attainment of maximum positive Lyapunov exponents.

  12. 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.

  13. Gamma streaming experiments for validation of Monte Carlo code

    International Nuclear Information System (INIS)

    Thilagam, L.; Mohapatra, D.K.; Subbaiah, K.V.; Iliyas Lone, M.; Balasubramaniyan, V.

    2012-01-01

    In-homogeneities in shield structures lead to considerable amount of leakage radiation (streaming) increasing the radiation levels in accessible areas. Development works on experimental as well as computational methods for quantifying this streaming radiation are still continuing. Monte Carlo based radiation transport code, MCNP is usually a tool for modeling and analyzing such problems involving complex geometries. In order to validate this computational method for streaming analysis, it is necessary to carry out some experimental measurements simulating these inhomogeneities like ducts and voids present in the bulk shields for typical cases. The data thus generated will be analysed by simulating the experimental set up employing MCNP code and optimized input parameters for the code in finding solutions for similar radiation streaming problems will be formulated. Comparison of experimental data obtained from radiation streaming experiments through ducts will give a set of thumb rules and analytical fits for total radiation dose rates within and outside the duct. The present study highlights the validation of MCNP code through the gamma streaming experiments carried out with the ducts of various shapes and dimensions. Over all, the present study throws light on suitability of MCNP code for the analysis of gamma radiation streaming problems for all duct configurations considered. In the present study, only dose rate comparisons have been made. Studies on spectral comparison of streaming radiation are in process. Also, it is planned to repeat the experiments with various shield materials. Since the penetrations and ducts through bulk shields are unavoidable in an operating nuclear facility the results on this kind of radiation streaming simulations and experiments will be very useful in the shield structure optimization without compromising the radiation safety

  14. Aeroelasticity of morphing wings using neural networks

    Science.gov (United States)

    Natarajan, Anand

    In this dissertation, neural networks are designed to effectively model static non-linear aeroelastic problems in adaptive structures and linear dynamic aeroelastic systems with time varying stiffness. The use of adaptive materials in aircraft wings allows for the change of the contour or the configuration of a wing (morphing) in flight. The use of smart materials, to accomplish these deformations, can imply that the stiffness of the wing with a morphing contour changes as the contour changes. For a rapidly oscillating body in a fluid field, continuously adapting structural parameters may render the wing to behave as a time variant system. Even the internal spars/ribs of the aircraft wing which define the wing stiffness can be made adaptive, that is, their stiffness can be made to vary with time. The immediate effect on the structural dynamics of the wing, is that, the wing motion is governed by a differential equation with time varying coefficients. The study of this concept of a time varying torsional stiffness, made possible by the use of active materials and adaptive spars, in the dynamic aeroelastic behavior of an adaptable airfoil is performed here. Another type of aeroelastic problem of an adaptive structure that is investigated here, is the shape control of an adaptive bump situated on the leading edge of an airfoil. Such a bump is useful in achieving flow separation control for lateral directional maneuverability of the aircraft. Since actuators are being used to create this bump on the wing surface, the energy required to do so needs to be minimized. The adverse pressure drag as a result of this bump needs to be controlled so that the loss in lift over the wing is made minimal. The design of such a "spoiler bump" on the surface of the airfoil is an optimization problem of maximizing pressure drag due to flow separation while minimizing the loss in lift and energy required to deform the bump. One neural network is trained using the CFD code FLUENT to

  15. Validation and testing of the VAM2D computer code

    International Nuclear Information System (INIS)

    Kool, J.B.; Wu, Y.S.

    1991-10-01

    This document describes two modeling studies conducted by HydroGeoLogic, Inc. for the US NRC under contract no. NRC-04089-090, entitled, ''Validation and Testing of the VAM2D Computer Code.'' VAM2D is a two-dimensional, variably saturated flow and transport code, with applications for performance assessment of nuclear waste disposal. The computer code itself is documented in a separate NUREG document (NUREG/CR-5352, 1989). The studies presented in this report involve application of the VAM2D code to two diverse subsurface modeling problems. The first one involves modeling of infiltration and redistribution of water and solutes in an initially dry, heterogeneous field soil. This application involves detailed modeling over a relatively short, 9-month time period. The second problem pertains to the application of VAM2D to the modeling of a waste disposal facility in a fractured clay, over much larger space and time scales and with particular emphasis on the applicability and reliability of using equivalent porous medium approach for simulating flow and transport in fractured geologic media. Reflecting the separate and distinct nature of the two problems studied, this report is organized in two separate parts. 61 refs., 31 figs., 9 tabs

  16. Integrated Validation System for a Thermal-hydraulic System Code, TASS/SMR-S

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee-Kyung; Kim, Hyungjun; Kim, Soo Hyoung; Hwang, Young-Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Hyeon-Soo [Chungnam National University, Daejeon (Korea, Republic of)

    2015-10-15

    Development including enhancement and modification of thermal-hydraulic system computer code is indispensable to a new reactor, SMART. Usually, a thermal-hydraulic system code validation is achieved by a comparison with the results of corresponding physical effect tests. In the reactor safety field, a similar concept, referred to as separate effect tests has been used for a long time. But there are so many test data for comparison because a lot of separate effect tests and integral effect tests are required for a code validation. It is not easy to a code developer to validate a computer code whenever a code modification is occurred. IVS produces graphs which shown the comparison the code calculation results with the corresponding test results automatically. IVS was developed for a validation of TASS/SMR-S code. The code validation could be achieved by a comparison code calculation results with corresponding test results. This comparison was represented as a graph for convenience. 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. The code developer could gain a confidence about his code modification easily and fast and could be free from tedious and long validation work. The popular software introduced in IVS supplies better usability and portability.

  17. Development validation and use of computer codes for inelastic analysis

    International Nuclear Information System (INIS)

    Jobson, D.A.

    1983-01-01

    A finite element scheme is a system which provides routines so carry out the operations which are common to all finite element programs. The list of items that can be provided as standard by the finite element scheme is surprisingly large and the list provided by the UNCLE finite element scheme is unusually comprehensive. This presentation covers the following: construction of the program, setting up a finite element mesh, generation of coordinates, incorporating boundary and load conditions. Program validation was done by creep calculations performed using CAUSE code. Program use is illustrated by calculating a typical inelastic analysis problem. This includes computer model of the PFR intermediate heat exchanger

  18. Three-Dimensional Aeroelastic and Aerothermoelastic Behavior in Hypersonic Flow

    Science.gov (United States)

    McNamara, Jack J.; Friedmann, Peretz P.; Powell, Kenneth G.; Thuruthimattam, Biju J.; Bartels, Robert E.

    2005-01-01

    The aeroelastic and aerothermoelastic behavior of three-dimensional configurations in hypersonic flow regime are studied. The aeroelastic behavior of a low aspect ratio wing, representative of a fin or control surface on a generic hypersonic vehicle, is examined using third order piston theory, Euler and Navier-Stokes aerodynamics. The sensitivity of the aeroelastic behavior generated using Euler and Navier-Stokes aerodynamics to parameters governing temporal accuracy is also examined. Also, a refined aerothermoelastic model, which incorporates the heat transfer between the fluid and structure using CFD generated aerodynamic heating, is used to examine the aerothermoelastic behavior of the low aspect ratio wing in the hypersonic regime. Finally, the hypersonic aeroelastic behavior of a generic hypersonic vehicle with a lifting-body type fuselage and canted fins is studied using piston theory and Euler aerodynamics for the range of 2.5 less than or equal to M less than or equal to 28, at altitudes ranging from 10,000 feet to 80,000 feet. This analysis includes a study on optimal mesh selection for use with Euler aerodynamics. In addition to the aeroelastic and aerothermoelastic results presented, three time domain flutter identification techniques are compared, namely the moving block approach, the least squares curve fitting method, and a system identification technique using an Auto-Regressive model of the aeroelastic system. In general, the three methods agree well. The system identification technique, however, provided quick damping and frequency estimations with minimal response record length, and therefore o ers significant reductions in computational cost. In the present case, the computational cost was reduced by 75%. The aeroelastic and aerothermoelastic results presented illustrate the applicability of the CFL3D code for the hypersonic flight regime.

  19. 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.

  20. Strain actuated aeroelastic control

    Science.gov (United States)

    Lazarus, Kenneth B.

    1992-01-01

    Viewgraphs on strain actuated aeroelastic control are presented. Topics covered include: structural and aerodynamic modeling; control law design methodology; system block diagram; adaptive wing test article; bench-top experiments; bench-top disturbance rejection: open and closed loop response; bench-top disturbance rejection: state cost versus control cost; wind tunnel experiments; wind tunnel gust alleviation: open and closed loop response at 60 mph; wind tunnel gust alleviation: state cost versus control cost at 60 mph; wind tunnel command following: open and closed loop error at 60 mph; wind tunnel flutter suppression: open loop flutter speed; and wind tunnel flutter suppression: closed loop state cost curves.

  1. Validation of thermal hydraulic codes for fusion reactors safety

    International Nuclear Information System (INIS)

    Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.

    2006-01-01

    A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)

  2. Aeroelastic Flutter of Subsonic Aircraft Wing Section with Control Surface

    Directory of Open Access Journals (Sweden)

    Aeroelastic Flutter of Subsonic Aircraft Wing Section with Control Surface

    2015-12-01

    Full Text Available Aeroelastic flutter in aircraft mechanisms is unavoidable, essentially in the wing and control surface. In this work a three degree-of-freedom aeroelastic wing section with trailing edge flap is modeled numerically and theoretically. FLUENT code based on the steady finite volume is used for the prediction of the steady aerodynamic characteristics (lift, drag, pitching moment, velocity, and pressure distribution as well as the Duhamel formulation is used to model the aerodynamic loads theoretically. The system response (pitch, flap pitch and plunge was determined by integration the governing equations using MATLAB with a standard Runge–Kutta algorithm in conjunction with Henon’s method. The results are compared with previous experimental data. The results show that the aerodynamic loads and wing-flap system response are increased when increasing the flow speed. On the other hand the aeroelastic response led up to limit cycle oscillation when the flow equals or more than flutter speed.

  3. Spent reactor fuel benchmark composition data for code validation

    International Nuclear Information System (INIS)

    Bierman, S.R.

    1991-09-01

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

  4. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

  5. Field Validation of the Stability Limit of a Multi MW Turbine

    Science.gov (United States)

    Kallesøe, Bjarne S.; Kragh, Knud A.

    2016-09-01

    Long slender blades of modern multi-megawatt turbines exhibit a flutter like instability at rotor speeds above a critical rotor speed. Knowing the critical rotor speed is crucial to a safe turbine design. The flutter like instability can only be estimated using geometrically non-linear aeroelastic codes. In this study, the estimated rotor speed stability limit of a 7 MW state of the art wind turbine is validated experimentally. The stability limit is estimated using Siemens Wind Powers in-house aeroelastic code, and the results show that the predicted stability limit is within 5% of the experimentally observed limit.

  6. 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.

  7. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

  8. Validation of the TIARA code to tritium inventory data

    International Nuclear Information System (INIS)

    Billone, M.C.

    1994-03-01

    The TIARA code has been developed to predict tritium inventory in Li 2 O breeder ceramic and to predict purge exit flow rate and composition. Inventory predictions are based on models for bulk diffusion, surface desorption, solubility and precipitation. Parameters for these models are determined from the results of laboratory annealing studies on unirradiated and irradiated Li 2 O. Inventory data from in-reactor purge flow tests are used for model improvement, fine-tuning of model parameters and validation. In this current work, the inventory measurement near the purge inlet from the BEATRIX-II thin-ring sample is used to fine tune the surface desorption model parameters for T > 470 degrees C, and the inventory measurement near the midplane from VOM-15H is used to fine tune the moisture solubility model parameters. predictions are then validated to the remaining inventory data from EXOTIC-2 (1 point), SIBELIUS (3 axial points), VOM-15H (2 axial points), CRITIC-1 (4 axial points), BEATRIX-II thin ring (3 axial points) and BEATRIX-II thick pellet (5 radial points). Thus. of the 20 data points, two we re used for fine tuning model parameters and 18 were used for validation. The inventory data span the range of 0.05--1.44 wppm with an average of 0.48 wppm. The data pertain to samples whose end-of-life temperatures were in the range of 490--1000 degrees C. On the average, the TIARA predictions agree quite well with the data (< 0.02 wppm difference). However, the root-mean-square deviation is 0.44 wppm, mostly due to over-predictions for the SIBELIUS samples and the higher-temperature radial samples from the BEATRIX-11 thick-pellet

  9. 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

  10. Aeroelastic simulation using CFD based reduced order models

    International Nuclear Information System (INIS)

    Zhang, W.; Ye, Z.; Li, H.; Yang, Q.

    2005-01-01

    This paper aims at providing an accurate and efficient method for aeroelastic simulation. System identification is used to get the reduced order models of unsteady aerodynamics. Unsteady Euler codes are used to compute the output signals while 3211 multistep input signals are utilized. LS(Least Squares) method is used to estimate the coefficients of the input-output difference model. The reduced order models are then used in place of the unsteady CFD code for aeroelastic simulation. The aeroelastic equations are marched by an improved 4th order Runge-Kutta method that only needs to compute the aerodynamic loads one time at every time step. The computed results agree well with that of the direct coupling CFD/CSD methods. The computational efficiency is improved 1∼2 orders while still retaining the high accuracy. A standard aeroelastic computing example (isogai wing) with S type flutter boundary is computed and analyzed. It is due to the system has more than one neutral points at the Mach range of 0.875∼0.9. (author)

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

    Science.gov (United States)

    Springate, David A; Kontopantelis, Evangelos; Ashcroft, Darren M; Olier, Ivan; Parisi, Rosa; Chamapiwa, Edmore; Reeves, David

    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 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.

  12. First experimental validation on the core equilibrium code: HARMONIE

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.; Cozzani, M.; Gnuffi, M.

    1981-08-01

    The code HARMONIE calculates the mechanical equilibrium of a fast reactor. An experimental program of deformation, in air, of groups of subassemblies, was performed on a mock-up, in the Super Phenix 1- geometry. This program included three kinds of tests, all performed without and then with grease: on groups of 2 or 3 rings of subassemblies, subjected to a force acting upon flats or angles; on groups of 35 and 41 subassemblies, subjected to a force acting on the first row, then with 1 or 2 empty cells; and on groups with 1 or 2 bowed subassemblies or 1 enlarged one over flats. A preliminary test on the friction coefficient in air between two pads showed some dependance upon the pad surface condition with a scattering factor of 8. Two basic code hypotheses were validated: the rotation of the subassemblies around their axis was negligible after deformation of the group, and the choice of a mean Maxwell coefficient, between those of 1st and 2nd slope, led to very similar results to experimental. The agreement between tests and HARMONIE calculations was suitable, qualitatively for all the groups and quantitatively for regular groups of 3 rings at most. But the difference increased for larger groups of 35 or 41 subassemblies: friction between pads, neglected by HARMONIE, seems to be the main reason. Other reasons for these differences are: the influence of the loading order on the mock-up, and the initial contacts issued from the gap between foot and diagrid-insert, and from manufacture bowings

  13. Verification and Validation of The Tritium Transport Code TMAP7

    International Nuclear Information System (INIS)

    Glen R. Longhurst; James Ambrosek

    2004-01-01

    The TMAP Code was written at the Idaho National Engineering and Environmental Laboratory in the late 1980s as a tool for safety analysis of systems involving tritium. Since then it has been upgraded several times and has been used in numerous applications including experiments supporting fusion safety, predictions for advanced systems such as the International Thermonuclear Experimental Reactor (ITER), and estimates involving tritium production technologies. Its most recent upgrade to TMAP7 was accomplished in response to several needs. Prior versions had the capacity to deal with only a single trap for diffusing gaseous species in solid structures. TMAP7 includes up to three separate traps and up to 10 diffusing species. The original code had difficulty dealing with heteronuclear molecule formation such as HD and DT. That has been removed. Under pre-specified boundary enclosure conditions and solution-law dependent diffusion boundary conditions, such as Sieverts' law, TMAP7 automatically generates heteronuclear molecular partial pressures when solubilities and partial pressures of the homonuclear molecular species are provided for law-dependent diffusion boundary conditions. A further sophistication is the addition of non-diffusing surface species. Atoms such as oxygen or nitrogen or formation of hydroxyl radicals on metal surfaces are sometimes important in molecule formation with diffusing hydrogen isotopes but do not, themselves, diffuse appreciably in the material. TMAP7 will accommodate up to 30 such surface species, allowing the user to specify relationships between those surface concentrations and partial pressures of gaseous species above the surfaces or to form them dynamically by combining diffusion species or other surface species. Additionally, TMAP7 allows the user to include a surface binding energy and an adsorption barrier energy and includes asymmetrical diffusion between the surface sites and regular diffusion sites in the bulk. All of the

  14. 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

  15. Applications of potential theory computations to transonic aeroelasticity

    Science.gov (United States)

    Edwards, J. W.

    1986-01-01

    Unsteady aerodynamic and aeroelastic stability calculations based upon transonic small disturbance (TSD) potential theory are presented. Results from the two-dimensional XTRAN2L code and the three-dimensional XTRAN3S code are compared with experiment to demonstrate the ability of TSD codes to treat transonic effects. The necessity of nonisentropic corrections to transonic potential theory is demonstrated. Dynamic computational effects resulting from the choice of grid and boundary conditions are illustrated. Unsteady airloads for a number of parameter variations including airfoil shape and thickness, Mach number, frequency, and amplitude are given. Finally, samples of transonic aeroelastic calculations are given. A key observation is the extent to which unsteady transonic airloads calculated by inviscid potential theory may be treated in a locally linear manner.

  16. Analytical validation of the CACECO containment analysis code

    International Nuclear Information System (INIS)

    Peak, R.D.

    1979-08-01

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

  17. Aeroelastic Wingbox Stiffener Topology Optimization

    Science.gov (United States)

    Stanford, Bret K.

    2017-01-01

    This work considers an aeroelastic wingbox model seeded with run-out blade stiffeners along the skins. Topology optimization is conducted within the shell webs of the stiffeners, in order to add cutouts and holes for mass reduction. This optimization is done with a global-local approach in order to moderate the computational cost: aeroelastic loads are computed at the wing-level, but the topology and sizing optimization is conducted at the panel-level. Each panel is optimized separately under stress, buckling, and adjacency constraints, and periodically reassembled to update the trimmed aeroelastic loads. The resulting topology is baselined against a design with standard full-depth solid stiffener blades, and found to weigh 7.43% less.

  18. Optimization of wind turbine rotors - using advanced aerodynamic and aeroelastic models and numerical optimization

    DEFF Research Database (Denmark)

    Døssing, Mads

    of very large machines introduces new problems in the practical design, and optimization tools are necessary. These must combine the dynamic eects of both aerodynamics and structure in an integrated optimization environment. This is referred to as aeroelastic optimization. The Ris DTU optimization...... software HAWTOPT has been used in this project. The quasi-steady aerodynamic module have been improved with a corrected blade element momentum method. A structure module has also been developed which lays out the blade structural properties. This is done in a simplied way allowing fast conceptual design...... studies and with focus on the overall properties relevant for the aeroelastic properties. Aeroelastic simulations in the time domain were carried out using the aeroelastic code HAWC2. With these modules coupled to HAWTOPT, optimizations have been made. In parallel with the developments of the mentioned...

  19. Research in Aeroelasticity EFP-2006[Wind turbines

    Energy Technology Data Exchange (ETDEWEB)

    Bak, C.

    2007-07-15

    shear on power and the interaction with cyclic pitch showed that there is a considerable uncertainty in modeling this rather common inflow case for a rotor. The implementation of the BEM model to handle this flow case can be carried out in different ways and the results of two advanced models have not so far been validated. In the new Aeroelastic Research Programme EFP2007 the investigations will be continued and the aspect of influence of individual pitch control on the blades will also be considered. (au)

  20. Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE

    International Nuclear Information System (INIS)

    Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa

    2009-01-01

    The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)

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

    International Nuclear Information System (INIS)

    Leskovar, M.; Mavko, B.

    1998-01-01

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

  2. Use of operational data for the validation of the SOPHT thermal-hydraulic code

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S F; Martin, G; Shoukas, L; Siddiqui, Z; Phillips, B [Ontario Hydro, Bowmanville, ON (Canada). Darlington Nuclear Generating Station

    1996-12-31

    The primary objective of this paper is to describe the validation process of the SOPHT and MINI-SOPHT codes with the use of reactor operational data. The secondary objective is to illustrative the effectiveness of the code as a performance monitoring tool by discussing the discoveries that were made during the validation process. (author). 2 refs.

  3. Aeroelastic modal dynamics of wind turbines including anisotropic effects

    Energy Technology Data Exchange (ETDEWEB)

    Fisker Skjoldan, P.

    2011-03-15

    Several methods for aeroelastic modal analysis of a rotating wind turbine are developed and used to analyse the modal dynamics of two simplified models and a complex model in isotropic and anisotropic conditions. The Coleman transformation is used to enable extraction of the modal frequencies, damping, and periodic mode shapes of a rotating wind turbine by describing the rotor degrees of freedom in the inertial frame. This approach is valid only for an isotropic system. Anisotropic systems, e.g., with an unbalanced rotor or operating in wind shear, are treated with the general approaches of Floquet analysis or Hill's method which do not provide a unique reference frame for observing the modal frequency, to which any multiple of the rotor speed can be added. This indeterminacy is resolved by requiring that the periodic mode shape be as constant as possible in the inertial frame. The modal frequency is thus identified as the dominant frequency in the response of a pure excitation of the mode observed in the inertial frame. A modal analysis tool based directly on the complex aeroelastic wind turbine code BHawC is presented. It uses the Coleman approach in isotropic conditions and the computationally efficient implicit Floquet analysis in anisotropic conditions. The tool is validated against system identifications with the partial Floquet method on the nonlinear BHawC model of a 2.3 MW wind turbine. System identification results show that nonlinear effects on the 2.3 MW turbine in most cases are small, but indicate that the controller creates nonlinear damping. In isotropic conditions the periodic mode shape contains up to three harmonic components, but in anisotropic conditions it can contain an infinite number of harmonic components with frequencies that are multiples of the rotor speed. These harmonics appear in calculated frequency responses of the turbine. Extreme wind shear changes the modal damping when the flow is separated due to an interaction between

  4. Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

    International Nuclear Information System (INIS)

    Brus, N.A.; Ioussoupov, O.E.

    2000-01-01

    This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

  5. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

    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)

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

    International Nuclear Information System (INIS)

    Cevolani, S.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    1995-12-01

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

  9. Generator dynamics in aeroelastic analysis and simulations

    DEFF Research Database (Denmark)

    Larsen, Torben J.; Hansen, Morten Hartvig; Iov, F.

    2003-01-01

    This report contains a description of a dynamic model for a doubly-fed induction generator. The model has physical input parameters (voltage, resistance, reactance etc.) and can be used to calculate rotor and stator currents, hence active and reactivepower. A perturbation method has been used...... to reduce the original generator model equations to a set of equations which can be solved with the same time steps as a typical aeroelastic code. The method is used to separate the fast transients of the modelfrom the slow variations and deduce a reduced order expression for the slow part. Dynamic effects...... of the first order terms in the model as well as the influence on drive train eigenfrequencies and damping has been investigated. Load response during timesimulation of wind turbine response have been compared to simulations with a traditional static generator model based entirely on the slip angle. A 2 MW...

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

    International Nuclear Information System (INIS)

    Bishop, W.E.; Lee, A.G.

    1998-01-01

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

  11. Research in Aeroelasticity EFP-2007-II

    DEFF Research Database (Denmark)

    is demonstrated. For attached flow over thin airfoils (18%) 2D computations provide good results while a combination of Detached Eddy Simulation and laminar/ turbulent transition modeling improve the results in stalled conditions for a thick airfoil. • The unsteady flow in the nacelle region of a wind turbine......This report contains results from the EFP-2007-II project "Program for Research in Applied Aeroelasticity". The main results can be summed up into the following bullets: • 2D CFD was used to investigate tower shadow effects on both upwind and downwind turbines, and was used to validate the tower...... is dominated by large flow gradients caused by unsteady shedding of vortices from the root sections of the blades. • The averaged nacelle wind speed compares well to the freestream wind speed, whereas the nacelle flow angle is highly sensitive to vertical positioning and tilt in the inflow. • The trailing edge...

  12. Aeroelastic modal dynamics of wind turbines including anisotropic effects

    DEFF Research Database (Denmark)

    Skjoldan, Peter Fisker

    frequency is thus identified as the dominant frequency in the response of a pure excitation of the mode observed in the inertial frame. A modal analysis tool based directly on the complex aeroelastic wind turbine code BHawC is presented. It uses the Coleman approach in isotropic conditions......Several methods for aeroelastic modal analysis of a rotating wind turbine are developed and used to analyse the modal dynamics of two simplified models and a complex model in isotropic and anisotropic conditions. The Coleman transformation is used to enable extraction of the modal frequencies...... of Floquet analysis or Hill's method which do not provide a unique reference frame for observing the modal frequency, to which any multiple of the rotor speed can be added. This indeterminacy is resolved by requiring that the periodic mode shape be as constant as possible in the inertial frame. The modal...

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

    Energy Technology Data Exchange (ETDEWEB)

    Herber, S.C.; Allelein, H.J. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6); Friege, N. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Kasselmann, S. [Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6)

    2012-11-01

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

  14. Toward a CFD nose-to-tail capability - Hypersonic unsteady Navier-Stokes code validation

    Science.gov (United States)

    Edwards, Thomas A.; Flores, Jolen

    1989-01-01

    Computational fluid dynamics (CFD) research for hypersonic flows presents new problems in code validation because of the added complexity of the physical models. This paper surveys code validation procedures applicable to hypersonic flow models that include real gas effects. The current status of hypersonic CFD flow analysis is assessed with the Compressible Navier-Stokes (CNS) code as a case study. The methods of code validation discussed to beyond comparison with experimental data to include comparisons with other codes and formulations, component analyses, and estimation of numerical errors. Current results indicate that predicting hypersonic flows of perfect gases and equilibrium air are well in hand. Pressure, shock location, and integrated quantities are relatively easy to predict accurately, while surface quantities such as heat transfer are more sensitive to the solution procedure. Modeling transition to turbulence needs refinement, though preliminary results are promising.

  15. 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.

  16. Study of experimental validation for combustion analysis of GOTHIC code

    International Nuclear Information System (INIS)

    Lee, J. Y.; Yang, S. Y.; Park, K. C.; Jeong, S. H.

    2001-01-01

    In this study, present lumped and subdivided GOTHIC6 code analyses of the premixed hydrogen combustion experiment at the Seoul National University and comparison with the experiment results. The experimental facility has 16367 cc free volume and rectangular shape. And the test was performed with unit equivalence ratio of the hydrogen and air, and with various location of igniter position. Using the lumped and mechanistic combustion model in GOTHIC6 code, the experiments were simulated with the same conditions. In the comparison between experiment and calculated results, the GOTHIC6 prediction of the combustion response does not compare well with the experiment results. In the point of combustion time, the lumped combustion model of GOTHIC6 code does not simulate the physical phenomena of combustion appropriately. In the case of mechanistic combustion model, the combustion time is predicted well, but the induction time of calculation data is longer than the experiment data remarkably. Also, the laminar combustion model of GOTHIC6 has deficiency to simulate combustion phenomena unless control the user defined value appropriately. And the pressure is not a proper variable that characterize the three dimensional effect of combustion

  17. Validation Study of CODES Dragonfly Network Model with Theta Cray XC System

    Energy Technology Data Exchange (ETDEWEB)

    Mubarak, Misbah [Argonne National Lab. (ANL), Argonne, IL (United States); Ross, Robert B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-05-31

    This technical report describes the experiments performed to validate the MPI performance measurements reported by the CODES dragonfly network simulation with the Theta Cray XC system at the Argonne Leadership Computing Facility (ALCF).

  18. A Systematic Method for Verification and Validation of Gyrokinetic Microstability Codes

    Energy Technology Data Exchange (ETDEWEB)

    Bravenec, Ronald [Fourth State Research, Austin, TX (United States)

    2017-11-14

    My original proposal for the period Feb. 15, 2014 through Feb. 14, 2017 called for an integrated validation and verification effort carried out by myself with collaborators. The validation component would require experimental profile and power-balance analysis. In addition, it would require running the gyrokinetic codes varying the input profiles within experimental uncertainties to seek agreement with experiment before discounting a code as invalidated. Therefore, validation would require a major increase of effort over my previous grant periods which covered only code verification (code benchmarking). Consequently, I had requested full-time funding. Instead, I am being funded at somewhat less than half time (5 calendar months per year). As a consequence, I decided to forego the validation component and to only continue the verification efforts.

  19. Context discovery using attenuated Bloom codes: model description and validation

    NARCIS (Netherlands)

    Liu, F.; Heijenk, Geert

    A novel approach to performing context discovery in ad-hoc networks based on the use of attenuated Bloom filters is proposed in this report. In order to investigate the performance of this approach, a model has been developed. This document describes the model and its validation. The model has been

  20. A modern course in aeroelasticity

    CERN Document Server

    Dowell, Earl H

    2015-01-01

    This book cover the basics of aeroelasticity or the dynamics of fluid-structure interaction. While the field began in response to the rapid development of aviation, it has now expanded into many branches of engineering and scientific disciplines and treat physical phenomena from aerospace engineering, bioengineering, civil engineering, and mechanical engineering in addition to drawing the attention of mathematicians and physicists.   The basic questions addressed are dynamic stability and response of fluid structural systems as revealed  by both linear and nonlinear mathematical models and correlation with experiment. The use of scaled models and full scale experiments and tests play a key role where theory is not considered sufficiently reliable.  In this new edition the more recent literature on nonlinear aeroelasticity has been brought up to date and the opportunity has been taken to correct the inevitable typographical errors that the authors and our readers have found to date. The early chapters of t...

  1. RELAP5-3D code validation for RBMK phenomena

    International Nuclear Information System (INIS)

    Fisher, J.E.

    1999-01-01

    The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena

  2. 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.

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

    International Nuclear Information System (INIS)

    Rowley, C.W.; Simpson, G.T.; Young, R.K.

    1990-01-01

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

  4. Validation of the containment code Sirius: interpretation of an explosion experiment on a scale model

    International Nuclear Information System (INIS)

    Blanchet, Y.; Obry, P.; Louvet, J.; Deshayes, M.; Phalip, C.

    1979-01-01

    The explicit 2-D axisymmetric Langrangian code SIRIUS, developed at the CEA/DRNR, Cadarache, deals with transient compressive flows in deformable primary tanks with more or less complex internal component geometries. This code has been subjected to a two-year intensive validation program on scale model experiments and a number of improvements have been incorporated. This paper presents a recent calculation of one of these experiments using the SIRIUS code, and the comparison with experimental results shows the encouraging possibilities of this Lagrangian code

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

    International Nuclear Information System (INIS)

    Haapalehto, T.

    1995-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Wren, D.J.; Popov, N.; Snell, V.G.

    2004-01-01

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

  7. 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

  8. Automation of RELAP5 input calibration and code validation using genetic algorithm

    International Nuclear Information System (INIS)

    Phung, Viet-Anh; Kööp, Kaspar; Grishchenko, Dmitry; Vorobyev, Yury; Kudinov, Pavel

    2016-01-01

    Highlights: • Automated input calibration and code validation using genetic algorithm is presented. • Predictions generally overlap experiments for individual system response quantities (SRQs). • It was not possible to predict simultaneously experimental maximum flow rate and oscillation period. • Simultaneous consideration of multiple SRQs is important for code validation. - Abstract: Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called “user effects”. The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the

  9. Automation of RELAP5 input calibration and code validation using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Vorobyev, Yury, E-mail: yura3510@gmail.com [National Research Center “Kurchatov Institute”, Kurchatov square 1, Moscow 123182 (Russian Federation); Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden)

    2016-04-15

    Highlights: • Automated input calibration and code validation using genetic algorithm is presented. • Predictions generally overlap experiments for individual system response quantities (SRQs). • It was not possible to predict simultaneously experimental maximum flow rate and oscillation period. • Simultaneous consideration of multiple SRQs is important for code validation. - Abstract: Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called “user effects”. The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the

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

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Malet, J.; Gelain, T.; Degrees du Lou, O.; Daru, V.

    2011-01-01

    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, CO 2 and SF 6 ) 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)

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

    International Nuclear Information System (INIS)

    Maheras, S.J.; Pippen, H.K.

    1995-05-01

    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

  13. In-vessel core degradation code validation matrix

    International Nuclear Information System (INIS)

    Haste, T.J.; Adroguer, B.; Gauntt, R.O.; Martinez, J.A.; Ott, L.J.; Sugimoto, J.; Trambauer, K.

    1996-01-01

    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.)

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

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Kim, Sangji

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Oberkampf, W.L.

    1993-01-01

    The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ''calibrated code,'' ''validated code,'' and a ''validation experiment'' is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance

  16. Effects of extreme wind shear on aeroelastic modal damping of wind turbines

    DEFF Research Database (Denmark)

    Skjoldan, P.F.; Hansen, Morten Hartvig

    2013-01-01

    Wind shear is an important contributor to fatigue loads on wind turbines. Because it causes an azimuthal variation in angle of attack, it can also affect aerodynamic damping. In this paper, a linearized model of a wind turbine, based on the nonlinear aeroelastic code BHawC, is used to investigate...

  17. 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.

  18. An approach to verification and validation of MHD codes for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Smolentsev, S., E-mail: sergey@fusion.ucla.edu [University of California, Los Angeles (United States); Badia, S. [Centre Internacional de Mètodes Numèrics en Enginyeria, Barcelona (Spain); Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Bhattacharyay, R. [Institute for Plasma Research, Gandhinagar, Gujarat (India); Bühler, L. [Karlsruhe Institute of Technology (Germany); Chen, L. [University of Chinese Academy of Sciences, Beijing (China); Huang, Q. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui (China); Jin, H.-G. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Krasnov, D. [Technische Universität Ilmenau (Germany); Lee, D.-W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Mas de les Valls, E. [Centre Internacional de Mètodes Numèrics en Enginyeria, Barcelona (Spain); Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Mistrangelo, C. [Karlsruhe Institute of Technology (Germany); Munipalli, R. [HyPerComp, Westlake Village (United States); Ni, M.-J. [University of Chinese Academy of Sciences, Beijing (China); Pashkevich, D. [St. Petersburg State Polytechnical University (Russian Federation); Patel, A. [Universitat Politècnica de Catalunya – Barcelona Tech (Spain); Pulugundla, G. [University of California, Los Angeles (United States); Satyamurthy, P. [Bhabha Atomic Research Center (India); Snegirev, A. [St. Petersburg State Polytechnical University (Russian Federation); Sviridov, V. [Moscow Power Engineering Institute (Russian Federation); Swain, P. [Bhabha Atomic Research Center (India); and others

    2015-11-15

    Highlights: • Review of status of MHD codes for fusion applications. • Selection of five benchmark problems. • Guidance for verification and validation of MHD codes for fusion applications. - Abstract: We propose a new activity on verification and validation (V&V) of MHD codes presently employed by the fusion community as a predictive capability tool for liquid metal cooling applications, such as liquid metal blankets. The important steps in the development of MHD codes starting from the 1970s are outlined first and then basic MHD codes, which are currently in use by designers of liquid breeder blankets, are reviewed. A benchmark database of five problems has been proposed to cover a wide range of MHD flows from laminar fully developed to turbulent flows, which are of interest for fusion applications: (A) 2D fully developed laminar steady MHD flow, (B) 3D laminar, steady developing MHD flow in a non-uniform magnetic field, (C) quasi-two-dimensional MHD turbulent flow, (D) 3D turbulent MHD flow, and (E) MHD flow with heat transfer (buoyant convection). Finally, we introduce important details of the proposed activities, such as basic V&V rules and schedule. The main goal of the present paper is to help in establishing an efficient V&V framework and to initiate benchmarking among interested parties. The comparison results computed by the codes against analytical solutions and trusted experimental and numerical data as well as code-to-code comparisons will be presented and analyzed in companion paper/papers.

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

    International Nuclear Information System (INIS)

    Hermann, O.W.

    2000-01-01

    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

  20. Gap conductance model validation in the TASS/SMR-S code

    International Nuclear Information System (INIS)

    Ahn, Sang-Jun; Yang, Soo-Hyung; Chung, Young-Jong; Bae, Kyoo-Hwan; Lee, Won-Jae

    2011-01-01

    An advanced integral pressurized water reactor, SMART (System-Integrated Modular Advanced ReacTor) has been developed by KAERI (Korea Atomic Energy Research and Institute). The purposes of the SMART are sea water desalination and an electricity generation. For the safety evaluation and performance analysis of the SMART, TASS/SMR-S (Transient And Setpoint Simulation/System-integrated Modular Reactor) code, has been developed. In this paper, the gap conductance model for the calculation of gap conductance has been validated by using another system code, MARS code, and experimental results. In the validation, the behaviors of fuel temperature and gap width are selected as the major parameters. According to the evaluation results, the TASS/SMR-S code predicts well the behaviors of fuel temperatures and gap width variation, compared to the MARS calculation results and experimental data. (author)

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

    Science.gov (United States)

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

    2006-01-01

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

  2. Validation of system codes RELAP5 and SPECTRA for natural convection boiling in narrow channels

    Energy Technology Data Exchange (ETDEWEB)

    Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; Slootman, M.L.F.; Wiersema, H.T.

    2016-10-15

    Highlights: • Computer codes RELAP5/Mod3.3 and SPECTRA 3.61 validated for boiling in narrow channels. • Validated codes can be used for LOCA analyses in research reactors. • Code validation based on natural convection boiling in narrow channels experiments. - Abstract: Safety analyses of LOCA scenarios in nuclear power plants are performed with so called thermal–hydraulic system codes, such as RELAP5. Such codes are validated for typical fuel geometries applied in nuclear power plants. The question considered by this article is if the codes can be applied for LOCA analyses in research reactors, in particular exceeding CHF in very narrow channels. In order to answer this question, validation calculations were performed with two thermal–hydraulic system codes: RELAP and SPECTRA. The validation was based on natural convection boiling in narrow channels experiments, performed by Prof. Monde et al. in the years 1990–2000. In total 42 vertical tube and annulus experiments were simulated with both codes. A good agreement of the calculated values with the measured data was observed. The main conclusions are: • The computer codes RELAP5/Mod 3.3 (US NRC version) and SPECTRA 3.61 have been validated for natural convection boiling in narrow channels using experiments of Monde. The dimensions applied in the experiments were performed for a range that covers the values observed in typical research reactors. Therefore it is concluded that both codes are validated and can be used for LOCA analyses in research reactors, including natural convection boiling. The applicability range of the present validation is: hydraulic diameters of 1.1 ⩽ D{sub hyd} ⩽ 9.0 mm, heated lengths of 0.1 ⩽ L ⩽ 1.0 m, pressures of 0.10 ⩽ P ⩽ 0.99 MPa. In most calculations the burnout was predicted to occur at lower power than that observed in the experiments. In several cases the burnout was observed at higher power. The overprediction was not larger than 16% in RELAP and 15% in

  3. Sizing and scaling requirements of a large-scale physical model for code validation

    International Nuclear Information System (INIS)

    Khaleel, R.; Legore, T.

    1990-01-01

    Model validation is an important consideration in application of a code for performance assessment and therefore in assessing the long-term behavior of the engineered and natural barriers of a geologic repository. Scaling considerations relevant to porous media flow are reviewed. An analysis approach is presented for determining the sizing requirements of a large-scale, hydrology physical model. The physical model will be used to validate performance assessment codes that evaluate the long-term behavior of the repository isolation system. Numerical simulation results for sizing requirements are presented for a porous medium model in which the media properties are spatially uncorrelated

  4. Aero-Elastic Optimization of a 10 MW Wind Turbine

    DEFF Research Database (Denmark)

    Zahle, Frederik; Tibaldi, Carlo; Verelst, David Robert

    2015-01-01

    This article describes a multi-disciplinary optimization and analysis tool for wind turbines that is based on the open-source framework OpenMDAO. Interfaces to several simulation codes have been implemented which allows for a wide variety of problem formulations and combinations of models....... In this article concurrent aeroelastic optimization of a 10 MW wind turbine rotor is carried out with respect to material distribution distribution and planform. The optimizations achieve up to 13% mass reduction while maintaining the same power production compared to the baseline DTU 10MW RWT....

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yoon Hee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    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.

  6. DEVELOPMENT AND VALIDATION OF A NUCLEAR FUEL CYCLE ANALYSIS TOOL: A FUTURE CODE

    Directory of Open Access Journals (Sweden)

    S.K. KIM

    2013-10-01

    Full Text Available 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.

  7. Implementation of QR Code and Digital Signature to Determine the Validity of KRS and KHS Documents

    Directory of Open Access Journals (Sweden)

    Fatich Fazlur Rochman

    2017-05-01

    Full Text Available Universitas Airlangga students often find it difficult to verify the mark that came out in the Kartu Hasil Studi (KHS is called Study Result Card or courses taken in the Kartu Rencana Studi (KRS is called Study Plan Card, if there are changes to the data on the system used Universitas Airlangga. This complicated KRS and KHS verification process happened because the KRS and KHS documents that owned by student is easier to counterfeit than the data in the system. Implementation digital signature and QR Code technology as a solution that can prove the validity of KRS or KHS. The KRS and KHS validation system developed by Digital Signature and QR Code. QR Code is a type of matrix code that was developed as a code that allows its contents to be decoded at high speed while the Digital Signature has a function as a marker on the data to ensure that the data is the original data. The verification process was divided into two types are reading the Digital Signature and printing document that works by scanning the data from QR Code. The application of the system is carried out were the addition of the QR Code on KRS and KHS, required a readiness of human resources. 

  8. Aeroelastic characteristics of composite bearingless rotor blades

    Science.gov (United States)

    Bielawa, R. L.

    1976-01-01

    Owing to the inherent unique structural features of composite bearingless rotors, various assumptions upon which conventional rotor aeroelastic analyses are formulated, are violated. Three such features identified are highly nonlinear and time-varying structural twist, structural redundancy in bending and torsion, and for certain configurations a strongly coupled low frequency bending-torsion mode. An examination of these aeroelastic considerations and appropriate formulations required for accurate analyses of such rotor systems is presented. Also presented are test results from a dynamically scaled model rotor and complementary analytic results obtained with the appropriately reformulated aeroelastic analysis.

  9. A methodology for aeroelastic constraint analysis in a conceptual design environment

    Science.gov (United States)

    de Baets, Peter Wilfried Gaston

    The objective of this study is the infusion of aeroelastic constraint knowledge into the design space. The mapping of such aeroelastic information in the conceptual design space has long been a desire of the design community. The conceptual design phase of an aircraft is a multidisciplinary environment and has the most influence on the future design of the vehicle. However, sufficient results cannot he obtained in a timely enough manner to materially contribute to early design decisions. Furthermore, the natural division of the engineering team into specialty groups is not well supported by the monolithic aerodynamic-structures codes typically used in modern aeroelastic analysis. The research examines how the Bi-Level Integrated System Synthesis decomposition technique can be adapted to perform as the conceptual aeroelastic design tool. The study describes a comprehensive solution of the aeroelastic coupled problem cast in this decomposition format and implemented in an integrated framework. The method is supported by application details of a proof of concept high speed vehicle. Physics-based codes such as finite element and an aerodynamic panel method are used to model the high-definition geometric characteristics of the vehicle. A synthesis and sizing code was added to referee the conflicts that arise between the two disciplines. This research's novelty lies in four points. First is the use of physics-based tools at the conceptual design phase to calculate the aeroelastic properties. Second is the projection of flutter and divergence velocity constraint lines in a power loading versus wing loading graph. Third is the aeroelastic assessment time reduction, which has moved from a matter of years to months. Lastly, this assessment allowed verification of the impact of changing velocity, altitude, and angle of attack on the aeroelastic properties. This then allowed identification of robust design space with respect to these three mission properties. The method

  10. Validation and application of the system code ATHLET-CD for BWR severe accident analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Marcello, Valentino, E-mail: valentino.marcello@kit.edu; Imke, Uwe; Sanchez, Victor

    2016-10-15

    Highlights: • We present the application of the system code ATHLET-CD code for BWR safety analyses. • Validation of core in-vessel models is performed based on KIT CORA experiments. • A SB-LOCA scenario is simulated on a generic German BWR plant up to vessel failure. • Different core reflooding possibilities are investigated to mitigate the accident consequences. • ATHLET-CD modelling features reflect the current state of the art of severe accident codes. - Abstract: This paper is aimed at the validation and application of the system code ATHLET-CD for the simulation of severe accident phenomena in Boiling Water Reactors (BWR). The corresponding models for core degradation behaviour e.g., oxidation, melting and relocation of core structural components are validated against experimental data available from the CORA-16 and -17 bundle tests. Model weaknesses are discussed along with needs for further code improvements. With the validated ATHLET-CD code, calculations are performed to assess the code capabilities for the prediction of in-vessel late phase core behaviour and reflooding of damaged fuel rods. For this purpose, a small break LOCA scenario for a generic German BWR with postulated multiple failures of the safety systems was selected. In the analysis, accident management measures represented by cold water injection into the damaged reactor core are addressed to investigate the efficacy in avoiding or delaying the failure of the reactor pressure vessel. Results show that ATHLET-CD is applicable to the description of BWR plant behaviour with reliable physical models and numerical methods adopted for the description of key in-vessel phenomena.

  11. Validation and verification of the ORNL Monte Carlo codes for nuclear safety analysis

    International Nuclear Information System (INIS)

    Emmett, M.B.

    1993-01-01

    The process of ensuring the quality of computer codes can be very time consuming and expensive. The Oak Ridge National Laboratory (ORNL) Monte Carlo codes all predate the existence of quality assurance (QA) standards and configuration control. The number of person-years and the amount of money spent on code development make it impossible to adhere strictly to all the current requirements. At ORNL, the Nuclear Engineering Applications Section of the Computing Applications Division is responsible for the development, maintenance, and application of the Monte Carlo codes MORSE and KENO. The KENO code is used for doing criticality analyses; the MORSE code, which has two official versions, CGA and SGC, is used for radiation transport analyses. Because KENO and MORSE were very thoroughly checked out over the many years of extensive use both in the United States and in the international community, the existing codes were open-quotes baselined.close quotes This means that the versions existing at the time the original configuration plan is written are considered to be validated and verified code systems based on the established experience with them

  12. Development and validation of computer codes for analysis of PHWR containment behaviour

    International Nuclear Information System (INIS)

    Markandeya, S.G.; Haware, S.K.; Ghosh, A.K.; Venkat Raj, V.

    1997-01-01

    In order to ensure that the design intent of the containment of Indian Pressurised Heavy Water Reactors (IPHWRs) is met, both analytical and experimental studies are being pursued at BARC. As a part of analytical studies, computer codes for predicting the behaviour of containment under various accident scenarios are developed/adapted. These include codes for predicting 1) pressure, temperature transients in the containment following either Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB), 2) hydrogen behaviour in respect of its distribution, combustion and the performance of proposed mitigation systems, and 3) behaviour of fission product aerosols in the piping circuits of the primary heat transport system and in the containment. All these codes have undergone thorough validation using data obtained from in-house test facilities or from international sources. Participation in the International Standard Problem (ISP) exercises has also helped in validation of the codes. The present paper briefly describes some of these codes and the various exercises performed for their validation. (author)

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    2004-08-01

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

  15. 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...

  16. Further validation of liquid metal MHD code for unstructured grid based on OpenFOAM

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Jingchao; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; He, Qingyun; Ye, Minyou

    2015-11-15

    Highlights: • Specific correction scheme has been adopted to revise the calculating result for non-orthogonal meshes. • The developed MHD code based on OpenFOAM platform has been validated by benchmark cases under uniform and non-uniform magnetic field in round and rectangular ducts. • ALEX experimental results have been used to validate the MHD code based on OpenFOAM. - Abstract: In fusion liquid metal blankets, complex geometries involving contractions, expansions, bends, manifolds are very common. The characteristics of liquid metal flow in these geometries are significant. In order to extend the magnetohydrodynamic (MHD) solver developed on OpenFOAM platform to be applied in the complex geometry, the MHD solver based on unstructured meshes has been implemented. The adoption of non-orthogonal correction techniques in the solver makes it possible to process the non-orthogonal meshes in complex geometries. The present paper focused on the validation of the code under critical conditions. An analytical solution benchmark case and two experimental benchmark cases were conducted to validate the code. Benchmark case I is MHD flow in a circular pipe with arbitrary electric conductivity of the walls in a uniform magnetic field. Benchmark cases II and III are experimental cases of 3D laminar steady MHD flow under fringing magnetic field. In all these cases, the numerical results match well with the benchmark cases.

  17. Further validation of liquid metal MHD code for unstructured grid based on OpenFOAM

    International Nuclear Information System (INIS)

    Feng, Jingchao; Chen, Hongli; He, Qingyun; Ye, Minyou

    2015-01-01

    Highlights: • Specific correction scheme has been adopted to revise the calculating result for non-orthogonal meshes. • The developed MHD code based on OpenFOAM platform has been validated by benchmark cases under uniform and non-uniform magnetic field in round and rectangular ducts. • ALEX experimental results have been used to validate the MHD code based on OpenFOAM. - Abstract: In fusion liquid metal blankets, complex geometries involving contractions, expansions, bends, manifolds are very common. The characteristics of liquid metal flow in these geometries are significant. In order to extend the magnetohydrodynamic (MHD) solver developed on OpenFOAM platform to be applied in the complex geometry, the MHD solver based on unstructured meshes has been implemented. The adoption of non-orthogonal correction techniques in the solver makes it possible to process the non-orthogonal meshes in complex geometries. The present paper focused on the validation of the code under critical conditions. An analytical solution benchmark case and two experimental benchmark cases were conducted to validate the code. Benchmark case I is MHD flow in a circular pipe with arbitrary electric conductivity of the walls in a uniform magnetic field. Benchmark cases II and III are experimental cases of 3D laminar steady MHD flow under fringing magnetic field. In all these cases, the numerical results match well with the benchmark cases.

  18. Non-linear aeroelastic prediction for aircraft applications

    Science.gov (United States)

    de C. Henshaw, M. J.; Badcock, K. J.; Vio, G. A.; Allen, C. B.; Chamberlain, J.; Kaynes, I.; Dimitriadis, G.; Cooper, J. E.; Woodgate, M. A.; Rampurawala, A. M.; Jones, D.; Fenwick, C.; Gaitonde, A. L.; Taylor, N. V.; Amor, D. S.; Eccles, T. A.; Denley, C. J.

    2007-05-01

    in this domain. This is set within the context of a generic industrial process and the requirements of UK and US aeroelastic qualification. A range of test cases, from simple small DOF cases to full aircraft, have been used to evaluate and validate the non-linear methods developed and to make comparison with the linear methods in everyday use. These have focused mainly on aerodynamic non-linearity, although some results for structural non-linearity are also presented. The challenges associated with time domain (coupled computational fluid dynamics-computational structural model (CFD-CSM)) methods have been addressed through the development of grid movement, fluid-structure coupling, and control surface movement technologies. Conclusions regarding the accuracy and computational cost of these are presented. The computational cost of time-domain methods, despite substantial improvements in efficiency, remains high. However, significant advances have been made in reduced order methods, that allow non-linear behaviour to be modelled, but at a cost comparable with that of the regular linear methods. Of particular note is a method based on Hopf bifurcation that has reached an appropriate maturity for deployment on real aircraft configurations, though only limited results are presented herein. Results are also presented for dynamically linearised CFD approaches that hold out the possibility of non-linear results at a fraction of the cost of time coupled CFD-CSM methods. Local linearisation approaches (higher order harmonic balance and continuation method) are also presented; these have the advantage that no prior assumption of the nature of the aeroelastic instability is required, but currently these methods are limited to low DOF problems and it is thought that these will not reach a level of maturity appropriate to real aircraft problems for some years to come. Nevertheless, guidance on the most likely approaches has been derived and this forms the basis for ongoing

  19. CFD Code Validation against Stratified Air-Water Flow Experimental Data

    Directory of Open Access Journals (Sweden)

    F. Terzuoli

    2008-01-01

    Full Text Available Pressurized thermal shock (PTS modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV lifetime is the cold water emergency core cooling (ECC injection into the cold leg during a loss of coolant accident (LOCA. Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX, and a research code (NEPTUNE CFD. The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.

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

    International Nuclear Information System (INIS)

    Macek, J.

    2001-01-01

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

  1. CFD Code Validation against Stratified Air-Water Flow Experimental Data

    International Nuclear Information System (INIS)

    Terzuoli, F.; Galassi, M.C.; Mazzini, D.; D'Auria, F.

    2008-01-01

    Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mecanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code (NEPTUNE CFD). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling

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

    International Nuclear Information System (INIS)

    Reyes H, A.; Ortiz R, J. M.; Reyes A, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

    2014-08-01

    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 6 LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: 252 Cf and 239 PuBe 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)

  3. Multifidelity Robust Aeroelastic Design, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Nielsen Engineering & Research (NEAR) proposes a new method to generate mathematical models of wind-tunnel models and flight vehicles for robust aeroelastic...

  4. Research on verification and validation strategy of detonation fluid dynamics code of LAD2D

    Science.gov (United States)

    Wang, R. L.; Liang, X.; Liu, X. Z.

    2017-07-01

    The verification and validation (V&V) is an important approach in the software quality assurance of code in complex engineering application. Reasonable and efficient V&V strategy can achieve twice the result with half the effort. This article introduces the software-Lagrangian adaptive hydrodynamics code in 2D space (LAD2D), which is self-developed software in detonation CFD with plastic-elastic structure. The V&V strategy of this detonation CFD code is presented based on the foundation of V&V methodology for scientific software. The basic framework of the module verification and the function validation is proposed, composing the detonation fluid dynamics model V&V strategy of LAD2D.

  5. Independent verification and validation testing of the FLASH computer code, Versiion 3.0

    International Nuclear Information System (INIS)

    Martian, P.; Chung, J.N.

    1992-06-01

    Independent testing of the FLASH computer code, Version 3.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at various Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Verification tests, and validation tests, were used to determine the operational status of the FLASH computer code. These tests were specifically designed to test: correctness of the FORTRAN coding, computational accuracy, and suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: blind testing, independent applications, and graduated difficulty of test cases. Both quantitative and qualitative testing was performed through evaluating relative root mean square values and graphical comparisons of the numerical, analytical, and experimental data. Four verification test were used to check the computational accuracy and correctness of the FORTRAN coding, and three validation tests were used to check the suitability to simulating actual conditions. These tests cases ranged in complexity from simple 1-D saturated flow to 2-D variably saturated problems. The verification tests showed excellent quantitative agreement between the FLASH results and analytical solutions. The validation tests showed good qualitative agreement with the experimental data. Based on the results of this testing, it was concluded that the FLASH code is a versatile and powerful two-dimensional analysis tool for fluid flow. In conclusion, all aspects of the code that were tested, except for the unit gradient bottom boundary condition, were found to be fully operational and ready for use in hydrological and environmental studies

  6. Design validation of the ITER EC upper launcher according to codes and standards

    Energy Technology Data Exchange (ETDEWEB)

    Spaeh, Peter, E-mail: peter.spaeh@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Gagliardi, Mario [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); F4E, Fusion for Energy, Joint Undertaking, Barcelona (Spain); Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Weinhorst, Bastian [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2015-10-15

    Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

  7. Design validation of the ITER EC upper launcher according to codes and standards

    International Nuclear Information System (INIS)

    Spaeh, Peter; Aiello, Gaetano; Gagliardi, Mario; Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro; Weinhorst, Bastian

    2015-01-01

    Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

  8. System code improvements for modelling passive safety systems and their validation

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

  9. 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.

  10. Uncertainty propagation applied to multi-scale thermal-hydraulics coupled codes. A step towards validation

    Energy Technology Data Exchange (ETDEWEB)

    Geffray, Clotaire Clement

    2017-03-20

    The work presented here constitutes an important step towards the validation of the use of coupled system thermal-hydraulics and computational fluid dynamics codes for the simulation of complex flows in liquid metal cooled pool-type facilities. First, a set of methods suited for uncertainty and sensitivity analysis and validation activities with regards to the specific constraints of the work with coupled and expensive-to-run codes is proposed. Then, these methods are applied to the ATHLET - ANSYS CFX model of the TALL-3D facility. Several transients performed at this latter facility are investigated. The results are presented, discussed and compared to the experimental data. Finally, assessments of the validity of the selected methods and of the quality of the model are offered.

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

    Science.gov (United States)

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

    2015-07-01

    International Classification of Disease, Ninth Revision (ICD-9) diagnosis codes have not been validated for identifying cases of missed abortion where a pregnancy is no longer viable but the cervical os remains closed. Our goal was to assess whether ICD-9 code "632" for missed abortion has high sensitivity and positive predictive value (PPV) in identifying patients in the emergency department (ED) with cases of stable early pregnancy failure (EPF). We studied females ages 13-50 years presenting to the ED of an urban academic medical center. We approached our analysis from two perspectives, evaluating both the sensitivity and PPV of ICD-9 code "632" in identifying patients with stable EPF. All patients with chief complaints "pregnant and bleeding" or "pregnant and cramping" over a 12-month period were identified. We randomly reviewed two months of patient visits and calculated the sensitivity of ICD-9 code "632" for true cases of stable miscarriage. To establish the PPV of ICD-9 code "632" for capturing missed abortions, we identified patients whose visits from the same time period were assigned ICD-9 code "632," and identified those with actual cases of stable EPF. We reviewed 310 patient records (17.6% of 1,762 sampled). Thirteen of 31 patient records assigned ICD-9 code for missed abortion correctly identified cases of stable EPF (sensitivity=41.9%), and 140 of the 142 patients without EPF were not assigned the ICD-9 code "632"(specificity=98.6%). Of the 52 eligible patients identified by ICD-9 code "632," 39 cases met the criteria for stable EPF (PPV=75.0%). ICD-9 code "632" has low sensitivity for identifying stable EPF, but its high specificity and moderately high PPV are valuable for studying cases of stable EPF in epidemiologic studies using administrative data.

  12. Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

    2014-12-01

    The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V and V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement. (author)

  13. Research in aeroelasticity EFP-2005

    DEFF Research Database (Denmark)

    2006-01-01

    In the Energy Research Project ”Program for Research in Applied Aeroelasticity” (EFP2005), Risø National Laboratory (Risø) and the Technical University of Denmark (DTU) have applied and further developed the tools in the aeroelastic design complex. Themain results from the project are: Adding...... a winglet to a wind turbine blade for minimizing the induced drag of the blade led to the biggest increase in power of 1.4%. Transient wind loads during pitch motion are determined using CFD. Compared to theNREL/NASA Ames test, reasonably good agreement is seen. A general method was developed...... for the determination of 3D angle of attack for rotating blades from either measurements or numerical computations using CFD. A model of the far wake behind windturbines was developed for stability studies of the tip vortices in the far wake. Investigating the blade root region showed that the power efficiency, CP...

  14. Integral large scale experiments on hydrogen combustion for severe accident code validation-HYCOM

    International Nuclear Information System (INIS)

    Breitung, W.; Dorofeev, S.; Kotchourko, A.; Redlinger, R.; Scholtyssek, W.; Bentaib, A.; L'Heriteau, J.-P.; Pailhories, P.; Eyink, J.; Movahed, M.; Petzold, K.-G.; Heitsch, M.; Alekseev, V.; Denkevits, A.; Kuznetsov, M.; Efimenko, A.; Okun, M.V.; Huld, T.; Baraldi, D.

    2005-01-01

    A joint research project was carried out in the EU Fifth Framework Programme, concerning hydrogen risk in a nuclear power plant. The goals were: Firstly, to create a new data base of results on hydrogen combustion experiments in the slow to turbulent combustion regimes. Secondly, to validate the partners CFD and lumped parameter codes on the experimental data, and to evaluate suitable parameter sets for application calculations. Thirdly, to conduct a benchmark exercise by applying the codes to the full scale analysis of a postulated hydrogen combustion scenario in a light water reactor containment after a core melt accident. The paper describes the work programme of the project and the partners activities. Significant progress has been made in the experimental area, where test series in medium and large scale facilities have been carried out with the focus on specific effects of scale, multi-compartent geometry, heat losses and venting. The data were used for the validation of the partners CFD and lumped parameter codes, which included blind predictive calculations and pre- and post-test intercomparison exercises. Finally, a benchmark exercise was conducted by applying the codes to the full scale analysis of a hydrogen combustion scenario. The comparison and assessment of the results of the validation phase and of the challenging containment calculation exercise allows a deep insight in the quality, capabilities and limits of the CFD and the lumped parameter tools which are currently in use at various research laboratories

  15. Validation of a Subchannel Analysis Code MATRA Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  16. CSNI Integral Test Facility Matrices for Validation of Best-Estimate Thermal-Hydraulic Computer Codes

    International Nuclear Information System (INIS)

    Glaeser, H.

    2008-01-01

    Internationally agreed Integral Test Facility (ITF) matrices for validation of realistic thermal hydraulic system computer codes were established. ITF development is mainly for Pressurised Water Reactors (PWRs) and Boiling Water Reactors (BWRs). A separate activity was for Russian Pressurised Water-cooled and Water-moderated Energy Reactors (WWER). Firstly, the main physical phenomena that occur during considered accidents are identified, test types are specified, and test facilities suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. In this paper some specific examples from the ITF matrices will also be provided. The matrices will be a guide for code validation, will be a basis for comparisons of code predictions performed with different system codes, and will contribute to the quantification of the uncertainty range of code model predictions. In addition to this objective, the construction of such a matrix is an attempt to record information which has been generated around the world over the last years, so that it is more accessible to present and future workers in that field than would otherwise be the case.

  17. Validation of the WIMSD4M cross-section generation code with benchmark results

    International Nuclear Information System (INIS)

    Deen, J.R.; Woodruff, W.L.; Leal, L.E.

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D 2 O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented

  18. Validation of the WIMSD4M cross-section generation code with benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Leal, L.E. [Oak Ridge National Lab., TN (United States)

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.

  19. Large leak sodium-water reaction code SWACS and its validation

    International Nuclear Information System (INIS)

    Miyake, O.; Shindo, Y.; Hiroi, H.; Tanabe, H.; Sato, M.

    1982-01-01

    A computer code SWACS for analyzing the large leak accident of an LMFBR steam generators has been developed and validated. Five tests data obtained by SWAT-3 test facility were compared with code results. In each of SWAT-3 tests, a double-ended guillotine rupture of one tube was simulated in a helical coil steam generator model with 1/2.5 scaled test vessel to the prototype SG. The analytical results, including an initial pressure spike, a propagated pressure in a secondary system, and a quasi-steady pressure, indicate that the overall large-leak event could be predicted in reasonably good agreement

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

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1995-01-01

    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

  1. Validation of coding algorithms for the identification of patients hospitalized for alcoholic hepatitis using administrative data.

    Science.gov (United States)

    Pang, Jack X Q; Ross, Erin; Borman, Meredith A; Zimmer, Scott; Kaplan, Gilaad G; Heitman, Steven J; Swain, Mark G; Burak, Kelly W; Quan, Hude; Myers, Robert P

    2015-09-11

    Epidemiologic studies of alcoholic hepatitis (AH) have been hindered by the lack of a validated International Classification of Disease (ICD) coding algorithm for use with administrative data. Our objective was to validate coding algorithms for AH using a hospitalization database. The Hospital Discharge Abstract Database (DAD) was used to identify consecutive adults (≥18 years) hospitalized in the Calgary region with a diagnosis code for AH (ICD-10, K70.1) between 01/2008 and 08/2012. Medical records were reviewed to confirm the diagnosis of AH, defined as a history of heavy alcohol consumption, elevated AST and/or ALT (34 μmol/L, and elevated INR. Subgroup analyses were performed according to the diagnosis field in which the code was recorded (primary vs. secondary) and AH severity. Algorithms that incorporated ICD-10 codes for cirrhosis and its complications were also examined. Of 228 potential AH cases, 122 patients had confirmed AH, corresponding to a positive predictive value (PPV) of 54% (95% CI 47-60%). PPV improved when AH was the primary versus a secondary diagnosis (67% vs. 21%; P codes for ascites (PPV 75%; 95% CI 63-86%), cirrhosis (PPV 60%; 47-73%), and gastrointestinal hemorrhage (PPV 62%; 51-73%) had improved performance, however, the prevalence of these diagnoses in confirmed AH cases was low (29-39%). In conclusion the low PPV of the diagnosis code for AH suggests that caution is necessary if this hospitalization database is used in large-scale epidemiologic studies of this condition.

  2. A validation study of the BURNUP and associated options of the MONTE CARLO neutronics code MONK5W

    International Nuclear Information System (INIS)

    Howard, E.A.

    1985-11-01

    This is a report on the validation of the burnup option of the Monte Carlo Neutronics Code MONK5W, together with the associated facilities which allow for control rod movements and power changes. The validation uses reference solutions produced by the Deterministic Neutronics Code LWR-WIMS for a 2D model which represents a whole reactor calculation with control rod movements. (author)

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

    International Nuclear Information System (INIS)

    El bakkari, B.; El Bardouni, T.; Merroun, O.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Chakir, E.

    2009-01-01

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

  4. Development and validation of ALEPH Monte Carlo burn-up code

    International Nuclear Information System (INIS)

    Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.

    2011-01-01

    The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)

  5. Validation of the 3D finite element transport theory code EVENT for shielding applications

    International Nuclear Information System (INIS)

    Warner, Paul; Oliveira, R.E. de

    2000-01-01

    This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P N ) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the code are compared with predictions of the Monte Carlo code MCBEND and also with the results from benchmark shielding experiments. In particular, results are presented for the ASPIS experimental configuration for both neutron and gamma ray calculations using the BUGLE 96 nuclear data library. (author)

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

    International Nuclear Information System (INIS)

    Palmer, I.; Rossiter, G.; White, R.J.

    2000-01-01

    The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)

  7. Examples of Use of SINBAD Database for Nuclear Data and Code Validation

    Science.gov (United States)

    Kodeli, Ivan; Žerovnik, Gašper; Milocco, Alberto

    2017-09-01

    The SINBAD database currently contains compilations and evaluations of over 100 shielding benchmark experiments. The SINBAD database is widely used for code and data validation. Materials covered include: Air, N. O, H2O, Al, Be, Cu, graphite, concrete, Fe, stainless steel, Pb, Li, Ni, Nb, SiC, Na, W, V and mixtures thereof. Over 40 organisations from 14 countries and 2 international organisations have contributed data and work in support of SINBAD. Examples of the use of the database in the scope of different international projects, such as the Working Party on Evaluation Cooperation of the OECD and the European Fusion Programme demonstrate the merit and possible usage of the database for the validation of modern nuclear data evaluations and new computer codes.

  8. 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; Quan, Hude; Kaplan, Gilaad G; Butalia, Sonia; Rabi, Doreen

    2016-08-05

    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. 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. 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). 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. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

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

    International Nuclear Information System (INIS)

    Brittain, I.; Bryce, W.M.; O'Mahoney, R.; Richards, C.G.; Gibson, I.H.; Porter, W.H.L.; Fell, J.

    1984-03-01

    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.)

  10. Validation of a CFD code for Unsteady Flows with cyclic boundary Conditions

    International Nuclear Information System (INIS)

    Kim, Jong-Tae; Kim, Sang-Baik; Lee, Won-Jae

    2006-01-01

    Currently Lilac code is under development to analyze thermo-hydraulics of a high-temperature gas-cooled reactor (GCR). Interesting thermo-hydraulic phenomena in a nuclear reactor are usually unsteady and turbulent. The analysis of the unsteady flows by using a three dimension CFD code is time-consuming if the flow domain is very large. Hopefully, flow domains commonly encountered in the nuclear thermo-hydraulics is periodic. So it is better to use the geometrical characteristics in order to reduce the computational resources. To get the benefits from reducing the computation domains especially for the calculations of unsteady flows, the cyclic boundary conditions are implemented in the parallelized CFD code LILAC. In this study, the parallelized cyclic boundary conditions are validated by solving unsteady laminar and turbulent flows past a circular cylinder

  11. Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU Fuel

    International Nuclear Information System (INIS)

    Tait, J.C.; Gauld, I.; Kerr, A.H.

    1994-10-01

    The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO 2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. 38 refs., 4 figs., 5 tabs

  12. The COSIMA-experiments, a data base for validation of two-phase flow computer codes

    International Nuclear Information System (INIS)

    Class, G.; Meyder, R.; Stratmanns, E.

    1985-12-01

    The report presents an overview on the large data base generated with COSIMA. The data base is to be used to validate and develop computer codes for two-phase flow. In terms of fuel rod behavior it was found that during blowdown under realistic conditions only small strains are reached. For clad rupture extremely high rod internal pressure is necessary. Additionally important results were found in the behavior of a fuel rod simulator and on the effect of thermocouples attached on the cladding outer surface. Post-test calculations, performed with the codes RELAP and DRUFAN show a good agreement with the experiments. This however can be improved if the phase separation models in the codes would be updated. (orig./HP) [de

  13. Validation and uncertainty analysis of the Athlet thermal-hydraulic computer code

    International Nuclear Information System (INIS)

    Glaeser, H.

    1995-01-01

    The computer code ATHLET is being developed by GRS as an advanced best-estimate code for the simulation of breaks and transients in Pressurized Water Reactor (PWRs) and Boiling Water Reactor (BWRs) including beyond design basis accidents. A systematic validation of ATHLET is based on a well balanced set of integral and separate effects tests emphasizing the German combined Emergency Core Cooling (ECC) injection system. When using best estimate codes for predictions of reactor plant states during assumed accidents, qualification of the uncertainty in these calculations is highly desirable. A method for uncertainty and sensitivity evaluation has been developed by GRS where the computational effort is independent of the number of uncertain parameters. (author)

  14. Validation of the ORIGEN-S code for predicting radionuclide inventories in used CANDU fuel

    International Nuclear Information System (INIS)

    Tait, J.C.; Gauld, I.; Kerr, A.H.

    1995-01-01

    The safety assessment being conducted by AECL Research for the concept of deep geological disposal of used CANDU UO 2 fuel requires the calculation of radionuclide inventories in the fuel to provide source terms for radionuclide release. This report discusses the validation of selected actinide and fission-product inventories calculated using the ORIGEN-S code coupled with the WIMS-AECL lattice code, using data from analytical measurements of radioisotope inventories in Pickering CANDU reactor fuel. The recent processing of new ENDF/B-VI cross-section data has allowed the ORIGEN-S calculations to be performed using the most up-to-date nuclear data available. The results indicate that the code is reliably predicting actinide and the majority of fission-product inventories to within the analytical uncertainty. ((orig.))

  15. Validation of Mean Drift Forces Computed with the BEM Code NEMOH

    DEFF Research Database (Denmark)

    Thomsen, Jonas Bjerg

    This report covers a simple investigation of mean drift forces found by use of the boundary element method code NEMOH. The results from NEMOH are compared to analytical results from literature and to numerical values found from the commercial software package WADAM by DNV-GL. The work was conduct...... under the project ”Mooring Solutions for Large Wave Energy Converters”, during", Work Package 4: Full Dynamic Analysis". The validation compares results from a simple sphere and from a vertical cylinder....

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

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J., E-mail: J.H.Choi@iaea.org [International Atomic Energy Agency, Vienna (Austria)

    2013-07-01

    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)

  17. National assessment of validity of coding of acute mastoiditis: a standardised reassessment of 1966 records.

    Science.gov (United States)

    Stalfors, J; Enoksson, F; Hermansson, A; Hultcrantz, M; Robinson, Å; Stenfeldt, K; Groth, A

    2013-04-01

    To investigate the internal validity of the diagnosis code used at discharge after treatment of acute mastoiditis. Retrospective national re-evaluation study of patient records 1993-2007 and make comparison with the original ICD codes. All ENT departments at university hospitals and one large county hospital department in Sweden. A total of 1966 records were reviewed for patients with ICD codes for in-patient treatment of acute (529), chronic (44) and unspecified mastoiditis (21) and acute otitis media (1372). ICD codes were reviewed by the authors with a defined protocol for the clinical diagnosis of acute mastoiditis. Those not satisfying the diagnosis were given an alternative diagnosis. Of 529 records with ICD coding for acute mastoiditis, 397 (75%) were found to meet the definition of acute mastoiditis used in this study, while 18% were not diagnosed as having any type of mastoiditis after review. Review of the in-patients treated for acute media otitis identified an additional 60 cases fulfilling the definition of acute mastoiditis. Overdiagnosis was common, and many patients with a diagnostic code indicating acute mastoiditis had been treated for external otitis or otorrhoea with transmyringeal drainage. The internal validity of the diagnosis acute mastoiditis is dependent on the use of standardised, well-defined criteria. Reliability of diagnosis is fundamental for the comparison of results from different studies. Inadequate reliability in the diagnosis of acute mastoiditis also affects calculations of incidence rates and statistical power and may also affect the conclusions drawn from the results. © 2013 Blackwell Publishing Ltd.

  18. Validation of ASTEC v1.0 computer code against FPT2 test

    International Nuclear Information System (INIS)

    Mladenov, I.; Tusheva, P.; Kalchev, B.; Dimov, D.; Ivanov, I.

    2005-01-01

    The aim of the work is by various nodalization schemes of the model to investigate the ASTEC v1.0 computer code sensitivity and to validate the code against PHEBUS - FPT2 experiment. This code is used for severe accident analysis. The aim corresponds to the main technical objective of the experiment which is to contribute to the validation of models and computer codes to be used for the calculation of the source term in case of a severe accident in a Light Water Reactor. The objective's scope of the FPT2 is large - separately for the bundle, the experimental circuit and the containment. Additional objectives are to characterize aerosol sizing and deposition processes, and also potential FP poisoning effects on hydrogen recombiner coupons exposed to containment atmospheric conditions representative of a LWR severe accident. The analyses of the results of the performed calculations show a good accordance with the reference case calculations, and then with the experimental data. Some differences in the calculations for the thermal behavior appear locally during the oxidation phase and the heat-up phase. There is very good confirmation regarding the volatile and semi-volatile fission products release from the fuel pellets. Important for analysis of the process is the final axial distribution of the mass of fuel relocation obtained at the end of the calculation

  19. Study of the microstructure of neutron irradiated beryllium for the validation of the ANFIBE code

    International Nuclear Information System (INIS)

    Rabaglino, E.; Ferrero, C.; Reimann, J.; Ronchi, C.; Schulenberg, T.

    2002-01-01

    The behaviour of beryllium under fast neutron irradiation is a key issue of the helium cooled pebble bed tritium breeding blanket, due to the production of large quantities of helium and of a non-negligible amount of tritium. To optimise the design, a reliable prediction of swelling due to helium bubbles and of tritium inventory during normal and off-normal operation of a fusion power reactor is needed. The ANFIBE code (ANalysis of Fusion Irradiated BEryllium) is being developed to meet this need. The code has to be applied in a range of irradiation conditions where no experimental data are available, therefore a detailed gas kinetics model, and a specific and particularly careful validation strategy are needed. The validation procedure of the first version of the code was based on macroscopic data of swelling and tritium release. This approach is, however, incomplete, since a verification of the microscopic behaviour of the gas in the metal is necessary to obtain a reliable description of swelling. This paper discusses a general strategy for a thorough validation of the gas kinetics models in ANFIBE. The microstructure characterisation of weakly irradiated beryllium pebbles, with different visual examination techniques, is then presented as an example of the application of this strategy. In particular, the advantage of developing 3D techniques, such as X-ray microtomography, is demonstrated

  20. Nonlinear aeroelastic behavior of compliant airfoils

    International Nuclear Information System (INIS)

    Thwapiah, G; Campanile, L F

    2010-01-01

    Since the beginning of aviation and up to the present time, airfoils have always been built as rigid structures. They are designed to fly under their divergence speed in order to avoid static aeroelastic instabilities and the resulting large deformations, which are not compatible with the typically low compliance of such airfoils. In recent years, research on airfoil morphing has generated interest in innovative ideas like the use of compliant systems, i.e. systems built to allow for large deformations without failure, in airfoil construction. Such systems can operate in the neighborhood of divergence and take advantage of large aeroelastic servo-effects. This, in turn, allows compact, advanced actuators to control the airfoil's deformation and loads, and hence complement or even replace conventional flaps. In order to analyze and design such compliant, active aeroelastic structures a nonlinear approach to static aeroelasticity is needed, which takes into account the effect of large deformations on aerodynamics and structure. Such an analytical approach is presented in this paper and applied to a compliant passive airfoil as the preliminary step in the realization of a piezoelectrically driven, active aeroelastic airfoil. Wind tunnel test results are also presented and compared with the analytic prediction. The good agreement and the observed behavior in the wind tunnel give confidence in the potential of this innovative idea

  1. Nonlinear aeroelastic behavior of compliant airfoils

    Science.gov (United States)

    Thwapiah, G.; Campanile, L. F.

    2010-03-01

    Since the beginning of aviation and up to the present time, airfoils have always been built as rigid structures. They are designed to fly under their divergence speed in order to avoid static aeroelastic instabilities and the resulting large deformations, which are not compatible with the typically low compliance of such airfoils. In recent years, research on airfoil morphing has generated interest in innovative ideas like the use of compliant systems, i.e. systems built to allow for large deformations without failure, in airfoil construction. Such systems can operate in the neighborhood of divergence and take advantage of large aeroelastic servo-effects. This, in turn, allows compact, advanced actuators to control the airfoil's deformation and loads, and hence complement or even replace conventional flaps. In order to analyze and design such compliant, active aeroelastic structures a nonlinear approach to static aeroelasticity is needed, which takes into account the effect of large deformations on aerodynamics and structure. Such an analytical approach is presented in this paper and applied to a compliant passive airfoil as the preliminary step in the realization of a piezoelectrically driven, active aeroelastic airfoil. Wind tunnel test results are also presented and compared with the analytic prediction. The good agreement and the observed behavior in the wind tunnel give confidence in the potential of this innovative idea.

  2. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L.; Xolocostli M, J. V.; Gomez T, A. M.

    2016-09-01

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S_N, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO_2 cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  3. Validity of the Italian Code of Ethics for everyday nursing practice.

    Science.gov (United States)

    Gobbi, Paola; Castoldi, Maria Grazia; Alagna, Rosa Anna; Brunoldi, Anna; Pari, Chiara; Gallo, Annamaria; Magri, Miriam; Marioni, Lorena; Muttillo, Giovanni; Passoni, Claudia; Torre, Anna La; Rosa, Debora; Carnevale, Franco A

    2016-12-07

    The research question for this study was as follows: Is the Code of Ethics for Nurses in Italy (Code) a valid or useful decision-making instrument for nurses faced with ethical problems in their daily clinical practice? Focus groups were conducted to analyze specific ethical problems through 11 case studies. The analysis was conducted using sections of the Code as well as other relevant documents. Each focus group had a specific theme and nurses participated freely in the discussions according to their respective clinical competencies. The executive administrative committee of the local nursing licensing council provided approval for conducting this project. Measures were taken to protect the confidentiality of consenting participants. The answer to the research question posed for this investigation was predominantly positive. Many sections of the Code were useful for discussion and identifying possible solutions for the ethical problems presented in the 11 cases. We concluded that the Code of Ethics for Nurses in Italy can be a valuable aid in daily practice in most clinical situations that can give rise to ethical problems. © The Author(s) 2016.

  4. Improvements and validation of the transient analysis code MOREL for molten salt reactors

    International Nuclear Information System (INIS)

    Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

    2017-01-01

    The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (author)

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

    DEFF Research Database (Denmark)

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

    2016-01-01

    BACKGROUND: Large health care databases are a valuable source of infectious disease epidemiology if diagnoses are valid. The aim of this study was to investigate the accuracy of the recorded diagnosis coding of herpes simplex encephalitis (HSE) in the Danish National Patient Registry (DNPR...... (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 of 75.9% (95% CI: 56.5-89.7), thereby representing the highest PPV. The estimated sensitivity was 95.5%. CONCLUSION: The PPVs of the ICD-10 diagnosis coding for adult HSE in the DNPR were relatively low. Hence...

  6. Validation of the Thermal-Hydraulic Model in the SACAP Code with the ISP Tests

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon-Ho; Kim, Dong-Min; Park, Chang-Hwan [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    In safety viewpoint, the pressure of the containment is the important parameter, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In Korea, there have been an extensive efforts to develop the computer code which can analyze the severe accident behavior of the pressurized water reactor. The development has been done in a modularized manner and SACAP(Severe Accident Containment Analysis Package) code is now under final stage of development. SACAP code adopts LP(Lumped Parameter) model and is applicable to analyze the synthetic behavior of the containment during severe accident occurred by thermal-hydraulic transient, combustible gas burn, direct containment heating by high pressure melt ejection, steam explosion and molten core-concrete interaction. The analyses of a number of ISP(International Standard Problem) experiments were done as a part of the SACAP code V and V(verification and validation). In this paper, the SACAP analysis results for ISP-35 NUPEC and ISP-47 TOSQAN are presented including comparison with other existing NPP simulation codes. In this paper, we selected and analyzed ISP-35 NUPEC, ISP-47 TOSQAN in order to confirm the computational performance of SACAP code currently under development. Now the multi-node analysis for the ISP-47 is under process. As a result of simulation, SACAP predicts well the thermal-hydraulic variables such as temperature, pressure, etc. Also, we verify that SACAP code is properly equipped to analyze the gas distribution and condensation.

  7. Verification and validation of the PLTEMP/ANL code for thermal hydraulic analysis of experimental and test reactors

    International Nuclear Information System (INIS)

    Kalimullah, M.; Olson, A.O.; Feldman, E.E.; Hanan, N.; Dionne, B.

    2012-01-01

    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.

  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. Preliminary validation of RELAP5/Mod4.0 code for LBE cooled NACIE facility

    Energy Technology Data Exchange (ETDEWEB)

    Kumari, Indu; Khanna, Ashok, E-mail: akhanna@iitk.ac.in

    2017-04-01

    Highlights: • Detail discussion of thermo physical properties of Lead Bismuth Eutectic incorporated in the code RELAP5/Mod4.0 included. • Benchmarking of LBE properties in RELAP5/Mod4.0 against literature. • NACIE facility for three different power levels (10.8, 21.7 and 32.5 kW) under natural circulation considered for benchmarking. • Preliminary validation of the LBE properties against experimental data. • NACIE facility for power level 22.5 kW considered for validation. - Abstract: The one-dimensional thermal hydraulic computer code RELAP5 was developed for thermal hydraulic study of light water reactor as well as for nuclear research reactors. The purpose of this work is to evaluate the code RELAP5/Mod4.0 for analysis of research reactors. This paper consists of three major sections. The first section presents detailed discussions on thermo-physical properties of Lead Bismuth Eutectic (LBE) incorporated in RELAP5/Mod4.0 code. In the second section, benchmarking of RELAP5/Mod4.0 has been done with the Natural Circulation Experimental (NACIE) facility in comparison with Barone’s simulations using RELAP5/Mod3.3. Three different power levels (10.8 kW, 21.7 kW and 32.5 kW) under natural circulation conditions are considered. Results obtained for LBE temperatures, temperature difference across heat section, pin surface temperatures, mass flow rates and heat transfer coefficients in heat section heat exchanger are in agreement with Barone’s simulation results within 7% of average relative error. Third section presents validation of RELAP5/Mod4.0 against the experimental data of NACIE facility performed by Tarantino et al. test number 21 at power of 22.5 kW comparing the profiles of temperatures, mass flow rate and velocity of LBE. Simulation and experimental results agree within 7% of average relative error.

  11. Aeroelastic Stability of Suspension Bridges using CFD

    DEFF Research Database (Denmark)

    Stærdahl, Jesper Winther; Sørensen, Niels; Nielsen, Søren R.K.

    2007-01-01

    using CFD models and the aeroelastic stability boundary has been successfully determined when comparing two-dimensional flow situations using wind tunnel test data and CFD methods for the flow solution and two-degrees-of-freedom structural models in translation perpendicular to the flow direction......In recent years large span suspension bridges with very thin and slender profiles have been built without proportional increasing torsional and bending stiffness. As a consequence large deformations at the mid-span can occur with risk of aeroelastic instability and structural failure. Analysis...... of aeroelastic stability also named flutter stability is mostly based on semi-empirical engineering models, where model specific parameters, the so-called flutter derivatives, need calibration from wind tunnel tests or numerical methods. Several papers have been written about calibration of flutter derivatives...

  12. Aeroelastic analysis of large horizontal wind turbine baldes?

    Institute of Scientific and Technical Information of China (English)

    Di TANG; Zhiliang LU; Tongqing GUO

    2016-01-01

    A nonlinear aeroelastic analysis method for large horizontal wind turbines is described. A vortex wake method and a nonlinear ?nite element method (FEM) are coupled in the approach. The vortex wake method is used to predict wind turbine aero-dynamic loads of a wind turbine, and a three-dimensional (3D) shell model is built for the rotor. Average aerodynamic forces along the azimuth are applied to the structural model, and the nonlinear static aeroelastic behaviors are computed. The wind rotor modes are obtained at the static aeroelastic status by linearizing the coupled equations. The static aeroelastic performance and dynamic aeroelastic responses are calculated for the NH1500 wind turbine. The results show that structural geometrical nonlinearities signi?cantly reduce displacements and vibration amplitudes of the wind turbine blades. Therefore, structural geometrical nonlinearities cannot be neglected both in the static aeroelastic analysis and dynamic aeroelastic analysis.

  13. First validation of the new continuous energy version of the MORET5 Monte Carlo code

    International Nuclear Information System (INIS)

    Miss, Joachim; Bernard, Franck; Forestier, Benoit; Haeck, Wim; Richet, Yann; Jacquet, Olivier

    2008-01-01

    The 5.A.1 version is the next release of the MORET Monte Carlo code dedicated to criticality and reactor calculations. This new version combines all the capabilities that are already available in the multigroup version with many new and enhanced features. The main capabilities of the previous version are the powerful association of a deterministic and Monte Carlo approach (like for instance APOLLO-MORET), the modular geometry, five source sampling techniques and two simulation strategies. The major advance in MORET5 is the ability to perform calculations either a multigroup or a continuous energy simulation. Thanks to these new developments, we now have better control over the whole process of criticality calculations, from reading the basic nuclear data to the Monte Carlo simulation itself. Moreover, this new capability enables us to better validate the deterministic-Monte Carlo multigroup calculations by performing continuous energy calculations with the same code, using the same geometry and tracking algorithms. The aim of this paper is to describe the main options available in this new release, and to present the first results. Comparisons of the MORET5 continuous-energy results with experimental measurements and against another continuous-energy Monte Carlo code are provided in terms of validation and time performance. Finally, an analysis of the interest of using a unified energy grid for continuous energy Monte Carlo calculations is presented. (authors)

  14. Development and validation of corium oxidation model for the VAPEX code

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, V.I.; Davydov, M.V.; Melikhov, O.I.; Borovkova, E.M.

    2011-01-01

    In light water reactor core melt accidents, the molten fuel (corium) can be brought into contact with coolant water in the course of the melt relocation in-vessel and ex-vessel as well as in an accident mitigation action of water addition. Mechanical energy release from such an interaction is of interest in evaluating the structural integrity of the reactor vessel as well as of the containment. Usually, the source for the energy release is considered to be the rapid transfer of heat from the molten fuel to the water ('vapor explosion'). When the fuel contains a chemically reactive metal component, there could be an additional source for the energy release, which is the heat release and hydrogen production due to the metal-water chemical reaction. In Electrogorsk Research and Engineering Center the computer code VAPEX (VAPor EXplosion) has been developed for analysis of the molten fuel coolant interaction. Multifield approach is used for modeling of dynamics of following phases: water, steam, melt jet, melt droplets, debris. The VAPEX code was successfully validated on FARO experimental data. Hydrogen generation was observed in FARO tests even though corium didn't contain metal component. The reason for hydrogen generation was not clear, so, simplified empirical model of hydrogen generation was implemented in the VAPEX code to take into account input of hydrogen into pressure increase. This paper describes new more detailed model of hydrogen generation due to the metal-water chemical reaction and results of its validation on ZREX experiments. (orig.)

  15. Aeroelastic instability problems for wind turbines

    DEFF Research Database (Denmark)

    Hansen, Morten Hartvig

    2007-01-01

    This paper deals with the aeroelostic instabilities that have occurred and may still occur for modem commercial wind turbines: stall-induced vibrations for stall-turbines, and classical flutter for pitch-regulated turbines. A review of previous works is combined with derivations of analytical...... stiffness and chordwise position of the center of gravity along the blades are the main parameters for flutter. These instability characteristics are exemplified by aeroelastic stability analyses of different wind turbines. The review of each aeroelastic instability ends with a list of current research...... issues that represent unsolved aeroelostic instability problems for wind turbines. Copyright (c) 2007 John Wiley & Sons, Ltd....

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

    Directory of Open Access Journals (Sweden)

    Lieberman Rebecca M

    2008-04-01

    Full Text Available 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 hypoglycemia visits. Methods This was a multicenter, retrospective cohort study using a structured medical record review at three academic emergency departments from July 1, 2005 to June 30, 2006. We prospectively derived a coding algorithm to identify hypoglycemia visits using ICD-9-CM codes (250.3, 250.8, 251.0, 251.1, 251.2, 270.3, 775.0, 775.6, and 962.3. We confirmed hypoglycemia cases by chart review identified by candidate ICD-9-CM codes during the study period. The case definition for hypoglycemia was documented blood glucose 3.9 mmol/l or emergency physician charted diagnosis of hypoglycemia. We evaluated individual components and calculated the positive predictive value. Results We reviewed 636 charts identified by the candidate ICD-9-CM codes and confirmed 436 (64% cases of hypoglycemia by chart review. Diabetes with other specified manifestations (250.8, often excluded in prior hypoglycemia analyses, identified 83% of hypoglycemia visits, and unspecified hypoglycemia (251.2 identified 13% of hypoglycemia visits. The absence of any predetermined co-diagnosis codes improved the positive predictive value of code 250.8 from 62% to 92%, while excluding only 10 (2% true hypoglycemia visits. Although prior analyses included only the first-listed ICD-9 code, more than one-quarter of identified hypoglycemia visits were outside this primary diagnosis field. Overall, the proposed algorithm had 89% positive predictive value (95% confidence interval, 86–92 for

  17. 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.)

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

    International Nuclear Information System (INIS)

    Vaidya, A.M.; Maheshwari, N.K.; Vijayan, P.K.; Saha, D.

    2010-01-01

    A 2D axi-symmetric code named NAFA (Version 1.0) is developed for studying the pipe flow under various conditions. It can handle laminar/ turbulent flows, with or without heat transfer, under sub-critical/super-critical conditions. The code solves for momentum, energy equations with standard k-ε turbulence model (with standard wall functions). It solves pipe flow subjected to 'velocity inlet', 'wall', 'axis' and 'pressure outlet' boundary conditions. It is validated for several cases by comparing its results with experimental data/analytical solutions/correlations. The code has excellent convergence characteristics as verified from fall of equation residual in each case. It has proven capability of generating mesh independent results for laminar as well as turbulent flows. The code is applied to supercritical flows. For supercritical flows, the effect of mesh size on prediction of heat transfer coefficient is studied. With grid refinement, the Y + reduces and reaches the limiting value of 11.63. Hence the accuracy is found to increase with grid refinement. NAFA is able to qualitatively predict the effect of heat flux and operating pressure on heat transfer coefficient. The heat transfer coefficient matches well with experimental values under various conditions. (author)

  19. Independent validation testing of the FLAME computer code, Version 1.0

    International Nuclear Information System (INIS)

    Martian, P.; Chung, J.N.

    1992-07-01

    Independent testing of the FLAME computer code, Version 1.0, was conducted to determine if the code is ready for use in hydrological and environmental studies at Department of Energy sites. This report describes the technical basis, approach, and results of this testing. Validation tests, (i.e., tests which compare field data to the computer generated solutions) were used to determine the operational status of the FLAME computer code and were done on a qualitative basis through graphical comparisons of the experimental and numerical data. These tests were specifically designed to check: (1) correctness of the FORTRAN coding, (2) computational accuracy, and (3) suitability to simulating actual hydrologic conditions. This testing was performed using a structured evaluation protocol which consisted of: (1) independent applications, and (2) graduated difficulty of test cases. Three tests ranging in complexity from simple one-dimensional steady-state flow field problems under near-saturated conditions to two-dimensional transient flow problems with very dry initial conditions

  20. Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests

    International Nuclear Information System (INIS)

    Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.

    1993-09-01

    PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs

  1. Experimental validation for combustion analysis of GOTHIC code in 2-dimensional combustion chamber

    International Nuclear Information System (INIS)

    Lee, J. W.; Yang, S. Y.; Park, K. C.; Jung, S. H.

    2002-01-01

    In this study, the prediction capability of GOTHIC code for hydrogen combustion phenomena was validated with the results of two-dimensional premixed hydrogen combustion experiment executed by Seoul National University. The experimental chamber has about 24 liter free volume (1x0.024x1 m 3 ) and 2-dimensional rectangular shape. The test were preformed with 10% hydrogen/air gas mixture and conducted with combination of two igniter positions (top center, top corner) and two boundary conditions (bottom full open, bottom right half open). Using the lumped parameter and mechanistic combustion model in GOTHIC code, the SNU experiments were simulated under the same conditions. The GOTHIC code prediction of the hydrogen combustion phenomena did not compare well with the experimental results. In case of lumped parameter simulation, the combustion time was predicted appropriately. But any other local information related combustion phenomena could not be obtained. In case of mechanistic combustion analysis, the physical combustion phenomena of gas mixture were not matched experimental ones. In boundary open cases, the GOTHIC predicted very long combustion time and the flame front propagation could not simulate appropriately. Though GOTHIC showed flame propagation phenomenon in adiabatic calculation, the induction time of combustion was still very long compare with experimental results. Also, it was found that the combustion model of GOTHIC code had some weak points in low concentration of hydrogen combustion simulation

  2. Validation of TEMP: A finite line heat transfer code for geologic repositories for nuclear waste

    International Nuclear Information System (INIS)

    Atterbury, W.G.; Hetteburg, J.R.; Wurm, K.J.

    1987-09-01

    TEMP is a FORTRAN computer code for calculating temperatures in a geologic repository for nuclear waste. A previous report discusses the structure, usage, verification, and benchmarking of TEMP V1.0 (Wurm et al., 1987). This report discusses modifications to the program in the development of TEMP V1.1 and documents the validation of TEMP. The development of TEMP V1.1 from TEMP V1.0 consisted of two major efforts. The first was to recode several of the subroutines to improve logic flow and to allow for geometry-independent temperature calculation routines which, in turn, allowed for the addition of the geometry-independent validation option. The validation option provides TEMP with the ability to model any geometry of temperature sources with any step-wise heat release rate. This capability allows TEMP to model the geometry and heat release characteristics of the validation problems. The validation of TEMP V1.1 consists of the comparison of TEMP to three in-ground heater tests. The three tests chosen were Avery Island, Louisiana, Site A; Avery Island, Louisiana, Site C; and Asse Mine, Federal Republic of Germany, Site 2. TEMP shows marginal comparison with the two Avery Island sites and good comparison with the Asse Mine Site. 8 refs., 25 figs., 14 tabs

  3. Aero-elastic stability of airfoil flow using 2-D CFD

    Energy Technology Data Exchange (ETDEWEB)

    Johansen, J [Risoe National Lab., Roskilde (Denmark)

    1999-03-01

    A three degrees-of-freedom structural dynamics model has been coupled to a two-dimensional incompressible CFD code. The numerical investigation considers aero-elastic stability for two different airfoils; the NACA0012 and the LM 2 18 % airfoils. Stable and unstable configurations and limit cycle oscillations are predicted in accordance with literature for the first airfoil. An attempt to predict stall induced edge-wise vibrations on a wind turbine airfoil fails using this two-dimensional approach. (au)

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

    International Nuclear Information System (INIS)

    Lee, A.G.; Wilkin, G.B.

    1996-03-01

    During the 'Workshop on R and D needs' at the 3rd Meeting of the International Group on Research Reactors (IGORR-III), the participants agreed that it would be useful to compile a survey of the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods various organizations use to verify and validate their codes and libraries. Five organizations, Atomic Energy of Canada Limited (AECL, Canada), China Institute of Atomic Energy (CIAE, People's Republic of China), Japan Atomic Energy Research Institute (JAERI, Japan), Oak Ridge National Laboratories (ORNL, USA), and Siemens (Germany) responded to the survey. The results of the survey are compiled in this report. (author) 36 refs., 3 tabs

  5. Off-take Model of the SPACE Code and Its Validation

    International Nuclear Information System (INIS)

    Oh, Myung Taek; Park, Chan Eok; Sohn, Jong Joo

    2011-01-01

    Liquid entrainment and vapor pull-through models of horizontal pipe have been implemented in the SPACE code. The model of SPACE accounts for the phase separation phenomena and computes the flux of mass and energy through an off-take attached to a horizontal pipe when stratified conditions occur in the horizontal pipe. This model is referred to as the off-take model. The importance of predicting the fluid conditions through an off-take in a small-break LOCA has been well known. In this case, the occurrence of the stratification can affect the break node void fraction and thus the break flow discharged from the primary system. In order to validate the off-take model newly developed for the SPACE code, a simulation of the HDU experiments has been performed. The main feature of the off-take model and its application results will be presented in this paper

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

    International Nuclear Information System (INIS)

    Yadav, D.H.; Datta, D.; Malhotra, P.K.; Ghadge, S.G.; Bajaj, S.S.

    2005-01-01

    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

  7. Phenomenological modeling of critical heat flux: The GRAMP code and its validation

    International Nuclear Information System (INIS)

    Ahmad, M.; Chandraker, D.K.; Hewitt, G.F.; Vijayan, P.K.; Walker, S.P.

    2013-01-01

    Highlights: ► Assessment of CHF limits is vital for LWR optimization and safety analysis. ► Phenomenological modeling is a valuable adjunct to pure empiricism. ► It is based on empirical representations of the (several, competing) phenomena. ► Phenomenological modeling codes making ‘aggregate’ predictions need careful assessment against experiments. ► The physical and mathematical basis of a phenomenological modeling code GRAMP is presented. ► The GRAMP code is assessed against measurements from BARC (India) and Harwell (UK), and the Look Up Tables. - Abstract: Reliable knowledge of the critical heat flux is vital for the design of light water reactors, for both safety and optimization. The use of wholly empirical correlations, or equivalently “Look Up Tables”, can be very effective, but is generally less so in more complex cases, and in particular cases where the heat flux is axially non-uniform. Phenomenological models are in principle more able to take into account of a wider range of conditions, with a less comprehensive coverage of experimental measurements. These models themselves are in part based upon empirical correlations, albeit of the more fundamental individual phenomena occurring, rather than the aggregate behaviour, and as such they too require experimental validation. In this paper we present the basis of a general-purpose phenomenological code, GRAMP, and then use two independent ‘direct’ sets of measurement, from BARC in India and from Harwell in the United Kingdom, and the large dataset embodied in the Look Up Tables, to perform a validation exercise on it. Very good agreement between predictions and experimental measurements is observed, adding to the confidence with which the phenomenological model can be used. Remaining important uncertainties in the phenomenological modeling of CHF, namely the importance of the initial entrained fraction on entry to annular flow, and the influence of the heat flux on entrainment rate

  8. Indications for spine surgery: validation of an administrative coding algorithm to classify degenerative diagnoses

    Science.gov (United States)

    Lurie, Jon D.; Tosteson, Anna N.A.; Deyo, Richard A.; Tosteson, Tor; Weinstein, James; Mirza, Sohail K.

    2014-01-01

    Study Design Retrospective analysis of Medicare claims linked to a multi-center clinical trial. Objective The Spine Patient Outcomes Research Trial (SPORT) provided a unique opportunity to examine the validity of a claims-based algorithm for grouping patients by surgical indication. SPORT enrolled patients for lumbar disc herniation, spinal stenosis, and degenerative spondylolisthesis. We compared the surgical indication derived from Medicare claims to that provided by SPORT surgeons, the “gold standard”. Summary of Background Data Administrative data are frequently used to report procedure rates, surgical safety outcomes, and costs in the management of spinal surgery. However, the accuracy of using diagnosis codes to classify patients by surgical indication has not been examined. Methods Medicare claims were link to beneficiaries enrolled in SPORT. The sensitivity and specificity of three claims-based approaches to group patients based on surgical indications were examined: 1) using the first listed diagnosis; 2) using all diagnoses independently; and 3) using a diagnosis hierarchy based on the support for fusion surgery. Results Medicare claims were obtained from 376 SPORT participants, including 21 with disc herniation, 183 with spinal stenosis, and 172 with degenerative spondylolisthesis. The hierarchical coding algorithm was the most accurate approach for classifying patients by surgical indication, with sensitivities of 76.2%, 88.1%, and 84.3% for disc herniation, spinal stenosis, and degenerative spondylolisthesis cohorts, respectively. The specificity was 98.3% for disc herniation, 83.2% for spinal stenosis, and 90.7% for degenerative spondylolisthesis. Misclassifications were primarily due to codes attributing more complex pathology to the case. Conclusion Standardized approaches for using claims data to accurately group patients by surgical indications has widespread interest. We found that a hierarchical coding approach correctly classified over 90

  9. Validity of administrative database code algorithms to identify vascular access placement, surgical revisions, and secondary patency.

    Science.gov (United States)

    Al-Jaishi, Ahmed A; Moist, Louise M; Oliver, Matthew J; Nash, Danielle M; Fleet, Jamie L; Garg, Amit X; Lok, Charmaine E

    2018-03-01

    We assessed the validity of physician billing codes and hospital admission using International Classification of Diseases 10th revision codes to identify vascular access placement, secondary patency, and surgical revisions in administrative data. We included adults (≥18 years) with a vascular access placed between 1 April 2004 and 31 March 2013 at the University Health Network, Toronto. Our reference standard was a prospective vascular access database (VASPRO) that contains information on vascular access type and dates of placement, dates for failure, and any revisions. We used VASPRO to assess the validity of different administrative coding algorithms by calculating the sensitivity, specificity, and positive predictive values of vascular access events. The sensitivity (95% confidence interval) of the best performing algorithm to identify arteriovenous access placement was 86% (83%, 89%) and specificity was 92% (89%, 93%). The corresponding numbers to identify catheter insertion were 84% (82%, 86%) and 84% (80%, 87%), respectively. The sensitivity of the best performing coding algorithm to identify arteriovenous access surgical revisions was 81% (67%, 90%) and specificity was 89% (87%, 90%). The algorithm capturing arteriovenous access placement and catheter insertion had a positive predictive value greater than 90% and arteriovenous access surgical revisions had a positive predictive value of 20%. The duration of arteriovenous access secondary patency was on average 578 (553, 603) days in VASPRO and 555 (530, 580) days in administrative databases. Administrative data algorithms have fair to good operating characteristics to identify vascular access placement and arteriovenous access secondary patency. Low positive predictive values for surgical revisions algorithm suggest that administrative data should only be used to rule out the occurrence of an event.

  10. Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

    International Nuclear Information System (INIS)

    T-M Sembiring; S-Pinem; P-H Liem

    2015-01-01

    The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

  11. Validation and optimisation of an ICD-10-coded case definition for sepsis using administrative health data

    Science.gov (United States)

    Jolley, Rachel J; Jetté, Nathalie; Sawka, Keri Jo; Diep, Lucy; Goliath, Jade; Roberts, Derek J; Yipp, Bryan G; Doig, Christopher J

    2015-01-01

    Objective Administrative health data are important for health services and outcomes research. We optimised and validated in intensive care unit (ICU) patients an International Classification of Disease (ICD)-coded case definition for sepsis, and compared this with an existing definition. We also assessed the definition's performance in non-ICU (ward) patients. Setting and participants All adults (aged ≥18 years) admitted to a multisystem ICU with general medicosurgical ICU care from one of three tertiary care centres in the Calgary region in Alberta, Canada, between 1 January 2009 and 31 December 2012 were included. Research design Patient medical records were randomly selected and linked to the discharge abstract database. In ICU patients, we validated the Canadian Institute for Health Information (CIHI) ICD-10-CA (Canadian Revision)-coded definition for sepsis and severe sepsis against a reference standard medical chart review, and optimised this algorithm through examination of other conditions apparent in sepsis. Measures Sensitivity (Sn), specificity (Sp), positive predictive value (PPV) and negative predictive value (NPV) were calculated. Results Sepsis was present in 604 of 1001 ICU patients (60.4%). The CIHI ICD-10-CA-coded definition for sepsis had Sn (46.4%), Sp (98.7%), PPV (98.2%) and NPV (54.7%); and for severe sepsis had Sn (47.2%), Sp (97.5%), PPV (95.3%) and NPV (63.2%). The optimised ICD-coded algorithm for sepsis increased Sn by 25.5% and NPV by 11.9% with slightly lowered Sp (85.4%) and PPV (88.2%). For severe sepsis both Sn (65.1%) and NPV (70.1%) increased, while Sp (88.2%) and PPV (85.6%) decreased slightly. Conclusions This study demonstrates that sepsis is highly undercoded in administrative data, thus under-ascertaining the true incidence of sepsis. The optimised ICD-coded definition has a higher validity with higher Sn and should be preferentially considered if used for surveillance purposes. PMID:26700284

  12. Optimization of wind turbine rotors - using advanced aerodynamic and aeroelastic models and numerical optimization

    Energy Technology Data Exchange (ETDEWEB)

    Doessing, M.

    2011-05-15

    During the last decades the annual energy produced by wind turbines has increased dramatically and wind turbines are now available in the 5MW range. Turbines in this range are constantly being developed and it is also being investigated whether turbines as large as 10-20MW are feasible. The design of very large machines introduces new problems in the practical design, and optimization tools are necessary. These must combine the dynamic effects of both aerodynamics and structure in an integrated optimization environment. This is referred to as aeroelastic optimization. The Risoe DTU optimization software HAWTOPT has been used in this project. The quasi-steady aerodynamic module have been improved with a corrected blade element momentum method. A structure module has also been developed which lays out the blade structural properties. This is done in a simplified way allowing fast conceptual design studies and with focus on the overall properties relevant for the aeroelastic properties. Aeroelastic simulations in the time domain were carried out using the aeroelastic code HAWC2. With these modules coupled to HAWTOPT, optimizations have been made. In parallel with the developments of the mentioned numerical modules, focus has been on analysis and a fundamental understanding of the key parameters in wind turbine design. This has resulted in insight and an effective design methodology is presented. Using the optimization environment a 5MW wind turbine rotor has been optimized for reduced fatigue loads due to apwise bending moments. Among other things this has indicated that airfoils for wind turbine blades should have a high lift coefficient. The design methodology proved to be stable and a help in the otherwise challenging task of numerical aeroelastic optimization. (Author)

  13. Real-time validation of receiver state information in optical space-time block code systems.

    Science.gov (United States)

    Alamia, John; Kurzweg, Timothy

    2014-06-15

    Free space optical interconnect (FSOI) systems are a promising solution to interconnect bottlenecks in high-speed systems. To overcome some sources of diminished FSOI performance caused by close proximity of multiple optical channels, multiple-input multiple-output (MIMO) systems implementing encoding schemes such as space-time block coding (STBC) have been developed. These schemes utilize information pertaining to the optical channel to reconstruct transmitted data. The STBC system is dependent on accurate channel state information (CSI) for optimal system performance. As a result of dynamic changes in optical channels, a system in operation will need to have updated CSI. Therefore, validation of the CSI during operation is a necessary tool to ensure FSOI systems operate efficiently. In this Letter, we demonstrate a method of validating CSI, in real time, through the use of moving averages of the maximum likelihood decoder data, and its capacity to predict the bit error rate (BER) of the system.

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

    International Nuclear Information System (INIS)

    Yamaguchi, Akira; 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)

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

    International Nuclear Information System (INIS)

    Palmiotti, G.; Salvatores, M.; Aliberti, G.

    2007-01-01

    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

  16. Concurrent validation of an inertial measurement system to quantify kicking biomechanics in four football codes.

    Science.gov (United States)

    Blair, Stephanie; Duthie, Grant; Robertson, Sam; Hopkins, William; Ball, Kevin

    2018-05-17

    Wearable inertial measurement systems (IMS) allow for three-dimensional analysis of human movements in a sport-specific setting. This study examined the concurrent validity of a IMS (Xsens MVN system) for measuring lower extremity and pelvis kinematics in comparison to a Vicon motion analysis system (MAS) during kicking. Thirty footballers from Australian football (n = 10), soccer (n = 10), rugby league and rugby union (n = 10) clubs completed 20 kicks across four conditions. Concurrent validity was assessed using a linear mixed-modelling approach, which allowed the partition of between and within-subject variance from the device measurement error. Results were expressed in raw and standardised units for assessments of differences in means and measurement error, and interpreted via non-clinical magnitude-based inferences. Trivial to small differences were found in linear velocities (foot and pelvis), angular velocities (knee, shank and thigh), sagittal joint (knee and hip) and segment angle (shank and pelvis) means (mean difference: 0.2-5.8%) between the IMS and MAS in Australian football, soccer and the rugby codes. Trivial to small measurement errors (from 0.1 to 5.8%) were found between the IMS and MAS in all kinematic parameters. The IMS demonstrated acceptable levels of concurrent validity compared to a MAS when measuring kicking biomechanics across the four football codes. Wearable IMS offers various benefits over MAS, such as, out-of-laboratory testing, larger measurement range and quick data output, to help improve the ecological validity of biomechanical testing and the timing of feedback. The results advocate the use of IMS to quantify biomechanics of high-velocity movements in sport-specific settings. Copyright © 2018 Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Jeltsov, Marti, E-mail: marti@safety.sci.kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Karbojian, Aram, E-mail: karbojan@kth.se; Villanueva, Walter, E-mail: walter@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2015-08-15

    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.

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

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1990-01-01

    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)

  19. Validation of the MCNP-DSP Monte Carlo code for calculating source-driven noise parameters of subcritical systems

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.

    1995-01-01

    This paper describes calculations performed to validate the modified version of the MCNP code, the MCNP-DSP, used for: the neutron and photon spectra of the spontaneous fission of californium 252; the representation of the detection processes for scattering detectors; the timing of the detection process; and the calculation of the frequency analysis parameters for the MCNP-DSP code

  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

    Objective: To determine the relative validity of the pre-coded food diary applied in the Danish National Survey of Dietary Habits and Physical Activity. Design: A cross-over study among seventy-two adults (aged 20 to 69 years) recording diet by means of a pre-coded food diary over 4 d and a 4 d...

  1. On the way to reliable aeroelastic load simulation on VAWT's

    DEFF Research Database (Denmark)

    Larsen, Torben J.; Aagaard Madsen, Helge

    2013-01-01

    In this paper a method for an implementation of a 2D actuator cylinder flow model of an Vertical Axis Wind Turbine (VAWT) is presented. The model is implemented in a full aeroelastic code including consideration of structural dynamics, dynamic inflow, tower shadow and dynamic stall, which is needed...... for a full load analysis relating to eg. certification of a VAWT turbine. Further on, principal load cases according to the IEC61400-1 are simulated for a fictitious 5MW VAWT turbine in it’s simplest 2 bladed Darrieus configuration. The IEC61400-1 load cases, originally developed for Horizontal Axis Wind...... Turbines (HAWT’s), are discussed regarding the application to VAWT’s. Further on a small section regarding aerodynamic flow in curved motion is included....

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

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Ikeda, Yujiro; 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)

  3. 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)

  4. Experiences using IAEA Code of practice for radiation sterilization of tissue allografts: Validation and routine control

    Energy Technology Data Exchange (ETDEWEB)

    Hilmy, N. [Batan Research Tissue Bank (BRTB), Centre for Research and Development of Isotopes and Radiation Technology, P.O. Box 7002, JKSKL, Jakarta 12070 (Indonesia)], E-mail: nazly@batan.go.id; Febrida, A.; Basril, A. [Batan Research Tissue Bank (BRTB), Centre for Research and Development of Isotopes and Radiation Technology, P.O. Box 7002, JKSKL, Jakarta 12070 (Indonesia)

    2007-11-15

    Problems of tissue allografts in using International Standard (ISO) 11137 for validation of radiation sterilization dose (RSD) are limited and low numbers of uniform samples per production batch, those are products obtained from one donor. Allograft is a graft transplanted between two different individuals of the same species. The minimum number of uniform samples needed for verification dose (VD) experiment at the selected sterility assurance level (SAL) per production batch according to the IAEA Code is 20, i.e., 10 for bio-burden determination and the remaining 10 for sterilization test. Three methods of the IAEA Code have been used for validation of RSD, i.e., method A1 that is a modification of method 1 of ISO 11137:1995, method B (ISO 13409:1996), and method C (AAMI TIR 27:2001). This paper describes VD experiments using uniform products obtained from one cadaver donor, i.e., cancellous bones, demineralized bone powders and amnion grafts from one life donor. Results of the verification dose experiments show that RSD is 15.4 kGy for cancellous and demineralized bone grafts and 19.2 kGy for amnion grafts according to method A1 and 25 kGy according to methods B and C.

  5. Development and validation of gui based input file generation code for relap

    International Nuclear Information System (INIS)

    Anwar, M.M.; Khan, A.A.; Chughati, I.R.; Chaudri, K.S.; Inyat, M.H.; Hayat, T.

    2009-01-01

    Reactor Excursion and Leak Analysis Program (RELAP) is a widely acceptable computer code for thermal hydraulics modeling of Nuclear Power Plants. It calculates thermal- hydraulic transients in water-cooled nuclear reactors by solving approximations to the one-dimensional, two-phase equations of hydraulics in an arbitrarily connected system of nodes. However, the preparation of input file and subsequent analysis of results in this code is a tedious task. The development of a Graphical User Interface (GUI) for preparation of the input file for RELAP-5 is done with the validation of GUI generated Input File. The GUI is developed in Microsoft Visual Studio using Visual C Sharp (C) as programming language. The Nodalization diagram is drawn graphically and the program contains various component forms along with the starting data form, which are launched for properties assignment to generate Input File Cards serving as GUI for the user. The GUI is provided with Open / Save function to store and recall the Nodalization diagram along with Components' properties. The GUI generated Input File is validated for several case studies and individual component cards are compared with the originally required format. The generated Input File of RELAP is found consistent with the requirement of RELAP. The GUI provided a useful platform for simulating complex hydrodynamic problems efficiently with RELAP. (author)

  6. Validation of two-phase flow code THYC on VATICAN experiment

    International Nuclear Information System (INIS)

    Maurel, F.; Portesse, A.; Rimbert, P.; Thomas, B.

    1997-01-01

    As part of a comprehensive program for THYC validation (THYC is a 3-dimensional two-phase flow computer code for PWR core configuration), an experimental project > has been initiated by the Direction des Etudes et Recherches of Electricite de France. Two mock-ups tested in Refrigerant-114, VATICAN-1 (with simple space grids) and VATICAN-2 (with mixing grids) were set up to investigate void fraction distributions using a single beam gamma densitometer. First, experiments were conducted with the VATICAN-1 mock-up. A set of constitutive laws to be used in rod bundles was determined but some doubts still remain for friction losses closure laws for oblique flow over tubes. From VATICAN-2 tests, calculations were performed using the standard set of correlations. Comparison with the experimental data shows an underprediction of void fraction by THYC in disturbed regions. Analyses highlight the poor treatment of axial relative velocity in these regions. A fitting of the radial and axial relative velocity values in the disturbed region improves the prediction of void fraction by the code but without any physical explanation. More analytical experiments should be carried out to validate friction losses closure laws for oblique flows and relative velocity downstream of a mixing grid. (author)

  7. Validation of two-phase flow code THYC on VATICAN experiment

    Energy Technology Data Exchange (ETDEWEB)

    Maurel, F.; Portesse, A.; Rimbert, P.; Thomas, B. [EDF/DER, Dept. TTA, 78 - Chatou (France)

    1997-12-31

    As part of a comprehensive program for THYC validation (THYC is a 3-dimensional two-phase flow computer code for PWR core configuration), an experimental project <> has been initiated by the Direction des Etudes et Recherches of Electricite de France. Two mock-ups tested in Refrigerant-114, VATICAN-1 (with simple space grids) and VATICAN-2 (with mixing grids) were set up to investigate void fraction distributions using a single beam gamma densitometer. First, experiments were conducted with the VATICAN-1 mock-up. A set of constitutive laws to be used in rod bundles was determined but some doubts still remain for friction losses closure laws for oblique flow over tubes. From VATICAN-2 tests, calculations were performed using the standard set of correlations. Comparison with the experimental data shows an underprediction of void fraction by THYC in disturbed regions. Analyses highlight the poor treatment of axial relative velocity in these regions. A fitting of the radial and axial relative velocity values in the disturbed region improves the prediction of void fraction by the code but without any physical explanation. More analytical experiments should be carried out to validate friction losses closure laws for oblique flows and relative velocity downstream of a mixing grid. (author)

  8. The data requirements for the verification and validation of a fuel performance code - the transuranus perspective

    International Nuclear Information System (INIS)

    Schubert, A.; Di Marcello, V.; Rondinella, V.; Van De Laar, J.; Van Uffelen, P.

    2013-01-01

    In general, the verification and validation (V and V) of a fuel performance code like TRANSURANUS consists of three basic steps: a) verifying the correctness and numerical stability of the sub-models; b) comparing the sub-models with experimental data; c) comparing the results of the integral fuel performance code with experimental data Only the second and third steps of the V and V rely on experimental information. This scheme can be further detailed according to the physical origin of the data: on one hand, in-reactor ('in-pile') experimental data are generated in the course of the irradiation; on the other hand ex-reactor ('out-of-pile') experimental data are obtained for instance from various postirradiation examinations (PIE) or dedicated experiments with fresh samples. For both categories, we will first discuss the V and V of sub-models of TRANSURANUS related to separate aspects of the fuel behaviour: this includes the radial variation of the composition and fissile isotopes, the thermal properties of the fuel (e.g. thermal conductivity, melting temperature, etc.), the mechanical properties of fuel and cladding (e.g. elastic constants, creep properties), as well as the models for the fission product behaviour. Secondly, the integral code verification will be addressed as it treats various aspects of the fuel behaviour, including the geometrical changes in the fuel and the gas pressure and composition of the free volume in the rod. (authors)

  9. Validation of the metal fuel version of the SAS4A accident analysis code

    International Nuclear Information System (INIS)

    Tentner, A.M.

    1991-01-01

    This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

  10. Validation study of computer code SPHINCS for sodium fire safety evaluation of fast reactor

    International Nuclear Information System (INIS)

    Yamaguchi, Akira; Tajima, Yuji

    2003-01-01

    A computer code SPHINCS solves coupled phenomena of thermal hydraulics and sodium fire based on a multi-zone model. It deals with an arbitrary number of rooms, each of which is connected mutually by doorways and penetrations. With regard to the combustion phenomena, a flame sheet model and a liquid droplet combustion model are used for pool and spray fires, respectively, with the chemical equilibrium model based on the Gibbs free energy minimization method. The chemical reaction and mass and heat transfer are solved interactively. A specific feature of SPHINCS is detailed representation of thermalhydraulics of a sodium pool and a steel liner, which is placed on the floor to prevent sodium-concrete contact. The authors analyzed a series of pool combustion experiments, in which gas and liner temperatures are measured in detail. It has been found that good agreement is obtained and the SPHINCS code has been validated with regard to pool combustion phenomena. Further research needs are identified for pool spreading modeling considering thermal deformation of steel liner and measurement of pool fluidity property as a mixture of liquid sodium and reaction products. The SPHINCS code is to be used mainly in the safety evaluation of the consequence of a sodium fire accident in a liquid metal cooled fast reactor as well as fire safety analysis in general

  11. Development and validation of a criticality calculation scheme based on French deterministic transport codes

    International Nuclear Information System (INIS)

    Santamarina, A.

    1991-01-01

    A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  13. Validation of CATHARE 3D code against UPTF TRAM C3 transients

    International Nuclear Information System (INIS)

    Glantz, Tony; Freitas, Roberto

    2007-01-01

    Within the nuclear reactor safety analysis, one of the events that could potentially lead to a recriticality accident in case of a Small Break LOCA (SBLOCA) in a pressurized water reactor (PWR) is a boron dilution scenario followed by a coolant mixing transient. Some UPTF experiments can be interpreted as generic boron dilution experiments. In fact, the UPTF experiments were originally designed to conduct separate effects studies focused on multi-dimensional thermal hydraulic phenomena. But, in the case of experimental program TRAM, some studies are realized on the boron mixing: tests C3. Some of these tests have been used for the validation and assessment of the 3D module of CATHARE code. Results are very satisfying; CATHARE 3D code is able to reproduce correctly the main features of the UPTF TRAM C3 tests, the temperature mixing in the cold leg, the formation of a strong stratification in the upper downcomer, the perfect mixing temperature in the lower downcomer and the strong stratification in the lower plenum. These results are also compared with the CFX-5 and TRIO-U codes results on these tests. (author)

  14. Validations of BWR nuclear design code using ABWR MOX numerical benchmark problems

    International Nuclear Information System (INIS)

    Takano, Shou; Sasagawa, Masaru; Yamana, Teppei; Ikehara, Tadashi; Yanagisawa, Naoki

    2017-01-01

    BWR core design code package (the HINES assembly code and the PANACH core simulator), being used for full MOX-ABWR core design, has been benchmarked against the high-fidelity numerical solutions as references, for the purpose of validating its capability of predicting the BWR core design parameters systematically from UO 2 to 100% MOX cores. The reference solutions were created by whole core critical calculations using MCNPs with the precisely modeled ABWR cores both in hot and cold conditions at BOC and EOC of the equilibrium cycle. A Doppler-Broadening Rejection Correction (DCRB) implemented MCNP5-1.4 with ENDF/B-VII.0 was mainly used to evaluate the core design parameters, except for effective delayed neutron fraction (β eff ) and prompt neutron lifetime (l) with MCNP6.1. The discrepancies in the results between the design codes HINES-PANACH and MCNPs for the core design parameters such as the bundle powers, hot pin powers, control rod worth, boron worth, void reactivity, Doppler reactivity, β eff and l, are almost within target accuracy, leading to the conclusion that HINES-PANACH has sufficient fidelity for application to full MOX-ABWR core design. (author)

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

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Arndt, S.; Klein-Hessling, W.; Schwarz, S.; Spengler, C.; Weber, G.

    2006-01-01

    For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of a severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants. COCOSYS is composed for three main modules, which are separate executable files. The communication is realized via PVM (parallel virtual machine). The thermal hydraulic main module (THY) contains several specific models relevant for the simulation of severe accidents. Beside the usual capabilities to calculate the gas distribution and thermal behavior inside the containment, there are special models for the simulation of Hydrogen deflagration, pressure suppression systems etc. Further detailed models exist for the simulation of safety systems, like catalytic recombiners (PAR's), safety relief valves (used in WWR-440/V-230 type plants), ice condenser model, pump and spray system models for the complete simulation of cooling systems. The aerosol and fission product part (AFP) describes the aerosol behavior of nonsoluble and as well as hygroscopic aerosols, iodine chemistry and fission transport. Further the decay process of nuclides is considered using ORIGIN like routines. The corium concrete interaction (CCI) main module is based on an improved version of WECHSL extended by the ChemApp module for the

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

    Energy Technology Data Exchange (ETDEWEB)

    Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Skifton, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stoots, Carl [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kim, Eung Soo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Conder, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

  17. Water evaporation over sump surface in nuclear containment studies: CFD and LP codes validation on TOSQAN tests

    Energy Technology Data Exchange (ETDEWEB)

    Malet, J., E-mail: jeanne.malet@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Degrees du Lou, O. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France); Arts et Métiers ParisTech, DynFluid Lab. EA92, 151, boulevard de l’Hôpital, 75013 Paris (France); Gelain, T. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA BP 68, 91192 Gif-sur-Yvette (France)

    2013-10-15

    Highlights: • Simulations of evaporative TOSQAN sump tests are performed. • These tests are under air–steam gas conditions with addition of He, CO{sub 2} and SF{sub 6}. • ASTEC-CPA LP and TONUS-CFD codes with UDF for sump model are used. • Validation of sump models of both codes show good results. • The code–experiment differences are attributed to turbulent gas mixing modeling. -- Abstract: 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 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 seven tests are air–steam tests, as well as tests with other non-condensable gases (He, CO{sub 2} and SF{sub 6}) under steady and transient conditions (two depressurization tests). The results show a good agreement between codes and experiments, indicating a good behavior of the sump models in both codes. The sump model developed as User-Defined Functions (UDF) for TONUS is considered as well validated and is ‘ready-to-use’ for all CFD codes in which such UDF can be added. The remaining discrepancies between codes and experiments are caused by turbulent transport and gas mixing, especially in the presence of non-condensable gases other than air, so that code validation on this important topic for hydrogen safety analysis is still recommended.

  18. Research in aeroelasticity EFP-2007

    Energy Technology Data Exchange (ETDEWEB)

    Bak, C.

    2008-07-15

    This report contains results from the EFP2007 project 'Program for Research in Applied Aeroelasticity'. The main results from this project are: 1) The rotor aerodynamics were computed using different types of models with focus on the flow around the tip. The results showed similar trend for all models. 2) Comparison of 3D CFD computations with and without inflow shear showed that the integrated rotor thrust and power were largely identical in the two situations. 3) The influence of tower shadow with and without inflow shear showed significant differences compared to BEMcomputations, which gives cause for further investigation. 4) 3D CFD computations showed that the flow in the region of the nacelle anemometer measured the flow angle in the wake with errors up to as much as 7 deg. relative to the freestream flow angle. 5) As long as the flow over a blade remains attached there is little difference between 2-D and 3-D flow. However, at separation an increased lift is observed close to the rotational axis. 6) A correlation based transition model has been implemented in the incompressible EllipSys2D/3D Navier-Stokes solver. Computations on airfoils and rotors showed good agreement and distinct improvement in the drag predictions compared to using fully turbulent computations. 7) Comparing the method of Dynamic Wake Meandering (DWM) and IEC, the IECmodel seems conservative regarding fatigue and extreme loads for the yaw, driving torque and flapwise bending, whereas the loads on tower and blade torsion are non-conservative. 8) An experimental method for measuring transition point and energy spectra in airfoil boundary layers using microphones has been developed. 9) A robust and automatic method for detecting transition based on microphone measurement on airfoil surfaces has been developed. 10) Transition points and the corresponding instabilities have clearly been observed in airfoil boundary layers. 11) Predictions of the transition points on airfoils using

  19. Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Mochizuki, H.; Okuno, H.; Miyoshi, Y.

    2004-01-01

    This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models. (authors)

  20. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  1. Validation of the CATHARE2 code against experimental data from Brayton-cycle plants

    International Nuclear Information System (INIS)

    Bentivoglio, Fabrice; Tauveron, Nicolas; Geffraye, Genevieve; Gentner, Herve

    2008-01-01

    In recent years the Commissariat a l'Energie Atomique (CEA) has commissioned a wide range of feasibility studies of future-advanced nuclear reactors, in particular gas-cooled reactors (GCR). The thermohydraulic behaviour of these systems is a key issue for, among other things, the design of the core, the assessment of thermal stresses, and the design of decay heat removal systems. These studies therefore require efficient and reliable simulation tools capable of modelling the whole reactor, including the core, the core vessel, piping, heat exchangers and turbo-machinery. CATHARE2 is a thermal-hydraulic 1D reference safety code developed and extensively validated for the French pressurized water reactors. It has been recently adapted to deal also with gas-cooled reactor applications. In order to validate CATHARE2 for these new applications, CEA has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE2 is being validated against existing experimental data. And in particular from the German power plants Oberhausen I and II. These facilities have both been operated by the German utility Energie Versorgung Oberhausen (E.V.O.) and their power conversion systems resemble to the high-temperature reactor concepts: Oberhausen I is a 13.75-MWe Brayton-cycle air turbine plant, and Oberhausen II is a 50-MWe Brayton-cycle helium turbine plant. The paper presents these two plants, the adopted CATHARE2 modelling and a comparison between experimental data and code results for both steady state and transient cases

  2. Validity of International Classification of Diseases (ICD) coding for dengue infections in hospital discharge records in Malaysia.

    Science.gov (United States)

    Woon, Yuan-Liang; Lee, Keng-Yee; Mohd Anuar, Siti Fatimah Zahra; Goh, Pik-Pin; Lim, Teck-Onn

    2018-04-20

    Hospitalization due to dengue illness is an important measure of dengue morbidity. However, limited studies are based on administrative database because the validity of the diagnosis codes is unknown. We validated the International Classification of Diseases, 10th revision (ICD) diagnosis coding for dengue infections in the Malaysian Ministry of Health's (MOH) hospital discharge database. This validation study involves retrospective review of available hospital discharge records and hand-search medical records for years 2010 and 2013. We randomly selected 3219 hospital discharge records coded with dengue and non-dengue infections as their discharge diagnoses from the national hospital discharge database. We then randomly sampled 216 and 144 records for patients with and without codes for dengue respectively, in keeping with their relative frequency in the MOH database, for chart review. The ICD codes for dengue were validated against lab-based diagnostic standard (NS1 or IgM). The ICD-10-CM codes for dengue had a sensitivity of 94%, modest specificity of 83%, positive predictive value of 87% and negative predictive value 92%. These results were stable between 2010 and 2013. However, its specificity decreased substantially when patients manifested with bleeding or low platelet count. The diagnostic performance of the ICD codes for dengue in the MOH's hospital discharge database is adequate for use in health services research on dengue.

  3. Computer code ENDSAM for random sampling and validation of the resonance parameters covariance matrices of some major nuclear data libraries

    International Nuclear Information System (INIS)

    Plevnik, Lucijan; Žerovnik, Gašper

    2016-01-01

    Highlights: • Methods for random sampling of correlated parameters. • Link to open-source code for sampling of resonance parameters in ENDF-6 format. • Validation of the code on realistic and artificial data. • Validation of covariances in three major contemporary nuclear data libraries. - Abstract: Methods for random sampling of correlated parameters are presented. The methods are implemented for sampling of resonance parameters in ENDF-6 format and a link to the open-source code ENDSAM is given. The code has been validated on realistic data. Additionally, consistency of covariances of resonance parameters of three major contemporary nuclear data libraries (JEFF-3.2, ENDF/B-VII.1 and JENDL-4.0u2) has been checked.

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

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.; Panayotov, D.; Ilieva, B.

    2001-08-01

    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.) [de

  5. Modification and validation of the natural heat convection and subcooled void formation models in the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Alhabit, F.; Ghazi, N.

    2008-01-01

    Two new modifications have been included in the current PARET code that is widely applied in the dynamic and safety analysis of research reactors. A new model was implemented for the simulation of void formation in the subcooled boiling regime, the other modification dealt with the implementation of a new approach to improve the prediction of heat transfer coefficient under natural circulation condition. The modified code was successfully validated using adequate single effect tests covering the physical phenomena of interest for both natural circulation and subcooled void formation at low pressure and low heat flux. The validation results indicate significant improvement of the code compared to the default version. Additionally, to simplify the code application an interactive user interface was developed enabling pre and post-processing of the code predictions. (author)

  6. Lessons learned in the verification, validation and application of a coupled heat and fluid flow code

    International Nuclear Information System (INIS)

    Tsang, C.F.

    1986-01-01

    A summary is given of the authors recent studies in the verification, validation and application of a coupled heat and fluid flow code. Verification has been done against eight analytic and semi-analytic solutions. These solutions include those involving thermal buoyancy flow and fracture flow. Comprehensive field validation studies over a period of four years are discussed. The studies are divided into three stages: (1) history matching, (2) double-blind prediction and confirmation, (3) design optimization. At each stage, parameter sensitivity studies are performed. To study the applications of mathematical models, a problem proposed by the International Energy Agency (IEA) is solved using this verified and validated numerical model as well as two simpler models. One of the simpler models is a semi-analytic method assuming the uncoupling of the heat and fluid flow processes. The other is a graphical method based on a large number of approximations. Variations are added to the basic IEA problem to point out the limits of ranges of applications of each model. A number of lessons are learned from the above investigations. These are listed and discussed

  7. An overview of selected NASP aeroelastic studies at the NASA Langley Research Center

    Science.gov (United States)

    Spain, Charles V.; Soistmann, David L.; Parker, Ellen C.; Gibbons, Michael D.; Gilbert, Michael G.

    1990-01-01

    Following an initial discussion of the NASP flight environment, the results of recent aeroelastic testing of NASP-type highly swept delta-wing models in Langley's Transonic Dynamics Tunnel (TDT) are summarized. Subsonic and transonic flutter characteristics of a variety of these models are described, and several analytical codes used to predict flutter of these models are evaluated. These codes generally provide good, but conservative predictions of subsonic and transonic flutter. Also, test results are presented on a nonlinear transonic phenomena known as aileron buzz which occurred in the wind tunnel on highly swept delta wings with full-span ailerons. An analytical procedure which assesses the effects of hypersonic heating on aeroelastic instabilities (aerothermoelasticity) is also described. This procedure accurately predicted flutter of a heated aluminum wing on which experimental data exists. Results are presented on the application of this method to calculate the flutter characteristics of a fine-element model of a generic NASP configuration. Finally, it is demonstrated analytically that active controls can be employed to improve the aeroelastic stability and ride quality of a generic NASP vehicle flying at hypersonic speeds.

  8. A Rapid Aeroelasticity Optimization Method Based on the Stiffness characteristics

    OpenAIRE

    Yuan, Zhe; Huo, Shihui; Ren, Jianting

    2018-01-01

    A rapid aeroelasticity optimization method based on the stiffness characteristics was proposed in the present study. Large time expense in static aeroelasticity analysis based on traditional time domain aeroelasticity method is solved. Elastic axis location and torsional stiffness are discussed firstly. Both torsional stiffness and the distance between stiffness center and aerodynamic center have a direct impact on divergent velocity. The divergent velocity can be adjusted by changing the cor...

  9. Aeroelastic modelling without the need for excessive computing power

    Energy Technology Data Exchange (ETDEWEB)

    Infield, D. [Loughborough Univ., Centre for Renewable Energy Systems Technology, Dept. of Electronic and Electrical Engineering, Loughborough (United Kingdom)

    1996-09-01

    The aeroelastic model presented here was developed specifically to represent a wind turbine manufactured by Northern Power Systems which features a passive pitch control mechanism. It was considered that this particular turbine, which also has low solidity flexible blades, and is free yawing, would provide a stringent test of modelling approaches. It was believed that blade element aerodynamic modelling would not be adequate to properly describe the combination of yawed flow, dynamic inflow and unsteady aerodynamics; consequently a wake modelling approach was adopted. In order to keep computation time limited, a highly simplified, semi-free wake approach (developed in previous work) was used. a similarly simple structural model was adopted with up to only six degrees of freedom in total. In order to take account of blade (flapwise) flexibility a simple finite element sub-model is used. Good quality data from the turbine has recently been collected and it is hoped to undertake model validation in the near future. (au)

  10. Combined Structural Optimization and Aeroelastic Analysis of a Vertical Axis Wind Turbine

    DEFF Research Database (Denmark)

    Roscher, Björn; Ferreira, Carlos Simao; Bernhammer, Lars O.

    2015-01-01

    Floating offshore wind energy poses challenges on the turbine design. A possible solution is vertical axis wind turbines, which are possibly easier to scale-up and require less components (lower maintenance) and a smaller floating structure than horizontal axis wind turbines. This paper presents...... a structural optimization and aeroelastic analysis of an optimized Troposkein vertical axis wind turbine to minimize the relation between the rotor mass and the swept area. The aeroelastic behavior of the different designs has been analyzed using a modified version of the HAWC2 code with the Actuator Cylinder...... model to compute the aerodynamics of the vertical axis wind turbine. The combined shape and topology optimization of a vertical axis wind turbine show a minimum mass to area ratio of 1.82 kg/m2 for blades with varying blade sections from a NACA 0040 at the attachment points to a NACA 0015...

  11. A combined aeroelastic-aeroacoustic model for wind turbine noise: Verification and analysis of field measurements

    DEFF Research Database (Denmark)

    Bertagnolio, Franck; Aagaard Madsen, Helge; Fischer, Andreas

    2017-01-01

    In this paper, semi-empirical engineering models for the three main wind turbine aerodynamic noise sources, namely, turbulent inflow, trailing edge and stall noise, are introduced. They are implemented into the in-house aeroelastic code HAWC2 commonly used for wind turbine load calculations...... and design. The results of the combined aeroelastic and aeroacoustic model are compared with field noise measurements of a 500kW wind turbine. Model and experimental data are in fairly good agreement in terms of noise levels and directivity. The combined model allows separating the various noise sources...... and highlights a number of mechanisms that are difficult to differentiate when only the overall noise from a wind turbine is measured....

  12. Computational Models for Nonlinear Aeroelastic Systems, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Clear Science Corp. and Duke University propose to develop and demonstrate new and efficient computational methods of modeling nonlinear aeroelastic systems. The...

  13. Variable Fidelity Aeroelastic Toolkit - Structural Model, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed innovation is a methodology to incorporate variable fidelity structural models into steady and unsteady aeroelastic and aeroservoelastic analyses in...

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

    International Nuclear Information System (INIS)

    Yum, Soo Been; Lee, Eo Hwak; Lee, Dong Won; Park, Goon Cherl

    2013-01-01

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

  15. Introduction of the ASP3D Computer Program for Unsteady Aerodynamic and Aeroelastic Analyses

    Science.gov (United States)

    Batina, John T.

    2005-01-01

    A new computer program has been developed called ASP3D (Advanced Small Perturbation 3D), which solves the small perturbation potential flow equation in an advanced form including mass-consistent surface and trailing wake boundary conditions, and entropy, vorticity, and viscous effects. The purpose of the program is for unsteady aerodynamic and aeroelastic analyses, especially in the nonlinear transonic flight regime. The program exploits the simplicity of stationary Cartesian meshes with the movement or deformation of the configuration under consideration incorporated into the solution algorithm through a planar surface boundary condition. The new ASP3D code is the result of a decade of developmental work on improvements to the small perturbation formulation, performed while the author was employed as a Senior Research Scientist in the Configuration Aerodynamics Branch at the NASA Langley Research Center. The ASP3D code is a significant improvement to the state-of-the-art for transonic aeroelastic analyses over the CAP-TSD code (Computational Aeroelasticity Program Transonic Small Disturbance), which was developed principally by the author in the mid-1980s. The author is in a unique position as the developer of both computer programs to compare, contrast, and ultimately make conclusions regarding the underlying formulations and utility of each code. The paper describes the salient features of the ASP3D code including the rationale for improvements in comparison with CAP-TSD. Numerous results are presented to demonstrate the ASP3D capability. The general conclusion is that the new ASP3D capability is superior to the older CAP-TSD code because of the myriad improvements developed and incorporated.

  16. Contribution to the validation of the Apollo code library for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; Van der Gucht, C.; Vanuxeem, J.

    1988-03-01

    The neutron nuclear data which are needed by reactor physicists to perform core calculation are brought together in the evaluated files. The files are processed to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data which are sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. We show in this paper, how the use of these integral experiments and the application of the tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the Apollo code library. For this purpose 60 buckling measurements (34 for uranium fuel multiplying media and 26 for plutonium fuel multiplying media) and 42 spent fuel analysis were used. Small modifications of the initial data are proposed. The final values are compared which recent recommended values of microscopic data and the agreement is good [fr

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

    International Nuclear Information System (INIS)

    Cantrel, L.; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-01-01

    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

  18. Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro

    2001-01-01

    Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within ±10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the 92 Mo(n, 2n) 91g Mo reaction in FENDL, and lack of activation cross section data, e.g., the 138 Ba(n, 2n) 137m Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

  19. Experimental validation of decay heat calculation codes and associated nuclear data libraries for fusion energy

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-01-01

    Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within {+-}10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the {sup 92}Mo(n, 2n){sup 91g}Mo reaction in FENDL, and lack of activation cross section data, e.g., the {sup 138}Ba(n, 2n){sup 137m}Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)

  20. Validation of Code ASTEC with LIVE-L1 Experimental Results

    International Nuclear Information System (INIS)

    Bachrata, Andrea

    2008-01-01

    The severe accidents with core melting are considered at the design stage of project at Generation 3+ of Nuclear Power Plants (NPP). Moreover, there is an effort to apply the severe accident management to the operated NPP. The one of main goals of severe accidents mitigation is corium localization and stabilization. The two strategies that fulfil this requirement are: the in-vessel retention (e.g. AP-600, AP- 1000) and the ex-vessel retention (e.g. EPR). To study the scenario of in-vessel retention, a large experimental program and the integrated codes have been developed. The LIVE-L1 experimental facility studied the formation of melt pools and the melt accumulation in the lower head using different cooling conditions. Nowadays, a new European computer code ASTEC is being developed jointly in France and Germany. One of the important steps in ASTEC development in the area of in-vessel retention of corium is its validation with LIVE-L1 experimental results. Details of the experiment are reported. Results of the ASTEC (module DIVA) application to the analysis of the test are presented. (author)

  1. Validation of a Computational Fluid Dynamics (CFD) Code for Supersonic Axisymmetric Base Flow

    Science.gov (United States)

    Tucker, P. Kevin

    1993-01-01

    The ability to accurately and efficiently calculate the flow structure in the base region of bodies of revolution in supersonic flight is a significant step in CFD code validation for applications ranging from base heating for rockets to drag for protectives. The FDNS code is used to compute such a flow and the results are compared to benchmark quality experimental data. Flowfield calculations are presented for a cylindrical afterbody at M = 2.46 and angle of attack a = O. Grid independent solutions are compared to mean velocity profiles in the separated wake area and downstream of the reattachment point. Additionally, quantities such as turbulent kinetic energy and shear layer growth rates are compared to the data. Finally, the computed base pressures are compared to the measured values. An effort is made to elucidate the role of turbulence models in the flowfield predictions. The level of turbulent eddy viscosity, and its origin, are used to contrast the various turbulence models and compare the results to the experimental data.

  2. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-01-01

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  3. Investigations on precursor measures for aeroelastic flutter

    Science.gov (United States)

    Venkatramani, J.; Sarkar, Sunetra; Gupta, Sayan

    2018-04-01

    Wind tunnel experiments carried out on a pitch-plunge aeroelastic system in the presence of fluctuating flows reveal that flutter instability is presaged by a regime of intermittency. It is observed that as the flow speed gradually increases towards the flutter speed, there appears intermittent bursts of periodic oscillations which become more frequent as the wind speed increases and eventually the dynamics transition into fully developed limit cycle oscillations, marking the onset of flutter. The signature from these intermittent oscillations are exploited to develop measures that forewarn a transition to flutter and can serve as precursors. This study investigates a suite of measures that are obtained directly from the time history of measurements and are hence model independent. The dependence of these precursors on the size of the measured data set and the time required for their computation is investigated. These measures can be useful in structural health monitoring of aeroelastic structures.

  4. NeoCASS: An integrated tool for structural sizing, aeroelastic analysis and MDO at conceptual design level

    Science.gov (United States)

    Cavagna, Luca; Ricci, Sergio; Travaglini, Lorenzo

    2011-11-01

    This paper presents a design framework called NeoCASS (Next generation Conceptual Aero-Structural Sizing Suite), developed at the Department of Aerospace Engineering of Politecnico di Milano in the frame of SimSAC (Simulating Aircraft Stability And Control Characteristics for Use in Conceptual Design) project, funded by EU in the context of 6th Framework Program. It enables the creation of efficient low-order, medium fidelity models particularly suitable for structural sizing, aeroelastic analysis and optimization at the conceptual design level. The whole methodology is based on the integration of geometry construction, aerodynamic and structural analysis codes that combine depictive, computational, analytical, and semi-empirical methods, validated in an aircraft design environment. The work here presented aims at including the airframe and its effect from the very beginning of the conceptual design. This aspect is usually not considered in this early phase. In most cases, very simplified formulas and datasheets are adopted, which implies a low level of detail and a poor accuracy. Through NeoCASS, a preliminar distribution of stiffness and inertias can be determined, given the initial layout. The adoption of empirical formulas is reduced to the minimum in favor of simple numerical methods. This allows to consider the aeroelastic behavior and performances, as well, improving the accuracy of the design tools during the iterative steps and lowering the development costs and reducing the time to market. The result achieved is a design tool based on computational methods for the aero-structural analysis and Multi-Disciplinary Optimization (MDO) of aircraft layouts at the conceptual design stage. A complete case study regarding the TransoniCRuiser aircraft, including validation of the results obtained using industrial standard tools like MSC/NASTRAN and a CFD (Computational Fluid Dynamics) code, is reported. As it will be shown, it is possible to improve the degree of

  5. Validation matrix for the assessment of thermal-hydraulic codes for VVER LOCA and transients. A report by the OECD support group on the VVER thermal-hydraulic code validation matrix

    International Nuclear Information System (INIS)

    2001-06-01

    This report deals with an internationally agreed experimental test facility matrix for the validation of best estimate thermal-hydraulic computer codes applied for the analysis of VVER reactor primary systems in accident and transient conditions. Firstly, the main physical phenomena that occur during the considered accidents are identified, test types are specified, and test facilities that supplement the CSNI CCVMs and are suitable for reproducing these aspects are selected. Secondly, a list of selected experiments carried out in these facilities has been set down. The criteria to achieve the objectives are outlined. The construction of VVER Thermal-Hydraulic Code Validation Matrix follows the logic of the CSNI Code Validation Matrices (CCVM). Similar to the CCVM it is an attempt to collect together in a systematic way the best sets of available test data for VVER specific 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 in countries operating VVER reactors 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. (authors)

  6. ASTEC code development, validation and applications for severe accident management within the CESAM European project - 15392

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Chatelard, P.; Chevalier-Jabet, K.; Nowack, H.; Herranz, L.E.; Pascal, G.; Sanchez-Espinoza, V.H.

    2015-01-01

    ASTEC, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European research on the domain. The CESAM project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. Within the CESAM project 3 main types of research activities are performed: -) further validation of ASTEC models important for SAM, in particular for the phenomena being of importance in the Fukushima-Daichi accidents, such as reflooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pools behaviour; -) modelling improvements, especially for BWR or based on the feedback of validation tasks; and -) ASTEC applications to severe accident scenarios in European NPPs in order to assess prevention and mitigation measures. An important step will be reached with the next major ASTEC V2.1 version planned to be delivered in the first part of 2015. Its main improvements will concern the possibility to simulate in details the core degradation of BWR and PHWR and a model of reflooding of severely degraded cores. A new user-friendly Graphic User Interface will be available for plant analyses

  7. Computer code validation study of PWR core design system, CASMO-3/MASTER-α

    International Nuclear Information System (INIS)

    Lee, K. H.; Kim, M. H.; Woo, S. W.

    1999-01-01

    In this paper, the feasibility of CASMO-3/MASTER-α 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-α 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-α 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-α was validated for commercial PWR core

  8. International integral experiments databases in support of nuclear data and code validation

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Gado, Janos; Hunter, Hamilton; Kodeli, Ivan; Salvatores, Massimo; Sartori, Enrico

    2002-01-01

    The OECD/NEA Nuclear Science Committee (NSC) has identified the need to establish international databases containing all the important experiments that are available for sharing among the specialists. The NSC has set up or sponsored specific activities to achieve this. The aim is to preserve them in an agreed standard format in computer accessible form, to use them for international activities involving validation of current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques. The databases so far established or in preparation related to nuclear data validation cover the following areas: SINBAD - A Radiation Shielding Experiments database encompassing reactor shielding, fusion blanket neutronics, and accelerator shielding. ICSBEP - International Criticality Safety Benchmark Experiments Project Handbook, with more than 2500 critical configurations with different combination of materials and spectral indices. IRPhEP - International Reactor Physics Experimental Benchmarks Evaluation Project. The different projects are described in the following including results achieved, work in progress and planned. (author)

  9. Validation of the DRAGON/DONJON code package for MNR using the IAEA 10 MW benchmark problem

    International Nuclear Information System (INIS)

    Day, S.E.; Garland, W.J.

    2000-01-01

    The first step in developing a framework for reactor physics analysis is to establish the appropriate and proven reactor physics codes. The chosen code package is tested, by executing a benchmark problem and comparing the results to the accepted standards. The IAEA 10 MW Benchmark problem is suitable for static reactor physics calculations on plate-fueled research reactor systems and has been used previously to validate codes for the McMaster Nuclear (MNR). The flexible and advanced geometry capabilities of the DRAGON transport theory code make it a desirable tool, and the accompanying DONJON diffusion theory code also has useful features applicable to safety analysis work at MNR. This paper describes the methodology used to benchmark the DRAGON/DONJON code package against this problem and the results herein extend the domain of validation of this code package. The results are directly applicable to MNR and are relevant to a reduced-enrichment fuel program. The DRAGON transport code models, used in this study, are based on the 1-D infinite slab approximation whereas the DONJON diffusion code models are defined in 3-D Cartesian geometry. The cores under consideration are composed of HEU (93% enrichment), MEU (45% enrichment) and LEU (20% enrichment) fuel and are examined in a fresh state, as well as at beginning-of-life (BOL) and end-of-life (EOL) exposures. The required flux plots and flux-ratio plots are included, as are transport theory code k∞and diffusion theory code k eff results. In addition to this, selected isotope atom densities are charted as a function of fuel burnup. Results from this analysis are compared to and are in good agreement with previously published results. (author)

  10. TASS code topical report. V.2 TASS code validation report for the non-LOCA transient analysis of the CE and Westinghouse type plants

    International Nuclear Information System (INIS)

    Sim, Suk K.; Chang, W. P.; Kim, K. D.; Lee, S. J.; Kim, H. C.; Yoon, H. Y.

    1997-02-01

    The development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the CE and Westinghouse type operating reactors as well as the PWR plants under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the FSAR transients, loss of AC power transient plant data, load rejection and startup test data for the reference plants as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best FSAR transient and shows its capability in simulating plant transient analyses. (author). 12 refs., 32 tabs., 132 figs

  11. 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.

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

    International Nuclear Information System (INIS)

    Hussein, M.S; Lewis, B.J.; Bonin, H.W.

    2013-01-01

    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 k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    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 k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} 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)

  14. Aeroelastic Research Programme EFP-2000; Forskning i Aeroelasticitet - EFP-2000

    Energy Technology Data Exchange (ETDEWEB)

    Aagaard Madsen, H. (ed.)

    2001-07-01

    The report presents the main results achieved within (Program for forskning i aeroelasticitet EFP-2000), which is a project carried out in collaboration between Risoe, DTU and the wind turbine industry. The project period has been 2000-2001 and it is the fourth period of a five years research program on aeroelasticity initiated in 1997. Within the present period the project has comprised the following six milestones: a.) status on 2D and 3D CFD computations b.) implementation of improved aerodynamic and structural sub models in the aeroelastic codes FLEX4 and HAWC c.) design of an airfoil family with high maximum lift d.) determination of the potential in prediction of dynamic stability e.) analysis of the uncertainty in computation of design loads f.) guidelines for optimised blade dynamics Within the project important results have been obtained and in particular within the following three main ares: 1) verification, development and application of 2D and 3D CFD computation on airfoils an rotors; 2) dynamic stability of a complete wind turbine structure; 3) importance of non-linearity's related to big blade deflections. The development of rotor computations with the 3D CFD code EllipSys3D has been an important research area since the start of the aeroelastic research programme in 1997, where initial results of 3D computations on a rotor were presented. However, first within the present project a verification of these 3D rotor computations has been possible. A blind test of rotor codes was carried out by NREL in USA using experimental data from a comprehensive wind tunnel experiment on a 10 m rotor. Out of about 20 different codes EllipSys3D gave results with the best correlation with the experimental data and in particular the 3D effect on the airfoil characteristics was well predicted. Within the research area on dynamic stability a simple, linear structural model has been developed enabling the computation of a Cambell diagram within a few seconds. Such a

  15. Code it rite the first time : automated invoice processing solution designed to ensure validity to field ticket coding

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, G.

    2010-03-15

    An entrepreneur who ran 55 rigs for a major oilfield operator in Calgary has developed a solution for the oil industry that reduces field ticketing errors from 40 per cent to almost none. The Code-Rite not only simplifies field ticketing but can eliminate weeks of trying to balance authorization for expenditure (AFE) numbers. A service provider who wants a field ticket signed for billing purposes following a service call to a well site receives all pertinent information on a barcode that includes AFE number, location, routing, approval authority and mailing address. Attaching the label to the field ticket provides all the invoicing information needed. This article described the job profile, education and life experiences and opportunities that led the innovator to develop this technology that solves an industry-wide problem. Code-Rite is currently being used by 3 large upstream oil and gas operators and plans are underway to automate the entire invoice processing system. 1 fig.

  16. Real-time simulation of aeroelastic rotor loads for horizontal axis wind turbines

    International Nuclear Information System (INIS)

    Marnett, M; Wellenberg, S; Schröder, W

    2014-01-01

    Wind turbine drivetrain research and test facilities with hardware-in-the-loop capabilities require a robust and accurate aeroelastic real-time rotor simulation environment. Recent simulation environments do not guarantee a computational response at real-time. Which is why a novel simulation tool has been developed. It resolves the physical time domain of the turbulent wind spectra and the operational response of the turbine at real-time conditions. Therefore, there is a trade-off between accuracy of the physical models and the computational costs. However, the study shows the possibility to preserve the necessary computational accuracy while simultaneously granting dynamic interaction with the aeroelastic rotor simulation environment. The achieved computational costs allow a complete aeroelastic rotor simulation at a resolution frequency of 100 Hz on standard computer platforms. Results obtained for the 5-MW reference wind turbine by the National Renewable Energy Laboratory (NREL) are discussed and compared to NREL's fatigue, aerodynamics, structures, and turbulence (FAST)- Code. The rotor loads show a convincing match. The novel simulation tool is applied to the wind turbine drivetrain test facility at the Center for Wind Power Drives (CWD), RWTH Aachen University to show the real-time hardware-in-the-loop capabilities

  17. Preliminary design of a small air loop for system analysis and validation of Cathare code

    International Nuclear Information System (INIS)

    Marchand, M.; Saez, M.; Tauveron, N.; Tenchine, D.; Germain, T.; Geffraye, G.; Ruby, G.P.

    2007-01-01

    The French Atomic Energy Commission (Cea) is carrying on the design of a Small Air Loop for System Analysis (SALSA), devoted to the study of gas cooled nuclear reactors behaviour in normal and incidental/accidental operating conditions. The reduced size of the SALSA components compared to a full-scale reactor and air as gaseous coolant instead of Helium will allow an easy management of the loop. The main purpose of SALSA will be the validation of the associated thermal hydraulic safety simulation codes, like CATHARE. The main goal of this paper is to present the methodology used to define the characteristics of the loop. In a first step, the study has been focused on a direct-cycle system for the SALSA loop with few global constraints using a similarity analysis to support the definition and design of the loop. Similarity requirements have been evaluated to determine the scale factors which have to be applied to the SALSA loop components. The preliminary conceptual design of the SALSA plant with a definition of each component has then be carried out. The whole plant has been modelled using the CATHARE code. Calculations of the SALSA steady-state in nominal conditions and of different plant transients in direct-cycle have been made. The first system results obtained on the global behaviour of the loop confirm that SALSA can be representative of a Gas-Cooled nuclear reactor with some minor design modifications. In a second step, the current prospects focus on the SALSA loop capability to reproduce correctly the heat transfer occurring in specific incidental situations. Heat decay removal by natural convection is a crucial point of interest. The first results show that the behaviour and the efficiency of the loop are strongly influenced by the definition of the main parameters for each component. A complete definition of SALSA is under progress. (authors)

  18. Validation of VHTRC calculation benchmark of critical experiment using the MCB code

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2016-01-01

    Full Text Available The calculation benchmark problem Very High Temperature Reactor Critical (VHTR a pin-in-block type core critical assembly has been investigated with the Monte Carlo Burnup (MCB code in order to validate the latest version of Nuclear Data Library based on ENDF format. Executed benchmark has been made on the basis of VHTR benchmark available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments. This benchmark is useful for verifying the discrepancies in keff values between various libraries and experimental values. This allows to improve accuracy of the neutron transport calculations that may help in designing the high performance commercial VHTRs. Almost all safety parameters depend on the accuracy of neutron transport calculation results that, in turn depend on the accuracy of nuclear data libraries. Thus, evaluation of the libraries applicability to VHTR modelling is one of the important subjects. We compared the numerical experiment results with experimental measurements using two versions of available nuclear data (ENDF-B-VII.1 and JEFF-3.2 prepared for required temperatures. Calculations have been performed with the MCB code which allows to obtain very precise representation of complex VHTR geometry, including the double heterogeneity of a fuel element. In this paper, together with impact of nuclear data, we discuss also the impact of different lattice modelling inside the fuel pins. The discrepancies of keff have been successfully observed and show good agreement with each other and with the experimental data within the 1 σ range of the experimental uncertainty. Because some propagated discrepancies observed, we proposed appropriate corrections in experimental constants which can improve the reactivity coefficient dependency. Obtained results confirm the accuracy of the new Nuclear Data Libraries.

  19. Apar-T: code, validation, and physical interpretation of particle-in-cell results

    Science.gov (United States)

    Melzani, Mickaël; Winisdoerffer, Christophe; Walder, Rolf; Folini, Doris; Favre, Jean M.; Krastanov, Stefan; Messmer, Peter

    2013-10-01

    We present the parallel particle-in-cell (PIC) code Apar-T and, more importantly, address the fundamental question of the relations between the PIC model, the Vlasov-Maxwell theory, and real plasmas. First, we present four validation tests: spectra from simulations of thermal plasmas, linear growth rates of the relativistic tearing instability and of the filamentation instability, and nonlinear filamentation merging phase. For the filamentation instability we show that the effective growth rates measured on the total energy can differ by more than 50% from the linear cold predictions and from the fastest modes of the simulation. We link these discrepancies to the superparticle number per cell and to the level of field fluctuations. Second, we detail a new method for initial loading of Maxwell-Jüttner particle distributions with relativistic bulk velocity and relativistic temperature, and explain why the traditional method with individual particle boosting fails. The formulation of the relativistic Harris equilibrium is generalized to arbitrary temperature and mass ratios. Both are required for the tearing instability setup. Third, we turn to the key point of this paper and scrutinize the question of what description of (weakly coupled) physical plasmas is obtained by PIC models. These models rely on two building blocks: coarse-graining, i.e., grouping of the order of p ~ 1010 real particles into a single computer superparticle, and field storage on a grid with its subsequent finite superparticle size. We introduce the notion of coarse-graining dependent quantities, i.e., quantities depending on p. They derive from the PIC plasma parameter ΛPIC, which we show to behave as ΛPIC ∝ 1/p. We explore two important implications. One is that PIC collision- and fluctuation-induced thermalization times are expected to scale with the number of superparticles per grid cell, and thus to be a factor p ~ 1010 smaller than in real plasmas, a fact that we confirm with

  20. 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.

  1. The aeroelasticity research project 2004[Wind turbines]; Forskning i aeroelasticitet EFP-2004

    Energy Technology Data Exchange (ETDEWEB)

    Bak, C.

    2005-05-01

    The report presents the results of the project ''Programme for Applied Aeroelasticity'', the Danish Energy Research Programme 2004. The main results are: 1) Based on an analysis of the NREL/NASA experiment with a wind turbine in a wind tunnel a new model is formulated for 3D corrections of profile data for aeroelastic codes. Use of the model on three rotors suggests that the load distribution is determined more correctly than in existing 3D models. 2) A near-wake model, originally developed for aerodynamic loads on helicopter rotors, is implemented for calculating dynamic induction on wind turbine rotors. The model has several advantages to the other normally used model BEM. 3) A detailed comparison of the aeroelastic models FLEX5 and HAWC shows that there are no model differences that can result in large differences in the calculated loads. The comparison shows that differences in the calculated loads are due to the use of the models. 4) A model for pitch-servo dynamics on a modern wind turbine is formed and implemented in HAWC2. The conclusion from analysis of the importance of the pitch-servo characteristics showed that coupling between structure/aerodynamics and pitch actuator may be of importance, especially for the loads on the actuator itself. Also large deflections are coupled to the pitch moment and thus also to torsion of the wing and wing bearing. 5) An un-linear stability analysis has been performed in which periodic loads are included and compared to a linear analysis used in HAWCStab. For a profile with near zero aerodynamic damping in one oscillation direction, the aerodynamic force in this direction depends mostly of the square on the profile's speed. The linear damping is changed only a little by the profile's forced oscillation. It is assumed that the present HAWCStab can predict the mean aeroelastic damping for turbines' oscillations in operation. (LN)

  2. Validity of Principal Diagnoses in Discharge Summaries and ICD-10 Coding Assessments Based on National Health Data of Thailand.

    Science.gov (United States)

    Sukanya, Chongthawonsatid

    2017-10-01

    This study examined the validity of the principal diagnoses on discharge summaries and coding assessments. Data were collected from the National Health Security Office (NHSO) of Thailand in 2015. In total, 118,971 medical records were audited. The sample was drawn from government hospitals and private hospitals covered by the Universal Coverage Scheme in Thailand. Hospitals and cases were selected using NHSO criteria. The validity of the principal diagnoses listed in the "Summary and Coding Assessment" forms was established by comparing data from the discharge summaries with data obtained from medical record reviews, and additionally, by comparing data from the coding assessments with data in the computerized ICD (the data base used for reimbursement-purposes). The summary assessments had low sensitivities (7.3%-37.9%), high specificities (97.2%-99.8%), low positive predictive values (9.2%-60.7%), and high negative predictive values (95.9%-99.3%). The coding assessments had low sensitivities (31.1%-69.4%), high specificities (99.0%-99.9%), moderate positive predictive values (43.8%-89.0%), and high negative predictive values (97.3%-99.5%). The discharge summaries and codings often contained mistakes, particularly the categories "Endocrine, nutritional, and metabolic diseases", "Symptoms, signs, and abnormal clinical and laboratory findings not elsewhere classified", "Factors influencing health status and contact with health services", and "Injury, poisoning, and certain other consequences of external causes". The validity of the principal diagnoses on the summary and coding assessment forms was found to be low. The training of physicians and coders must be strengthened to improve the validity of discharge summaries and codings.

  3. 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)

  4. Validity of congenital malformation diagnostic codes recorded in Québec's administrative databases.

    Science.gov (United States)

    Blais, Lucie; Bérard, Anick; Kettani, Fatima-Zohra; Forget, Amélie

    2013-08-01

    To assess the validity of the diagnostic codes of congenital malformations (CMs) recorded in two of Québec's administrative databases. A cohort of pregnancies and infants born to asthmatic and non-asthmatic women in 1990-2002 was reconstructed using Québec's administrative databases. From this cohort, we selected 269 infants with a CM and 144 without CM born to asthmatic women, together with 284 and 138 infants, respectively, born to non-asthmatic women. The diagnoses of CMs recorded in the databases were compared with the diagnoses written by the physicians in the infants' medical charts. The positive predictive values (PPV) and negative predictive values (NPV) for all, major, and several specific CMs were estimated. The PPVs for all CMs and major CMs were 82.2% (95% confidence interval (CI): 78.5%-85.9%) and 78.1% (74.1%-82.1%), respectively, in the asthmatic group and were 79.2% (75.4%-83.1%) and 69.0% (64.6%-73.4%), respectively, in the non-asthmatic group. PPVs >80% were found for several specific CMs, including cardiac, cleft, and limb CMs in both groups. The NPV for any CM was 88.2% (95% CI: 85.1%-91.3%) in the asthmatic group and 94.2% (92.2%-96.2%) in the non-asthmatic group. Québec's administrative databases are valid tools for epidemiological research of CMs. The results were similar between infants born to women with and without asthma. Copyright © 2013 John Wiley & Sons, Ltd.

  5. VULCAN: An Open-source, Validated Chemical Kinetics Python Code for Exoplanetary Atmospheres

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, Shang-Min; Grosheintz, Luc; Kitzmann, Daniel; Heng, Kevin [University of Bern, Center for Space and Habitability, Sidlerstrasse 5, CH-3012, Bern (Switzerland); Lyons, James R. [Arizona State University, School of Earth and Space Exploration, Bateman Physical Sciences, Tempe, AZ 85287-1404 (United States); Rimmer, Paul B., E-mail: shang-min.tsai@space.unibe.ch, E-mail: kevin.heng@csh.unibe.ch, E-mail: jimlyons@asu.edu [University of St. Andrews, School of Physics and Astronomy, St. Andrews, KY16 9SS (United Kingdom)

    2017-02-01

    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 and Helling. It reproduces the models of HD 189733b and HD 209458b by Moses et al., which employ a network with nearly 1600 reactions. We also use VULCAN to examine 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 approximation and find that it is accurate for methane but breaks down for acetylene, because the disequilibrium abundance of acetylene is not directly determined by transport-induced quenching, but is rather indirectly controlled by the disequilibrium abundance of methane. Therefore we suggest that the quenching approximation should be used with caution and must always be checked against a chemical kinetics calculation. A one-dimensional model atmosphere with 100 layers, computed using VULCAN, typically takes several minutes to complete. VULCAN is part of the Exoclimes Simulation Platform (ESP; exoclime.net) and publicly available at https://github.com/exoclime/VULCAN.

  6. Development of an aeroelastic methodology for surface morphing rotors

    Science.gov (United States)

    Cook, James R.

    transmission of force and deflection information to achieve an aeroelastic coupling updated at each time step. The method is validated first by comparing the integrated aerodynamic work at CFD and CSD nodes to verify work conservation across the interface. Second, the method is verified by comparing the sectional blade loads and deflections of a rotor in hover and in forward flight with experimental data. Finally, stability analyses for pitch/plunge flutter and camber flutter are performed with comprehensive CSD/low-order-aerodynamics and tightly coupled CFD/CSD simulations and compared to analytical solutions of Peters' thin airfoil theory to verify proper aeroelastic behavior. The effects of simple harmonic camber actuation are examined and compared to the response predicted by Peters' finite-state (F-S) theory. In anticipation of active rotor experiments inside enclosed facilities, computational simulations are performed to evaluate the capability of CFD for accurately simulating flow inside enclosed volumes. A computational methodology for accurately simulating a rotor inside a test chamber is developed to determine the influence of test facility components and turbulence modeling and performance predictions. A number of factors that influence the physical accuracy of the simulation, such as temporal resolution, grid resolution, and aeroelasticity are also evaluated.

  7. Validation of one-dimensional module of MARS 2.1 computer code by comparison with the RELAP5/MOD3.3 developmental assessment results

    International Nuclear Information System (INIS)

    Lee, Y. J.; Bae, S. W.; Chung, B. D.

    2003-02-01

    This report records the results of the code validation for the one-dimensional module of the MARS 2.1 thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 code development assessment problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS 2.1 code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The results suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

  8. Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir [Westinghouse Electric Sweden AB, Vaesteraas, SE-721 63 (Sweden)

    2008-07-01

    This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (author)

  9. Validation of Westinghouse integrated code POLCA-T against OECD NEACRP-L-335 rod ejection benchmark

    International Nuclear Information System (INIS)

    Panayotov, Dobromir

    2008-01-01

    This paper describes the work performed and results obtained in the validation of the POLCA-T code against NEACRP PWR Rod Ejection Transients Benchmark. Presented work is a part of the POLCA-T licensing Assessment Data Base for BWR Control Rod Drop Accident (CRDA) Application. The validation against a PWR Rod Ejection Accidents (REA) Benchmark is relevant for the validation of the code for BWR CRDA, as the analyses of both transients require identical phenomena to be modelled. All six benchmark cases have been analyzed in the presented work. Initial state steady-state calculations including boron search, control rod worth, and final state power search have been performed by POLCA7 code. Initial state boron adjustment and steady-state CR worth as well as the transient analyses were performed by POLCA-T code. Benchmark results including 3D transient power distributions are compared with reference PANTHER solutions and published results of other codes. Given the similarity of the kinetics modelling for a BWR CRDA and a PWR REA and the fact that POLCA-T accurately predicts the local transient power and thus, the resulting fuel enthalpy, it is concluded that POLCA-T is a state-of-art tool also for BWR CRDA analysis. (author)

  10. Validation of the TASS/SMR-S Code for the PRHRS Condensation Heat Transfer Model

    International Nuclear Information System (INIS)

    Jun, In Sub; Yang, Soo Hyoung; Chung, Young Jong; Lee, Won Jae

    2011-01-01

    When some accidents or events are occurred in the SMART, the secondary system is used to remove the core decay heat for the long time such as a feedwater system. But if the feedwater system can't remove the residual core heat because of its malfunction, the core decay heat is removed using the Passive Residual Heat Removal System (PRHRS). The PRHRS is passive type safety system adopted to enhance the safety of the SMART. It can fundamentally eliminate the uncertainty of operator action. TASS/SMR-S (Transient And Setpoint Simulation/ System-integrated Modular Reactor-Safety) code has various heat transfer models reflecting the design features of the SMART. One of the heat transfer models is the PRHRS condensation heat transfer model. The role of this model is to calculate the heat transfer coefficient in the heat exchanger (H/X) tube side using the relevant heat transfer correlations for all of the heat transfer modes. In this paper, the validation of the condensation heat transfer model was carried out using the POSTECH H/X heat transfer test

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

    International Nuclear Information System (INIS)

    Perkins, B.; Travis, B.; DePoorter, G.

    1985-01-01

    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

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

    International Nuclear Information System (INIS)

    1996-07-01

    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

  13. Aeroelastic stability of full-span tiltrotor aircraft model in forward flight

    Directory of Open Access Journals (Sweden)

    Zhiquan LI

    2017-12-01

    Full Text Available The existing full-span models of the tiltrotor aircraft adopted the rigid blade model without considering the coupling relationship among the elastic blade, wing and fuselage. To overcome the limitations of the existing full-span models and improve the precision of aeroelastic analysis of tiltrotor aircraft in forward flight, the aeroelastic stability analysis model of full-span tiltrotor aircraft in forward flight has been presented in this paper by considering the coupling among elastic blade, wing, fuselage and various components. The analytical model is validated by comparing with the calculation results and experimental data in the existing references. The influence of some structural parameters, such as the fuselage degrees of freedom, relative displacement between the hub center and the gravity center, and nacelle length, on the system stability is also investigated. The results show that the fuselage degrees of freedom decrease the critical stability velocity of tiltrotor aircraft, and the variation of the structural parameters has great influence on the system stability, and the instability form of system can change between the anti-symmetric and symmetric wing motions of vertical and chordwise bending. Keywords: Aeroelastic stability, Forward flight, Full-span model, Modal analysis, Tiltrotor aircraft

  14. Mapped Chebyshev Pseudo-Spectral Method for Dynamic Aero-Elastic Problem of Limit Cycle Oscillation

    Science.gov (United States)

    Im, Dong Kyun; Kim, Hyun Soon; Choi, Seongim

    2018-05-01

    A mapped Chebyshev pseudo-spectral method is developed as one of the Fourier-spectral approaches and solves nonlinear PDE systems for unsteady flows and dynamic aero-elastic problem in a given time interval, where the flows or elastic motions can be periodic, nonperiodic, or periodic with an unknown frequency. The method uses the Chebyshev polynomials of the first kind for the basis function and redistributes the standard Chebyshev-Gauss-Lobatto collocation points more evenly by a conformal mapping function for improved numerical stability. Contributions of the method are several. It can be an order of magnitude more efficient than the conventional finite difference-based, time-accurate computation, depending on the complexity of solutions and the number of collocation points. The method reformulates the dynamic aero-elastic problem in spectral form for coupled analysis of aerodynamics and structures, which can be effective for design optimization of unsteady and dynamic problems. A limit cycle oscillation (LCO) is chosen for the validation and a new method to determine the LCO frequency is introduced based on the minimization of a second derivative of the aero-elastic formulation. Two examples of the limit cycle oscillation are tested: nonlinear, one degree-of-freedom mass-spring-damper system and two degrees-of-freedom oscillating airfoil under pitch and plunge motions. Results show good agreements with those of the conventional time-accurate simulations and wind tunnel experiments.

  15. Application of advanced validation concepts to oxide fuel performance codes: LIFE-4 fast-reactor and FRAPCON thermal-reactor fuel performance codes

    Energy Technology Data Exchange (ETDEWEB)

    Unal, C., E-mail: cu@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Williams, B.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Yacout, A. [Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Higdon, D.M. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-10-15

    Highlights: ► The application of advanced validation techniques (sensitivity, calibration and prediction) to nuclear performance codes FRAPCON and LIFE-4 is the focus of the paper. ► A sensitivity ranking methodology narrows down the number of selected modeling parameters from 61 to 24 for FRAPCON and from 69 to 35 for LIFE-4. ► Fuel creep, fuel thermal conductivity, fission gas transport/release, crack/boundary, and fuel gap conductivity models of LIFE-4 are identified for improvements. ► FRAPCON sensitivity results indicated the importance of the fuel thermal conduction and the fission gas release models. -- Abstract: Evolving nuclear energy programs expect to use enhanced modeling and simulation (M and S) capabilities, using multiscale, multiphysics modeling approaches, to reduce both cost and time from the design through the licensing phases. Interest in the development of the multiscale, multiphysics approach has increased in the last decade because of the need for predictive tools for complex interacting processes as a means of eliminating the limited use of empirically based model development. Complex interacting processes cannot be predicted by analyzing each individual component in isolation. In most cases, the mathematical models of complex processes and their boundary conditions are nonlinear. As a result, the solutions of these mathematical models often require high-performance computing capabilities and resources. The use of multiscale, multiphysics (MS/MP) models in conjunction with high-performance computational software and hardware introduces challenges in validating these predictive tools—traditional methodologies will have to be modified to address these challenges. The advanced MS/MP codes for nuclear fuels and reactors are being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the US Department of Energy (DOE) – Nuclear Energy (NE). This paper does not directly address challenges in calibration/validation

  16. GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

    International Nuclear Information System (INIS)

    Ramachandran, S.; Krause, M.; Nguyen, T.

    2003-01-01

    This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

  17. Validation of CBZ code system for post-irradiation examination analysis and sensitivity analysis of (n,γ) branching ratio

    International Nuclear Information System (INIS)

    Kawamoto, Yosuke; Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

    2013-01-01

    A code system CBZ is being developed in Hokkaido University. In order to validate it, PIE data, which are nuclide composition data of a spent fuel, have been analyzed with CBZ. The validity is evaluated as ratios of the calculation values to the experimental ones (C/E ratios). Differences between experimental values and calculation ones are smaller than 20% except some nuclides. Thus this code system is validated. Additionally, we evaluate influence of change of (n,γ) branching ratio on inventories of fission products and actinides. As a result, branching ratios of Sb-121, Pm-147, and Am-241 influence inventories of several nuclides. We perform PIE analysis using different (n,γ) branching ratio data from the ORIGEN-2 library, JNDC-Ver.2, and JEFF-3.1A, and find that differences in (n,γ) branching ratios between different nuclear libraries have a non-negligible influence on inventories of several nuclides. (author)

  18. Aeroelastic Optimization of MW Wind Turbines

    DEFF Research Database (Denmark)

    Hansen, Morten Hartvig; Zahle, Frederik

    This report contains the results from the Energy Development and Demonstration Project “Aeroelastic Optimization of MW wind turbine” (AeroOpt). The project has had the following five Work Packages: 1. Geometric non-linear, anisotropic beamelement forHAWC2 2. Closed-loop eigenvalue analysis...... of controlled wind turbines 3. Resonant wave excitation of lateral tower bending modes 4. Development of next generation aerodynamic design tools 5. Advanced design and verification of airfoils The purposes of these Work Packages are briefly described in the Preface and a summary of the results are given...

  19. Studying aeroelastic oscillations with tensoresistor and Arduino

    Science.gov (United States)

    Demenkov, Maxim

    2018-05-01

    We describe a modification of the Flexy device, originally developed at the Slovak University of Technology. With our version of it, constructed at the Institute of Control Sciences, one can study aeroelastic oscillations (flutter) using cheap and freely available components. Flex sensor (tensoresistor) changes its electrical resistance proportionally to its bending. The lightweight plastic plate (attached to the resistor) plays the role of a wing in the flow generated by a small fan. Both fan and tensoresistor are connected to an Arduino microcontroller and it is possible to obtain and analyze experimental data from the device on a personal computer.

  20. Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.

    2016-09-01

    The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

  1. Validation of DRAGON code in connection with WIMS-AECL/RFSP code system based on ENDF/B-VI library and two group model

    International Nuclear Information System (INIS)

    Hong, In Seob; Suk, Ho Chun; Kim, Soon Young; Jo, Chang Keun

    2002-06-01

    The major objective of this research is to validate the incremental cross section property of DRAGON code in connection with WIMS-AECL/DRAGON/RFSP code system with ENDF/B-VI library and full 2G calculation model. The direct comparison between the incremental cross section results calculated by DRAGON with ENDF/B-VI and ENDF/B-V and MULTICELL with ENDF/B-V indicate that there are not much differences between the incremental cross sections of DRAGON with ENDF/B-V and ENDF/B-VI, but there exists large discrepancies between the results of DRAGON and those of MULTICELL. In the analysis of the difference between calculated and measured reactivity worths of various types of control devices during Phase-B Post-Simulation of Wolsong Units 2, 3 and 4, WIMS-AECL/DRAGON/RFSP analysis well agrees with those of previous WIMS-AECL /MULTICELL/RFSP analysis within very small differences. From those results, we can conclude that DRAGON code can be used as a general purpose incremental cross section generation tool for not only the natural uranium fuel but also slightly enriched fuel such as RU or SEU, to cover the shortcomings of natural uranium based MULTICELL code

  2. Aeroelastic Optimization of a 10 MW Wind Turbine Blade with Active Trailing Edge Flaps

    DEFF Research Database (Denmark)

    Barlas, Athanasios; Tibaldi, Carlo; Zahle, Frederik

    2016-01-01

    This article presents the aeroelastic optimization of a 10MW wind turbine ‘smart blade’ equipped with active trailing edge flaps. The multi-disciplinary wind turbine analysis and optimization tool HawtOpt2 is utilized, which is based on the open-source framework Open-MDAO. The tool interfaces...... to several state-of-the art simulation codes, allowing for a wide variety of problem formulations and combinations of models. A simultaneous aerodynamic and structural optimization of a 10 MW wind turbine rotor is carried out with respect to material layups and outer shape. Active trailing edge flaps...

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

    International Nuclear Information System (INIS)

    Tkatschenko, I.; Studer, E.; Paillere, H.

    2005-01-01

    During a severe accident in a Pressurized Water Reactor (PWR), the formation of a combustible gas mixture in the complex geometry of the reactor depends on the understanding of hydrogen production, the complex 3D thermal-hydraulics flow due to gas/steam injection, natural convection, heat transfer by condensation on walls and effect of mitigation devices. Numerical simulation of such flows may be performed either by Lumped Parameter (LP) or by Computational Fluid Dynamics (CFD) codes. Advantages and drawbacks of LP and CFD codes are well-known. LP codes are mainly developed for full size containment analysis but they need improvements, especially since they are not able to accurately predict the local gas mixing within the containment. CFD codes require a process of validation on well-instrumented experimental data before they can be used with a high degree of confidence. The MISTRA coupled effect test facility has been built at CEA to fulfil this validation objective: with numerous measurement points in the gaseous volume - temperature, gas concentration, velocity and turbulence - and with well controlled boundary conditions. As illustration of both experimental and simulation areas of this topic, a recent example in the use of MISTRA test data is presented for the case of the International Standard Problem ISP47. The proposed experimental work in the MISTRA facility provides essential data to fill the gaps in the modelling/validation of computational tools. (author)

  4. Helicopter rotor dynamics and aeroelasticity - Some key ideas and insights

    Science.gov (United States)

    Friedmann, Peretz P.

    1990-01-01

    Four important current topics in helicopter rotor dynamics and aeroelasticity are discussed: (1) the role of geometric nonlinearities in rotary-wing aeroelasticity; (2) structural modeling, free vibration, and aeroelastic analysis of composite rotor blades; (3) modeling of coupled rotor/fuselage areomechanical problems and their active control; and (4) use of higher-harmonic control for vibration reduction in helicopter rotors in forward flight. The discussion attempts to provide an improved fundamental understanding of the current state of the art. In this way, future research can be focused on problems which remain to be solved instead of producing marginal improvements on problems which are already understood.

  5. 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.

  6. Development and validation of GWHEAD, a three-dimensional groundwater head computer code

    International Nuclear Information System (INIS)

    Beckmeyer, R.R.; Root, R.W.; Routt, K.R.

    1980-03-01

    A computer code has been developed to solve the groundwater flow equation in three dimensions. The code has finite-difference approximations solved by the strongly implicit solution procedure. Input parameters to the code include hydraulic conductivity, specific storage, porosity, accretion (recharge), and initial hydralic head. These parameters may be input as varying spatially. The hydraulic conductivity may be input as isotropic or anisotropic. The boundaries either may permit flow across them or may be impermeable. The code has been used to model leaky confined groundwater conditions and spherical flow to a continuous point sink, both of which have exact analytical solutions. The results generated by the computer code compare well with those of the analytical solutions. The code was designed to be used to model groundwater flow beneath fuel reprocessing and waste storage areas at the Savannah River Plant

  7. Aeroelastic Modeling of a Nozzle Startup Transient

    Science.gov (United States)

    Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen

    2014-01-01

    Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a tightly coupled aeroelastic modeling algorithm by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed under the framework of modal analysis. Transient aeroelastic nozzle startup analyses at sea level were performed, and the computed transient nozzle fluid-structure interaction physics presented,

  8. 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.)

  9. 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.)

  10. MARS-KS code validation activity through the atlas domestic standard problem

    International Nuclear Information System (INIS)

    Choi, K. Y.; Kim, Y. S.; Kang, K. H.; Park, H. S.; Cho, S.

    2012-01-01

    The 2 nd Domestic Standard Problem (DSP-02) exercise using the ATLAS integral effect test data was executed to transfer the integral effect test data to domestic nuclear industries and to contribute to improving the safety analysis methodology for PWRs. A small break loss of coolant accident of a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. Ten calculation results using MARS-KS code were collected, major prediction results were described qualitatively and code prediction accuracy was assessed quantitatively using the FFTBM. In addition, special code assessment activities were carried out to find out the area where the model improvement is required in the MARS-KS code. The lessons from this DSP-02 and recommendations to code developers are described in this paper. (authors)

  11. An Aeroelastic Evaluation of the Flexible Thermal Protection System for an Inatable Aerodynamic Decelerator

    Science.gov (United States)

    Goldman, Benjamin D.

    asymmetric flutter at high dynamic pressures relative to the single shell models. Parameter studies also examine the effects of tension, shear modulus reduction, and elastic support stiffness. Limitations of a linear structural model and piston theory aerodynamics prompted a more elaborate evaluation of the flight configuration. Using nonlinear Donnell conical shell theory for the FTPS structure, the pressure buckling and aeroelastic limit cycle oscillations were studied for a single elastically-supported conical shell. While piston theory was used initially, a time-dependent correction factor was derived using transform methods and potential flow theory to calculate more accurately the low Mach number supersonic flow. Three conical shell geometries were considered: a 3-meter diameter 70° shell, a 3.7-meter 70° shell, and a 6-meter diameter 70° shell. The 6-meter configuration was loaded statically and the results were compared with an experimental load test of a 6-meter HIAD vehicle. Though agreement between theoretical and experimental strains was poor, circumferential wrinkling phenomena observed during the experiments was captured by the theory and axial deformations were qualitatively similar in shape. With piston theory aerodynamics, the nonlinear flutter dynamic pressures of the 3-meter configuration were in agreement with the values calculated using linear theory, and the limit cycle amplitudes were generally on the order of the shell thickness. Pre-buckling pressure loads and the aerodynamic pressure correction factor were studied for all geometries, and these effects resulted in significantly lower flutter boundaries compared with piston theory alone. In the final phase of this work, the existing linear and nonlinear FTPS shell models were coupled with NASA's FUN3D Reynolds Averaged Navier Stokes CFD code, allowing for the most physically realistic flight predictions. For the linear shell structural model, the elastically-supported shell natural modes were

  12. Research in aeroelasticity EFP-2002; Forskning i aeroelasticitet EFP-2002

    Energy Technology Data Exchange (ETDEWEB)

    Bak, Christian (ed.)

    2004-02-01

    This report contains results from the Energy Research Project 'Application, demonstration and further development of advanced aerodynamic and aeroelastic models' (EFP 2002), covering the time from July 1 2002 to December 31 2003. The partners in the project are Risoe National Labo-ratory (Risoe), The Technical University of Denmark (DTU), Bonus Energy A/S, LM Glasfiber A/S, NEG Micon A/S og Vestas Wind Systems A/S. In the project, Risoe and DTU have de-monstrated the application of their advanced computational methods on several different mega-Watt-size wind turbine designs. Compared to traditional methods the advanced methods have among other results shown: 1) that the aerodynamics at the blade tip for a wind turbine cannot be analysed correctly for a non-rotating blade. 2) that the drag coefficient distribution on a rotor in stand still according to Computational Fluid Dynamics should be increased from the blade root towards the blade tip. 3) that the maximum 2D lift coefficient in airfoil characteristics should be reduced at the blade tip and should be increased significantly on the inner part of the rotor. The drag coefficients should in general be increased for all sections on the blade, when the flow is separating. 4) that the choice of airfoil characteristics, aerodynamical as well as structural, are impor-tant for the loads, the noise and the design of a wind turbine. 5) that blade edgewise vibrations in stand still computed with an aeroelastic code are most critical around 40 deg. and 140 deg. angles of attack and that these vibrations depend completely on the given values of lift and drag. 6) that the energy production decreases in the case of large deflections of the blades. 7) that the blade flap eigenfrequency increases in the case of large deflections. 8) that there is an increased coupling between blade edge and blade torsional frequency in the case of large deflections. 9) that an overview of the dynamics for a wind turbine design can be

  13. Aeroelastic Simulation Tool for Inflatable Ballute Aerocapture, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — This project will develop a much-needed multidisciplinary analysis tool for predicting the impact of aeroelastic effects on the functionality of inflatable...

  14. Static aeroelastic behavior of an adaptive laminated piezoelectric composite wing

    Science.gov (United States)

    Weisshaar, T. A.; Ehlers, S. M.

    1990-01-01

    The effect of using an adaptive material to modify the static aeroelastic behavior of a uniform wing is examined. The wing structure is idealized as a laminated sandwich structure with piezoelectric layers in the upper and lower skins. A feedback system that senses the wing root loads applies a constant electric field to the piezoelectric actuator. Modification of pure torsional deformaton behavior and pure bending deformation are investigated, as is the case of an anisotropic composite swept wing. The use of piezoelectric actuators to create an adaptive structure is found to alter static aeroelastic behavior in that the proper choice of the feedback gain can increase or decrease the aeroelastic divergence speed. This concept also may be used to actively change the lift effectiveness of a wing. The ability to modify static aeroelastic behavior is limited by physical limitations of the piezoelectric material and the manner in which it is integrated into the parent structure.

  15. Computational Models for Nonlinear Aeroelastic Systems, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Clear Science Corp. and Duke University propose to develop and demonstrate a new and efficient computational method of modeling nonlinear aeroelastic systems. The...

  16. Unified Nonlinear Flight Dynamics and Aeroelastic Simulator Tool, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — ZONA Technology, Inc. (ZONA) proposes a R&D effort to develop a Unified Nonlinear Flight Dynamics and Aeroelastic Simulator (UNFDAS) Tool that will combine...

  17. Validation of the standalone implementation of the dynamic wake meandering model for power production

    DEFF Research Database (Denmark)

    Keck, Rolf-Erik Henrik Jussi

    2015-01-01

    This paper presents validation for using the standalone implementation of the dynamic wake meandering (DWM) model to conduct numerical simulations of power production of rows of wind turbines. The standalone DWM model is an alternative formulation of the conventional DWM model that does not require...... information exchange with an aeroelastic code. As a consequence, the standalone DWM model has significantly shorter computational times and lower demands on the user environment. The drawback of the standalone DWM model is that it does not have the capability to predict turbine loads. Instead, it should...

  18. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    International Nuclear Information System (INIS)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha

    2016-01-01

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  19. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  20. Chiari malformation Type I surgery in pediatric patients. Part 1: validation of an ICD-9-CM code search algorithm.

    Science.gov (United States)

    Ladner, Travis R; Greenberg, Jacob K; Guerrero, Nicole; Olsen, Margaret A; Shannon, Chevis N; Yarbrough, Chester K; Piccirillo, Jay F; Anderson, Richard C E; Feldstein, Neil A; Wellons, John C; Smyth, Matthew D; Park, Tae Sung; Limbrick, David D

    2016-05-01

    OBJECTIVE Administrative billing data may facilitate large-scale assessments of treatment outcomes for pediatric Chiari malformation Type I (CM-I). Validated International Classification of Diseases, Ninth Revision, Clinical Modification (ICD-9-CM) code algorithms for identifying CM-I surgery are critical prerequisites for such studies but are currently only available for adults. The objective of this study was to validate two ICD-9-CM code algorithms using hospital billing data to identify pediatric patients undergoing CM-I decompression surgery. METHODS The authors retrospectively analyzed the validity of two ICD-9-CM code algorithms for identifying pediatric CM-I decompression surgery performed at 3 academic medical centers between 2001 and 2013. Algorithm 1 included any discharge diagnosis code of 348.4 (CM-I), as well as a procedure code of 01.24 (cranial decompression) or 03.09 (spinal decompression or laminectomy). Algorithm 2 restricted this group to the subset of patients with a primary discharge diagnosis of 348.4. The positive predictive value (PPV) and sensitivity of each algorithm were calculated. RESULTS Among 625 first-time admissions identified by Algorithm 1, the overall PPV for CM-I decompression was 92%. Among the 581 admissions identified by Algorithm 2, the PPV was 97%. The PPV for Algorithm 1 was lower in one center (84%) compared with the other centers (93%-94%), whereas the PPV of Algorithm 2 remained high (96%-98%) across all subgroups. The sensitivity of Algorithms 1 (91%) and 2 (89%) was very good and remained so across subgroups (82%-97%). CONCLUSIONS An ICD-9-CM algorithm requiring a primary diagnosis of CM-I has excellent PPV and very good sensitivity for identifying CM-I decompression surgery in pediatric patients. These results establish a basis for utilizing administrative billing data to assess pediatric CM-I treatment outcomes.

  1. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 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, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

  2. VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2015-10-01

    Full Text Available ABSTRACT VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS. The coupled neutronic and thermal-hydraulic (T/H code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR ejection at peripheral core using a full core geometry model, the C1 and C2 cases.  By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM and the improved quasistatic method (IQS. All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4% for C2 case.  All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. Keywords: nodal method, coupled neutronic and thermal-hydraulic code, PWR, transient case, control rod ejection.   ABSTRAK VALIDASI MODEL GEOMETRI TERAS PENUH PAKET PROGRAM NODAL3 DALAM PROBLEM BENCHMARK GAYUT WAKTU PWR. Paket program kopel neutronik dan termohidraulika (T/H, NODAL3, telah divalidasi dengan beberapa kasus benchmark statis PWR dan kasus benchmark gayut waktu PWR NEACRP.  Akan tetapi, paket program NODAL3 belum divalidasi dalam kasus benchmark gayut waktu akibat penarikan sebuah perangkat batang kendali (CR di tepi teras menggunakan model geometri teras penuh, yaitu kasus C1 dan C2. Dengan penelitian ini, akurasi paket program

  3. 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.

  4. Acceptance and validation test report for HANSF code version 1.3.2

    International Nuclear Information System (INIS)

    PIEPHO, M.G.

    2001-01-01

    The HANSF code, Version 1.3.2, is a stand-along code that runs only in DOS. As a result, it runs on any Windows' platform, since each Windows(trademark) platform can create a DOS-prompt window and execute HANSF in the DOS window. The HANSF code is proprietary to Fauske and Associates, Inc., (FAI) of Burr Ridge, IL, the developers of the code. The SNF Project has a license from FAI to run the HANSF code on any computer for only work related to SNF Project. The SNF Project owns the MCO.FOR routine, which is the main routine in HANSF for CVDF applications. The HANSF code calculates physical variables such as temperature, pressure, oxidation rates due to chemical reactions of uranium metal/fuel with water or oxygen. The code is used by the Spent Nuclear Fuel (SNF) Project at Hanford; for example, the report Thermal Analysis of Cold Vacuum Drying of Spent Nuclear Fuel (HNF-SD-SNF-CN-023). The primary facilities of interest are the K-Basins, Cold Vacuum Drying Facility (CVDF), Canister Storage Building (CSB) and T Plant. The overall Summary is presented in Section 2.0, Variances in Section 3.0, Comprehensive Assessment in Section 4.0, Results in Section 5.0, Evaluation in Section 6.0, and Summary of Activities in Section 7.0

  5. Validation of thermohydraulic codes by comparison of experimental results with computer simulations

    International Nuclear Information System (INIS)

    Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

    1989-01-01

    The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

  6. Research in aeroelasticity[Wind turbines

    Energy Technology Data Exchange (ETDEWEB)

    Bak, C.

    2006-05-15

    In the Energy Research Project 'Program for Research in Applied Aeroelasticity' (EFP2005), Risoe National Laboratory (Risoe) and the Technical University of Denmark (DTU) have applied and further developed the tools in the aeroelastic design complex. The main results from the project are: 1) Adding a winglet to a wind turbine blade for minimizing the induced drag of the blade led to the biggest increase in power of 1.4%. 2) Transient wind loads during pitch motion are determined using CFD. Compared to the NREL/NASA Ames test, reasonably good agreement is seen. 3) A general method was developed for the determination of 3D angle of attack for rotating blades from either measurements or numerical computations using CFD. 4) A model of the far wake behind wind turbines was developed for stability studies of the tip vortices in the far wake. 5) Investigating the blade root region showed that the power efficiency, CP, locally can be increased significantly beyond the Betz limit, but that the global CP for the rotor cannot exceed the Betz limit. When including tip losses and a minimum blade drag coefficient, a maximum rotor CP in the range of 0.51-0.52 was obtained. 6) A new airfoil family was designed and a 3D airfoil design tool was developed. Compared to the Risoe-B1 family, the new airfoil family showed similar or improved aerodynamic and structural characteristics. 7) Four different airfoils were analyzed to reveal the differences between 2D and 3D CFD. The major conclusions are the dependency of computational results to transition modelling, and the ability of 3D DES calculations to realistically simulate the turbulent wake of an airfoil in stall. 8) The capability of a theory for simulation of Gaussian turbulence driven gust events was demonstrated by emulating a violent shear gust event from a complex site. An asymptotic model for the PDF of the largest excursion from the mean level, during an arbitrary recurrence period, has been derived for a stochastic

  7. Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

    Science.gov (United States)

    2014-03-27

    Vehicle Code System (VCS), the Monte Carlo Adjoint SHielding (MASH), and the Monte Carlo n- Particle ( MCNP ) code. Of the three, the oldest and still most...widely utilized radiation transport code is MCNP . First created at Los Alamos National Laboratory (LANL) in 1957, the code simulated neutral...particle types, and previous versions of MCNP were repeatedly validated using both simple and complex 10 geometries [12, 13]. Much greater discussion and

  8. Indicial lift response function: an empirical relation for finite‐thickness airfoils, and effects on aeroelastic simulations

    DEFF Research Database (Denmark)

    Bergami, Leonardo; Gaunaa, Mac; Heinz, Joachim Christian

    2013-01-01

    The aeroelastic response of wind turbines is often simulated in the time domain by using indicial response techniques. Unsteady aerodynamics in attached flow are usually based on Jones's approximation of the flat plate indicial response, although the response for finite‐thickness airfoils differs...... from the flat plate one. The indicial lift response of finite‐thickness airfoils is simulated with a panel code, and an empirical relation is outlined connecting the airfoil indicial response to its geometric characteristics. The effects of different indicial approximations are evaluated on a 2D...... of equivalent fatigue loads, ultimate loads, and stability limits. The agreement with CFD computations of a 2D profile in harmonic motion is improved by the indicial function accounting for the finite‐thickness of the airfoil. Concerning the full wind turbine aeroelastic behavior, the differences between...

  9. Using clinician text notes in electronic medical record data to validate transgender-related diagnosis codes.

    Science.gov (United States)

    Blosnich, John R; Cashy, John; Gordon, Adam J; Shipherd, Jillian C; Kauth, Michael R; Brown, George R; Fine, Michael J

    2018-04-04

    Transgender individuals are vulnerable to negative health risks and outcomes, but research remains limited because data sources, such as electronic medical records (EMRs), lack standardized collection of gender identity information. Most EMR do not include the gold standard of self-identified gender identity, but International Classification of Diseases (ICDs) includes diagnostic codes indicating transgender-related clinical services. However, it is unclear if these codes can indicate transgender status. The objective of this study was to determine the extent to which patients' clinician notes in EMR contained transgender-related terms that could corroborate ICD-coded transgender identity. Data are from the US Department of Veterans Affairs Corporate Data Warehouse. Transgender patients were defined by the presence of ICD9 and ICD10 codes associated with transgender-related clinical services, and a 3:1 comparison group of nontransgender patients was drawn. Patients' clinician text notes were extracted and searched for transgender-related words and phrases. Among 7560 patients defined as transgender based on ICD codes, the search algorithm identified 6753 (89.3%) with transgender-related terms. Among 22 072 patients defined as nontransgender without ICD codes, 246 (1.1%) had transgender-related terms; after review, 11 patients were identified as transgender, suggesting a 0.05% false negative rate. Using ICD-defined transgender status can facilitate health services research when self-identified gender identity data are not available in EMR.

  10. Aeroelastic tailoring of composite aircraft wings

    Science.gov (United States)

    Mihaila-Andres, Mihai; Larco, Ciprian; Rosu, Paul-Virgil; Rotaru, Constantin

    2017-07-01

    The need of a continuously increasing size and performance of aerospace structures has settled the composite materials as the preferred materials in aircraft structures. Apart from the clear capacity to reduce the structural weight and with it the manufacture cost and the fuel consumption while preserving proper airworthiness, the prospect of tailoring a structure using the unique directional stiffness properties of composite materials allows an aerospace engineer to optimize aircraft structures to achieve particular design objectives. This paper presents a brief review of what is known as the aeroelastic tailoring of airframes with the intent of understanding the evolution of this research topic and at the same time providing useful references for further studies.

  11. Aeroelastic optimization of MW wind turbines

    Energy Technology Data Exchange (ETDEWEB)

    Hartvig Hansen, M.; Zahle, F.

    2011-12-15

    This report contains the results from the Energy Development and Demonstration Project ''Aeroelastic Optimization of MW wind turbine'' (AeroOpt). The project has had the following five Work Packages: 1. Geometric non-linear, anisotropic beam element for HAWC2. 2. Closed-loop eigenvalue analysis of controlled wind turbines. 3. Resonant wave excitation of lateral tower bending modes. 4. Development of next generation aerodynamic design tools. 5. Advanced design and verification of airfoils. The purposes of these Work Packages are briefly described in the Preface and a summary of the results are given in Section 2. Thereafter, the results from each Work Package are described in eight subsequent chapters. (Author)

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

    International Nuclear Information System (INIS)

    Kiteley, J.C.; Carver, M.B.; Zhou, Q.N.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

    1993-01-01

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

  15. Validation of the computer code system ATHLET / ATHLET-CD. Final report; Validierung des Rechenprogrammsystems ATHLET / ATHLET-CD. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Austregesilo, H.; Bals, C.; Erdmann, W.; Horche, W.; Krzykacz-Hausmann, B.; Pointner, W.; Schoeffel, P.; Skorek, T.; Weber, S.; Wielenberg, A.

    2010-04-15

    In the frame of the reactor safety project RS1173, sponsored by the German Federal Ministry of Economics and Technology, analyses of international integral and separate effects tests have been performed for the validation of the code system ATHLET/ATHLET-CD. The work mainly comprised post-test calculations of selected experiments and the contributions to the working groups accompanying the experimental programs. For the assessment of the thermal-hydraulic models in ATHLET 8 integral tests and 4 separate effect tests have been considered. Together with the corroboration of the existing models, the validation analyses were mainly dedicated to the assessment of the modelling of non-condensable gases and their influence on two-phase natural circulation and on the primary heat removal through steam generators, as well as of the simulation of multi-dimensional flow processes. The validation calculations with respect to the simulation of multi-dimensional one- and two-phase flows aimed to investigate the range of applicability and limitations of the method of parallel channels in connection with the separate momentum equations for water and steam current used in ATHLET as well as to assess the status of the coupled version ATHLET/FLUBOX-3D. The ATHLET-CD validation analyses included the post-test calculations of 9 bundle tests, and was mainly focussed on the assessment of the improved and new models for core degradation, including the models for oxidation, melt formation and relocation for BWR components, as well as of the modelling of fission products and aerosol transport within the primary circuit taking into account chemical reactions within the module SOPHAEROS. As an additional contribution to code validation, the GRS methodology of uncertainty and sensitivity analysis was applied exemplarily to two validation calculations, one with ATHLET and one with ATHLET-CD. The results of these uncertainty analyses endorse the capability of the code system to reproduce

  16. Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C

    2006-10-15

    In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions.

  17. Validation Calculations for the Application of MARS Code to the Safety Analysis of Research Reactors

    International Nuclear Information System (INIS)

    Park, Cheol; Kim, H.; Chae, H. T.; Lim, I. C.

    2006-10-01

    In order to investigate the applicability of MARS code to the accident analysis of the HANARO and other RRs, the following test data were simulated. Test data of the HANARO design and operation, Test data of flow instability and void fraction from published documents, IAEA RR transient data in TECDOC-643, Brazilian IEA-R1 experimental data. For the simulation of the HANARO data with finned rod type fuels at low pressure and low temperature conditions, MARS code, developed for the transient analysis of power reactors, was modified. Its prediction capability was assessed against the experimental data for the HANARO. From the assessment results, it can be said that the modified MARS code could be used for analyzing the thermal hydraulic transient of the HANARO. Some other simulations such as flow instability test and reactor transients were also done for the application of MARS code to RRs with plate type fuels. In the simulation for these cases, no modification was made. The results of simulated cases show that the MARS code can be used to the transient analysis of RRs with careful considerations. In particular, it seems that an improvement on a void model may be necessary for dealing with the phenomena in high void conditions

  18. Confirm Content Validity and Sender Authenticity for Text Messages by Using QR Code

    Directory of Open Access Journals (Sweden)

    Firas Mohammed Aswad

    2018-05-01

    Full Text Available In light of the information revolution taking place in the modern world, therefore it becomes necessary and important to save this electronic messages. So we offered this technique to ensure the safety of the content of the messages and authenticity of the sender through  networks communication by converting the message's symbols to numbers , each one of this symbols (letters, numbers, symbols will converted into three digits, the first digit represents the ASCII code of the symbol , the second digit represents the frequency of this symbol in the message (the number of times this symbol is appear in the message, and the third digit represents the total number of the locations of the symbol (calculates the symbol location from the first symbol in the message to this symbol itself and blanks also calculated too .The digital signature of the sender will converted to numbers like the symbols of message we explained it before, and this numbers of the digital signature will gathering together to produce three numbers only, this number will gathering with each numbers of the message's symbols, the final  numbers will converted to QR Code , the QR Code will placed with the message and sent to the recipient. The recipient returns the steps of the sender (produce QR Code from the received message and compared it the received QR Codes, if it is match or not. The recipient will ensure that the content is secure, and confirms the authenticity of the sender.

  19. Experiments for the validation of computer codes uses to assess the protection factors afforded by dwellings

    International Nuclear Information System (INIS)

    Le Grand, J.; Roux, Y.; Kerlau, G.

    1988-09-01

    Two experimental campaigns were carried out to verify: 1) the method of assessing the mean kerma in a household used in the computer code BILL calculating the protection factor afforded by dwellings; 2) in what conditions the kerma calculated in cubic meshes of a given size (code PIECE) agreed with TLD measurements. To that purpose, a house was built near the caesium 137 source of the Ecosystem irradiator located at the Cadarache Nuclear Research Center. During the first campaign, four experiments with different house characteristics were conducted. Some 50 TLSs locations describing the inhabitable volume were defined in order to obtain the mean kerma. 16 locations were considered outside the house. During the second campaign a cobalt 60 source was installed on the side. Only five measurement locations were defined, each with 6 TLDs. The results of dosimetric measurements are presented and compared with the calculations of the two computer codes. The effects of wall heterogeneity were also studied [fr

  20. Validation of a new library of nuclear constants of the WIMS code

    International Nuclear Information System (INIS)

    Aguilar H, F.

    1991-10-01

    The objective of the present work is to reproduce the experimental results of the thermal reference problems (benchmarks) TRX-1, TRX-2 and BAPL-1 to BAPL-3 with the WIMS code. It was proceeded in two stages, the first one consisted on using the original library of the code, while in the second one, a library that only contains the present elements in the benchmarks: H 1 , O 16 , Al 27 , U 235 and U 238 was generated. To generate the present nuclear data in the WIMS library, it was used the ENDF/B-IV database and the Data processing system of Nuclear Data NJOY, the library was generated using the FIXER code. (Author)

  1. Validation of Simulation Codes for Future Systems: Motivations, Approach and the Role of Nuclear Data

    International Nuclear Information System (INIS)

    G. Palmiotti; M. Salvatores; G. Aliberti

    2007-01-01

    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

  2. EVLncRNAs: a manually curated database for long non-coding RNAs validated by low-throughput experiments

    Science.gov (United States)

    Zhao, Huiying; Yu, Jiafeng; Guo, Chengang; Dou, Xianghua; Song, Feng; Hu, Guodong; Cao, Zanxia; Qu, Yuanxu

    2018-01-01

    Abstract Long non-coding RNAs (lncRNAs) play important functional roles in various biological processes. Early databases were utilized to deposit all lncRNA candidates produced by high-throughput experimental and/or computational techniques to facilitate classification, assessment and validation. As more lncRNAs are validated by low-throughput experiments, several databases were established for experimentally validated lncRNAs. However, these databases are small in scale (with a few hundreds of lncRNAs only) and specific in their focuses (plants, diseases or interactions). Thus, it is highly desirable to have a comprehensive dataset for experimentally validated lncRNAs as a central repository for all of their structures, functions and phenotypes. Here, we established EVLncRNAs by curating lncRNAs validated by low-throughput experiments (up to 1 May 2016) and integrating specific databases (lncRNAdb, LncRANDisease, Lnc2Cancer and PLNIncRBase) with additional functional and disease-specific information not covered previously. The current version of EVLncRNAs contains 1543 lncRNAs from 77 species that is 2.9 times larger than the current largest database for experimentally validated lncRNAs. Seventy-four percent lncRNA entries are partially or completely new, comparing to all existing experimentally validated databases. The established database allows users to browse, search and download as well as to submit experimentally validated lncRNAs. The database is available at http://biophy.dzu.edu.cn/EVLncRNAs. PMID:28985416

  3. Experimental validation of the DPM Monte Carlo code using minimally scattered electron beams in heterogeneous media

    International Nuclear Information System (INIS)

    Chetty, Indrin J.; Moran, Jean M.; Nurushev, Teamor S.; McShan, Daniel L.; Fraass, Benedick A.; Wilderman, Scott J.; Bielajew, Alex F.

    2002-01-01

    A comprehensive set of measurements and calculations has been conducted to investigate the accuracy of the Dose Planning Method (DPM) Monte Carlo code for electron beam dose calculations in heterogeneous media. Measurements were made using 10 MeV and 50 MeV minimally scattered, uncollimated electron beams from a racetrack microtron. Source distributions for the Monte Carlo calculations were reconstructed from in-air ion chamber scans and then benchmarked against measurements in a homogeneous water phantom. The in-air spatial distributions were found to have FWHM of 4.7 cm and 1.3 cm, at 100 cm from the source, for the 10 MeV and 50 MeV beams respectively. Energy spectra for the electron beams were determined by simulating the components of the microtron treatment head using the code MCNP4B. Profile measurements were made using an ion chamber in a water phantom with slabs of lung or bone-equivalent materials submerged at various depths. DPM calculations are, on average, within 2% agreement with measurement for all geometries except for the 50 MeV incident on a 6 cm lung-equivalent slab. Measurements using approximately monoenergetic, 50 MeV, 'pencil-beam'-type electrons in heterogeneous media provide conditions for maximum electronic disequilibrium and hence present a stringent test of the code's electron transport physics; the agreement noted between calculation and measurement illustrates that the DPM code is capable of accurate dose calculation even under such conditions. (author)

  4. Validation of a thermal-hydraulic system code on a simple example

    International Nuclear Information System (INIS)

    Kopecek, Vit; Zacha, Pavel

    2014-01-01

    A mathematical model of a U tube was set up and the analytical solution was calculated and used in the assessment of the numerical solutions obtained by using the RELAP5 mod3.3 and TRACE V5 thermal hydraulics codes. A good agreement between the 2 types of calculation was obtained.

  5. Development and validation of the fast doppler broadening module coupled within RMC code

    International Nuclear Information System (INIS)

    Yu Jiankai; Liang Jin'gang; Yu Ganglin; Wang Kan

    2015-01-01

    It is one of the efficient approach to reduce the memory consumption in Monte Carlo based reactor physical simulations by using the On-the-fly Doppler broadening for temperature dependent nuclear cross sections. RXSP is a nuclear cross sections processing code being developed by REAL team in Department of Engineering Physics in Tsinghua University, which has an excellent performance in Doppler broadening the temperature dependent continuous energy neutron cross sections. To meet the dual requirements of both accuracy and efficiency during the Monte Carlo simulations with many materials and many temperatures in it, this work enables the capability of on-the-fly pre-Doppler broadening cross sections during the neutron transport by coupling the Fast Doppler Broaden module in RXSP code embedded in the RMC code also being developed by REAL team in Tsinghua University. Additionally, the original OpenMP-based parallelism has been successfully converted into the MPI-based framework, being fully compatible with neutron transport in RMC code, which has achieved a vast parallel efficiency improvement. This work also provides a flexible approach to solve Monte Carlo based full core depletion calculation with many temperatures feedback in many isotopes. (author)

  6. Code Validation of CFD Heat Transfer Models for Liquid Rocket Engine Combustion Devices

    National Research Council Canada - National Science Library

    Coy, E. B

    2007-01-01

    .... The design of the rig and its capabilities are described. A second objective of the test rig is to provide CFD validation data under conditions relevant to liquid rocket engine thrust chambers...

  7. Validation of the REL2005 code package on Gd-poisoned PWR type assemblies through the CAMELEON experimental program

    International Nuclear Information System (INIS)

    Blaise, Patrick; Vidal, Jean-Francois; Santamarina, Alain

    2009-01-01

    This paper details the validation of Gd-poisoned 17x17 PWR lattices, through several configurations of the CAMELEON experimental program, by using the newly qualified REL2005 French code package. After a general presentation of the CAMELEON program that took place in the EOLE critical Facility in Cadarache, one describes the new REL2005 code package relying on the deterministic transport code APOLLO2.8 based on characteristics method (MOC), and its new CEA2005 library based on the latest JEFF-3.1.1 nuclear data evaluation. For critical masses, the average Calculation-to-Experiment C/E's on the k eff are (136 ± 80) pcm and (300 ± 76) pcm for the reference 281 groups MOC and optimized 26 groups MOC schemes respectively. These values include also a drastic improvement of about 250 pcm due to the change in the library from JEF2.2 to JEFF3.1. For pin-by-pin radial power distributions, reference and REL2005 results are very close, with maximum discrepancies of the order of 2%, i.e., in the experimental uncertainty limits. The Optimized REL2005 code package allows to predict the reactivity worth of the Gd-clusters (averaged on 9 experimental configurations) to be C/E Δρ(Gd clusters) = +1.3% ± 2.3%. (author)

  8. Verification and Validation of the k-kL Turbulence Model in FUN3D and CFL3D Codes

    Science.gov (United States)

    Abdol-Hamid, Khaled S.; Carlson, Jan-Renee; Rumsey, Christopher L.

    2015-01-01

    The implementation of the k-kL turbulence model using multiple computational uid dy- namics (CFD) codes is reported herein. The k-kL model is a two-equation turbulence model based on Abdol-Hamid's closure and Menter's modi cation to Rotta's two-equation model. Rotta shows that a reliable transport equation can be formed from the turbulent length scale L, and the turbulent kinetic energy k. Rotta's equation is well suited for term-by-term mod- eling and displays useful features compared to other two-equation models. An important di erence is that this formulation leads to the inclusion of higher-order velocity derivatives in the source terms of the scale equations. This can enhance the ability of the Reynolds- averaged Navier-Stokes (RANS) solvers to simulate unsteady ows. The present report documents the formulation of the model as implemented in the CFD codes Fun3D and CFL3D. Methodology, veri cation and validation examples are shown. Attached and sepa- rated ow cases are documented and compared with experimental data. The results show generally very good comparisons with canonical and experimental data, as well as matching results code-to-code. The results from this formulation are similar or better than results using the SST turbulence model.

  9. Assessment of heat transfer correlations for supercritical water in the frame of best-estimate code validation

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Espinoza, Victor H. Sanchez; Schneider, Niko; Hurtado, Antonio

    2009-01-01

    Within the frame of the Generation IV international forum six innovative reactor concepts are the subject of comprehensive investigations. In some projects supercritical water will be considered as coolant, moderator (as for the High Performance Light Water Reactor) or secondary working fluid (one possible option for Liquid Metal-cooled Fast Reactors). Supercritical water is characterized by a pronounced change of the thermo-physical properties when crossing the pseudo-critical line, which goes hand in hand with a change in the heat transfer (HT) behavior. Hence, it is essential to estimate, in a proper way, the heat-transfer coefficient and subsequently the wall temperature. The scope of this paper is to present and discuss the activities at the Institute for Reactor Safety (IRS) related to the implementation of correlations for wall-to-fluid HT at supercritical conditions in Best-Estimate codes like TRACE as well as its validation. It is important to validate TRACE before applying it to safety analyses of HPLWR or of other reactor systems. In the past 3 decades various experiments have been performed all over the world to reveal the peculiarities of wall-to-fluid HT at supercritical conditions. Several different heat transfer phenomena such as HT enhancement (due to higher Prandtl numbers in the vicinity of the pseudo-critical point) or HT deterioration (due to strong property variations) were observed. Since TRACE is a component based system code with a finite volume method the resolution capabilities are limited and not all physical phenomena can be modeled properly. But Best -Estimate system codes are nowadays the preferred option for safety related investigations of full plants or other integral systems. Thus, the increase of the confidence in such codes is of high priority. In this paper, the post-test analysis of experiments with supercritical parameters will be presented. For that reason various correlations for the HT, which considers the characteristics

  10. Numerical techniques for the improved performance of a finite element approach to wind turbine aeroelastics

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, M.B. [Renewable Energy Systems Ltd., Hemel Hempstead (United Kingdom)

    1996-09-01

    It is possible to compute the aeroelastic response of a horizontal axis wind turbine comprising; Structural: rotor substructure 144 dof, tower substructure 48 dof, induction, synchronous or variable speed, and gearbox. Aerodynamic: 3 blades (10 elements per blade), dynamic stall, and 6 different aerofoil types with combination of fixed or pitching elements. Control: stall or power regulation or speed control and shutdowns, wind shear, and tower shadow. Turbulence: 8 radial points, 32 circumferential, and 3 components. On a DEC Alpha Workstation the code will simulate the response inclose to real-time. As the code is presently formulated deflections from the initial starting point have to be small and therefore its ability to fully analyse very flexible structures is limited. (EG)

  11. Parallel scalability and efficiency of vortex particle method for aeroelasticity analysis of bluff bodies

    Science.gov (United States)

    Tolba, Khaled Ibrahim; Morgenthal, Guido

    2018-01-01

    This paper presents an analysis of the scalability and efficiency of a simulation framework based on the vortex particle method. The code is applied for the numerical aerodynamic analysis of line-like structures. The numerical code runs on multicore CPU and GPU architectures using OpenCL framework. The focus of this paper is the analysis of the parallel efficiency and scalability of the method being applied to an engineering test case, specifically the aeroelastic response of a long-span bridge girder at the construction stage. The target is to assess the optimal configuration and the required computer architecture, such that it becomes feasible to efficiently utilise the method within the computational resources available for a regular engineering office. The simulations and the scalability analysis are performed on a regular gaming type computer.

  12. Development, verification and validation of the fuel channel behaviour computer code FACTAR

    Energy Technology Data Exchange (ETDEWEB)

    Westbye, C J; Brito, A C; MacKinnon, J C; Sills, H E; Langman, V J [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thermal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate loss of coolant accident conditions including transition and large break LOCA`s (loss of coolant accidents) with emergency coolant injection assumed available. FACTAR`s predictions of fuel temperature and sheath failure times are used to subsequent assessment of fission product releases and fuel string expansion. This paper discusses the origin and development history of FACTAR, presents the mathematical models and solution technique, the detailed quality assurance procedures that are followed during development, and reports the future development of the code. (author). 27 refs., 3 figs.

  13. RELAP-7 Software Verification and Validation Plan - Requirements Traceability Matrix (RTM) Part 2: Code Assessment Strategy, Procedure, and RTM Update

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jun Soo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Choi, Yong Joon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This document addresses two subjects involved with the RELAP-7 Software Verification and Validation Plan (SVVP): (i) the principles and plan to assure the independence of RELAP-7 assessment through the code development process, and (ii) the work performed to establish the RELAP-7 assessment plan, i.e., the assessment strategy, literature review, and identification of RELAP-7 requirements. Then, the Requirements Traceability Matrices (RTMs) proposed in previous document (INL-EXT-15-36684) are updated. These RTMs provide an efficient way to evaluate the RELAP-7 development status as well as the maturity of RELAP-7 assessment through the development process.

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

    International Nuclear Information System (INIS)

    Daverio, Hernando J.

    2003-01-01

    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)

  15. RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

    International Nuclear Information System (INIS)

    Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

    2003-01-01

    This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

  16. Validation of favor code linear elastic fracture solutions for finite-length flaw geometries

    International Nuclear Information System (INIS)

    Dickson, T.L.; Keeney, J.A.; Bryson, J.W.

    1995-01-01

    One of the current tasks within the US Nuclear Regulatory Commission (NRC)-funded Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is the continuing development of the FAVOR (Fracture, analysis of Vessels: Oak Ridge) computer code. FAVOR performs structural integrity analyses of embrittled nuclear reactor pressure vessels (RPVs) with stainless steel cladding, to evaluate compliance with the applicable regulatory criteria. Since the initial release of FAVOR, the HSST program has continued to enhance the capabilities of the FAVOR code. ABAQUS, a nuclear quality assurance certified (NQA-1) general multidimensional finite element code with fracture mechanics capabilities, was used to generate a database of stress-intensity-factor influence coefficients (SIFICs) for a range of axially and circumferentially oriented semielliptical inner-surface flaw geometries applicable to RPVs with an internal radius (Ri) to wall thickness (w) ratio of 10. This database of SIRCs has been incorporated into a development version of FAVOR, providing it with the capability to perform deterministic and probabilistic fracture analyses of RPVs subjected to transients, such as pressurized thermal shock (PTS), for various flaw geometries. This paper discusses the SIFIC database, comparisons with other investigators, and some of the benchmark verification problem specifications and solutions

  17. Development and validation of a low-frequency modeling code for high-moment transmitter rod antennas

    Science.gov (United States)

    Jordan, Jared Williams; Sternberg, Ben K.; Dvorak, Steven L.

    2009-12-01

    The goal of this research is to develop and validate a low-frequency modeling code for high-moment transmitter rod antennas to aid in the design of future low-frequency TX antennas with high magnetic moments. To accomplish this goal, a quasi-static modeling algorithm was developed to simulate finite-length, permeable-core, rod antennas. This quasi-static analysis is applicable for low frequencies where eddy currents are negligible, and it can handle solid or hollow cores with winding insulation thickness between the antenna's windings and its core. The theory was programmed in Matlab, and the modeling code has the ability to predict the TX antenna's gain, maximum magnetic moment, saturation current, series inductance, and core series loss resistance, provided the user enters the corresponding complex permeability for the desired core magnetic flux density. In order to utilize the linear modeling code to model the effects of nonlinear core materials, it is necessary to use the correct complex permeability for a specific core magnetic flux density. In order to test the modeling code, we demonstrated that it can accurately predict changes in the electrical parameters associated with variations in the rod length and the core thickness for antennas made out of low carbon steel wire. These tests demonstrate that the modeling code was successful in predicting the changes in the rod antenna characteristics under high-current nonlinear conditions due to changes in the physical dimensions of the rod provided that the flux density in the core was held constant in order to keep the complex permeability from changing.

  18. 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)

  19. Manual versus automated coding of free-text self-reported medication data in the 45 and Up Study: a validation study.

    Science.gov (United States)

    Gnjidic, Danijela; Pearson, Sallie-Anne; Hilmer, Sarah N; Basilakis, Jim; Schaffer, Andrea L; Blyth, Fiona M; Banks, Emily

    2015-03-30

    Increasingly, automated methods are being used to code free-text medication data, but evidence on the validity of these methods is limited. To examine the accuracy of automated coding of previously keyed in free-text medication data compared with manual coding of original handwritten free-text responses (the 'gold standard'). A random sample of 500 participants (475 with and 25 without medication data in the free-text box) enrolled in the 45 and Up Study was selected. Manual coding involved medication experts keying in free-text responses and coding using Anatomical Therapeutic Chemical (ATC) codes (i.e. chemical substance 7-digit level; chemical subgroup 5-digit; pharmacological subgroup 4-digit; therapeutic subgroup 3-digit). Using keyed-in free-text responses entered by non-experts, the automated approach coded entries using the Australian Medicines Terminology database and assigned corresponding ATC codes. Based on manual coding, 1377 free-text entries were recorded and, of these, 1282 medications were coded to ATCs manually. The sensitivity of automated coding compared with manual coding was 79% (n = 1014) for entries coded at the exact ATC level, and 81.6% (n = 1046), 83.0% (n = 1064) and 83.8% (n = 1074) at the 5, 4 and 3-digit ATC levels, respectively. The sensitivity of automated coding for blank responses was 100% compared with manual coding. Sensitivity of automated coding was highest for prescription medications and lowest for vitamins and supplements, compared with the manual approach. Positive predictive values for automated coding were above 95% for 34 of the 38 individual prescription medications examined. Automated coding for free-text prescription medication data shows very high to excellent sensitivity and positive predictive values, indicating that automated methods can potentially be useful for large-scale, medication-related research.

  20. Application of software quality assurance methods in validation and maintenance of reactor analysis computer codes

    International Nuclear Information System (INIS)

    Reznik, L.

    1994-01-01

    Various computer codes employed at Israel Electricity Company for preliminary reactor design analysis and fuel cycle scoping calculations have been often subject to program source modifications. Although most changes were due to computer or operating system compatibility problems, a number of significant modifications were due to model improvement and enhancements of algorithm efficiency and accuracy. With growing acceptance of software quality assurance requirements and methods, a program of implementing extensive testing of modified software has been adopted within the regular maintenance activities. In this work survey has been performed of various software quality assurance methods of software testing which belong mainly to the two major categories of implementation ('white box') and specification-based ('black box') testing. The results of this survey exhibits a clear preference of specification-based testing. In particular the equivalence class partitioning method and the boundary value method have been selected as especially suitable functional methods for testing reactor analysis codes.A separate study of software quality assurance methods and techniques has been performed in this work objective to establish appropriate pre-test software specification methods. Two methods of software analysis and specification have been selected as the most suitable for this purpose: The method of data flow diagrams has been shown to be particularly valuable for performing the functional/procedural software specification while the entities - relationship diagrams has been approved to be efficient for specifying software data/information domain. Feasibility of these two methods has been analyzed in particular for software uncertainty analysis and overall code accuracy estimation. (author). 14 refs

  1. Burn-up function of fuel management code for aqueous homogeneous reactors and its validation

    International Nuclear Information System (INIS)

    Wang Liangzi; Yao Dong; Wang Kan

    2011-01-01

    Fuel Management Code for Aqueous Homogeneous Reactors (FMCAHR) is developed based on the Monte Carlo transport method, to analyze the physics characteristics of aqueous homogeneous reactors. FMCAHR has the ability of doing resonance treatment, searching for critical rod heights, thermal hydraulic parameters calculation, radiolytic-gas bubbles' calculation and bum-up calculation. This paper introduces the theory model and scheme of its burn-up function, and then compares its calculation results with benchmarks and with DRAGON's burn-up results, which confirms its bum-up computing precision and its applicability in the bum-up calculation and analysis for aqueous solution reactors. (authors)

  2. Validation of the LH antenna code ALOHA against Tore Supra experiments

    International Nuclear Information System (INIS)

    Hillairet, J.; Ekedahl, A.; Kocan, M.; Gunn, J. P.; Goniche, M.

    2009-01-01

    Comparisons between ALOHA code predictions and experimental measurements of reflection coefficients for the two different Lower Hybrid Current Drive (LHCD) antennas (named C2 and C3) in Tore Supra are presented. A large variation of density in front of the antennas was obtained by varying the distance between the plasma and the antennas. Low power ( 2 ) was used in order to avoid non-linear effects on the wave coupling. Results obtained with ALOHA are in good agreement with the experimental measurements for both Tore Supra antennas and show that ALOHA is an efficient LH predictive tool.

  3. Unstructed Navier-Stokes Analysis of Wind-Tunnel Aeroelastic Effects on TCA Model 2

    Science.gov (United States)

    Frink, Neal T.; Allison, Dennis O.; Parikh, Paresh C.

    1999-01-01

    The aim of this work is to demonstrate a simple technique which accounts for aeroelastic deformations experienced by HSR wind-tunnel models within CFD computations. With improved correlations, CFD can become a more effective tool for augmenting the post-test understanding of experimental data. The present technique involves the loose coupling of a low-level structural representation within the ELAPS code, to an unstructured Navier-Stokes flow solver, USM3Dns. The ELAPS model is initially calibrated against bending characteristics of the wind-tunnel model. The strength of this method is that, with a single point calibration of a simple structural representation, the static aeroelastic effects can be accounted for in CFD calculations across a range of test conditions. No prior knowledge of the model deformation during the wind-on test is required. This approach has been successfully applied to the high aspect-ratio planforms of subsonic transports. The current challenge is to adapt the procedure to low aspect-ratio planforms typical of HSR configurations.

  4. Aeroelastic response and blade loads of a composite rotor in forward flight

    Science.gov (United States)

    Smith, Edward C.; Chopra, Inderjit

    1992-01-01

    The aeroelastic response, blade and hub loads, and shaft-fixed aeroelastic stability is investigated for a helicopter with elastically tailored composite rotor blades. A new finite element based structural analysis including nonclassical effects such as transverse shear, torsion related warping and inplane elasticity is integrated with the University of Maryland Advanced Rotorcraft Code. The structural dynamics analysis is correlated against both experimental data and detailed finite element results. Correlation of rotating natural frequencies of coupled composite box-beams is generally within 5-10 percent. The analysis is applied to a soft-inplane hingeless rotor helicopter in free flight propulsive trim. For example, lag mode damping can be increased 300 percent over a range of thrust conditions and forward speeds. The influence of unsteady aerodynamics on the blade response and vibratory hub loads is also investigated. The magnitude and phase of the flap response is substantially altered by the unsteady aerodynamic effects. Vibratory hub loads increase up to 30 percent due to unsteady aerodynamic effects.

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

    International Nuclear Information System (INIS)

    Kedziur, F.

    1980-03-01

    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.) [de

  6. Validation of activity determination codes and nuclide vectors by using results from processing of retired components and operational waste

    International Nuclear Information System (INIS)

    Lundgren, Klas; Larsson, Arne

    2012-01-01

    Decommissioning studies for nuclear power reactors are performed in order to assess the decommissioning costs and the waste volumes as well as to provide data for the licensing and construction of the LILW repositories. An important part of this work is to estimate the amount of radioactivity in the different types of decommissioning waste. Studsvik ALARA Engineering has performed such assessments for LWRs and other nuclear facilities in Sweden. These assessments are to a large content depending on calculations, senior experience and sampling on the facilities. The precision in the calculations have been found to be relatively high close to the reactor core. Of natural reasons the precision will decline with the distance. Even if the activity values are lower the content of hard to measure nuclides can cause problems in the long term safety demonstration of LLW repositories. At the same time Studsvik is processing significant volumes of metallic and combustible waste from power stations in operation and in decommissioning phase as well as from other nuclear facilities such as research and waste treatment facilities. Combining the unique knowledge in assessment of radioactivity inventory and the large data bank the waste processing represents the activity determination codes can be validated and the waste processing analysis supported with additional data. The intention with this presentation is to highlight how the European nuclear industry jointly could use the waste processing data for validation of activity determination codes. (authors)

  7. Validation and verification of MCNP6 against intermediate and high-energy experimental data and results by other codes

    International Nuclear Information System (INIS)

    Mashnik, Stepan G.

    2011-01-01

    MCNP6, the latest and most advanced LANL transport code representing a recent merger of MCNP5 and MCNPX, has been Validated and Verified (V and V) against a variety of intermediate and high-energy experimental data and against results by different versions of MCNPX and other codes. In the present work, we V and V MCNP6 using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.02 and LAQGSM03.03. We found that MCNP6 describes reasonably well various reactions induced by particles and nuclei at incident energies from 18 MeV to about 1 TeV per nucleon measured on thin and thick targets and agrees very well with similar results obtained with MCNPX and calculations by CEM03.02, LAQGSM03.01 (03.03), INCL4 + ABLA, and Bertini INC + Dresner evaporation, EPAX, ABRABLA, HIPSE, and AMD, used as stand alone codes. Most of several computational bugs and more serious physics problems observed in MCNP6/X during our V and V have been fixed; we continue our work to solve all the known problems before MCNP6 is distributed to the public. (author)

  8. Experimental validation for combustion analysis of GOTHIC 6.1b code in 2-dimensional premixed combustion experiments

    International Nuclear Information System (INIS)

    Lee, J. Y.; Lee, J. J.; Park, K. C.

    2003-01-01

    In this study, the prediction capability of GOTHIC code for hydrogen combustion phenomena was validated with the results of two-dimensional premixed hydrogen combustion experiment executed by Seoul National University. In the experimental results, we could confirm the propagation characteristics of hydrogen flame such as buoyancy effect, flame front shape etc.. The combustion time of the tests was about 0.1 sec.. In the GOTHIC analyses results, the GOTHIC code could predict the overall hydrogen flame propagation characteristics but the buoyancy effect and flame shape did not compare well with the experimental results. Especially, in case of the flame propagate to the dead-end, GOTHIC predicted the flame did not affected by the flow and this cause quite different results in flame propagation from experimental results. Moreover the combustion time of the analyses was about 1 sec. which is ten times longer than the experimental result. To obtain more reasonable analysis results, it is necessary that combustion model parameters in GOTHIC code apply appropriately and hydrogen flame characteristics be reflected in solving governing equations

  9. Development of a FBR fuel pin bundle deformation analysis code 'BAMBOO' . Development of a dispersion model and its validation

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ukai, Shigeharu; Asaga, Takeo

    2002-03-01

    Bundle Duct Interaction (BDI) is one of the life limiting factors of a FBR fuel subassembly. Under the BDI condition, the fuel pin dispersion would occur mainly by the deviation of the wire position due to the irradiation. In this study the effect of the dispersion on the bundle deformation was evaluated by using the BAMBOO code and following results were obtained. (1) A new contact analysis model was introduced in BAMBOO code. This model considers the contact condition at the axial position other than the nodal point of the beam element that composes the fuel pin. This improvement made it possible in the bundle deformation analysis to cause fuel pin dispersion due to the deviations of the wire position. (2) This model was validated with the results of the out-of-pile compression test with the wire deviation. The calculated pin-to-duct and pin-to-pin clearances with the dispersion model almost agreed with the test results. Therefore it was confirmed that the BAMBOO code reasonably predicts the bundle deformation with the dispersion. (3) In the dispersion bundle the pin-to-pin clearances widely scattered. And the minimum pin-to-duct clearance increased or decreased depending on the dispersion condition compared to the no-dispersion bundle. This result suggests the possibility that the considerable dispersion would affect the thermal integrity of the bundle. (author)

  10. Model Reduction of Nonlinear Aeroelastic Systems Experiencing Hopf Bifurcation

    KAUST Repository

    Abdelkefi, Abdessattar

    2013-06-18

    In this paper, we employ the normal form to derive a reduced - order model that reproduces nonlinear dynamical behavior of aeroelastic systems that undergo Hopf bifurcation. As an example, we consider a rigid two - dimensional airfoil that is supported by nonlinear springs in the pitch and plunge directions and subjected to nonlinear aerodynamic loads. We apply the center manifold theorem on the governing equations to derive its normal form that constitutes a simplified representation of the aeroelastic sys tem near flutter onset (manifestation of Hopf bifurcation). Then, we use the normal form to identify a self - excited oscillator governed by a time - delay ordinary differential equation that approximates the dynamical behavior while reducing the dimension of the original system. Results obtained from this oscillator show a great capability to predict properly limit cycle oscillations that take place beyond and above flutter as compared with the original aeroelastic system.

  11. Aeroelastic behavior of composite rotor blades with swept tips

    Science.gov (United States)

    Yuan, Kuo-An; Friedmann, Peretz P.; Venkatesan, Comandur

    1992-01-01

    This paper presents an analytical study of the aeroelastic behavior of composite rotor blades with straight and swept tips. The blade is modeled by beam type finite elements. A single finite element is used to model the swept tip. The nonlinear equations of motion for the finite element model are derived using Hamilton's principle and based on a moderate deflection theory and accounts for: arbitrary cross-sectional shape, pretwist, generally anisotropic material behavior, transverse shears and out-of-plane warping. Numerical results illustrating the effects of tip sweep, anhedral and composite ply orientation on blade aeroelastic behavior are presented. It is shown that composite ply orientation has a substantial effect on blade stability. At low thrust conditions, certain ply orientations can cause instability in the lag mode. The flap-torsion coupling associated with tip sweep can also induce aeroelastic instability in the blade. This instability can be removed by appropriate ply orientation in the composite construction.

  12. Aeroelastic Tailoring of a Plate Wing with Functionally Graded Materials

    Science.gov (United States)

    Dunning, Peter D.; Stanford, Bret K.; Kim, H. Alicia; Jutte, Christine V.

    2014-01-01

    This work explores the use of functionally graded materials for the aeroelastic tailoring of a metallic cantilevered plate-like wing. Pareto trade-off curves between dynamic stability (flutter) and static aeroelastic stresses are obtained for a variety of grading strategies. A key comparison is between the effectiveness of material grading, geometric grading (i.e., plate thickness variations), and using both simultaneously. The introduction of material grading does, in some cases, improve the aeroelastic performance. This improvement, and the physical mechanism upon which it is based, depends on numerous factors: the two sets of metallic material parameters used for grading, the sweep of the plate, the aspect ratio of the plate, and whether the material is graded continuously or discretely.

  13. Benchmarking (Code2Code) of the 1Hs 3-Bladed Onshore VAWT

    DEFF Research Database (Denmark)

    Galinos, Christos; Schmidt Paulsen, Uwe

    This study is part of the Inflow project. In this report the Nenuphar’s onshore 3-bladed Vertical Axis Wind Turbine (VAWT) prototype (1HS) is modelled in HAWC2 aeroelastic code. In the first part the model properties are summarized. Then the analysis is focused on the rotor performance and various...

  14. Validation of Heat Transfer and Film Cooling Capabilities of the 3-D RANS Code TURBO

    Science.gov (United States)

    Shyam, Vikram; Ameri, Ali; Chen, Jen-Ping

    2010-01-01

    The capabilities of the 3-D unsteady RANS code TURBO have been extended to include heat transfer and film cooling applications. The results of simulations performed with the modified code are compared to experiment and to theory, where applicable. Wilcox s k-turbulence model has been implemented to close the RANS equations. Two simulations are conducted: (1) flow over a flat plate and (2) flow over an adiabatic flat plate cooled by one hole inclined at 35 to the free stream. For (1) agreement with theory is found to be excellent for heat transfer, represented by local Nusselt number, and quite good for momentum, as represented by the local skin friction coefficient. This report compares the local skin friction coefficients and Nusselt numbers on a flat plate obtained using Wilcox's k-model with the theory of Blasius. The study looks at laminar and turbulent flows over an adiabatic flat plate and over an isothermal flat plate for two different wall temperatures. It is shown that TURBO is able to accurately predict heat transfer on a flat plate. For (2) TURBO shows good qualitative agreement with film cooling experiments performed on a flat plate with one cooling hole. Quantitatively, film effectiveness is under predicted downstream of the hole.

  15. Development and Validation of a Momentum Integral Numerical Analysis Code for Liquid Metal Fast Reactor

    International Nuclear Information System (INIS)

    Chen, Xiangyi; Suh, Kune Y.

    2016-01-01

    In this work, this benchmark problem is conducted to assess the precision of the upgraded in-house code MINA. Comparison of the results from different best estimate codes employed by various grid spacer pressure drop correlations is carried out to suggest the best one. By modifying In's method, it presents good agreement with the experiment data which is shown in Figure 7. The reason for the failure of the prediction in previous work is caused by the utilization of Rehme's method which is categorized into four groups according to different fitting strategy. Through comparison of drag coefficients calculated by four groups of Rheme's method, equivalent drag coefficient calculated by In's method and experiment data shown in Figure 8, we can conclude that Rehme's method considerably underestimate the drag coefficients in grid spacers used in HELIOS and In's method give a reasonable prediction. Starting from the core inlet, the accumulated pressure losses are presented in figure 9 along the accumulated length of the forced convection flow path; the good agreement of the prediction from MINA with the experiment result shows MINA has very good capability in integrated momentum analysis makes it robust in the future design scoping method development of LFR.

  16. Validation of Printed Circuit Heat Exchanger Design Code KAIST{sub H}XD

    Energy Technology Data Exchange (ETDEWEB)

    Baik, Seungjoon; Kim, Seong Gu; Lee, Jekyoung; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been suggested for the SFR due to the relatively mild sodium-CO{sub 2} interaction. The S-CO{sub 2} power conversion cycle can achieve not only high safety but also high efficiency with SFR core thermal condition. However, due to the dramatic property change near the critical point, the inlet pressure and temperature conditions of compressor can have significant effect on the overall cycle efficiency. To maintain the inlet condition of compressor, a sensitive precooler control system is required for stable operation. Therefore understanding the precooler performance is essential for the S-CO{sub 2} power conversion system. According to experimental result, designed PCHE showed high effectiveness in various operating regions. Comparing the experimental and the design data, heat transfer performance estimation showed less than 6% error. On the other hand, the pressure drop estimation showed large gap. The water side pressure drop showed 50-70% under estimation. Because the form losses were not included in the design code, water side pressure drop estimation result seems reliable. However, the CO{sub 2} side showed more than 70% over estimation in the pressure drop from the code. The authors suspect that the differences may have occurred by the channel corner shape. The real channel has round corners and smooth edge, but the correlation is based on the sharp edged zig-zag channel. Further studies are required to understand and interpret the results correctly in the future.

  17. Development and Validation of a Momentum Integral Numerical Analysis Code for Liquid Metal Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiangyi; Suh, Kune Y. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this work, this benchmark problem is conducted to assess the precision of the upgraded in-house code MINA. Comparison of the results from different best estimate codes employed by various grid spacer pressure drop correlations is carried out to suggest the best one. By modifying In's method, it presents good agreement with the experiment data which is shown in Figure 7. The reason for the failure of the prediction in previous work is caused by the utilization of Rehme's method which is categorized into four groups according to different fitting strategy. Through comparison of drag coefficients calculated by four groups of Rheme's method, equivalent drag coefficient calculated by In's method and experiment data shown in Figure 8, we can conclude that Rehme's method considerably underestimate the drag coefficients in grid spacers used in HELIOS and In's method give a reasonable prediction. Starting from the core inlet, the accumulated pressure losses are presented in figure 9 along the accumulated length of the forced convection flow path; the good agreement of the prediction from MINA with the experiment result shows MINA has very good capability in integrated momentum analysis makes it robust in the future design scoping method development of LFR.

  18. Validation of PV-RPM Code in the System Advisor Model.

    Energy Technology Data Exchange (ETDEWEB)

    Klise, Geoffrey Taylor [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lavrova, Olga [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Freeman, Janine [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-04-01

    This paper describes efforts made by Sandia National Laboratories (SNL) and the National Renewable Energy Laboratory (NREL) to validate the SNL developed PV Reliability Performance Model (PV - RPM) algorithm as implemented in the NREL System Advisor Model (SAM). The PV - RPM model is a library of functions that estimates component failure and repair in a photovoltaic system over a desired simulation period. The failure and repair distributions in this paper are probabilistic representations of component failure and repair based on data collected by SNL for a PV power plant operating in Arizona. The validation effort focuses on whether the failure and repair dist ributions used in the SAM implementation result in estimated failures that match the expected failures developed in the proof - of - concept implementation. Results indicate that the SAM implementation of PV - RPM provides the same results as the proof - of - concep t implementation, indicating the algorithms were reproduced successfully.

  19. On various validity criteria for the configuration average in collisional-radiative codes

    Energy Technology Data Exchange (ETDEWEB)

    Poirier, M [Commissariat a l' Energie Atomique, Service ' Photons, Atomes et Molecules' , Centre d' Etudes de Saclay, F91191 Gif-sur-Yvette Cedex (France)

    2008-01-28

    The characterization of out-of-local-thermal-equilibrium plasmas requires the use of collisional-radiative kinetic equations. This leads to the solution of large linear systems, for which statistical treatments such as configuration average may bring considerable simplification. In order to check the validity of this procedure, a criterion based on the comparison between a partial-rate systems and the Saha-Boltzmann solution is discussed in detail here. Several forms of this criterion are discussed. The interest of these variants is that they involve each type of relevant transition (collisional or radiative), which allows one to check separately the influence of each of these processes on the configuration-average validity. The method is illustrated by a charge-distribution analysis in carbon and neon plasmas. Finally, it is demonstrated that when the energy dispersion of every populated configuration is smaller than the electron thermal energy, the proposed criterion is fulfilled in each of its forms.

  20. Validation of main nuclear libraries used in thorium reactors using the Serpent code

    International Nuclear Information System (INIS)

    Faga, Lucas J.

    2017-01-01

    The purpose of this work is to validate the library of the Serpent standard database for systems containing U-233, U-235, Th-232, Pu-239 and Pu-240. The project will support the other projects of the newly created study group of Nuclear Engineering Center (CEN) of Instituto de Pesquisas Energéticas e Nucleares (IPEN), linked to the study of several types of reactors and their application in thorium cycles, a subject that gains more and more visibility, due to strong and potential promises of energy revolution. The results obtained at the end of the simulations were satisfactory, with the multiplication factors being effective close to 100 PCM of the values provided by the benchmarks, as expected for a validated library. The minimum distance between these values was 2 PCM and the maximum of 280 PCM. The final analysis demonstrates that the ENDF / B-VII library has validated nuclear data for the isotopes of interest and may be used in future thorium study group projects

  1. The database 'EDUD Base' for validation of neutron-physics codes used to analyze the WWER-440 cores

    International Nuclear Information System (INIS)

    Rocek, J.; Belac, J.; Miasnikov, A.

    2003-01-01

    The program and data system EDUDBase for validation of reactor computing codes was developed at NRI. It is designed for validation and evaluation of the precision of different computer codes used for WWER core analyses. The main goal of this database is to provide data for comparison with calculation results of tested codes and tools for statistical analysis or differences between the calculation results and the test data. The benchmark data sets are based on in-core measurements performed on WWER-440 reactors of Dukovany NPP. The initial data from NPP are verified, errors and inaccuracies are eliminated and data are transferred to a form, which is suitable for comparison with results of calculations. A special reduced operating history data set is created for each operating cycle ('Benchmark Operation History') to be used as an input data for calculation. It contains values of some integral quantities for each time point: effective time, integral thermal power, boron concentration, position of working group control assemblies (group 6) and inlet coolant temperature. At present, sets are available for all completed cycles up to: (unit/cycle) 1/17, 2/16, 3/15, 4/15. Power distribution is described for approx. 40 time steps during each operating cycle. 2D-power distributions are transferred into 60-degree core symmetry sector of reactor core. At present, such data sets are available only for later cycles starting with: (unit/cycle) 1/7, 2/6, 3/5, 4/5 (in other words last II cycles for each unit) (Authors)

  2. Use of Sensitivity and Uncertainty Analysis to Select Benchmark Experiments for the Validation of Computer Codes and Data

    International Nuclear Information System (INIS)

    Elam, K.R.; Rearden, B.T.

    2003-01-01

    Sensitivity and uncertainty analysis methodologies under development at Oak Ridge National Laboratory were applied to determine whether existing benchmark experiments adequately cover the area of applicability for the criticality code and data validation of PuO 2 and mixed-oxide (MOX) powder systems. The study examined three PuO 2 powder systems and four MOX powder systems that would be useful for establishing mass limits for a MOX fuel fabrication facility. Using traditional methods to choose experiments for criticality analysis validation, 46 benchmark critical experiments were identified as applicable to the PuO 2 powder systems. However, only 14 experiments were thought to be within the area of applicability for dry MOX powder systems.The applicability of 318 benchmark critical experiments, including the 60 experiments initially identified, was assessed. Each benchmark and powder system was analyzed using the Tools for Sensitivity and UNcertainty Analysis Methodology Implementation (TSUNAMI) one-dimensional (TSUNAMI-1D) or TSUNAMI three-dimensional (TSUNAMI-3D) sensitivity analysis sequences, which will be included in the next release of the SCALE code system. This sensitivity data and cross-section uncertainty data were then processed with TSUNAMI-IP to determine the correlation of each application to each experiment in the benchmarking set. Correlation coefficients are used to assess the similarity between systems and determine the applicability of one system for the code and data validation of another.The applicability of most of the experiments identified using traditional methods was confirmed by the TSUNAMI analysis. In addition, some PuO 2 and MOX powder systems were determined to be within the area of applicability of several other benchmarks that would not have been considered using traditional methods. Therefore, the number of benchmark experiments useful for the validation of these systems exceeds the number previously expected. The TSUNAMI analysis

  3. Aeroelastic research programme EFP-2001[YAW;STALL]; Forskning i aeroelasticitet EFP-2001

    Energy Technology Data Exchange (ETDEWEB)

    Aagaard Madsen, H. (ed.)

    2002-12-01

    The project covers the one year period from mid 2001 to mid 2002 and is the last part of a 5 years research programme on aeroelasticity. The overall objectives of the project are to improve the load and design basis for wind turbines and to ensure in collaboration with industry a continu-ously running process on development of new designs and solution of actual problems. Specifi-cally the main objectives for the present period are the following: a) development of a design tool for analysis of dynamic stability b) investigations of blade tip aerodynamics and blade tip design on basis of 3D CFD computa-tions c) publication of an airfoil catalogue d) load reduction using new control strategies e) aeroacoustic modelling of noise propagation During the present project period the computer code HAWCModal has been finished. The code computes the modal characteristics for a turbine as function of rotational speed. It is based on the structural modelling in the aeroelastic code HAWC and uses the same input files. The computed eigen frequencies are shown in a Campbell diagram and the corresponding modal forms can be shown graphically for an operating turbine. Finally, the structural damping is also computed by the code. HAWCModal is the basis for the stability analysis tool HAWCStab which is now under devel-opment. With HAWCStab the aeroelastic stability of a turbine can be analysed. The complex aerodynamics at three different blade tip shapes have been analysed with the three-dimensional CFD code EllipSys3D. The tip vortex was visualised and the lift and drag coef-ficients in the tip region were analysed in order to study the influence of the tip geometry on the performance and aerodynamic damping. An airfoil catalogue containing computations on 28 different airfoils for wind turbine applica-tion in comparison with experimental data has been developed and is available via the internet. Besides the main themes of the project as mentioned above there have been research

  4. Validation of the Serpent 2-DYNSUB code sequence using the Special Power Excursion Reactor Test III (SPERT III)

    International Nuclear Information System (INIS)

    Knebel, Miriam; Mercatali, Luigi; Sanchez, Victor; Stieglitz, Robert; Macian-Juan, Rafael

    2016-01-01

    Highlights: • Full few-group cross section tables created by Monte Carlo lattice code Serpent 2. • Serpent 2 group constant methodology verified for HFP static and transient cases. • Serpent 2-DYNSUB tool chainvalidated using SPERT III REA experiments. • Serpent 2-DYNSUB tool chain suitable to model RIAs in PWRs. - Abstract: The Special Power Excursion Reactor Test III (SPERT III) is studied using the Serpent 2-DYNSUB code sequence in order to validate it for modeling reactivity insertion accidents (RIA) in PWRs. The SPERT III E-core was a thermal research reactor constructed to analyze reactor dynamics. Its configuration resembles a commercial PWR on terms of fuel type, choice of moderator, coolant flow and system pressure. The initial conditions of the rod ejection accident experiments (REA) performed cover cold startup, hot startup, hot standby and operating power scenarios. Eight of these experiments were analyzed in detail. Firstly, multi-dimensional nodal diffusion cross section tables were created for the three-dimensional reactor simulator DYNSUB employing the Monte Carlo neutron transport code Serpent 2. In a second step, DYNSUB stationary simulations were compared to Monte Carlo reference three-dimensional full scale solutions obtained with Serpent 2 (cold startup conditions) and Serpent 2/SUBCHANFLOW (operating power conditions) with a good agreement being observed. The latter tool is an internal coupling of Serpent 2 and the sub-channel thermal-hydraulics code SUBCHANFLOW. Finally, DYNSUB was utilized to study the eight selected transient experiments. Results were found to match measurements well. As the selected experiments cover much of the possible transient (delayed super-critical, prompt super-critical and super-prompt critical excursion) and initial conditions (cold and hot as well as zero, little and full power reactor states) one expects in commercial PWRs, the obtained results give confidence that the Serpent 2-DYNSUB tool chain is

  5. New Flutter Analysis Technique for Time-Domain Computational Aeroelasticity

    Science.gov (United States)

    Pak, Chan-Gi; Lung, Shun-Fat

    2017-01-01

    A new time-domain approach for computing flutter speed is presented. Based on the time-history result of aeroelastic simulation, the unknown unsteady aerodynamics model is estimated using a system identification technique. The full aeroelastic model is generated via coupling the estimated unsteady aerodynamic model with the known linear structure model. The critical dynamic pressure is computed and used in the subsequent simulation until the convergence of the critical dynamic pressure is achieved. The proposed method is applied to a benchmark cantilevered rectangular wing.

  6. Validation of extended magnetohydrodynamic simulations of the HIT-SI3 experiment using the NIMROD code

    Science.gov (United States)

    Morgan, K. D.; Jarboe, T. R.; Hossack, A. C.; Chandra, R. N.; Everson, C. J.

    2017-12-01

    The HIT-SI3 experiment uses a set of inductively driven helicity injectors to apply a non-axisymmetric current drive on the edge of the plasma, driving an axisymmetric spheromak equilibrium in a central confinement volume. These helicity injectors drive a non-axisymmetric perturbation that oscillates in time, with relative temporal phasing of the injectors modifying the mode structure of the applied perturbation. A set of three experimental discharges with different perturbation spectra are modelled using the NIMROD extended magnetohydrodynamics code, and comparisons are made to both magnetic and fluid measurements. These models successfully capture the bulk dynamics of both the perturbation and the equilibrium, though disagreements related to the pressure gradients experimentally measured exist.

  7. Validation of film dryout model in a three-fluid code FIDAS

    International Nuclear Information System (INIS)

    Sugawara, Satoru

    1989-11-01

    Analytical prediction model of critical heat flux (CHF) has been developed on the basis of film dryout criterion due to droplets deposition and entrainment in annular mist flow. CHF in round tubes were analyzed by the Film Dryout Analysis Code in Subchannels, FIDAS, which is based on the three-fluid, three-field and newly developed film dryout model. Predictions by FIDAS were compared with the world-wide experimental data on CHF obtained in water and Freon for uniformly and non-uniformly heated tubes under vertical upward flow condition. Furthermore, CHF prediction capability of FIDAS was compared with those of other film dryout models for annular flow and Katto's CHF correlation. The predictions of FIDAS are in sufficient agreement with the experimental CHF data, and indicate better agreement than the other film dryout models and empirical correlation of Katto. (author)

  8. Experimental benchmark and code validation for airfoils equipped with passive vortex generators

    International Nuclear Information System (INIS)

    Baldacchino, D; Ferreira, C; Florentie, L; Timmer, N; Van Zuijlen, A; Manolesos, M; Chaviaropoulos, T; Diakakis, K; Papadakis, G; Voutsinas, S; González Salcedo, Á; Aparicio, M; García, N R.; Sørensen, N N.; Troldborg, N

    2016-01-01

    Experimental results and complimentary computations for airfoils with vortex generators are compared in this paper, as part of an effort within the AVATAR project to develop tools for wind turbine blade control devices. Measurements from two airfoils equipped with passive vortex generators, a 30% thick DU97W300 and an 18% thick NTUA T18 have been used for benchmarking several simulation tools. These tools span low-to-high complexity, ranging from engineering-level integral boundary layer tools to fully-resolved computational fluid dynamics codes. Results indicate that with appropriate calibration, engineering-type tools can capture the effects of vortex generators and outperform more complex tools. Fully resolved CFD comes at a much higher computational cost and does not necessarily capture the increased lift due to the VGs. However, in lieu of the limited experimental data available for calibration, high fidelity tools are still required for assessing the effect of vortex generators on airfoil performance. (paper)

  9. Threats to Validity When Using Open-Ended Items in International Achievement Studies: Coding Responses to the PISA 2012 Problem-Solving Test in Finland

    Science.gov (United States)

    Arffman, Inga

    2016-01-01

    Open-ended (OE) items are widely used to gather data on student performance in international achievement studies. However, several factors may threaten validity when using such items. This study examined Finnish coders' opinions about threats to validity when coding responses to OE items in the PISA 2012 problem-solving test. A total of 6…

  10. Computational aeroelasticity using a pressure-based solver

    Science.gov (United States)

    Kamakoti, Ramji

    A computational methodology for performing fluid-structure interaction computations for three-dimensional elastic wing geometries is presented. The flow solver used is based on an unsteady Reynolds-Averaged Navier-Stokes (RANS) model. A well validated k-ε turbulence model with wall function treatment for near wall region was used to perform turbulent flow calculations. Relative merits of alternative flow solvers were investigated. The predictor-corrector-based Pressure Implicit Splitting of Operators (PISO) algorithm was found to be computationally economic for unsteady flow computations. Wing structure was modeled using Bernoulli-Euler beam theory. A fully implicit time-marching scheme (using the Newmark integration method) was used to integrate the equations of motion for structure. Bilinear interpolation and linear extrapolation techniques were used to transfer necessary information between fluid and structure solvers. Geometry deformation was accounted for by using a moving boundary module. The moving grid capability was based on a master/slave concept and transfinite interpolation techniques. Since computations were performed on a moving mesh system, the geometric conservation law must be preserved. This is achieved by appropriately evaluating the Jacobian values associated with each cell. Accurate computation of contravariant velocities for unsteady flows using the momentum interpolation method on collocated, curvilinear grids was also addressed. Flutter computations were performed for the AGARD 445.6 wing at subsonic, transonic and supersonic Mach numbers. Unsteady computations were performed at various dynamic pressures to predict the flutter boundary. Results showed favorable agreement of experiment and previous numerical results. The computational methodology exhibited capabilities to predict both qualitative and quantitative features of aeroelasticity.

  11. Validation and configuration management plan for the KE basins KE-PU spreadsheet code

    International Nuclear Information System (INIS)

    Harris, R.A.

    1996-01-01

    This report provides documentation of the spreadsheet KE-PU software that is used to verify compliance with the Operational Safety Requirement and Process Standard limit on the amount of plutonium in the KE-Basin sandfilter backwash pit. Included are: A summary of the verification of the method and technique used in KE-PU that were documented elsewhere, the requirements, plans, and results of validation tests that confirm the proper functioning of the software, the procedures and approvals required to make changes to the software, and the method used to maintain configuration control over the software

  12. Validation and application of the system code TRACE for safety related investigations of innovative nuclear energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim

    2011-12-19

    The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity

  13. Development of blow down and sodium-water reaction jet analysis codes-Validation by sodium-water reaction tests (SWAT-1R)

    International Nuclear Information System (INIS)

    Hiroshi Seino; Akikazu Kurihara; Isao Ono; Koji Jitsu

    2005-01-01

    Blow down analysis code (LEAP-BLOW) and sodium-water reaction jet analysis code (LEAP-JET) have been developed in order to improve the evaluation method on sodium-water reaction event in the steam generator (SG) of a sodium cooled fast breeder reactor (FBR). The validation analyses by these two codes were carried out using the data of Sodium-Water Reaction Test (SWAT-1R). The following main results have been obtained through this validation: (1) The calculational results by LEAP-BLOW such as internal pressure and water flow rate show good agreement with the results of the SWAT- 1R test. (2) The LEAP-JET code can qualitatively simulate the behavior of sodium-water reaction. However, it is found that the code has tendency to overestimate the maximum temperature of the reaction jet. (authors)

  14. Langmuir probe-based observables for plasma-turbulence code validation and application to the TORPEX basic plasma physics experiment

    International Nuclear Information System (INIS)

    Ricci, Paolo; Theiler, C.; Fasoli, A.; Furno, I.; Labit, B.; Mueller, S. H.; Podesta, M.; Poli, F. M.

    2009-01-01

    The methodology for plasma-turbulence code validation is discussed, with focus on the quantities to use for the simulation-experiment comparison, i.e., the validation observables, and application to the TORPEX basic plasma physics experiment [A. Fasoli et al., Phys. Plasmas 13, 055902 (2006)]. The considered validation observables are deduced from Langmuir probe measurements and are ordered into a primacy hierarchy, according to the number of model assumptions and to the combinations of measurements needed to form each of them. The lowest levels of the primacy hierarchy correspond to observables that require the lowest number of model assumptions and measurement combinations, such as the statistical and spectral properties of the ion saturation current time trace, while at the highest levels, quantities such as particle transport are considered. The comparison of the observables at the lowest levels in the hierarchy is more stringent than at the highest levels. Examples of the use of the proposed observables are applied to a specific TORPEX plasma configuration characterized by interchange-driven turbulence.

  15. Development and validation of an improved version of the DART code

    International Nuclear Information System (INIS)

    Taboada, H; Moscarda, M.V.; Markiewicz, M.; Estevez, E.; Rest, J.

    2002-01-01

    ANL/USDOE and CNEA Argentina have been participating within a SisterLab Program in the area of Low Enriched Uranium Advanced Fuels since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex we developed a visual version of DART named FASTDART for silicide and U-Mo fuels that was presented at the RERTR Meeting in Las Vegas, Nevada. This paper describes several major improvements in the FASTDART code: a thermal calculation subroutine, a fuel particle size distribution subroutine and several visual interfaces for thermal output plotting and particle size input. Using the power history, coolant regime data and fuel dimensions, the new thermal subroutine is able to calculate at each time step the maximum temperature along the z-longitudinal axis as a function of plate/rod morphology (corrosion oxide, cladding, meat, aluminide particle layer, each radial shell of a central fuel particle, and particle center). Calculated temperatures at each time step are coupled to the DART calculation kernel such that swelling processes, volume phase fractions and meat thermal conductivity are calculated synergistically. The new fuel particle size-distribution subroutine is essential in order to determine the evolution of the volume fraction of reaction product. This phase degrades the heat transport by a twofold mechanism: its appearance implies a diminution of aluminium phase and its thermal conductivity is lower than those of fuel and dispersant phase. The new version includes the capability of plotting thermal data output by means of the plate/rod temperature profile at a given irradiation step, and displaying the maximum temperature evolution of each layer. A comparison between the reaction layer thickness and matrix and fuel volume fractions of several RERTR-3 experiment

  16. Development and validation of a model TRIGA Mark III reactor with code MCNP5

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Aguilar H, F.

    2015-09-01

    The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

  17. Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

    International Nuclear Information System (INIS)

    Whitesides, G.H.; Stephens, M.E.

    1984-01-01

    During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

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

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim; Manes, Jorge Perez; Imke, Uwe; Escalante, Javier Jimenez; Espinoza, Victor Sanchez, E-mail: victor.sanchez@kit.edu

    2013-10-15

    Highlights: • Simulation of BFBT turbine and pump transients at multiple scales. • CFD, sub-channel and system codes are used for the comparative study. • Heat transfer models are compared to identify difference between the code predictions. • All three scales predict results in good agreement to experiment. • Sub cooled boiling models are identified as field for future research. -- Abstract: The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in the validation and qualification of modern thermo hydraulic simulations tools at various scales. In the present paper, the prediction capabilities of four codes from three different scales – NEPTUNE{sub C}FD as fine mesh computational fluid dynamics code, SUBCHANFLOW and COBRA-TF as sub channels codes and TRACE as system code – are assessed with respect to their two-phase flow modeling capabilities. The subject of the investigations is the well-known and widely used data base provided within the NUPEC BFBT benchmark related to BWRs. Void fraction measurements simulating a turbine and a re-circulation pump trip are provided at several axial levels of the bundle. The prediction capabilities of the codes for transient conditions with various combinations of boundary conditions are validated by comparing the code predictions with the experimental data. In addition, the physical models of the different codes are described and compared to each other in order to explain the different results and to identify areas for further improvements.

  20. Validation of the thermal code of RadTherm-IR, IR-Workbench, and F-TOM

    Science.gov (United States)

    Schwenger, Frédéric; Grossmann, Peter; Malaplate, Alain

    2009-05-01

    System assessment by image simulation requires synthetic scenarios that can be viewed by the device to be simulated. In addition to physical modeling of the camera, a reliable modeling of scene elements is necessary. Software products for modeling of target data in the IR should be capable of (i) predicting surface temperatures of scene elements over a long period of time and (ii) computing sensor views of the scenario. For such applications, FGAN-FOM acquired the software products RadTherm-IR (ThermoAnalytics Inc., Calumet, USA; IR-Workbench (OKTAL-SE, Toulouse, France). Inspection of the accuracy of simulation results by validation is necessary before using these products for applications. In the first step of validation, the performance of both "thermal solvers" was determined through comparison of the computed diurnal surface temperatures of a simple object with the corresponding values from measurements. CUBI is a rather simple geometric object with well known material parameters which makes it suitable for testing and validating object models in IR. It was used in this study as a test body. Comparison of calculated and measured surface temperature values will be presented, together with the results from the FGAN-FOM thermal object code F-TOM. In the second validation step, radiances of the simulated sensor views computed by RadTherm-IR and IR-Workbench will be compared with radiances retrieved from the recorded sensor images taken by the sensor that was simulated. Strengths and weaknesses of the models RadTherm-IR, IR-Workbench and F-TOM will be discussed.

  1. Radiation sterilization of tissue allografts: Requirements for validation and routine control. A code of practice

    International Nuclear Information System (INIS)

    2007-12-01

    These recommendations for the radiation sterilization of tissue allografts adopt the principles that the International Organization for Standardization (ISO) applies to the radiation sterilization of health care products. The approach has been adapted to take into account the special features associated with human tissues and the features that distinguish them from industrially produced sterile health care products. The approach as described here is not applicable if viral contamination is identified. Thus it is emphasized that the human donors of the tissues must be medically and serologically screened. To further support this screening it is recommended that autopsy reports be reviewed if available. This adaptation of established ISO methods can thus only be applied to sterilization of tissue allografts if the radiation sterilization described here is the terminal stage of a careful, detailed, documented sequence of procedures involving: donor selection; tissue retrieval; tissue banking general procedures; specific processing procedures; labelling; and distribution. The methods proposed here for the establishment of a sterilization dose are based on statistical approaches used for the sterilization of health care products and modified appropriately for the low numbers of tissue allograft samples typically available. This code of practice will be useful to tissue banking staff, surgeons using tissues for transplantation, regulators who oversee the safety of transplantation and radiation sterilization procedures, members of tissue banking associations, health service personnel in hospitals in which tissue transplantations are performed and inter-governmental organizations involved in transplantation issues, for example the World Health Organization. This publication was discussed extensively at an international meeting in Wrexham in the United Kingdom and was approved by the Technical Advisory Committee of the relevant IAEA project, which included the Chairpersons

  2. Qualitative and quantitative validation of the SINBAD code on complex HPGe gamma-ray spectra

    Energy Technology Data Exchange (ETDEWEB)

    Rohee, E.; Coulon, R.; Normand, S.; Carrel, F. [CEA, LIST, Laboratoire Capteurs et Architectures electroniques, F-91191 Gif-sur-Yvette, (France); Dautremer, T.; Barat, E.; Montagu, T. [CEA, LIST, Laboratoire Modelisation, Simulation et Systemes, F-91191 Gif-sur-Yvette, (France); Jammes, C. [CEA/DEN/SPEx/LDCI, Centre de Cadarache, F-13109 Saint-Paul-lez-Durance, (France)

    2015-07-01

    Radionuclides identification and quantification is a serious concern for many applications as safety or security of nuclear power plant or fuel cycle facility, CBRN risk identification, environmental radioprotection and waste measurements. High resolution gamma-ray spectrometry based on HPGe detectors is a performing solution for all these topics. During last decades, a great number of software has been developed to improve gamma spectra analysis. However, some difficulties remain in the analysis when photoelectric peaks are folded together with a high ratio between theirs amplitudes, when the Compton background is much larger compared to the signal of a single peak and when spectra are composed of a great number of peaks. This study deals with the comparison between conventional methods in radionuclides identification and quantification and the code called SINBAD ('Spectrometrie par Inference Non parametrique Bayesienne Deconvolutive'). For many years, SINBAD has been developed by CEA LIST for unfolding complex spectra from HPGe detectors. Contrary to conventional methods using fitting procedures, SINBAD uses a probabilistic approach with Bayesian inference to describe spectrum data. This conventional fitting method founded for example in Genie 2000 is compared with the nonparametric SINBAD approach regarding some key figures of merit as the peak centroid evaluation (identification) and peak surface evaluation (quantification). Unfriendly cases are studied for nuclides detection with closed gamma-rays energies and high photoelectric peak intensity differences. Tests are performed with spectra from the International Atomic Energy Agency (IAEA) for gamma spectra analysis software benchmark and with spectra acquired at the laboratory. Results show that SINBAD and Genie 2000 performances are quite similar with sometimes best results for SINBAD with the important difference that to achieve same performances the nonparametric method is user-friendly compared

  3. Initial validation of 4D-model for a clinical PET scanner using the Monte Carlo code gate

    International Nuclear Information System (INIS)

    Vieira, Igor F.; Lima, Fernando R.A.; Gomes, Marcelo S.; Vieira, Jose W.; Pacheco, Ludimila M.; Chaves, Rosa M.

    2011-01-01

    Building exposure computational models (ECM) of emission tomography (PET and SPECT) currently has several dedicated computing tools based on Monte Carlo techniques (SimSET, SORTEO, SIMIND, GATE). This paper is divided into two steps: (1) using the dedicated code GATE (Geant4 Application for Tomographic Emission) to build a 4D model (where the fourth dimension is the time) of a clinical PET scanner from General Electric, GE ADVANCE, simulating the geometric and electronic structures suitable for this scanner, as well as some phenomena 4D, for example, rotating gantry; (2) the next step is to evaluate the performance of the model built here in the reproduction of test noise equivalent count rate (NEC) based on the NEMA Standards Publication NU protocols 2-2007 for this tomography. The results for steps (1) and (2) will be compared with experimental and theoretical values of the literature showing actual state of art of validation. (author)

  4. Development and validation of a catalytic recombiner model for the containment code RALOC MOD4.0

    International Nuclear Information System (INIS)

    Rohde, J.; Klein-Hebling, W.; Chakraborty, A.K.

    1997-01-01

    This paper reports on the development of a catalytic recombiner model for the containment code RALOC MOD4.0 /KLH 95, KLH 96/ and the detailed validation work, carried out at GRS. The model was qualified by using the results of medium and large scale experiments, being performed in Germany /KAN 91/. The comparison of measured data with the calculations demonstrates, that this new model is suitable for real plant applications to investigate the overall effectiveness of a catalytic recombiner system under severe accident conditions for large dry containments of German PWR design. The results of such investigations will serve as the basis to work out some guidance for the determination of the system capacity needed and an optimal positioning of such devices in containments. (author)

  5. Initial validation of 4D-model for a clinical PET scanner using the Monte Carlo code gate

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Igor F.; Lima, Fernando R.A.; Gomes, Marcelo S., E-mail: falima@cnen.gov.b [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil); Vieira, Jose W.; Pacheco, Ludimila M. [Instituto Federal de Educacao, Ciencia e Tecnologia (IFPE), Recife, PE (Brazil); Chaves, Rosa M. [Instituto de Radium e Supervoltagem Ivo Roesler, Recife, PE (Brazil)

    2011-07-01

    Building exposure computational models (ECM) of emission tomography (PET and SPECT) currently has several dedicated computing tools based on Monte Carlo techniques (SimSET, SORTEO, SIMIND, GATE). This paper is divided into two steps: (1) using the dedicated code GATE (Geant4 Application for Tomographic Emission) to build a 4D model (where the fourth dimension is the time) of a clinical PET scanner from General Electric, GE ADVANCE, simulating the geometric and electronic structures suitable for this scanner, as well as some phenomena 4D, for example, rotating gantry; (2) the next step is to evaluate the performance of the model built here in the reproduction of test noise equivalent count rate (NEC) based on the NEMA Standards Publication NU protocols 2-2007 for this tomography. The results for steps (1) and (2) will be compared with experimental and theoretical values of the literature showing actual state of art of validation. (author)

  6. The validity of using ICD-9 codes and pharmacy records to identify patients with chronic obstructive pulmonary disease

    Directory of Open Access Journals (Sweden)

    Lee Todd A

    2011-02-01

    Full Text Available Abstract Background Administrative data is often used to identify patients with chronic obstructive pulmonary disease (COPD, yet the validity of this approach is unclear. We sought to develop a predictive model utilizing administrative data to accurately identify patients with COPD. Methods Sequential logistic regression models were constructed using 9573 patients with postbronchodilator spirometry at two Veterans Affairs medical centers (2003-2007. COPD was defined as: 1 FEV1/FVC Results 4564 of 9573 patients (47.7% had an FEV1/FVC Conclusion Commonly used definitions of COPD in observational studies misclassify the majority of patients as having COPD. Using multiple diagnostic codes in combination with pharmacy data improves the ability to accurately identify patients with COPD.

  7. Enclosure environment characterization testing for the base line validation of computer fire simulation codes

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1987-03-01

    This report describes a series of fire tests conducted under the direction of Sandia National Laboratories for the US Nuclear Regulatory Commission. The primary purpose of these tests was to provide data against which to validate computer fire environment simulation models to be used in the analysis of nuclear power plant enclosure fire situations. Examples of the data gathered during three of the tests are presented, though the primary objective of this report is to provide a timely description of the test effort itself. These tests were conducted in an enclosure measuring 60x40x20 feet constructed at the Factory Mutual Research Corporation fires test facility in Rhode Island. All of the tests utilized forced ventilation conditions. The ventilation system was designed to simulate typical nuclear power plant installation practices and ventilation rates. A total of 22 tests using simple gas burner, heptane pool, methanol pool, and PMMA solid fires was conducted. Four of these tests were conducted with a full-scale control room mockup in place. Parameters varied during testing were fire intensity, enclosure ventilation rate, and fire location. Data gathered include air temperatures, air velocities, radiative and convective heat flux levels, optical smoke densities, inner and outer enclosure surface temperatures, enclosure surface heat flux levels, and gas concentrations within the enclosure in the exhaust stream

  8. BIGHORN Computational Fluid Dynamics Theory, Methodology, and Code Verification & Validation Benchmark Problems

    Energy Technology Data Exchange (ETDEWEB)

    Xia, Yidong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andrs, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This document presents the theoretical background for a hybrid finite-element / finite-volume fluid flow solver, namely BIGHORN, based on the Multiphysics Object Oriented Simulation Environment (MOOSE) computational framework developed at the Idaho National Laboratory (INL). An overview of the numerical methods used in BIGHORN are discussed and followed by a presentation of the formulation details. The document begins with the governing equations for the compressible fluid flow, with an outline of the requisite constitutive relations. A second-order finite volume method used for solving the compressible fluid flow problems is presented next. A Pressure-Corrected Implicit Continuous-fluid Eulerian (PCICE) formulation for time integration is also presented. The multi-fluid formulation is being developed. Although multi-fluid is not fully-developed, BIGHORN has been designed to handle multi-fluid problems. Due to the flexibility in the underlying MOOSE framework, BIGHORN is quite extensible, and can accommodate both multi-species and multi-phase formulations. This document also presents a suite of verification & validation benchmark test problems for BIGHORN. The intent for this suite of problems is to provide baseline comparison data that demonstrates the performance of the BIGHORN solution methods on problems that vary in complexity from laminar to turbulent flows. Wherever possible, some form of solution verification has been attempted to identify sensitivities in the solution methods, and suggest best practices when using BIGHORN.

  9. Standard problem exercise to validate criticality codes for large arrays of packages of fissile materials

    International Nuclear Information System (INIS)

    Whitesides, G.E.; Stephens, M.E.

    1986-01-01

    A study has been conducted by an Office of Economic Cooperation and Development-Committee on the Safety of Nuclear Installations (OECD-CSNI) Working Group that examined computational methods used to compute k/sub eff/ for large greater than or equal to5 3 arrays of fissile material (in which each unit is a substantial fraction of a critical mass). Five fissile materials that might typically be transported were used in the study. The ''packages'' used for this exercise were simplified to allow studies unperturbed by the variety of structural materials which would exist in an actual package. The only material present other than the fissile material was a variation in the moderator (water) surrounding the fissile material. Consistent results were obtained from calculations using several computational methods. That is, when the bias demonstrated by each method for actual critical experiments was used to ''correct'' the results obtained for systems for which there were no experimental data, there was good agreement between the methods. Two major areas of concern were raised by this exercise. First, the lack of experimental data for arrays with size greater than 5 3 limits validation for large systems. Second, there is a distinct possibility that the comingling of two shipments of unlike units could result in a reduction of the safety margins. Additional experiments and calculations will be required to satisfactorily resolve the remaining questions regarding the safe transport of large arrays of fissile materials

  10. Developments and validation of large eddy simulation of turbulent flows in an industrial code

    International Nuclear Information System (INIS)

    Ackermann, C.

    2000-01-01

    Large Eddy Simulation, where large scales of the flow are resolved and sub-grid scales are modelled, is well adapted to the study of turbulent flow, in which geometry and/or heat transfer effects lead to unsteady phenomena. To obtain an improved numerical tool, simulations of elementary test cases, Homogeneous Isotropic Turbulence and Turbulent Plane Channel, were clone on both structured and unstructured grids, before moving to more complex geometries. This allowed the influence of the different physical and numerical parameters to be studied separately. On structured grids, the different properties of the numerical methods corresponding to our problem were identified, a new sub-grid model was elaborated and several laws of the wall tested: for this discretization, our numerical tool is yet validated. On unstructured grids, the construction of numerical methods with the same properties as on the structured grids is harder, especially for the convection scheme: several numerical schemes were tested, and sub-grid models and laws of the wall were adapted to unstructured grids. Simulations of the same elementary tests were clone: the results are relatively satisfactorily, even if they are not so good as the one obtained in structured grids, most probably because the numerical methods chosen cannot perfectly isolate the effects between the convection scheme, physical modelling and the mesh chosen. This work is the first stage towards the development of a practical Large Eddy Simulation tool for unstructured grid. (author) [fr

  11. State of the art in wind turbine aerodynamics and aeroelasticity

    DEFF Research Database (Denmark)

    Hansen, Martin Otto Laver; Sørensen, Jens Nørkær; Voutsinas, S

    2006-01-01

    A comprehensive review of wind turbine aeroelasticity is given. The aerodynamic part starts with the simple aerodynamic Blade Element Momentum Method and ends with giving a review of the work done applying CFD on wind turbine rotors. In between is explained some methods of intermediate complexity...

  12. Aeroelastic System Development Using Proper Orthogonal Decomposition and Volterra Theory

    Science.gov (United States)

    Lucia, David J.; Beran, Philip S.; Silva, Walter A.

    2003-01-01

    This research combines Volterra theory and proper orthogonal decomposition (POD) into a hybrid methodology for reduced-order modeling of aeroelastic systems. The out-come of the method is a set of linear ordinary differential equations (ODEs) describing the modal amplitudes associated with both the structural modes and the POD basis functions for the uid. For this research, the structural modes are sine waves of varying frequency, and the Volterra-POD approach is applied to the fluid dynamics equations. The structural modes are treated as forcing terms which are impulsed as part of the uid model realization. Using this approach, structural and uid operators are coupled into a single aeroelastic operator. This coupling converts a free boundary uid problem into an initial value problem, while preserving the parameter (or parameters) of interest for sensitivity analysis. The approach is applied to an elastic panel in supersonic cross ow. The hybrid Volterra-POD approach provides a low-order uid model in state-space form. The linear uid model is tightly coupled with a nonlinear panel model using an implicit integration scheme. The resulting aeroelastic model provides correct limit-cycle oscillation prediction over a wide range of panel dynamic pressure values. Time integration of the reduced-order aeroelastic model is four orders of magnitude faster than the high-order solution procedure developed for this research using traditional uid and structural solvers.

  13. Aeroelastic Loads Modeling for Composite Aircraft Design Support

    NARCIS (Netherlands)

    Baluch, H.A.

    2009-01-01

    With regard to the simulation of structural vibrations and consequent aeroelastic loads in aircraft components, the use of elastic axis e.a as reference of vibrations is quite common. The e.a decouples the bending and torsion degrees of freedom (D.o.F) during the dynamic analysis. The use of the e.a

  14. Gravo-Aeroelastic Scaling for Extreme-Scale Wind Turbines

    Energy Technology Data Exchange (ETDEWEB)

    Fingersh, Lee J [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Loth, Eric [University of Virginia; Kaminski, Meghan [University of Virginia; Qin, Chao [University of Virginia; Griffith, D. Todd [Sandia National Laboratories

    2017-06-09

    A scaling methodology is described in the present paper for extreme-scale wind turbines (rated at 10 MW or more) that allow their sub-scale turbines to capture their key blade dynamics and aeroelastic deflections. For extreme-scale turbines, such deflections and dynamics can be substantial and are primarily driven by centrifugal, thrust and gravity forces as well as the net torque. Each of these are in turn a function of various wind conditions, including turbulence levels that cause shear, veer, and gust loads. The 13.2 MW rated SNL100-03 rotor design, having a blade length of 100-meters, is herein scaled to the CART3 wind turbine at NREL using 25% geometric scaling and blade mass and wind speed scaled by gravo-aeroelastic constraints. In order to mimic the ultralight structure on the advanced concept extreme-scale design the scaling results indicate that the gravo-aeroelastically scaled blades for the CART3 are be three times lighter and 25% longer than the current CART3 blades. A benefit of this scaling approach is that the scaled wind speeds needed for testing are reduced (in this case by a factor of two), allowing testing under extreme gust conditions to be much more easily achieved. Most importantly, this scaling approach can investigate extreme-scale concepts including dynamic behaviors and aeroelastic deflections (including flutter) at an extremely small fraction of the full-scale cost.

  15. Contribution to finite element modelling of airfoil aeroelastic instabilities

    Czech Academy of Sciences Publication Activity Database

    Horáček, Jaromír; Sváček, P.; Růžička, M.; Feistauer, M.

    2007-01-01

    Roč. 1, č. 1 (2007), s. 43-52 ISSN 1802-680X. [Computational Mechanics 2007. Hrad Nečtiny, 05.11.2007-07.11.2007] R&D Projects: GA MPO FT-TA/026 Institutional research plan: CEZ:AV0Z20760514 Keywords : induced vibration * aeroelasticity * nonlinear vibrations Subject RIV: BI - Acoustics

  16. 14 CFR 25.629 - Aeroelastic stability requirements.

    Science.gov (United States)

    2010-01-01

    ... stability envelopes as follows: (1) For normal conditions without failures, malfunctions, or adverse conditions, all combinations of altitudes and speeds encompassed by the VD/MD versus altitude envelope... necessary by the Administrator. (b) Aeroelastic stability envelopes. The airplane must be designed to be...

  17. Parameter estimation of an aeroelastic aircraft using neural networks

    Indian Academy of Sciences (India)

    Many proposed model reduction procedures rely on numerical techniques andaor ... The capacity to act as general function approximator presents FFNNs as an alternative tool ... This paper investigates the aerodynamic modelling of an aeroelastic aircraft using ... the learning (training) process ± backpropagation of error.

  18. Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

    Science.gov (United States)

    Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

    The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

  19. Physical analysis and modelling of aerosols transport. implementation in a finite elements code. Experimental validation in laminar and turbulent flows

    International Nuclear Information System (INIS)

    Armand, Patrick

    1995-01-01

    The aim of this work consists in the Fluid Mechanics and aerosol Physics coupling. In the first part, the order of magnitude analysis of the particle dynamics is done. This particle is embedded in a non-uniform unsteady flow. Flow approximations around the inclusion are described. Corresponding aerodynamic drag formulae are expressed. Possible situations related to the problem data are extensively listed. In the second part, one studies the turbulent particles transport. Eulerian approach which is particularly well adapted to industrial codes is preferred in comparison with the Lagrangian methods. One chooses the two-fluid formalism in which career gas-particles slip is taken into account. Turbulence modelling gets through a k-epsilon modulated by the inclusions action on the flow. The model is implemented In a finite elements code. Finally, In the third part, one validates the modelling in laminar and turbulent cases. We compare simulations to various experiments (settling battery, inertial impaction in a bend, jets loaded with glass beads particles) which are taken in the literature or done by ourselves at the laboratory. The results are very close. It is a good point when it is thought of the particles transport model and associated software future use. (author) [fr

  20. Contributions to the validation of advanced codes for accident analysis calculations with 3-dimensional neutron kinetics. STC with the Ukraine. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Rohde, U.; Khalimonchuk, V.; Kuchin, A.; Seidel, A.

    2000-10-01

    In the frame of a project of scientific-technical cooperation funded by BMBF/BMWi, the coupled code ATHLET-DYN3D has been transferred to the Scientific and Technical Centre on Nuclear and Radiation Safety Kiev (Ukraine). This program code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermohydraulics code system ATHLET. For the purpose of validation of this coupled code, a measurement data base has been generated. In the data base suitable experimental data for operational transients from NPPs are collected. The data collection and documentation was performed in accordance with a directive about requirements to measurement data for code validation, which has been elaborated within the project. The validation calculations have been performed for two selected transients. The results of these calculations were compared with measurement values from the data base. The function of the code DYN3D was expanded with a subroutine for reactivity coefficients calculation. Using this modification of the code DYN3D, investigations of reactivity contributions on different operational processes can be performed. (orig.) [de

  1. Fast Trailed Vorticity Modeling for Wind Turbine Aerodynamics and its Influence on Aeroelastic Stability

    DEFF Research Database (Denmark)

    Pirrung, Georg

    In this work, an aerodynamic model for the use in aeroelastic wind turbine codes is presented. It consists of a simplified lifting line model covering the induction due to the trailed vorticity in the near wake, a 2D shed vorticity model and a far wake model using the well known blade element...... to earlier implementations, the model has been improved in several ways: Among other things, the need for model-specific user input has been removed, the effect of downwind convection of the trailed vorticity is modeled, the near wake induction is iterated to stabilize the computations and the numerical......-of-plane vibrations agrees much better with high fidelity models. Further, the trailed vorticity effects on the aerodynamic work are found to be of the same order of magnitude as the shed vorticity effects. The trailed vorticity effects are, however, mainly important close to the tip in the investigated cases, which...

  2. Experimental Definition and Validation of Protein Coding Transcripts in Chlamydomonas reinhardtii

    Energy Technology Data Exchange (ETDEWEB)

    Kourosh Salehi-Ashtiani; Jason A. Papin

    2012-01-13

    Algal fuel sources promise unsurpassed yields in a carbon neutral manner that minimizes resource competition between agriculture and fuel crops. Many challenges must be addressed before algal biofuels can be accepted as a component of the fossil fuel replacement strategy. One significant challenge is that the cost of algal fuel production must become competitive with existing fuel alternatives. Algal biofuel production presents the opportunity to fine-tune microbial metabolic machinery for an optimal blend of biomass constituents and desired fuel molecules. Genome-scale model-driven algal metabolic design promises to facilitate both goals by directing the utilization of metabolites in the complex, interconnected metabolic networks to optimize production of the compounds of interest. Using Chlamydomonas reinhardtii as a model, we developed a systems-level methodology bridging metabolic network reconstruction with annotation and experimental verification of enzyme encoding open reading frames. We reconstructed a genome-scale metabolic network for this alga and devised a novel light-modeling approach that enables quantitative growth prediction for a given light source, resolving wavelength and photon flux. We experimentally verified transcripts accounted for in the network and physiologically validated model function through simulation and generation of new experimental growth data, providing high confidence in network contents and predictive applications. The network offers insight into algal metabolism and potential for genetic engineering and efficient light source design, a pioneering resource for studying light-driven metabolism and quantitative systems biology. Our approach to generate a predictive metabolic model integrated with cloned open reading frames, provides a cost-effective platform to generate metabolic engineering resources. While the generated resources are specific to algal systems, the approach that we have developed is not specific to algae and

  3. Validation of the coupling of mesh models to GEANT4 Monte Carlo code for simulation of internal sources of photons

    International Nuclear Information System (INIS)

    Caribe, Paulo Rauli Rafeson Vasconcelos; Cassola, Vagner Ferreira; Kramer, Richard; Khoury, Helen Jamil

    2013-01-01

    The use of three-dimensional models described by polygonal meshes in numerical dosimetry enables more accurate modeling of complex objects than the use of simple solid. The objectives of this work were validate the coupling of mesh models to the Monte Carlo code GEANT4 and evaluate the influence of the number of vertices in the simulations to obtain absorbed fractions of energy (AFEs). Validation of the coupling was performed to internal sources of photons with energies between 10 keV and 1 MeV for spherical geometries described by the GEANT4 and three-dimensional models with different number of vertices and triangular or quadrilateral faces modeled using Blender program. As a result it was found that there were no significant differences between AFEs for objects described by mesh models and objects described using solid volumes of GEANT4. Since that maintained the shape and the volume the decrease in the number of vertices to describe an object does not influence so meant dosimetric data, but significantly decreases the time required to achieve the dosimetric calculations, especially for energies less than 100 keV

  4. Validation and application of help code used for design and review of cover of low and intermediate level radioactive waste disposal in near-surface facilities

    International Nuclear Information System (INIS)

    Fan Zhiwen; Gu Cunli; Zhang Jinsheng; Liu Xiuzhen

    1996-01-01

    The authors describes validation and application of HELP code used by the United States Environmental Protective Agency for design and review of cover of low and intermediate level radioactive waste disposal in near-surface facilities. The HELP code was validated using data of field aerated moisture movement test by China Institute for Radiation Protection. The results show that simulation of HELP code is reasonable. Effects of surface layer thickness and surface treatment on moisture distribution in a cover was simulated with HELP code in the conditions of south-west China. The simulation results demonstrated that surface plantation of a cover plays very important role in moisture distribution in the cover. Special attention should be paid in cover design. In humid area, radioactive waste disposal safety should take full consideration with functions of chemical barrier. It was recommended that engineering economy should be added in future cover research so as to achieve optimization of cover design

  5. Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

    International Nuclear Information System (INIS)

    Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras

    2003-01-01

    This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

  6. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    International Nuclear Information System (INIS)

    Chang Oh

    2006-01-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 900 C and operational fuel temperatures above 1250 C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR's higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gases (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  7. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  8. Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

    Energy Technology Data Exchange (ETDEWEB)

    Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

    2012-07-01

    The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

  9. Software requirements, design, and verification and validation for the FEHM application - a finite-element heat- and mass-transfer code

    International Nuclear Information System (INIS)

    Dash, Z.V.; Robinson, B.A.; Zyvoloski, G.A.

    1997-07-01

    The requirements, design, and verification and validation of the software used in the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media, are described. The test of the DOE Code Comparison Project, Problem Five, Case A, which verifies that FEHM has correctly implemented heat and mass transfer and phase partitioning, is also covered

  10. 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

  11. Test and validation of CFD codes for the simulation of accident-typical phenomena in the reactor containment

    International Nuclear Information System (INIS)

    Schramm, Berthold; Stewering, Joern; Sonnenkalb, Martin

    2014-03-01

    CFD (Computational Fluid Dynamic) simulation techniques have a growing relevance for the simulation and assessment of accidents in nuclear reactor containments. Some fluid dynamic problems like the calculation of the flow resistances in a complex geometry, turbulence calculations or the calculation of deflagrations could only be solved exactly for very simple cases. These fluid dynamic problems could not be represented by lumped parameter models and must be approximated numerically. Therefore CFD techniques are discussed by a growing international community in conferences like the CFD4NRS-conference. Also the number of articles with a CFD topic is increasing in professional journals like Nuclear Engineering and Design. CFD tools like GASFLOW or GOTHIC are already in use in European nuclear site licensing processes for future nuclear power plants like EPR or AP1000 and the results of these CFD tools are accepted by the authorities. For these reasons it seems to be necessary to build up national competences in the field of CFD techniques and it is important to validate and assess the existing CFD tools. GRS continues the work for the validation and assessment of CFD codes for the simulation of accident scenarios in a nuclear reactor containment within the framework of the BMWi sponsored project RS1500. The focus of this report is on the following topics: - Further validation of condensation models from GRS, FZJ and ANSYS and development of a new condensate model. - Validation of a new turbulence model which was developed by the University of Stuttgart in cooperation with ANSYS. - The formation and dissolution of light gas stratifications are analyzed by large scale experiments. These experiments were simulated by GRS. - The AREVA correlations for hydrogen recombiners (PARs) could be improved by GRS after the analysis of experimental data. Relevant experiments were simulated with this improved recombiner correlation. - Analyses on the simulation of H_2 deflagration

  12. Large-scale, multi-compartment tests in PANDA for LWR-containment analysis and code validation

    International Nuclear Information System (INIS)

    Paladino, Domenico; Auban, Olivier; Zboray, Robert

    2006-01-01

    The large-scale thermal-hydraulic PANDA facility has been used for the last years for investigating passive decay heat removal systems and related containment phenomena relevant for next-generation and current light water reactors. As part of the 5. EURATOM framework program project TEMPEST, a series of tests was performed in PANDA to experimentally investigate the distribution of hydrogen inside the containment and its effect on the performance of the Passive Containment Cooling System (PCCS) designed for the Economic Simplified Boiling Water Reactor (ESBWR). In a postulated severe accident, a large amount of hydrogen could be released in the Reactor Pressure Vessel (RPV) as a consequence of the cladding Metal- Water (M-W) reaction and discharged together with steam to the Drywell (DW) compartment. In PANDA tests, hydrogen was simulated by using helium. This paper illustrates the results of a TEMPEST test performed in PANDA and named as Test T1.2. In Test T1.2, the gas stratification (steam-helium) patterns forming in the large-scale multi-compartment PANDA DW, and the effect of non-condensable gas (helium) on the overall behaviour of the PCCS were identified. Gas mixing and stratification in a large-scale multi-compartment system are currently being further investigated in PANDA in the frame of the OECD project SETH. The testing philosophy in this new PANDA program is to produce data for code validation in relation to specific phenomena, such as: gas stratification in the containment, gas transport between containment compartments, wall condensation, etc. These types of phenomena are driven by buoyant high-momentum injections (jets) and/or low momentum injection (plumes), depending on the transient scenario. In this context, the new SETH tests in PANDA are particularly valuable to produce an experimental database for code assessment. This paper also presents an overview of the PANDA SETH tests and the major improvements in instrumentation carried out in the PANDA

  13. Analysis of the impact of correlated benchmark experiments on the validation of codes for criticality safety analysis

    International Nuclear Information System (INIS)

    Bock, M.; Stuke, M.; Behler, M.

    2013-01-01

    The validation of a code for criticality safety analysis requires the recalculation of benchmark experiments. The selected benchmark experiments are chosen such that they have properties similar to the application case that has to be assessed. A common source of benchmark experiments is the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments' (ICSBEP Handbook) compiled by the 'International Criticality Safety Benchmark Evaluation Project' (ICSBEP). In order to take full advantage of the information provided by the individual benchmark descriptions for the application case, the recommended procedure is to perform an uncertainty analysis. The latter is based on the uncertainties of experimental results included in most of the benchmark descriptions. They can be performed by means of the Monte Carlo sampling technique. The consideration of uncertainties is also being introduced in the supplementary sheet of DIN 25478 'Application of computer codes in the assessment of criticality safety'. However, for a correct treatment of uncertainties taking into account the individual uncertainties of the benchmark experiments is insufficient. In addition, correlations between benchmark experiments have to be handled correctly. For example, these correlations can arise due to different cases of a benchmark experiment sharing the same components like fuel pins or fissile solutions. Thus, manufacturing tolerances of these components (e.g. diameter of the fuel pellets) have to be considered in a consistent manner in all cases of the benchmark experiment. At the 2012 meeting of the Expert Group on 'Uncertainty Analysis for Criticality Safety Assessment' (UACSA) of the OECD/NEA a benchmark proposal was outlined that aimed for the determination of the impact on benchmark correlations on the estimation of the computational bias of the neutron multiplication factor (k eff ). The analysis presented here is based on this proposal. (orig.)

  14. Recent advance in nonlinear aeroelastic analysis and control of the aircraft

    Directory of Open Access Journals (Sweden)

    Xiang Jinwu

    2014-02-01

    Full Text Available A review on the recent advance in nonlinear aeroelasticity of the aircraft is presented in this paper. The nonlinear aeroelastic problems are divided into three types based on different research objects, namely the two dimensional airfoil, the wing, and the full aircraft. Different nonlinearities encountered in aeroelastic systems are discussed firstly, where the emphases is placed on new nonlinear model to describe tested nonlinear relationship. Research techniques, especially new theoretical methods and aeroelastic flutter control methods are investigated in detail. The route to chaos and the cause of chaotic motion of two-dimensional aeroelastic system are summarized. Various structural modeling methods for the high-aspect-ratio wing with geometric nonlinearity are discussed. Accordingly, aerodynamic modeling approaches have been developed for the aeroelastic modeling of nonlinear high-aspect-ratio wings. Nonlinear aeroelasticity about high-altitude long-endurance (HALE and fight aircrafts are studied separately. Finally, conclusions and the challenges of the development in nonlinear aeroelasticity are concluded. Nonlinear aeroelastic problems of morphing wing, energy harvesting, and flapping aircrafts are proposed as new directions in the future.

  15. Validation of One-Dimensional Module of MARS-KS1.2 Computer Code By Comparison with the RELAP5/MOD3.3/patch3 Developmental Assessment Results

    International Nuclear Information System (INIS)

    Bae, S. W.; Chung, B. D.

    2010-07-01

    This report records the results of the code validation for the one-dimensional module of the MARS-KS thermal hydraulics analysis code by means of result-comparison with the RELAP5/MOD3.3 computer code. For the validation calculations, simulations of the RELAP5 Code Developmental Assessment Problem, which consists of 22 simulation problems in 3 categories, have been selected. The results of the 3 categories of simulations demonstrate that the one-dimensional module of the MARS code and the RELAP5/MOD3.3 code are essentially the same code. This is expected as the two codes have basically the same set of field equations, constitutive equations and main thermal hydraulic models. The result suggests that the high level of code validity of the RELAP5/MOD3.3 can be directly applied to the MARS one-dimensional module

  16. Validation of a modified PENELOPE Monte Carlo code for applications in digital and dual-energy mammography

    Science.gov (United States)

    Del Lama, L. S.; Cunha, D. M.; Poletti, M. E.

    2017-08-01

    The presence and morphology of microcalcification clusters are the main point to provide early indications of breast carcinomas. However, the visualization of those structures may be jeopardized due to overlapping tissues even for digital mammography systems. Although digital mammography is the current standard for breast cancer diagnosis, further improvements should be achieved in order to address some of those physical limitations. One possible solution for such issues is the application of the dual-energy technique (DE), which is able to highlight specific lesions or cancel out the tissue background. In this sense, this work aimed to evaluate several quantities of interest in radiation applications and compare those values with works present in the literature to validate a modified PENELOPE code for digital mammography applications. For instance, the scatter-to-primary ratio (SPR), the scatter fraction (SF) and the normalized mean glandular dose (DgN) were evaluated by simulations and the resulting values were compared to those found in earlier studies. Our results present a good correlation for the evaluated quantities, showing agreement equal or better than 5% for the scatter and dosimetric-related quantities when compared to the literature. Finally, a DE imaging chain was simulated and the visualization of microcalcifications was investigated.

  17. Level-Set Topology Optimization with Aeroelastic Constraints

    Science.gov (United States)

    Dunning, Peter D.; Stanford, Bret K.; Kim, H. Alicia

    2015-01-01

    Level-set topology optimization is used to design a wing considering skin buckling under static aeroelastic trim loading, as well as dynamic aeroelastic stability (flutter). The level-set function is defined over the entire 3D volume of a transport aircraft wing box. Therefore, the approach is not limited by any predefined structure and can explore novel configurations. The Sequential Linear Programming (SLP) level-set method is used to solve the constrained optimization problems. The proposed method is demonstrated using three problems with mass, linear buckling and flutter objective and/or constraints. A constraint aggregation method is used to handle multiple buckling constraints in the wing skins. A continuous flutter constraint formulation is used to handle difficulties arising from discontinuities in the design space caused by a switching of the critical flutter mode.

  18. Quasi-Static Condensation of Aeroelastic Suspension Bridge Model

    DEFF Research Database (Denmark)

    Møller, Randi N.; Krenk, Steen; N. Svendsen, Martin

    2017-01-01

    For long span bridges the wind-induced dynamic response is a design driving factor and therefore continuously a subject for detailed analysis. Traditionally both buffeting and stability calculations have been considered in the frequency domain. However, this yields alimitation in accounting...... for turbulence when considering the stability limit and further it is not possible to account for non-linear effects. These limitations suggest to do simulations of the aeroelastic response of long span bridges in the time domain. For this it is of interest to have an efficient model while still maintaining...... sufficient accuracy. This contribution is on quasi-static reduction of an aeroelastic finite element model of a 3000m suspension bridge proposed for crossing Sulafjorden in Norway. The model is intended for stability limit calculation where the representation of higher modes is of less importance...

  19. Constructal Theory and Aeroelastic Design of Flexible Flying Wing Aircraft

    Directory of Open Access Journals (Sweden)

    Pezhman Mardanpour

    2017-07-01

    Full Text Available The aeroelastic behavior of high-aspect-ratio very flexible flying wing is highly affected by the geometric nonlinearities of the aircraft structure. This paper reviews the findings on how these nonlinearities influence the structural and flight dynamics, and it shows that the aeroelastic flight envelope could significantly be extended with proper choices of design parameters such as engine placement. Moreover, in order to investigate the physics behind the effects of design parameters, constructal theory of design is reviewed. The constructal theory advances the philosophy of design as science, it states that the better structural design emerges when stress flow strangulation is avoided. Furthermore, it shows that airplanes, through their evolution, have obeyed theoretical allometric rules that unite their designs.

  20. CFD and Aeroelastic Analysis of the MEXICO Wind Turbine

    International Nuclear Information System (INIS)

    Carrión, M; Woodgate, M; Steijl, R; Barakos, G; Gómez-Iradi, S; Munduate, X

    2014-01-01

    This paper presents an aerodynamic and aeroelastic analysis of the MEXICO wind turbine, using the compressible HMB solver of Liverpool. The aeroelasticity of the blade, as well as the effect of a low-Mach scheme were studied for the zero-yaw 15m/s wind case and steady- state computations. The wake developed behind the rotor was also extracted and compared with the experimental data, using the compressible solver and a low-Mach scheme. It was found that the loads were not sensitive to the Mach number effects, although the low-Mach scheme improved the wake predictions. The sensitivity of the results to the blade structural properties was also highlighted

  1. Aeroelastic analysis of an adaptive trailing edge with a smart elastic skin

    Science.gov (United States)

    Arena, Maurizio; Pecora, Rosario; Amoroso, Francesco; Noviello, Maria Chiara; Rea, Francesco; Concilio, Antonio

    2017-09-01

    Nowadays, the design choices of the new generation aircraft are moving towards the research and development of innovative technologies, aimed at improving performance as well as to minimize the environmental impact. In the current "greening" context, the morphing structures represent a very attractive answer to such requirements: both aerodynamic and structural advantages are ensured in several flight conditions, safeguarding the fuel consumption at the same time. An aeronautical intelligent system is therefore the outcome of combining complex smart materials and structures, assuring the best functionality level in the flight envelope. The Adaptive Trailing Edge Device (ATED) is a sub-project inside SARISTU (Smart Intelligent Aircraft Structures), an L2 level project of the 7th EU Framework programme coordinated by Airbus, aimed at developing technologies for realizing a morphing wing extremity addressed to improve the general aircraft performance and to reduce the fuel burning up to 5%. This specific study, divided into design, manufacturing and testing phases, involved universities, research centers and leading industries of the European consortium. The paper deals with the aeroelastic impact assessment of a full-scale morphing wing trailing edge on a Large Aeroplanes category aircraft. The FE (Finite Element) model of the technology demonstrator, located in the aileron region and manufactured within the project, was referenced to for the extrapolation of the structural properties of the whole adaptive trailing edge device placed in its actual location in the outer wing. The input FE models were processed within MSC-Nastran® environment to estimate stiffness and inertial distributions suitable to construct the aeroelastic stick-beam mock-up of the reference structure. Afterwards, a flutter analysis in simulated operative condition, have been carried out by means of Sandy®, an in-house code, according to meet the safety requirements imposed by the applicable

  2. Validation of fuel performance codes at the NRI Rez plc for Temelin and Dukovany NPPs fuel safety evaluations and operation support

    International Nuclear Information System (INIS)

    Valach, M.; Hejna, J.; Zymak, J.

    2003-05-01

    The report summarises the first phase of the FUMEX II related work performed in the period September 2002 - May 2003. An inventory of the PIN and FRAS codes family used and developed during previous years was made in light of their applicability (validity) in the domain of high burn-up and FUMEX II Project Experimental database. KOLA data were chosen as appropriate for the first step of both codes fixing (both tuned for VVER fuel originally). The modern requirements, expressed by adaptation of the UO 2 conductivity degradation from OECD HRP, RIM and FGR (athermal) modelling implementation into the PIN code and a diffusion FGR model development planned for embedding, into this code allow us to reasonably shadow or keep tight contact with top quality models as TRANSURANUS, COPERNIC, CYRANO, FEMAXI, FRAPCON3 or ENIGMA. Testing and validation runs with prepared input KOLA deck were made. FUMEX II exercise propose LOCA and RIA like transients, so we started development of those two codes coupling - denominated as PIN2FRAS code. Principles of the interface were tested, benchmarking on tentative RIA pulses on highly burned KOLA fuel are presented as the first achievement from our work. (author)

  3. Test and validation of the iterative code for the neutrons spectrometry and dosimetry: NSDUAZ; Prueba y validacion del codigo iterativo para la espectrometria y dosimetria de neutrones: NSDUAZ

    Energy Technology Data Exchange (ETDEWEB)

    Reyes H, A.; Ortiz R, J. M.; Reyes A, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R., E-mail: alfredo_reyesh@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Lopez Velarde 801, Col. Centro, 98000 Zacatecas (Mexico)

    2014-08-15

    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 {sup 6}LiI(Eu) with 7 polyethylene spheres with diameter of 0, 2, 3, 5, 8, 10 and 12. The count rates measured with two neutron sources: {sup 252}Cf and {sup 239}PuBe 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)

  4. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104)

    International Nuclear Information System (INIS)

    Kress, T.S.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time

  5. Aeroelastic stability analysis of a Darrieus wind turbine

    Science.gov (United States)

    Popelka, D.

    1982-02-01

    An aeroelastic stability analysis was developed for predicting flutter instabilities on vertical axis wind turbines. The analytical model and mathematical formulation of the problem are described as well as the physical mechanism that creates flutter in Darrieus turbines. Theoretical results are compared with measured experimental data from flutter tests of the Sandia 2 Meter turbine. Based on this comparison, the analysis appears to be an adequate design evaluation tool.

  6. Unsteady airfoil flows with application to aeroelastic stability

    Energy Technology Data Exchange (ETDEWEB)

    Johansen, Jeppe

    1999-09-01

    The present report describes numerical investigation of two-dimensional unsteady airfoil flows with application to aeroelastic stability. The report is divided in two parts. Part A describes the purely aerodynamic part, while Part B includes the aeroelastic part. In Part A a transition prediction algorithm based on a simplified version of the e{sup n} method is proposed. Laminar Boundary Layer instability data are stored in a database from which stability characteristics can be extracted by interpolation. Input to the database are laminar integral boundary layer parameters. These are computed from an integral boundary layer formulation coupled to a Navier-Stokes flow solver. Five different airfoils are considered at fixed angle of attack, and the flow is computed assuming both fully turbulent and transitional flow and compared with experimental data. Results indicate that using a transition model the drag prediction is improved considerably. Also the lift is slightly improved. At high angles of attack transition will affect leading edge separation which again will affect the overall vortex shedding. If the transition point is not properly predicted this will affect the whole hysteresis curve. The transition model developed in the present work showed more stable predictions compared to the empirical transition model. In Part B a simple three degrees-of-freedom (DOF) structural dynamics model is developed and coupled to the aerodynamics models from Part A. A 2nd order accurate time integration scheme is used to solve the equations of motion. Two airfoils are investigated. The aeroelastic models predict stable conditions well at low angle of attack. But at high angles of attack, and where unstable behaviour is expected, only the Navier-Stokes solver predict correct aeroelastic response. The semi-empirical dynamic stall model does not predict vortex shedding and moment correctly leading to an erroneous aerodynamic damping. (au) 5 tabs.; 55 ills., 52 refs.

  7. Aeroelastic stability analysis of a Darrieus wind turbine

    Energy Technology Data Exchange (ETDEWEB)

    Popelka, D.

    1982-02-01

    An aeroelastic stability analysis has been developed for predicting flutter instabilities on vertical axis wind turbines. The analytical model and mathematical formulation of the problem are described as well as the physical mechanism that creates flutter in Darrieus turbines. Theoretical results are compared with measured experimental data from flutter tests of the Sandia 2 Meter turbine. Based on this comparison, the analysis appears to be an adequate design evaluation tool.

  8. Refining the accuracy of validated target identification through coding variant fine-mapping in type 2 diabetes

    DEFF Research Database (Denmark)

    Mahajan, Anubha

    2018-01-01

    , compelling evidence for coding variant causality was obtained for only 16 signals. At 13 others, the associated coding variants clearly represent 'false leads' with potential to generate erroneous mechanistic inference. Coding variant associations offer a direct route to biological insight for complex...

  9. Identifying Psoriasis and Psoriatic Arthritis Patients in Retrospective Databases When Diagnosis Codes Are Not Available: A Validation Study Comparing Medication/Prescriber Visit-Based Algorithms with Diagnosis Codes.

    Science.gov (United States)

    Dobson-Belaire, Wendy; Goodfield, Jason; Borrelli, Richard; Liu, Fei Fei; Khan, Zeba M

    2018-01-01

    Using diagnosis code-based algorithms is the primary method of identifying patient cohorts for retrospective studies; nevertheless, many databases lack reliable diagnosis code information. To develop precise algorithms based on medication claims/prescriber visits (MCs/PVs) to identify psoriasis (PsO) patients and psoriatic patients with arthritic conditions (PsO-AC), a proxy for psoriatic arthritis, in Canadian databases lacking diagnosis codes. Algorithms were developed using medications with narrow indication profiles in combination with prescriber specialty to define PsO and PsO-AC. For a 3-year study period from July 1, 2009, algorithms were validated using the PharMetrics Plus database, which contains both adjudicated medication claims and diagnosis codes. Positive predictive value (PPV), negative predictive value (NPV), sensitivity, and specificity of the developed algorithms were assessed using diagnosis code as the reference standard. Chosen algorithms were then applied to Canadian drug databases to profile the algorithm-identified PsO and PsO-AC cohorts. In the selected database, 183,328 patients were identified for validation. The highest PPVs for PsO (85%) and PsO-AC (65%) occurred when a predictive algorithm of two or more MCs/PVs was compared with the reference standard of one or more diagnosis codes. NPV and specificity were high (99%-100%), whereas sensitivity was low (≤30%). Reducing the number of MCs/PVs or increasing diagnosis claims decreased the algorithms' PPVs. We have developed an MC/PV-based algorithm to identify PsO patients with a high degree of accuracy, but accuracy for PsO-AC requires further investigation. Such methods allow researchers to conduct retrospective studies in databases in which diagnosis codes are absent. Copyright © 2018 International Society for Pharmacoeconomics and Outcomes Research (ISPOR). Published by Elsevier Inc. All rights reserved.

  10. Continuous-time state-space unsteady aerodynamic modelling for efficient aeroelastic load analysis

    NARCIS (Netherlands)

    Werter, N.P.M.; De Breuker, R.; Abdalla, M.M.

    2015-01-01

    Over the years, wings have become lighter and more flexible, making them more prone to aeroelastic effects. Thus, aeroelasticity in design becomes more important. In order to determine the response of an aircraft to, for example, a gust, an unsteady aerodynamic model is required to determine the

  11. Nonlinear Aeroelastic Study of Stall Induced Oscillation in a Symmetric Airfoil

    NARCIS (Netherlands)

    Sarkar, S.; Bijl, H.

    2006-01-01

    In this paper the aeroelastic stability of a wind turbine rotor in the dynamic stall regime is investigated. Increased flexibility of modern turbine blades makes them more susceptible to aeroelastic instabilities. Complex oscillation modes like flap/lead-lag are of particular concern, which give way

  12. Development of an automatic validation system for simulation codes of the fusion research; Entwicklung eines automatischen Validierungssystems fuer Simulationscodes der Fusionsforschung

    Energy Technology Data Exchange (ETDEWEB)

    Galonska, Andreas

    2010-03-15

    In the present master thesis the development oa an automatic validation system for the simulation code ERO is documented. This 3D Monte-carlo code models the transport of impurities as well as plasma-wall interaction processes and has great importance for the fusion research. The validation system is based on JuBE (Julich Benchmarking Environment), the flexibility of which allows a slight extension of the system to other codes, for instance such, which are operated in the framework of the EU Task Force ITM (Integrated Tokamak Modelling). The chosen solution - JuBE and a special program for the ''intellectual'' comparison of actual and reference-edition data of ERO is described and founded. The use of this program and the configuration of JuBE are detailedly described. Simulations to different plasma experiments, which serve as reference cases for the automatic validation, are explained. The working of the system is illustrated by the description of a test case. This treats the failure localization and improvement in the parallelization of an important ERO module (tracking of physically eroded particle). It is demonstrated, how the system reacts in an erroneous validation and the subsequently performed error correction leads to a positive result. Finally a speed-up curve of the parallelization is established by means of the output data of JuBE.

  13. Three-dimensional all-speed CFD code for safety analysis of nuclear reactor containment: Status of GASFLOW parallelization, model development, validation and application

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Jianjun, E-mail: jianjun.xiao@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, John R., E-mail: jack_travis@comcast.com [Engineering and Scientific Software Inc., 3010 Old Pecos Trail, Santa Fe, NM 87505 (United States); Royl, Peter, E-mail: peter.royl@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Necker, Gottfried, E-mail: gottfried.necker@partner.kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Svishchev, Anatoly, E-mail: anatoly.svishchev@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Jordan, Thomas, E-mail: thomas.jordan@kit.edu [Institute of Nuclear and Energy Technologies, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2016-05-15

    Highlights: • 3-D scalable semi-implicit pressure-based CFD code for containment safety analysis. • Robust solution algorithm valid for all-speed flows. • Well validated and widely used CFD code for hydrogen safety analysis. • Code applied in various types of nuclear reactor containments. • Parallelization enables high-fidelity models in large scale containment simulations. - Abstract: GASFLOW is a three dimensional semi-implicit all-speed CFD code which can be used to predict fluid dynamics, chemical kinetics, heat and mass transfer, aerosol transportation and other related phenomena involved in postulated accidents in nuclear reactor containments. The main purpose of the paper is to give a brief review on recent GASFLOW code development, validations and applications in the field of nuclear safety. GASFLOW code has been well validated by international experimental benchmarks, and has been widely applied to hydrogen safety analysis in various types of nuclear power plants in European and Asian countries, which have been summarized in this paper. Furthermore, four benchmark tests of a lid-driven cavity flow, low Mach number jet flow, 1-D shock tube and supersonic flow over a forward-facing step are presented in order to demonstrate the accuracy and wide-ranging capability of ICE’d ALE solution algorithm for all-speed flows. GASFLOW has been successfully parallelized using the paradigms of Message Passing Interface (MPI) and domain decomposition. The parallel version, GASFLOW-MPI, adds great value to large scale containment simulations by enabling high-fidelity models, including more geometric details and more complex physics. It will be helpful for the nuclear safety engineers to better understand the hydrogen safety related physical phenomena during the severe accident, to optimize the design of the hydrogen risk mitigation systems and to fulfill the licensing requirements by the nuclear regulatory authorities. GASFLOW-MPI is targeting a high

  14. Validation study of SRAC2006 code system based on evaluated nuclear data libraries for TRIGA calculations by benchmarking integral parameters of TRX and BAPL lattices of thermal reactors

    International Nuclear Information System (INIS)

    Khan, M.J.H.; Sarker, M.M.; Islam, S.M.A.

    2013-01-01

    Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations

  15. Low-order aeroelastic models of wind turbines for controller design

    DEFF Research Database (Denmark)

    Sønderby, Ivan Bergquist

    Wind turbine controllers are used to optimize the performance of wind turbines such as to reduce power variations and fatigue and extreme loads on wind turbine components. Accurate tuning and design of modern controllers must be done using low-order models that accurately captures the aeroelastic...... response of the wind turbine. The purpose of this thesis is to investigate the necessary model complexity required in aeroelastic models used for controller design and to analyze and propose methods to design low-order aeroelastic wind turbine models that are suited for model-based control design....... The thesis contains a characterization of the dynamics that influence the open-loop aeroelastic frequency response of a modern wind turbine, based on a high-order aeroelastic wind turbine model. One main finding is that the transfer function from collective pitch to generator speed is affected by two low...

  16. Aeroelastic Tailoring of Transport Aircraft Wings: State-of-the-Art and Potential Enabling Technologies

    Science.gov (United States)

    Jutte, Christine; Stanford, Bret K.

    2014-01-01

    This paper provides a brief overview of the state-of-the-art for aeroelastic tailoring of subsonic transport aircraft and offers additional resources on related research efforts. Emphasis is placed on aircraft having straight or aft swept wings. The literature covers computational synthesis tools developed for aeroelastic tailoring and numerous design studies focused on discovering new methods for passive aeroelastic control. Several new structural and material technologies are presented as potential enablers of aeroelastic tailoring, including selectively reinforced materials, functionally graded materials, fiber tow steered composite laminates, and various nonconventional structural designs. In addition, smart materials and structures whose properties or configurations change in response to external stimuli are presented as potential active approaches to aeroelastic tailoring.

  17. Implementation, verification, and validation of the FPIN2 metal fuel pin mechanics model in the SASSYS/SAS4A LMR transient analysis codes

    International Nuclear Information System (INIS)

    Sofu, T.; Kramer, J.M.

    1994-01-01

    The metal fuel version of the FPIN2 code which provides a validated pin mechanics model is coupled with SASSYS/SAS4A Version 3.0 for single pin calculations. In this implementation, SASSY/SAS4A provides pin temperatures, and FPIN2 performs analysis of pin deformation and predicts the time and location of cladding failure. FPIN2 results are also used for the estimates of axial expansion of fuel and associated reactivity effects. The revalidation of the integrated SAS-FPIN2 code system is performed using TREAT tests

  18. Wing-Body Aeroelasticity Using Finite-Difference Fluid/Finite-Element Structural Equations on Parallel Computers

    Science.gov (United States)

    Byun, Chansup; Guruswamy, Guru P.; Kutler, Paul (Technical Monitor)

    1994-01-01

    In recent years significant advances have been made for parallel computers in both hardware and software. Now parallel computers have become viable tools in computational mechanics. Many application codes developed on conventional computers have been modified to benefit from parallel computers. Significant speedups in some areas have been achieved by parallel computations. For single-discipline use of both fluid dynamics and structural dynamics, computations have been made on wing-body configurations using parallel computers. However, only a limited amount of work has been completed in combining these two disciplines for multidisciplinary applications. The prime reason is the increased level of complication associated with a multidisciplinary approach. In this work, procedures to compute aeroelasticity on parallel computers using direct coupling of fluid and structural equations will be investigated for wing-body configurations. The parallel computer selected for computations is an Intel iPSC/860 computer which is a distributed-memory, multiple-instruction, multiple data (MIMD) computer with 128 processors. In this study, the computational efficiency issues of parallel integration of both fluid and structural equations will be investigated in detail. The fluid and structural domains will be modeled using finite-difference and finite-element approaches, respectively. Results from the parallel computer will be compared with those from the conventional computers using a single processor. This study will provide an efficient computational tool for the aeroelastic analysis of wing-body structures on MIMD type parallel computers.

  19. STARS: An Integrated, Multidisciplinary, Finite-Element, Structural, Fluids, Aeroelastic, and Aeroservoelastic Analysis Computer Program

    Science.gov (United States)

    Gupta, K. K.

    1997-01-01

    A multidisciplinary, finite element-based, highly graphics-oriented, linear and nonlinear analysis capability that includes such disciplines as structures, heat transfer, linear aerodynamics, computational fluid dynamics, and controls engineering has been achieved by integrating several new modules in the original STARS (STructural Analysis RoutineS) computer program. Each individual analysis module is general-purpose in nature and is effectively integrated to yield aeroelastic and aeroservoelastic solutions of complex engineering problems. Examples of advanced NASA Dryden Flight Research Center projects analyzed by the code in recent years include the X-29A, F-18 High Alpha Research Vehicle/Thrust Vectoring Control System, B-52/Pegasus Generic Hypersonics, National AeroSpace Plane (NASP), SR-71/Hypersonic Launch Vehicle, and High Speed Civil Transport (HSCT) projects. Extensive graphics capabilities exist for convenient model development and postprocessing of analysis results. The program is written in modular form in standard FORTRAN language to run on a variety of computers, such as the IBM RISC/6000, SGI, DEC, Cray, and personal computer; associated graphics codes use OpenGL and IBM/graPHIGS language for color depiction. This program is available from COSMIC, the NASA agency for distribution of computer programs.

  20. Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2010-05-15

    Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)

  1. Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment; Validierung des ATHLET-Codes 2.1A anhand des Einzeleffekt-Tests ECTHOR

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Andreas; Sarkadi, Peter; Schaffrath, Andreas [TUEV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

    2010-06-15

    Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)

  2. Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

    International Nuclear Information System (INIS)

    Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

    2015-01-01

    Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

  3. Measurements in Regions of Shock Wave/Turbulent Boundary Layer Interaction from Mach 3 to 10 for Open and Blind Code Evaluation/Validation

    Science.gov (United States)

    2013-03-01

    34Blind" Code Evaluation/Validation Michael S. Holden, Timothy P. Wadhams, Matthew G. MacLean, Aaron Dufrene CUBRC , Inc March 2013 Final...298 Back (Rev. 8/98) *Fellow, AIAA, Vice President-Hypersonics, CUBRC , 4455 Genesee Street, Buffalo, NY 14225 ** Member, AIAA, Project Engineers... CUBRC , 4455 Genesee Street, Buffalo, NY 14225 This work was supported by AFOSR Grant No. FA9550-11-1-0290 MEASUREMENTS IN REGIONS OF SHOCK WAVE

  4. The Physician Recommendation Coding System (PhyReCS): A Reliable and Valid Method to Quantify the Strength of Physician Recommendations During Clinical Encounters.

    Science.gov (United States)

    Scherr, Karen A; Fagerlin, Angela; Williamson, Lillie D; Davis, J Kelly; Fridman, Ilona; Atyeo, Natalie; Ubel, Peter A

    2017-01-01

    Physicians' recommendations affect patients' treatment choices. However, most research relies on physicians' or patients' retrospective reports of recommendations, which offer a limited perspective and have limitations such as recall bias. To develop a reliable and valid method to measure the strength of physician recommendations using direct observation of clinical encounters. Clinical encounters (n = 257) were recorded as part of a larger study of prostate cancer decision making. We used an iterative process to create the 5-point Physician Recommendation Coding System (PhyReCS). To determine reliability, research assistants double-coded 50 transcripts. To establish content validity, we used 1-way analyses of variance to determine whether relative treatment recommendation scores differed as a function of which treatment patients received. To establish concurrent validity, we examined whether patients' perceived treatment recommendations matched our coded recommendations. The PhyReCS was highly reliable (Krippendorf's alpha = 0.89, 95% CI [0.86, 0.91]). The average relative treatment recommendation score for each treatment was higher for individuals who received that particular treatment. For example, the average relative surgery recommendation score was higher for individuals who received surgery versus radiation (mean difference = 0.98, SE = 0.18, P recommendations matched coded recommendations 81% of the time. The PhyReCS is a reliable and valid way to capture the strength of physician recommendations. We believe that the PhyReCS would be helpful for other researchers who wish to study physician recommendations, an important part of patient decision making. © The Author(s) 2016.

  5. Aeroelastic Dynamics Simulation of Two BaffleBased Connected Shells

    Directory of Open Access Journals (Sweden)

    G. A. Shcheglov

    2015-01-01

    Full Text Available The present work is an extention study of aeroelastic vibrations of thin-walled structures with a spatial subsonic flow. An original algorithm for solving complex conjugated aeroelasticity problem, allowing to carry out direct numerical simulation of structural oscillations in the spatial flow of an incompressible medium are developed and tested. On the basis of this simulation study of the spectrum comes the driving forces acting on the flow in a spatial component elastic structure mounted on an impenetrable screen.Currently, updating the mathematical models of unsteady loads that act