WorldWideScience

Sample records for reactor model geometry

  1. Control rod interaction models for use in 2D and 3D reactor geometries

    International Nuclear Information System (INIS)

    Bannerman, R.C.

    1985-11-01

    Control rod interaction models are developed for use in two-dimensional and three-dimensional reactor geometries. These models allow the total worth of any combination of control rods inserted to be determined from the individual worths in conjunction with an algorithm representing interaction effects between them. The validity of the assumptions is demonstrated for fast and thermal systems showing modelling errors of 1#percent# or less in inserted control rod worths. The models are ideally suited for most reactor systems. (UK)

  2. Computational geometry for reactor applications

    International Nuclear Information System (INIS)

    Brown, F.B.; Bischoff, F.G.

    1988-01-01

    Monte Carlo codes for simulating particle transport involve three basic computational sections: a geometry package for locating particles and computing distances to regional boundaries, a physics package for analyzing interactions between particles and problem materials, and an editing package for determining event statistics and overall results. This paper describes the computational geometry methods in RACER, a vectorized Monte Carlo code used for reactor physics analysis, so that comparisons may be made with techniques used in other codes. The principal applications for RACER are eigenvalue calculations and power distributions associated with reactor core physics analysis. Successive batches of neutrons are run until convergence and acceptable confidence intervals are obtained, with typical problems involving >10 6 histories. As such, the development of computational geometry methods has emphasized two basic needs: a flexible but compact geometric representation that permits accurate modeling of reactor core details and efficient geometric computation to permit very large numbers of histories to be run. The current geometric capabilities meet these needs effectively, supporting a variety of very large and demanding applications

  3. Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

    International Nuclear Information System (INIS)

    Siefman, Daniel J.; Girardin, Gaëtan; Rais, Adolfo; Pautz, Andreas; Hursin, Mathieu

    2015-01-01

    Highlights: • Modeling of research reactors. • Serpent and PARCS coupling. • Lattice physics codes modeling techniques. - Abstract: This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of k eff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of k eff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients

  4. A model of gas cavity breakup behind a blockage in fast breeder reactor subassembly geometry

    International Nuclear Information System (INIS)

    Fukuzawa, Y.

    1980-05-01

    A semi-empirical model has been developed to describe the transient behaviour of a gas cavity due to breakup behind a blockage in Liquid Metal Fast Breeder Reactor subassembly geometry. The main mechanisms assumed for gas cavity breakup in the present model are as follows: The gas cavity is broken up by the pressure fluctuation at the interface due to turbulence in the liquid. The centrifugal force on the liquid opposes breakup. The model is able to describe experimental results on the transient behaviour of a gas cavity due to breakup after the termination of gas injection. On the basis of the present model the residence time of a gas cavity behind a blockage in sodium is predicted and the dependence of the residence time on blockage size is discussed. (orig.) [de

  5. RGG: Reactor geometry (and mesh) generator

    International Nuclear Information System (INIS)

    Jain, R.; Tautges, T.

    2012-01-01

    The reactor geometry (and mesh) generator RGG takes advantage of information about repeated structures in both assembly and core lattices to simplify the creation of geometry and mesh. It is released as open source software as a part of the MeshKit mesh generation library. The methodology operates in three stages. First, assembly geometry models of various types are generated by a tool called AssyGen. Next, the assembly model or models are meshed by using MeshKit tools or the CUBIT mesh generation tool-kit, optionally based on a journal file output by AssyGen. After one or more assembly model meshes have been constructed, a tool called CoreGen uses a copy/move/merge process to arrange the model meshes into a core model. In this paper, we present the current state of tools and new features in RGG. We also discuss the parallel-enabled CoreGen, which in several cases achieves super-linear speedups since the problems fit in available RAM at higher processor counts. Several RGG applications - 1/6 VHTR model, 1/4 PWR reactor core, and a full-core model for Monju - are reported. (authors)

  6. Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P{sub 3} nodal methods

    Energy Technology Data Exchange (ETDEWEB)

    Duerigen, Susan

    2013-05-15

    The superior advantage of a nodal method for reactor cores with hexagonal fuel assemblies discretized as cells consisting of equilateral triangles is its mesh refinement capability. In this thesis, a diffusion and a simplified P{sub 3} (or SP{sub 3}) neutron transport nodal method are developed based on trigonal geometry. Both models are implemented in the reactor dynamics code DYN3D. As yet, no other well-established nodal core analysis code comprises an SP{sub 3} transport theory model based on trigonal meshes. The development of two methods based on different neutron transport approximations but using identical underlying spatial trigonal discretization allows a profound comparative analysis of both methods with regard to their mathematical derivations, nodal expansion approaches, solution procedures, and their physical performance. The developed nodal approaches can be regarded as a hybrid NEM/AFEN form. They are based on the transverse-integration procedure, which renders them computationally efficient, and they use a combination of polynomial and exponential functions to represent the neutron flux moments of the SP{sub 3} and diffusion equations, which guarantees high accuracy. The SP{sub 3} equations are derived in within-group form thus being of diffusion type. On this basis, the conventional diffusion solver structure can be retained also for the solution of the SP{sub 3} transport problem. The verification analysis provides proof of the methodological reliability of both trigonal DYN3D models. By means of diverse hexagonal academic benchmark and realistic detailed-geometry full-transport-theory problems, the superiority of the SP{sub 3} transport over the diffusion model is demonstrated in cases with pronounced anisotropy effects, which is, e.g., highly relevant to the modeling of fuel assemblies comprising absorber material.

  7. Assessment of capability for modeling the core degradation in 2D geometry with ASTEC V2 integral code for VVER type of reactor

    International Nuclear Information System (INIS)

    Dimov, D.

    2011-01-01

    The ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out since 2004. The purpose of this analysis is to assess ASTEC code modelling of main phenomena arising during hypothetical severe accidents and particularly in-vessel degradation in 2D geometry. The investigation covers both early and late phase of degradation of reactor core as well as determination of corium which will enter the reactor cavity. The initial event is station back-out. In order to receive severe accident condition, failure of all active component of emergency core cooling system is apply. The analysis is focus on ICARE module of ASTEC code and particularly on so call MAGMA model. The aim of study is to determine the capability of the integral code to simulate core degradation and to determine the corium composition entering the reactor cavity. (author)

  8. MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR

    Directory of Open Access Journals (Sweden)

    Suwardi Suwardi

    2015-03-01

    Full Text Available MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*. The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m. MODEL SIMULASI VARIASI GEOMETRI DAN STRESS-STRAIN DARI PROTOTIP BAHAN BAKAR PIN BATAN SELAMA UJI IRADIASI DI REAKTOR RSG-GAS. Pusat Teknologi Bahan Bakar Nuklir (PTBBN telah menyiapkan tangkai (pin bahan bakar pendek perdana yang berisi pelet UO2 alam dalam kelongsong paduan zircaloy untuk dilakukan uji iradiasi daya naik. Penelitian ini merupakan bagian dari perancangan percobaan iradiasi pertama di PRTF (Power Ramp Test Fasility yang terpasang di reaktor serbaguna RSG-GAS berdaya 30 MW. Telah dilakukan pemodelan dan simulasi kinerja termal mekanikal pin selama iradiasi. Variasi geometri pelet dan kelongsong selama pengujian dimodelkan dengan memperhatikan fenomena ekspansi termal

  9. Documentation for MeshKit - Reactor Geometry (&mesh) Generator

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    This report gives documentation for using MeshKit’s Reactor Geometry (and mesh) Generator (RGG) GUI and also briefly documents other algorithms and tools available in MeshKit. RGG is a program designed to aid in modeling and meshing of complex/large hexagonal and rectilinear reactor cores. RGG uses Argonne’s SIGMA interfaces, Qt and VTK to produce an intuitive user interface. By integrating a 3D view of the reactor with the meshing tools and combining them into one user interface, RGG streamlines the task of preparing a simulation mesh and enables real-time feedback that reduces accidental scripting mistakes that could waste hours of meshing. RGG interfaces with MeshKit tools to consolidate the meshing process, meaning that going from model to mesh is as easy as a button click. This report is designed to explain RGG v 2.0 interface and provide users with the knowledge and skills to pilot RGG successfully. Brief documentation of MeshKit source code, tools and other algorithms available are also presented for developers to extend and add new algorithms to MeshKit. RGG tools work in serial and parallel and have been used to model complex reactor core models consisting of conical pins, load pads, several thousands of axially varying material properties of instrumentation pins and other interstices meshes.

  10. Compound and Geometry-Dependent Pre-Compound Models to Calculate the Nuclear Data for Fusion Reactors

    International Nuclear Information System (INIS)

    Jahn, Helmut

    2005-01-01

    Compound and geometry-dependent pre-compound nuclear reactions are very useful concepts of nuclear theory to calculate cross sections of neutrons of around 14 MeV and below scattered by nuclei of material of installations producing energy of nuclear fusion. If these concepts are used to discuss and improve the experimental data they have to be completed by DWBA-type contributions to the small-step region of the incident neutron which can account for the angular distribution of the scattered neutron because there is the difficulty to separate experimentally the incoming from the scattered beam. The angle integrated cross-section in this region can be shown to be accounted for the surface dependent components of Blanns geometry-dependent precompound mechanism of the statistical state density and level density contributions of the compound and precompound components beeing calculated according to the recent developments of Anzaldo using the analytic number theory. The experimental data have been taken from the results of Hermsdorf, Meister, Sassonov, Seeliger, Seidel, Shahin and of A.Takahashi

  11. Stochastic Modelling of River Geometry

    DEFF Research Database (Denmark)

    Sørensen, John Dalsgaard; Schaarup-Jensen, K.

    1996-01-01

    Numerical hydrodynamic river models are used in a large number of applications to estimate critical events for rivers. These estimates are subject to a number of uncertainties. In this paper, the problem to evaluate these estimates using probabilistic methods is considered. Stochastic models for ...... for river geometries are formulated and a coupling between hydraulic computational methods and numerical reliability methods is presented....

  12. Tropical geometry of statistical models.

    Science.gov (United States)

    Pachter, Lior; Sturmfels, Bernd

    2004-11-16

    This article presents a unified mathematical framework for inference in graphical models, building on the observation that graphical models are algebraic varieties. From this geometric viewpoint, observations generated from a model are coordinates of a point in the variety, and the sum-product algorithm is an efficient tool for evaluating specific coordinates. Here, we address the question of how the solutions to various inference problems depend on the model parameters. The proposed answer is expressed in terms of tropical algebraic geometry. The Newton polytope of a statistical model plays a key role. Our results are applied to the hidden Markov model and the general Markov model on a binary tree.

  13. Performance of a fine-grained parallel model for multi-group nodal-transport calculations in three-dimensional pin-by-pin reactor geometry

    International Nuclear Information System (INIS)

    Masahiro, Tatsumi; Akio, Yamamoto

    2003-01-01

    A production code SCOPE2 was developed based on the fine-grained parallel algorithm by the red/black iterative method targeting parallel computing environments such as a PC-cluster. It can perform a depletion calculation in a few hours using a PC-cluster with the model based on a 9-group nodal-SP3 transport method in 3-dimensional pin-by-pin geometry for in-core fuel management of commercial PWRs. The present algorithm guarantees the identical convergence process as that in serial execution, which is very important from the viewpoint of quality management. The fine-mesh geometry is constructed by hierarchical decomposition with introduction of intermediate management layer as a block that is a quarter piece of a fuel assembly in radial direction. A combination of a mesh division scheme forcing even meshes on each edge and a latency-hidden communication algorithm provided simplicity and efficiency to message passing to enhance parallel performance. Inter-processor communication and parallel I/O access were realized using the MPI functions. Parallel performance was measured for depletion calculations by the 9-group nodal-SP3 transport method in 3-dimensional pin-by-pin geometry with 340 x 340 x 26 meshes for full core geometry and 170 x 170 x 26 for quarter core geometry. A PC cluster that consists of 24 Pentium-4 processors connected by the Fast Ethernet was used for the performance measurement. Calculations in full core geometry gave better speedups compared to those in quarter core geometry because of larger granularity. Fine-mesh sweep and feedback calculation parts gave almost perfect scalability since granularity is large enough, while 1-group coarse-mesh diffusion acceleration gave only around 80%. The speedup and parallel efficiency for total computation time were 22.6 and 94%, respectively, for the calculation in full core geometry with 24 processors. (authors)

  14. Mixing In Jet-Stirred Reactors With Different Geometries

    KAUST Repository

    Ayass, Wassim W.

    2013-12-01

    This work offers a well-developed understanding of the mixing process inside Jet- Stirred Reactors (JSR’s) with different geometries. Due to the difficulty of manufacturing these JSR’s made in quartz, existing JSR configurations were assessed with certain modifications and optimal operating conditions were suggested for each reactor. The effect of changing the reactor volume, the nozzle diameter and shape on mixing were both studied. Two nozzle geometries were examined in this study, a crossed shape nozzle and an inclined shape nozzle. Overall, six reactor configurations were assessed by conducting tracer experiments - using the state-of-art technologies of high-speed cameras and laser absorption spectroscopy- and Computational Fluid Dynamics (CFD) simulations. The high-speed camera tracer experiment gives unique qualitative information – not present in the literature – about the actual flow field. On the other hand, when using the laser technique, a more quantitative analysis emerges with determining the experimental residence time distribution (RTD) curves of each reactor. Comparing these RTD curves with the ideal curve helped in eliminating two cases. Finally, the CFD simulations predict the RTD curves as well as the mixing levels of the JSR’s operated at different residence times. All of these performed studies suggested the use of an inclined nozzle configuration with a reactor diameter D of 40mm and a nozzle diameter d of 1mm as the optimal choice for low residence time operation. However, for higher residence times, the crossed configuration reactor with D=56mm and d=0.3mm gave a nearly perfect behavior.

  15. Geometry modeling for SAM-CE Monte Carlo calculations

    International Nuclear Information System (INIS)

    Steinberg, H.A.; Troubetzkoy, E.S.

    1980-01-01

    Three geometry packages have been developed and incorporated into SAM-CE, for representing in three dimensions the transport medium. These are combinatorial geometry - a general (non-lattice) system, complex combinatorial geometry - a very general system with lattice capability, and special reactor geometry - a special purpose system for light water reactor geometries. Their different attributes are described

  16. Application of finite element numerical technique to nuclear reactor geometries

    Energy Technology Data Exchange (ETDEWEB)

    Rouai, N M [Nuclear engineering department faculty of engineering Al-fateh universty, Tripoli (Libyan Arab Jamahiriya)

    1995-10-01

    Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs.

  17. Application of finite element numerical technique to nuclear reactor geometries

    International Nuclear Information System (INIS)

    Rouai, N. M.

    1995-01-01

    Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs

  18. Differential Geometry Based Multiscale Models

    Science.gov (United States)

    Wei, Guo-Wei

    2010-01-01

    Large chemical and biological systems such as fuel cells, ion channels, molecular motors, and viruses are of great importance to the scientific community and public health. Typically, these complex systems in conjunction with their aquatic environment pose a fabulous challenge to theoretical description, simulation, and prediction. In this work, we propose a differential geometry based multiscale paradigm to model complex macromolecular systems, and to put macroscopic and microscopic descriptions on an equal footing. In our approach, the differential geometry theory of surfaces and geometric measure theory are employed as a natural means to couple the macroscopic continuum mechanical description of the aquatic environment with the microscopic discrete atom-istic description of the macromolecule. Multiscale free energy functionals, or multiscale action functionals are constructed as a unified framework to derive the governing equations for the dynamics of different scales and different descriptions. Two types of aqueous macromolecular complexes, ones that are near equilibrium and others that are far from equilibrium, are considered in our formulations. We show that generalized Navier–Stokes equations for the fluid dynamics, generalized Poisson equations or generalized Poisson–Boltzmann equations for electrostatic interactions, and Newton's equation for the molecular dynamics can be derived by the least action principle. These equations are coupled through the continuum-discrete interface whose dynamics is governed by potential driven geometric flows. Comparison is given to classical descriptions of the fluid and electrostatic interactions without geometric flow based micro-macro interfaces. The detailed balance of forces is emphasized in the present work. We further extend the proposed multiscale paradigm to micro-macro analysis of electrohydrodynamics, electrophoresis, fuel cells, and ion channels. We derive generalized Poisson–Nernst–Planck equations that

  19. Differential geometry based multiscale models.

    Science.gov (United States)

    Wei, Guo-Wei

    2010-08-01

    Large chemical and biological systems such as fuel cells, ion channels, molecular motors, and viruses are of great importance to the scientific community and public health. Typically, these complex systems in conjunction with their aquatic environment pose a fabulous challenge to theoretical description, simulation, and prediction. In this work, we propose a differential geometry based multiscale paradigm to model complex macromolecular systems, and to put macroscopic and microscopic descriptions on an equal footing. In our approach, the differential geometry theory of surfaces and geometric measure theory are employed as a natural means to couple the macroscopic continuum mechanical description of the aquatic environment with the microscopic discrete atomistic description of the macromolecule. Multiscale free energy functionals, or multiscale action functionals are constructed as a unified framework to derive the governing equations for the dynamics of different scales and different descriptions. Two types of aqueous macromolecular complexes, ones that are near equilibrium and others that are far from equilibrium, are considered in our formulations. We show that generalized Navier-Stokes equations for the fluid dynamics, generalized Poisson equations or generalized Poisson-Boltzmann equations for electrostatic interactions, and Newton's equation for the molecular dynamics can be derived by the least action principle. These equations are coupled through the continuum-discrete interface whose dynamics is governed by potential driven geometric flows. Comparison is given to classical descriptions of the fluid and electrostatic interactions without geometric flow based micro-macro interfaces. The detailed balance of forces is emphasized in the present work. We further extend the proposed multiscale paradigm to micro-macro analysis of electrohydrodynamics, electrophoresis, fuel cells, and ion channels. We derive generalized Poisson-Nernst-Planck equations that are

  20. Pressure loss coefficient evaluation based on CFD analysis for simple geometries and PWR reactor vessel without geometry simplification

    International Nuclear Information System (INIS)

    Ko II, B.; Park, J. P.; Jeong, J. H.

    2008-01-01

    Nuclear vendors and utilities perform lots of simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes and most of them were developed based on 1-dimensional lumped parameter models. These thermal-hydraulic system analysis codes require user input for pressure loss coefficient, k-factor; since they numerically solve Euler-equation. In spite of its high impact on the safety analysis results, there has not been good validation method for the selection of loss coefficient. During the past decade, however; computers, parallel computation methods, and 3-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. It is believed to be beneficial to take advantage of advanced commercial CFD codes in safety analysis and design of NPP5. The present work aims to validate pressure loss coefficient evaluation for simple geometries and k-factor calculation for PWR based on CFD. The performances of standard k-ε model, RNG k-ε model, Reynolds stress model (RSM) on the simulation of pressure drop for simple geometry such as, or sudden-expansion, and sudden-contraction are evaluated. The calculated value was compared with pressure loss coefficient in handbook of hydraulic resistance. Then the present work carried out analysis for flow distribution in downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of lower plenum is used. The results give a clear figure about flow fields in the reactor vessel, which is one of major safety concerns. The calculated pressure drop across downcomer and lower plenum appears to be in good agreement

  1. Surrogate Modeling for Geometry Optimization

    DEFF Research Database (Denmark)

    Rojas Larrazabal, Marielba de la Caridad; Abraham, Yonas; Holzwarth, Natalie

    2009-01-01

    A new approach for optimizing the nuclear geometry of an atomic system is described. Instead of the original expensive objective function (energy functional), a small number of simpler surrogates is used.......A new approach for optimizing the nuclear geometry of an atomic system is described. Instead of the original expensive objective function (energy functional), a small number of simpler surrogates is used....

  2. S/sub N/ computational benchmark solutions for slab geometry models of a gas-cooled fast reactor (GCFR) lattice cell

    International Nuclear Information System (INIS)

    McCoy, D.R.

    1981-01-01

    S/sub N/ computational benchmark solutions are generated for a onegroup and multigroup fuel-void slab lattice cell which is a rough model of a gas-cooled fast reactor (GCFR) lattice cell. The reactivity induced by the extrusion of the fuel material into the voided region is determined for a series of partially extruded lattice cell configurations. A special modified Gauss S/sub N/ ordinate array design is developed in order to obtain eigenvalues with errors less than 0.03% in all of the configurations that are considered. The modified Gauss S/sub N/ ordinate array design has a substantially improved eigenvalue angular convergence behavior when compared to existing S/sub N/ ordinate array designs used in neutron streaming applications. The angular refinement computations are performed in some cases by using a perturbation theory method which enables one to obtain high order S/sub N/ eigenvalue estimates for greatly reduced computational costs

  3. 3D modeling and visualization software for complex geometries

    International Nuclear Information System (INIS)

    Guse, Guenter; Klotzbuecher, Michael; Mohr, Friedrich

    2011-01-01

    The reactor safety depends on reliable nondestructive testing of reactor components. For 100% detection probability of flaws and the determination of their size using ultrasonic methods the ultrasonic waves have to hit the flaws within a specific incidence and squint angle. For complex test geometries like testing of nozzle welds from the outside of the component these angular ranges can only be determined using elaborate mathematical calculations. The authors developed a 3D modeling and visualization software tool that allows to integrate and present ultrasonic measuring data into the 3D geometry. The software package was verified using 1:1 test samples (example: testing of the nozzle edge of the feedwater nozzle of a steam generator from the outside; testing of the reactor pressure vessel nozzle edge from the inside).

  4. A 3D transport-based core analysis code for research reactors with unstructured geometry

    International Nuclear Information System (INIS)

    Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

    2013-01-01

    Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

  5. Calculation models for a nuclear reactor

    International Nuclear Information System (INIS)

    Tashanii, Ahmed Ali

    2010-01-01

    Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)

  6. Differential geometry in string models

    International Nuclear Information System (INIS)

    Alvarez, O.

    1986-01-01

    In this article the author reviews the differential geometric approach to the quantization of strings. A seminal paper demonstrates the connection between the trace anomaly and the critical dimension. The role played by the Faddeev-Popov ghosts has been instrumental in much of the subsequent work on the quantization of strings. This paper discusses the differential geometry of two dimensional surfaces and its importance in the quantization of strings. The path integral quantization approach to strings will be carefully analyzed to determine the correct effective measure for string theories. The choice of measure for the path integral is determined by differential geometric considerations. Once the measure is determined, the manifest diffeomorphism invariance of the theory will have to be broken by using the Faddeev-Popov ansatz. The gauge fixed theory is studied in detail with emphasis on the role of conformal and gravitational anomalies. In the analysis, the path integral formulation of the gauge fixed theory requires summing over all the distinct complex structures on the manifold

  7. Particle bed reactor modeling

    Science.gov (United States)

    Sapyta, Joe; Reid, Hank; Walton, Lew

    The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

  8. Parameterized combinatorial geometry modeling in Moritz

    International Nuclear Information System (INIS)

    Van Riper, K.A.

    2005-01-01

    We describe the use of named variables as surface and solid body coefficients in the Moritz geometry editing program. Variables can also be used as material numbers, cell densities, and transformation values. A variable is defined as a constant or an arithmetic combination of constants and other variables. A variable reference, such as in a surface coefficient, can be a single variable or an expression containing variables and constants. Moritz can read and write geometry models in MCNP and ITS ACCEPT format; support for other codes will be added. The geometry can be saved with either the variables in place, for modifying the models in Moritz, or with the variables evaluated for use in the transport codes. A program window shows a list of variables and provides fields for editing them. Surface coefficients and other values that use a variable reference are shown in a distinctive style on object property dialogs; associated buttons show fields for editing the reference. We discuss our use of variables in defining geometry models for shielding studies in PET clinics. When a model is parameterized through the use of variables, changes such as room dimensions, shielding layer widths, and cell compositions can be quickly achieved by changing a few numbers without requiring knowledge of the input syntax for the transport code or the tedious and error prone work of recalculating many surface or solid body coefficients. (author)

  9. Advances in Spectral Nodal Methods applied to SN Nuclear Reactor Global calculations in Cartesian Geometry

    International Nuclear Information System (INIS)

    Barros, R.C.; Filho, H.A.; Oliveira, F.B.S.; Silva, F.C. da

    2004-01-01

    Presented here are the advances in spectral nodal methods for discrete ordinates (SN) eigenvalue problems in Cartesian geometry. These coarse-mesh methods are based on three ingredients: (i) the use of the standard discretized spatial balance SN equations; (ii) the use of the non-standard spectral diamond (SD) auxiliary equations in the multiplying regions of the domain, e.g. fuel assemblies; and (iii) the use of the non-standard spectral Green's function (SGF) auxiliary equations in the non-multiplying regions of the domain, e.g., the reflector. In slab-geometry the hybrid SD-SGF method generates numerical results that are completely free of spatial truncation errors. In X,Y-geometry, we obtain a system of two 'slab-geometry' SN equations for the node-edge average angular fluxes by transverse-integrating the X,Y-geometry SN equations separately in the y- and then in the x-directions within an arbitrary node of the spatial grid set up on the domain. In this paper, we approximate the transverse leakage terms by constants. These are the only approximations considered in the SD-SGF-constant nodal method, as the source terms, that include scattering and eventually fission events, are treated exactly. Moreover, we describe in this paper the progress of the approximate SN albedo boundary conditions for substituting the non-multiplying regions around the nuclear reactor core. We show numerical results to typical model problems to illustrate the accuracy of spectral nodal methods for coarse-mesh SN criticality calculations. (Author)

  10. Mixing In Jet-Stirred Reactors With Different Geometries

    KAUST Repository

    Ayass, Wassim W.

    2013-01-01

    analysis emerges with determining the experimental residence time distribution (RTD) curves of each reactor. Comparing these RTD curves with the ideal curve helped in eliminating two cases. Finally, the CFD simulations predict the RTD curves as well

  11. Geometries

    CERN Document Server

    Sossinsky, A B

    2012-01-01

    The book is an innovative modern exposition of geometry, or rather, of geometries; it is the first textbook in which Felix Klein's Erlangen Program (the action of transformation groups) is systematically used as the basis for defining various geometries. The course of study presented is dedicated to the proposition that all geometries are created equal--although some, of course, remain more equal than others. The author concentrates on several of the more distinguished and beautiful ones, which include what he terms "toy geometries", the geometries of Platonic bodies, discrete geometries, and classical continuous geometries. The text is based on first-year semester course lectures delivered at the Independent University of Moscow in 2003 and 2006. It is by no means a formal algebraic or analytic treatment of geometric topics, but rather, a highly visual exposition containing upwards of 200 illustrations. The reader is expected to possess a familiarity with elementary Euclidean geometry, albeit those lacking t...

  12. Geometry

    Indian Academy of Sciences (India)

    . In the previous article we looked at the origins of synthetic and analytic geometry. More practical minded people, the builders and navigators, were studying two other aspects of geometry- trigonometry and integral calculus. These are actually ...

  13. Geometry

    CERN Document Server

    Prasolov, V V

    2015-01-01

    This book provides a systematic introduction to various geometries, including Euclidean, affine, projective, spherical, and hyperbolic geometries. Also included is a chapter on infinite-dimensional generalizations of Euclidean and affine geometries. A uniform approach to different geometries, based on Klein's Erlangen Program is suggested, and similarities of various phenomena in all geometries are traced. An important notion of duality of geometric objects is highlighted throughout the book. The authors also include a detailed presentation of the theory of conics and quadrics, including the theory of conics for non-Euclidean geometries. The book contains many beautiful geometric facts and has plenty of problems, most of them with solutions, which nicely supplement the main text. With more than 150 figures illustrating the arguments, the book can be recommended as a textbook for undergraduate and graduate-level courses in geometry.

  14. Computational algebraic geometry of epidemic models

    Science.gov (United States)

    Rodríguez Vega, Martín.

    2014-06-01

    Computational Algebraic Geometry is applied to the analysis of various epidemic models for Schistosomiasis and Dengue, both, for the case without control measures and for the case where control measures are applied. The models were analyzed using the mathematical software Maple. Explicitly the analysis is performed using Groebner basis, Hilbert dimension and Hilbert polynomials. These computational tools are included automatically in Maple. Each of these models is represented by a system of ordinary differential equations, and for each model the basic reproductive number (R0) is calculated. The effects of the control measures are observed by the changes in the algebraic structure of R0, the changes in Groebner basis, the changes in Hilbert dimension, and the changes in Hilbert polynomials. It is hoped that the results obtained in this paper become of importance for designing control measures against the epidemic diseases described. For future researches it is proposed the use of algebraic epidemiology to analyze models for airborne and waterborne diseases.

  15. Measurements of plume geometry and argon-41 radiation field at the BR1 reactor in Mol, Belgium

    International Nuclear Information System (INIS)

    Drews, M.; Joergensen, H.; Lauritzen, Bent; Mikkelsen, T.; Aage, H.K.; Korsbech, U.; Bargholz, K.; Rojas-Palma, C.; Ammel, R. van

    2002-02-01

    An atmospheric dispersion experiment was conducted using a visible tracer along with the routine releases of 41 Ar from the BR1 air-cooled research reactor in Mol. In the experiment, simultaneous measurements of the radiation field from the 41 Ar decay, the meteorology, the 41 Ar source term and plume geometry were performed. The visible tracer was injected into the reactor emission stack, and the plume cross section determined by Lidar scanning of the released aerosols. The data collected in the exercise provide a valuable resource for atmospheric dispersion and dose rate modeling. (au)

  16. Fractal reactor: An alternative nuclear fusion system based on nature's geometry

    International Nuclear Information System (INIS)

    Siler, T. L.

    2007-01-01

    The author presents his concept of the Fractal Reactor, which explores the possibility of building a plasma fusion power reactor based on the real geometry of nature [fractals], rather than the virtual geometry that Euclid postulated around 330 BC; nearly every architect of our plasma fusion devices has been influenced by his three-dimensional geometry. The idealized points, lines, planes, and spheres of this classical geometry continue to be used to represent the natural world and to describe the properties of all geometrical objects, even though they neither accurately nor fully convey nature's structures and processes. The Fractal Reactor concept contrasts the current containment mechanisms of both magnetic and inertial containment systems for confining and heating plasmas. All of these systems are based on Euclidean geometry and use geometrical designs that, ultimately, are inconsistent with the Non-Euclidean geometry and irregular, fractal forms of nature (3). The author explores his premise that a controlled, thermonuclear fusion energy system might be more effective if it more closely embodies the physics of a star

  17. Optimization geometries of a vortex gliding-arc reactor for partial oxidation of methane

    International Nuclear Information System (INIS)

    Guofeng, Xu; Xinwei, Ding

    2012-01-01

    The effects of the geometry of gliding-arc reactor – such as distance between the electrodes, outlet diameter, and inlet position – on the reactor characteristics (methane conversion, hydrogen yield, and energy efficiency) have not been fully investigated. In this paper, AC gliding-arc reactors including the vortex flow configuration are designed to produce hydrogen from the methane by partial oxidation. The influence of vortex flow configuration on the reactor characteristics is also studied by varying the inlet position. When the inlet of the gliding-arc reactor is positioned close to the outlet, reverse vortex flow reactor (RVFR), the maximum energy efficiency reaches 50% and the yields of hydrogen and carbon monoxide are 40% and 65%, respectively. As the distance between electrodes increases from 5 mm to 15 mm, both hydrogen yield and energy efficiency increase approximately 10% for the RVFR. The energy efficiency and hydrogen yield are highest when the ratio of the outlet diameter to the inner diameter is 0.5 for the RVFR. Experimental results indicate that the flow field in the plasma reactor has an important influence on the reactor performance. Furthermore, hydrogen production increases as the number of feed gas flows in contact with the plasma zone increases. -- Highlights: ► Gliding-arc reactors were designed to produce hydrogen for studying the characteristics of the vortex flow reactor. ► Hydrogen yield of reverse vortex flow reactor was 10% higher than that of forward vortex flow reactor. ► Maximum energy efficiency was 50% for reverse vortex flow reactor. ► If discharge power was supplied to the reactors, the reactor performance increased with increasing distance between electrodes. ► Optimum ratio of the outlet and inner diameter was 1/2.

  18. Hydrogen combustion modelling in large-scale geometries

    International Nuclear Information System (INIS)

    Studer, E.; Beccantini, A.; Kudriakov, S.; Velikorodny, A.

    2014-01-01

    Hydrogen risk mitigation issues based on catalytic recombiners cannot exclude flammable clouds to be formed during the course of a severe accident in a Nuclear Power Plant. Consequences of combustion processes have to be assessed based on existing knowledge and state of the art in CFD combustion modelling. The Fukushima accidents have also revealed the need for taking into account the hydrogen explosion phenomena in risk management. Thus combustion modelling in a large-scale geometry is one of the remaining severe accident safety issues. At present day there doesn't exist a combustion model which can accurately describe a combustion process inside a geometrical configuration typical of the Nuclear Power Plant (NPP) environment. Therefore the major attention in model development has to be paid on the adoption of existing approaches or creation of the new ones capable of reliably predicting the possibility of the flame acceleration in the geometries of that type. A set of experiments performed previously in RUT facility and Heiss Dampf Reactor (HDR) facility is used as a validation database for development of three-dimensional gas dynamic model for the simulation of hydrogen-air-steam combustion in large-scale geometries. The combustion regimes include slow deflagration, fast deflagration, and detonation. Modelling is based on Reactive Discrete Equation Method (RDEM) where flame is represented as an interface separating reactants and combustion products. The transport of the progress variable is governed by different flame surface wrinkling factors. The results of numerical simulation are presented together with the comparisons, critical discussions and conclusions. (authors)

  19. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  20. IVA2 - a computer code for modelling of transient 3D-three phase three component flows using three velocity fields in cylindrical geometry with arbitrary internals including nuclear reactor PWR/BWR-core

    International Nuclear Information System (INIS)

    Kolev, N.I.

    1986-06-01

    This report contains a formal code description (description of the input data, contents of the COMMON blocks, functions of the IVA2/001 routines). In addition the nonformal description of the current IVA2/001 constitutive package and the reactor core model are given. (orig.) [de

  1. Comparative analysis of a fusion reactor blanket in cylindrical and toroidal geometry using Monte Carlo

    International Nuclear Information System (INIS)

    Chapin, D.L.

    1976-03-01

    Differences in neutron fluxes and nuclear reaction rates in a noncircular fusion reactor blanket when analyzed in cylindrical and toroidal geometry are studied using Monte Carlo. The investigation consists of three phases--a one-dimensional calculation using a circular approximation to a hexagonal shaped blanket; a two-dimensional calculation of a hexagonal blanket in an infinite cylinder; and a three-dimensional calculation of the blanket in tori of aspect ratios 3 and 5. The total blanket reaction rate in the two-dimensional model is found to be in good agreement with the circular model. The toroidal calculations reveal large variations in reaction rates at different blanket locations as compared to the hexagonal cylinder model, although the total reaction rate is nearly the same for both models. It is shown that the local perturbations in the toroidal blanket are due mainly to volumetric effects, and can be predicted by modifying the results of the infinite cylinder calculation by simple volume factors dependent on the blanket location and the torus major radius

  2. Geometry

    CERN Document Server

    Pedoe, Dan

    1988-01-01

    ""A lucid and masterly survey."" - Mathematics Gazette Professor Pedoe is widely known as a fine teacher and a fine geometer. His abilities in both areas are clearly evident in this self-contained, well-written, and lucid introduction to the scope and methods of elementary geometry. It covers the geometry usually included in undergraduate courses in mathematics, except for the theory of convex sets. Based on a course given by the author for several years at the University of Minnesota, the main purpose of the book is to increase geometrical, and therefore mathematical, understanding and to he

  3. Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)

    2012-07-01

    Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

  4. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  5. A Statistical Model for Synthesis of Detailed Facial Geometry

    OpenAIRE

    Golovinskiy, Aleksey; Matusik, Wojciech; Pfister, Hanspeter; Rusinkiewicz, Szymon; Funkhouser, Thomas

    2006-01-01

    Detailed surface geometry contributes greatly to the visual realism of 3D face models. However, acquiring high-resolution face geometry is often tedious and expensive. Consequently, most face models used in games, virtual reality, or computer vision look unrealistically smooth. In this paper, we introduce a new statistical technique for the analysis and synthesis of small three-dimensional facial features, such as wrinkles and pores. We acquire high-resolution face geometry for people across ...

  6. The VSEPR model of molecular geometry

    CERN Document Server

    Gillespie, Ronald J

    2012-01-01

    Valence Shell Electron Pair Repulsion (VSEPR) theory is a simple technique for predicting the geometry of atomic centers in small molecules and molecular ions. This authoritative reference was written by Istvan Hartiggai and the developer of VSEPR theory, Ronald J. Gillespie. In addition to its value as a text for courses in molecular geometry and chemistry, it constitutes a classic reference for professionals.Starting with coverage of the broader aspects of VSEPR, this volume narrows its focus to a succinct survey of the methods of structural determination. Additional topics include the appli

  7. Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Hang; Wang Jie; Laurien, E.

    2010-01-01

    The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)

  8. Stages As Models of Scene Geometry

    NARCIS (Netherlands)

    Nedović, V.; Smeulders, A.W.M.; Redert, A.; Geusebroek, J.M.

    2010-01-01

    Reconstruction of 3D scene geometry is an important element for scene understanding, autonomous vehicle and robot navigation, image retrieval, and 3D television. We propose accounting for the inherent structure of the visual world when trying to solve the scene reconstruction problem. Consequently,

  9. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

    International Nuclear Information System (INIS)

    Reyes F, M. C.; Del Valle G, E.; Filio L, C.

    2013-10-01

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  10. Stages as models of scene geometry.

    Science.gov (United States)

    Nedović, Vladimir; Smeulders, Arnold W M; Redert, André; Geusebroek, Jan-Mark

    2010-09-01

    Reconstruction of 3D scene geometry is an important element for scene understanding, autonomous vehicle and robot navigation, image retrieval, and 3D television. We propose accounting for the inherent structure of the visual world when trying to solve the scene reconstruction problem. Consequently, we identify geometric scene categorization as the first step toward robust and efficient depth estimation from single images. We introduce 15 typical 3D scene geometries called stages, each with a unique depth profile, which roughly correspond to a large majority of broadcast video frames. Stage information serves as a first approximation of global depth, narrowing down the search space in depth estimation and object localization. We propose different sets of low-level features for depth estimation, and perform stage classification on two diverse data sets of television broadcasts. Classification results demonstrate that stages can often be efficiently learned from low-dimensional image representations.

  11. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weissfloch, R

    1973-07-15

    The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

  12. Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Weissfloch, R.

    The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

  13. TU electric reactor model verification

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1989-01-01

    Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

  14. Probability-neighbor method of accelerating geometry treatment in reactor Monte Carlo code RMC

    International Nuclear Information System (INIS)

    She, Ding; Li, Zeguang; Xu, Qi; Wang, Kan; Yu, Ganglin

    2011-01-01

    Probability neighbor method (PNM) is proposed in this paper to accelerate geometry treatment of Monte Carlo (MC) simulation and validated in self-developed reactor Monte Carlo code RMC. During MC simulation by either ray-tracking or delta-tracking method, large amounts of time are spent in finding out which cell one particle is located in. The traditional way is to search cells one by one with certain sequence defined previously. However, this procedure becomes very time-consuming when the system contains a large number of cells. Considering that particles have different probability to enter different cells, PNM method optimizes the searching sequence, i.e., the cells with larger probability are searched preferentially. The PNM method is implemented in RMC code and the numerical results show that the considerable time of geometry treatment in MC calculation for complicated systems is saved, especially effective in delta-tracking simulation. (author)

  15. Study on modeling technology in digital reactor system

    International Nuclear Information System (INIS)

    Liu Xiaoping; Luo Yuetong; Tong Lili

    2004-01-01

    Modeling is the kernel part of a digital reactor system. As an extensible platform for reactor conceptual design, it is very important to study modeling technology and develop some kind of tools to speed up preparation of all classical computing models. This paper introduces the background of the project and basic conception of digital reactor. MCAM is taken as an example for modeling and its related technologies used are given. It is an interface program for MCNP geometry model developed by FDS team (ASIPP and HUT), and designed to run on windows system. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model. There have been two ways for MCAM to utilize CAD technology: (1) Making use of user interface technology in aid of generation of MCNP geometry model; (2) Making use of existing 3D CAD model to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM's major function. At last, several examples are given to demonstrate MCAM's various capabilities. (authors)

  16. The effect of reactor geometry on the synthesis of graphene materials in plasma jets

    Science.gov (United States)

    Shavelkina, M. B.; Amirov, R. H.; Shatalova, T. B.

    2017-05-01

    The possibility of synthesis of graphene and graphane (hydrogenated graphene) using the decomposition of hydrocarbons by thermal plasma has been investigated. Investigations of the influence of the plasma-forming gas on the efficiency of synthesis and the morphology of graphene materials were carried out. The synthesis products have been characterized by the methods of scanning microscopy, Raman spectroscopy and thermal analysis. It is found that the morphology of graphene materials is affected by the geometry of the reactor. It was demonstrated that the obtained graphene materials are uniformly distributed in the volume of plastic based on cyanate ester resins under mixing.

  17. A co-ordinate system for reactor physics calculations in hexagonal geometry

    International Nuclear Information System (INIS)

    Burte, D.P.

    1990-01-01

    A method for generating all the geometric information concerning typical reactor physics calculations for a basically hexagonal reactor core or its sector involving any of the possible symmetries is presented. The geometrically allowed symmetries for regular hexagons are discussed. The approach is based on the choice of a suitable co-ordinate system, viz. one using three coplanar (including one redundant) axes, each at 120 0 with its cyclically preceding one. A code named KEKULE' is developed for a 2-D, finite difference, one-group diffusion analysis of a hexagonal core using the approach. It can cater to a full hexagonal core as well as to any symmetric sectorial part of it. The main feature of the code is that the input concerning geometry is a bare minimum. It is hoped that the approach presented will be useful even for the calculations for hexagonal fuel assemblies. (author)

  18. Mathematical modelling of fluidized bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Werther, J [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1978-11-01

    Among the many fluidized bed models to be found in the literature, the two-phase model originally proposed by May has proved most suitable for accomodation of recent advances in flow mechanics: this model resolves the gas/solids fluidized bed into a bubble phase and a suspension phase surrounding the bubbles. Its limitation to slow reactions is a disadvantage. On the basis of the analogy between fluidized beds and gas/liquid systems, a general two-phase model that is valid for fast reactions has therefore been developed and its validity is confirmed by comparison with the experimental results obtained by others. The model describes mass transfer across the phase interface with the aid of the film theory known from gas/liquid reactor technology, and the reaction occurring in the suspension phase as a pseudo-homogeneous reaction. Since the dependence of the performance of fluidized bed reactors upon geometry is accounted for, the model can also be used for scale-up calculations. Its use is illustrated with the aid of design diagrams.

  19. 2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

    International Nuclear Information System (INIS)

    Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

    1969-01-01

    1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

  20. Two Step Procedure Using a 1-D Slab Spectral Geometry in a Pebble Bed Reactor Core Analysis

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Kim, Kang Seog; Noh, Jae Man; Joo, Hyung Kook

    2005-01-01

    A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted iteration between the spectrum calculation in a spectral zone and the global core calculation. In VSOP, the whole problem domain is divided into many spectral zones in which the fine group spectrum is calculated using bucklings for fast groups and albedos for thermal groups from the global core calculation. The resulting spectrum in each spectral zone is used to generate broad group cross sections of the spectral zone for the global core calculation. In this paper, we demonstrate a two step procedure in a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. The equivalent cross sections generated in this way include the effect of the spectral interaction between the core and the reflector. In the second step, we perform a diffusion calculation using the equivalent cross sections generated in the first step. A simple benchmark problem derived from the PMBR-400 Reactor was introduced to verify this approach. We compared the two step solutions with the Monte Carlo (MC) solutions for the problem

  1. Convex-based void filling method for CAD-based Monte Carlo geometry modeling

    International Nuclear Information System (INIS)

    Yu, Shengpeng; Cheng, Mengyun; Song, Jing; Long, Pengcheng; Hu, Liqin

    2015-01-01

    Highlights: • We present a new void filling method named CVF for CAD based MC geometry modeling. • We describe convex based void description based and quality-based space subdivision. • The results showed improvements provided by CVF for both modeling and MC calculation efficiency. - Abstract: CAD based automatic geometry modeling tools have been widely applied to generate Monte Carlo (MC) calculation geometry for complex systems according to CAD models. Automatic void filling is one of the main functions in the CAD based MC geometry modeling tools, because the void space between parts in CAD models is traditionally not modeled while MC codes such as MCNP need all the problem space to be described. A dedicated void filling method, named Convex-based Void Filling (CVF), is proposed in this study for efficient void filling and concise void descriptions. The method subdivides all the problem space into disjointed regions using Quality based Subdivision (QS) and describes the void space in each region with complementary descriptions of the convex volumes intersecting with that region. It has been implemented in SuperMC/MCAM, the Multiple-Physics Coupling Analysis Modeling Program, and tested on International Thermonuclear Experimental Reactor (ITER) Alite model. The results showed that the new method reduced both automatic modeling time and MC calculation time

  2. Almost-commutative geometries beyond the standard model

    International Nuclear Information System (INIS)

    Stephan, Christoph A

    2006-01-01

    In Iochum et al (2004 J. Math. Phys. 45 5003), Jureit and Stephan (2005 J. Math. Phys. 46 043512), Schuecker T (2005 Preprint hep-th/0501181) and Jureit et al (2005 J. Math. Phys. 46 072303), a conjecture is presented that almost-commutative geometries, with respect to sensible physical constraints, allow only the standard model of particle physics and electro-strong models as Yang-Mills-Higgs theories. In this paper, a counter-example will be given. The corresponding almost-commutative geometry leads to a Yang-Mills-Higgs model which consists of the standard model of particle physics and two new fermions of opposite electro-magnetic charge. This is the second Yang-Mills-Higgs model within noncommutative geometry, after the standard model, which could be compatible with experiments. Combined to a hydrogen-like composite particle, these new particles provide a novel dark matter candidate

  3. Computational modeling of geometry dependent phonon transport in silicon nanostructures

    Science.gov (United States)

    Cheney, Drew A.

    Recent experiments have demonstrated that thermal properties of semiconductor nanostructures depend on nanostructure boundary geometry. Phonons are quantized mechanical vibrations that are the dominant carrier of heat in semiconductor materials and their aggregate behavior determine a nanostructure's thermal performance. Phonon-geometry scattering processes as well as waveguiding effects which result from coherent phonon interference are responsible for the shape dependence of thermal transport in these systems. Nanoscale phonon-geometry interactions provide a mechanism by which nanostructure geometry may be used to create materials with targeted thermal properties. However, the ability to manipulate material thermal properties via controlling nanostructure geometry is contingent upon first obtaining increased theoretical understanding of fundamental geometry induced phonon scattering processes and having robust analytical and computational models capable of exploring the nanostructure design space, simulating the phonon scattering events, and linking the behavior of individual phonon modes to overall thermal behavior. The overall goal of this research is to predict and analyze the effect of nanostructure geometry on thermal transport. To this end, a harmonic lattice-dynamics based atomistic computational modeling tool was created to calculate phonon spectra and modal phonon transmission coefficients in geometrically irregular nanostructures. The computational tool is used to evaluate the accuracy and regimes of applicability of alternative computational techniques based upon continuum elastic wave theory. The model is also used to investigate phonon transmission and thermal conductance in diameter modulated silicon nanowires. Motivated by the complexity of the transmission results, a simplified model based upon long wavelength beam theory was derived and helps explain geometry induced phonon scattering of low frequency nanowire phonon modes.

  4. A 3D Geometry Model Search Engine to Support Learning

    Science.gov (United States)

    Tam, Gary K. L.; Lau, Rynson W. H.; Zhao, Jianmin

    2009-01-01

    Due to the popularity of 3D graphics in animation and games, usage of 3D geometry deformable models increases dramatically. Despite their growing importance, these models are difficult and time consuming to build. A distance learning system for the construction of these models could greatly facilitate students to learn and practice at different…

  5. Noncommutative geometry and the standard model vacuum

    International Nuclear Information System (INIS)

    Barrett, John W.; Dawe Martins, Rachel A.

    2006-01-01

    The space of Dirac operators for the Connes-Chamseddine spectral action for the standard model of particle physics coupled to gravity is studied. The model is extended by including right-handed neutrino states, and the S 0 -reality axiom is not assumed. The possibility of allowing more general fluctuations than the inner fluctuations of the vacuum is proposed. The maximal case of all possible fluctuations is studied by considering the equations of motion for the vacuum. While there are interesting nontrivial vacua with Majorana-type mass terms for the leptons, the conclusion is that the equations are too restrictive to allow solutions with the standard model mass matrix

  6. One dimensional reactor core model

    International Nuclear Information System (INIS)

    Kostadinov, V.; Stritar, A.; Radovo, M.; Mavko, B.

    1984-01-01

    The one dimensional model of neutron dynamic in reactor core was developed. The core was divided in several axial nodes. The one group neutron diffusion equation for each node is solved. Feedback affects of fuel and water temperatures is calculated. The influence of xenon, boron and control rods is included in cross section calculations for each node. The system of equations is solved implicitly. The model is used in basic principle Training Simulator of NPP Krsko. (author)

  7. The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

    International Nuclear Information System (INIS)

    Grundmann, U.

    1995-11-01

    The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

  8. Modelling of vapour explosion in stratified geometrie

    International Nuclear Information System (INIS)

    Picchi, St.

    1999-01-01

    When a hot liquid comes into contact with a colder volatile liquid, one can obtain in some conditions an explosive vaporization, told vapour explosion, whose consequences can be important on neighbouring structures. This explosion needs the intimate mixing and the fine fragmentation between the two liquids. In a stratified vapour explosion, these two liquids are initially superposed and separated by a vapor film. A triggering of the explosion can induce a propagation of this along the film. A study of experimental results and existent models has allowed to retain the following main points: - the explosion propagation is due to a pressure wave propagating through the medium; - the mixing is due to the development of Kelvin-Helmholtz instabilities induced by the shear velocity between the two liquids behind the pressure wave. The presence of the vapour in the volatile liquid explains experimental propagation velocity and the velocity difference between the two fluids at the pressure wave crossing. A first model has been proposed by Brayer in 1994 in order to describe the fragmentation and the mixing of the two fluids. Results of the author do not show explosion propagation. We have therefore built a new mixing-fragmentation model based on the atomization phenomenon that develops itself during the pressure wave crossing. We have also taken into account the transient aspect of the heat transfer between fuel drops and the volatile liquid, and elaborated a model of transient heat transfer. These two models have been introduced in a multi-components, thermal, hydraulic code, MC3D. Results of calculation show a qualitative and quantitative agreement with experimental results and confirm basic options of the model. (author)

  9. Pengembangan Perangkat Pembelajaran Geometri Ruang dengan Model Proving Theorem

    Directory of Open Access Journals (Sweden)

    Bambang Eko Susilo

    2016-03-01

    Full Text Available Kemampuan berpikir kritis dan kreatif mahasiswa masih lemah. Hal ini ditemukan pada mahasiswa yang mengambil mata kuliah Geometri Ruang yaitu dalam membuktikan soal-soal pembuktian (problem to proof. Mahasiswa masih menyelesaikan secara algoritmik atau prosedural sehingga diperlukan pengembangan perangkat pembelajaran Geometri Ruang berbasis kompetensi dan konservasi dengan model Proving Theorem. Dalam penelitian ini perangkat perkuliahan yang dikembangkan yaitu Silabus, Satuan Acara Perkuliahan (SAP, Kontrak Perkuliahan, Media Pembelajaran, Bahan Ajar, Tes UTS dan UAS serta Angket Karakter Konservasi telah dilaksanakan dengan baik dengan kriteria (1 validasi perangkat pembelajaran mata kuliah Geometri ruang berbasis kompetensi dan konservasi dengan model proving theorem berkategori baik dan layak digunakan dan (2 keterlaksanaan RPP pada pembelajaran yang dikembangkan secara keseluruhan berkategori baik.Critical and creative thinking abilities of students still weak. It is found in students who take Space Geometry subjects that is in solving problems to to prove. Students still finish in algorithmic or procedural so that the required the development of Space Geometry learning tools based on competency and conservation with Proving Theorem models. This is a research development which refers to the 4-D models that have been modified for the Space Geometry learning tools, second semester academic year 2014/2015. Instruments used include validation sheet, learning tools and character assessment questionnaire. In this research, the learning tools are developed, namely Syllabus, Lesson Plan, Lecture Contract, Learning Media, Teaching Material, Tests, and Character Conservation Questionnaire had been properly implemented with the criteria (1 validation of Space Geometry learning tools based on competency and conservation with Proving Theorem models categorized good and feasible to use, and (2 the implementation of Lesson Plan on learning categorized

  10. Hurwitz numbers, matrix models and enumerative geometry

    CERN Document Server

    Bouchard, Vincent

    2007-01-01

    We propose a new, conjectural recursion solution for Hurwitz numbers at all genera. This conjecture is based on recent progress in solving type B topological string theory on the mirrors of toric Calabi-Yau manifolds, which we briefly review to provide some background for our conjecture. We show in particular how this B-model solution, combined with mirror symmetry for the one-leg, framed topological vertex, leads to a recursion relation for Hodge integrals with three Hodge class insertions. Our conjecture in Hurwitz theory follows from this recursion for the framed vertex in the limit of infinite framing.

  11. Fuel management optimization in pressure water reactors with hexagonal geometry using hill climbing method

    International Nuclear Information System (INIS)

    Andres Diaz, J.; Quintero, Ruben; Melian, Manuel; Rosete, Alejandro

    2000-01-01

    In this work the general-purpose optimization method, Hill Climbing, was applied to the Fuel Management Optimization problem in PWR reactors, WWER type. They were carried out a series of experiments in order to study the performance of Hill Climbing. It was proven two starting point for initialize the search: a reload configuration by project and a reload configuration generated with the application of a minimal knowledge of the problem. It was also studied the effect of imposing constraints based on the physics of the reactor in order to reduce the number of possible solutions to be generated. The operator used in Hill Climbing was defined as a binary exchange of fuel assemblies. For the simulation of each generated configuration, the tridimensional simulator program SPPS-1 was used. It was formulated an objective function with power peaking constraint to guide the search. As results, a methodology ws proposed for the In-core Fuel Management Optimization in hexagonal geometry, and the feasibility of the application of the Hill Climbing to this type of problem was demonstrated. (author)

  12. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  13. Modelling and simulation of gas explosions in complex geometries

    Energy Technology Data Exchange (ETDEWEB)

    Saeter, Olav

    1998-12-31

    This thesis presents a three-dimensional Computational Fluid Dynamics (CFD) code (EXSIM94) for modelling and simulation of gas explosions in complex geometries. It gives the theory and validates the following sub-models : (1) the flow resistance and turbulence generation model for densely packed regions, (2) the flow resistance and turbulence generation model for single objects, and (3) the quasi-laminar combustion model. It is found that a simple model for flow resistance and turbulence generation in densely packed beds is able to reproduce the medium and large scale MERGE explosion experiments of the Commission of European Communities (CEC) within a band of factor 2. The model for a single representation is found to predict explosion pressure in better agreement with the experiments with a modified k-{epsilon} model. This modification also gives a slightly improved grid independence for realistic gas explosion approaches. One laminar model is found unsuitable for gas explosion modelling because of strong grid dependence. Another laminar model is found to be relatively grid independent and to work well in harmony with the turbulent combustion model. The code is validated against 40 realistic gas explosion experiments. It is relatively grid independent in predicting explosion pressure in different offshore geometries. It can predict the influence of ignition point location, vent arrangements, different geometries, scaling effects and gas reactivity. The validation study concludes with statistical and uncertainty analyses of the code performance. 98 refs., 96 figs, 12 tabs.

  14. DIGA/NSL new calculational model in slab geometry

    International Nuclear Information System (INIS)

    Makai, M.; Gado, J.; Kereszturi, A.

    1987-04-01

    A new calculational model is presented based on a modified finite-difference algorithm, in which the coefficients are determined by means of the so-called gamma matrices. The DIGA program determines the gamma matrices and the NSL program realizes the modified finite difference model. Both programs assume slab cell geometry, DIGA assumes 2 energy groups and 3 diffusive regions. The DIGA/NSL programs serve to study the new calculational model. (author)

  15. A survey on the geometry of production models in economics

    Directory of Open Access Journals (Sweden)

    Alina-Daniela Vîlcu

    2017-01-01

    Full Text Available In this article we survey selected recent results on the geometry of production models, focussing on the main production functions that are usually analyzed in economics, namely homogeneous, homothetic, quasi-sum and quasi-product production functions.

  16. Stochastic geometry, spatial statistics and random fields models and algorithms

    CERN Document Server

    2015-01-01

    Providing a graduate level introduction to various aspects of stochastic geometry, spatial statistics and random fields, this volume places a special emphasis on fundamental classes of models and algorithms as well as on their applications, for example in materials science, biology and genetics. This book has a strong focus on simulations and includes extensive codes in Matlab and R, which are widely used in the mathematical community. It can be regarded as a continuation of the recent volume 2068 of Lecture Notes in Mathematics, where other issues of stochastic geometry, spatial statistics and random fields were considered, with a focus on asymptotic methods.

  17. Physical model of reactor pulse

    International Nuclear Information System (INIS)

    Petrovic, A.; Ravnik, M.

    2004-01-01

    Pulse experiments have been performed at J. Stefan Institute TRIGA reactor since 1991. In total, more than 130 pulses have been performed. Extensive experimental information on the pulse physical characteristics has been accumulated. Fuchs-Hansen adiabatic model has been used for predicting and analysing the pulse parameters. The model is based on point kinetics equation, neglecting the delayed neutrons and assuming constant inserted reactivity in form of step function. Deficiencies of the Fuchs-Hansen model and systematic experimental errors have been observed and analysed. Recently, the pulse model was improved by including the delayed neutrons and time dependence of inserted reactivity. The results explain the observed non-linearity of the pulse energy for high pulses due to finite time of pulse rod withdrawal and the contribution of the delayed neutrons after the prompt part of the pulse. The results of the improved model are in good agreement with experimental results. (author)

  18. A Wear Geometry Model of Plain Woven Fabric Composites

    Directory of Open Access Journals (Sweden)

    Gu Dapeng

    2014-09-01

    Full Text Available The paper g describes a model meant for analysis of the wear geometry of plain woven fabric composites. The referred model consists of a mathematical description of plain woven fabric based on Peirce’s model coupled with a stratified method for the solution of the wear geometry. The evolutions of the wear area ratio of weft yarn, warp yarn and matrix resin on the worn surface are simulated by MatLab software in combination of warp and weft yarn diameters, warp and weft yarn-to-yarn distances, fabric structure phases (SPs. By comparing theoretical and experimental results from the PTFE/Kevlar fabric wear experiment, it can be concluded that the model can present a trend of the component area ratio variations along with the thickness of fabric, but has a inherently large error in quantitative analysis as an idealized model.

  19. Empirical intrinsic geometry for nonlinear modeling and time series filtering.

    Science.gov (United States)

    Talmon, Ronen; Coifman, Ronald R

    2013-07-30

    In this paper, we present a method for time series analysis based on empirical intrinsic geometry (EIG). EIG enables one to reveal the low-dimensional parametric manifold as well as to infer the underlying dynamics of high-dimensional time series. By incorporating concepts of information geometry, this method extends existing geometric analysis tools to support stochastic settings and parametrizes the geometry of empirical distributions. However, the statistical models are not required as priors; hence, EIG may be applied to a wide range of real signals without existing definitive models. We show that the inferred model is noise-resilient and invariant under different observation and instrumental modalities. In addition, we show that it can be extended efficiently to newly acquired measurements in a sequential manner. These two advantages enable us to revisit the Bayesian approach and incorporate empirical dynamics and intrinsic geometry into a nonlinear filtering framework. We show applications to nonlinear and non-Gaussian tracking problems as well as to acoustic signal localization.

  20. The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

    International Nuclear Information System (INIS)

    Duerigen, S.; Fridman, E.

    2011-01-01

    DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

  1. Physical modeling and numerical simulation of subcooled boiling in one- and three-dimensional representation of bundle geometry

    International Nuclear Information System (INIS)

    Bottoni, M.; Lyczkowski, R.; Ahuja, S.

    1995-01-01

    Numerical simulation of subcooled boiling in one-dimensional geometry with the Homogeneous Equilibrium Model (HEM) may yield difficulties related to the very low sonic velocity associated with the HEM. These difficulties do not arise with subcritical flow. Possible solutions of the problem include introducing a relaxation of the vapor production rate. Three-dimensional simulations of subcooled boiling in bundle geometry typical of fast reactors can be performed by using two systems of conservation equations, one for the HEM and the other for a Separated Phases Model (SPM), with a smooth transition between the two models

  2. MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport

    International Nuclear Information System (INIS)

    Huria, H.C.

    1985-01-01

    1 - Description of problem or function: MURLI is an integral transport theory code to calculate fluxes and reaction rates in one- dimensional cylindrical geometry lattice cells of a thermal reactor. For a specified buckling, it computes k-effective using few-group diffusion theory and a few-group collapsed set of Cross sections. The code can optionally be used to solve a first order differential equation for the number density of fissile, fertile and fission product nuclei as a function of time, and to recalculate fluxes, reaction rates and k-effective at different stages of burnup. A 27-group cross section data library is included. There are four pseudo-fission products each associated with the decay chains of plutonium and uranium isotopes in addition to Rh-105, Xe-135, Np-239, U-236, Am-241, Am-242 and Am-243. There is also data for one lumped pseudo-fission product. 2 - Method of solution: Multiple collision probabilities and escape probabilities are calculated for each cylindrical shell region assuming protons are born uniformly and isotropically over the entire region volume. The equations of integral transport theory can then be solved for neutron flux. The first order differential burnup equation is solved by a fourth order Runge-Kutta method. 3 - Restrictions on the complexity of the problem: There are maxima of 8 fissionable elements, 8 resonant elements, and 20 spatial regions

  3. Slab2 - Updated Subduction Zone Geometries and Modeling Tools

    Science.gov (United States)

    Moore, G.; Hayes, G. P.; Portner, D. E.; Furtney, M.; Flamme, H. E.; Hearne, M. G.

    2017-12-01

    The U.S. Geological Survey database of global subduction zone geometries (Slab1.0), is a highly utilized dataset that has been applied to a wide range of geophysical problems. In 2017, these models have been improved and expanded upon as part of the Slab2 modeling effort. With a new data driven approach that can be applied to a broader range of tectonic settings and geophysical data sets, we have generated a model set that will serve as a more comprehensive, reliable, and reproducible resource for three-dimensional slab geometries at all of the world's convergent margins. The newly developed framework of Slab2 is guided by: (1) a large integrated dataset, consisting of a variety of geophysical sources (e.g., earthquake hypocenters, moment tensors, active-source seismic survey images of the shallow slab, tomography models, receiver functions, bathymetry, trench ages, and sediment thickness information); (2) a dynamic filtering scheme aimed at constraining incorporated seismicity to only slab related events; (3) a 3-D data interpolation approach which captures both high resolution shallow geometries and instances of slab rollback and overlap at depth; and (4) an algorithm which incorporates uncertainties of contributing datasets to identify the most probable surface depth over the extent of each subduction zone. Further layers will also be added to the base geometry dataset, such as historic moment release, earthquake tectonic providence, and interface coupling. Along with access to several queryable data formats, all components have been wrapped into an open source library in Python, such that suites of updated models can be released as further data becomes available. This presentation will discuss the extent of Slab2 development, as well as the current availability of the model and modeling tools.

  4. Generalized complex geometry, generalized branes and the Hitchin sigma model

    International Nuclear Information System (INIS)

    Zucchini, Roberto

    2005-01-01

    Hitchin's generalized complex geometry has been shown to be relevant in compactifications of superstring theory with fluxes and is expected to lead to a deeper understanding of mirror symmetry. Gualtieri's notion of generalized complex submanifold seems to be a natural candidate for the description of branes in this context. Recently, we introduced a Batalin-Vilkovisky field theoretic realization of generalized complex geometry, the Hitchin sigma model, extending the well known Poisson sigma model. In this paper, exploiting Gualtieri's formalism, we incorporate branes into the model. A detailed study of the boundary conditions obeyed by the world sheet fields is provided. Finally, it is found that, when branes are present, the classical Batalin-Vilkovisky cohomology contains an extra sector that is related non trivially to a novel cohomology associated with the branes as generalized complex submanifolds. (author)

  5. Micromagnetic recording model of writer geometry effects at skew

    Science.gov (United States)

    Plumer, M. L.; Bozeman, S.; van Ek, J.; Michel, R. P.

    2006-04-01

    The effects of the pole-tip geometry at the air-bearing surface on perpendicular recording at a skew angle are examined through modeling and spin-stand test data. Head fields generated by the finite element method were used to record transitions within our previously described micromagnetic recording model. Write-field contours for a variety of square, rectangular, and trapezoidal pole shapes were evaluated to determine the impact of geometry on field contours. Comparing results for recorded track width, transition width, and media signal to noise ratio at 0° and 15° skew demonstrate the benefits of trapezoidal and reduced aspect-ratio pole shapes. Consistency between these modeled results and test data is demonstrated.

  6. Modeling a Packed Bed Reactor Utilizing the Sabatier Process

    Science.gov (United States)

    Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

    2017-01-01

    A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

  7. The AFEN Method in Cylindrical (r,θ ,z) Geometry for Pebble Bed Reactors -Incorporation of Acceleration and Discontinuity Factor

    International Nuclear Information System (INIS)

    Lee, Jaejun; Cho, Namzin

    2007-01-01

    Most existing methods of nuclear design analysis for pebble bed reactors (PBRs) are based on old finite difference solvers or on statistical methods. These methods require very long computer times. Therefore, there is strong desire of making available high fidelity coarse-mesh nodal computer codes. Recently, we extended the analytic function expansion nodal (AFEN) method developed quite extensively in Cartesian (x,y,z) geometry and in hexagonal-z geometry to the treatment of the full three dimensional cylindrical (r,θ,z) geometry for pebble bed reactors(PBRs). The AFEN methodology in this geometry as in hexagonal geometry is 'robust', due to the unique feature of the AFEN method that it does not use the transverse integration. This paper presents an acceleration scheme based on the coarse-group rebalance (CGR) concept and provides test results verifying the method and its implementation in the TOPS code. Also, we implemented discontinuity factors in the TOPS code and tested on benchmark problems. The TOPS results are in excellent agreement with those of the VENTURE code, using significantly less computer time

  8. External exposure model for various geometries of contaminated materials

    International Nuclear Information System (INIS)

    LePoire, D.; Kamboj, S.; Yu, C.

    1996-01-01

    A computational model for external exposure was developed for the U.S. Department of Energy's residual radioactive material guideline computer code (RESRAD) on the basis of dose coefficients from Federal Guidance Report No. 12 and the point-kernel method. This model includes the effects of different materials and exposure distances, as well as source geometry (cover thickness, source depth, area, and shape). A material factor is calculated on the basis of the point-kernel method using material-specific photon cross-section data and buildup factors. This present model was incorporated into RESRAD-RECYCLE (a RESRAD family code used for computing radiological impacts of metal recycling) and is being incorporated into RESRAD-BUILD (a DOE recommended code for computing impacts of building decontamination). The model was compared with calculations performed with the Monte Carlo N-Particle Code (MCNP) and the Microshield code for three different source geometries, three different radionuclides ( 234 U, 238 U, and 60 Co, representing low, medium, and high energy, respectively), and five different source materials (iron, copper, aluminum, water, and soil). The comparison shows that results of this model are in very good agreement with MCNP calculations (within 5% for 60 Co and within 30% for 238 U and 234 U for all materials and source geometries). Significant differences (greater than 100%) were observed with Microshield for thin 234 U sources

  9. Mathematical model of geometry and fibrous structure of the heart.

    Science.gov (United States)

    Nielsen, P M; Le Grice, I J; Smaill, B H; Hunter, P J

    1991-04-01

    We developed a mathematical representation of ventricular geometry and muscle fiber organization using three-dimensional finite elements referred to a prolate spheroid coordinate system. Within elements, fields are approximated using basis functions with associated parameters defined at the element nodes. Four parameters per node are used to describe ventricular geometry. The radial coordinate is interpolated using cubic Hermite basis functions that preserve slope continuity, while the angular coordinates are interpolated linearly. Two further nodal parameters describe the orientation of myocardial fibers. The orientation of fibers within coordinate planes bounded by epicardial and endocardial surfaces is interpolated linearly, with transmural variation given by cubic Hermite basis functions. Left and right ventricular geometry and myocardial fiber orientations were characterized for a canine heart arrested in diastole and fixed at zero transmural pressure. The geometry was represented by a 24-element ensemble with 41 nodes. Nodal parameters fitted using least squares provided a realistic description of ventricular epicardial [root mean square (RMS) error less than 0.9 mm] and endocardial (RMS error less than 2.6 mm) surfaces. Measured fiber fields were also fitted (RMS error less than 17 degrees) with a 60-element, 99-node mesh obtained by subdividing the 24-element mesh. These methods provide a compact and accurate anatomic description of the ventricles suitable for use in finite element stress analysis, simulation of cardiac electrical activation, and other cardiac field modeling problems.

  10. Moritz enhancements for visualization of complicated geometry models

    International Nuclear Information System (INIS)

    Van Riper, K. A.

    2009-01-01

    We describe new features implemented in the Moritz geometry editing and visualization program to enhance the accuracy and efficiency of viewing complex geometry models. The 3D display is based on OpenGL and requires conversion of the combinatorial surface and solid body geometry used by MCNP and other transport codes to a set of polygons. Calculation of those polygons can take many minutes for complex models. Once calculated, the polygons can be saved to a file and reused when the same or a derivative model is loaded; the file can be read and processed in under a second. A cell can be filled with a collection of other cells constituting a universe. A new option bypasses use of the filled cell's boundaries when calculating the polygons for the filling universe. This option, when applicable, speeds processing, improves the 3D image, and permits reuse of the universe's polygons when other cells are filled with transformed instances of the universe. Surfaces and solid bodies used in a cell description must be converted to polygons before calculating the polygonal representation of a cell; this conversion requires truncation of infinite surfaces. A new method for truncating transformed surfaces ensures the finite surface intersects the entire model. When a surface or solid body is processed in a cell description, an optional test detects when that object does not contribute additional polygons; if so, that object May be extraneous for the cell description. (authors)

  11. ORIGAMI -- The Oak Ridge Geometry Analysis and Modeling Interface

    International Nuclear Information System (INIS)

    Burns, T.J.

    1996-01-01

    A revised ''ray-tracing'' package which is a superset of the geometry specifications of the radiation transport codes MORSE, MASH (GIFT Versions 4 and 5), HETC, and TORT has been developed by ORNL. Two additional CAD-based formats are also included as part of the superset: the native format of the BRL-CAD system--MGED, and the solid constructive geometry subset of the IGES specification. As part of this upgrade effort, ORNL has designed an Xwindows-based utility (ORIGAMI) to facilitate the construction, manipulation, and display of the geometric models required by the MASH code. Since the primary design criterion for this effort was that the utility ''see'' the geometric model exactly as the radiation transport code does, ORIGAMI is designed to utilize the same ''ray-tracing'' package as the revised version of MASH. ORIGAMI incorporates the functionality of two previously developed graphical utilities, CGVIEW and ORGBUG, into a single consistent interface

  12. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  13. FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1980-01-01

    1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can

  14. Geometry Modeling Program Implementation of Human Hip Tissue

    Directory of Open Access Journals (Sweden)

    WANG Mo-nan

    2017-10-01

    Full Text Available Abstract:Aiming to design a simulate software of human tissue modeling and analysis,Visual Studio 2010 is selected as a development tool to develop a 3 D reconstruction software of human tissue with language C++.It can be used alone. It also can be a module of the virtual surgery systems. The system includes medical image segmentation modules and 3 D reconstruction modules,and can realize the model visualization. This software system has been used to reconstruct hip muscles,femur and hip bone accurately. The results show these geometry models can simulate the structure of hip tissues.

  15. Geometry Modeling Program Implementation of Human Hip Tissue

    Directory of Open Access Journals (Sweden)

    WANG Monan

    2017-04-01

    Full Text Available Aiming to design a simulate software of human tissue modeling and analysis,Visual Studio 2010 is selected as a development tool to develop a 3 D reconstruction software of human tissue with language C++.It can be used alone. It also can be a module of the virtual surgery systems. The system includes medical image segmentation modules and 3 D reconstruction modules,and can realize the model visualization. This software system has been used to reconstruct hip muscles,femur and hip bone accurately. The results show these geometry models can simulate the structure of hip tissues.

  16. Application of advanced model of radiative heat transfer in a rod geometry to QUENCH and PARAMETER tests

    International Nuclear Information System (INIS)

    Vasiliev, A.D.; Kobelev, G.V.; Astafieva, V.O.

    2007-01-01

    Radiative heat transfer is very important in different fields of mechanical engineering and related technologies including nuclear reactors, heat transfer in furnaces, aerospace, different high-temperature assemblies. In particular, in the course of a hypothetical severe accident at PWR-type nuclear reactor the temperatures inside the reactor vessel reach high values at which taking into account of radiative heat exchange between the structures of reactor (including core and other reactor vessel elements) gets important. Radiative heat transfer dominates the late phase of severe accident because radiative heat fluxes (proportional to T4, where T is the temperature) are generally considerably higher than convective and conductive heat fluxes in a system. In particular, heat transfer due to radiation determines the heating and degradation of the core and surrounding steel in-vessel structures and finally influences the composition, temperature and mass of materials pouring out of the reactor vessel after its loss of integrity. Existing models of radiative heat exchange use many limitations and approximations: approximate estimation of view factors and beam lengths; the geometry change in the course of the accident is neglected; the database for emissivities of materials is not complete; absorption/emission by steam-noncondensable medium is taken into account approximately. The module MRAD was developed in this paper to model the radiative heat exchange in rod-like geometry typical of PWR-type reactor. Radiative heat exchange is computed using dividing on zones (zonal method) as in existing radiation models implemented to severe accident numerical codes such as ICARE, SCDAP/RELAP, MELCOR but improved in following aspects: new approach to evaluation of view factors and mean beam length; detailed evaluation of gas absorptivity and emissivity; account of effective radiative thermal conductivity for the large core; account of geometry modification in the course of severe

  17. Modeling flow for modified concentric cylinder rheometer geometry

    Science.gov (United States)

    Ekeruche, Karen; Connelly, Kelly; Kavehpour, H. Pirouz

    2016-11-01

    Rheology experiments on biological fluids can be difficult when samples are limited in volume, sensitive to degradation, and delicate to extract from tissues. A probe-like geometry has been developed to perform shear creep experiments on biological fluids and to use the creep response to characterize fluid material properties. This probe geometry is a modified concentric cylinder setup, where the gap is large and we assume the inner cylinder rotates in an infinite fluid. To validate this assumption we perform shear creep tests with the designed probe on Newtonian and non-Newtonian fluids and vary the outer cylinder container diameter. We have also created a numerical model based on the probe geometry setup to compare with experimental results at different outer cylinder diameters. A creep test is modeled by applying rotation to the inner cylinder and solving for the deformation of the fluid throughout the gap. Steady state viscosity values are calculated from creep compliance curves and compared between experimental and numerical results.

  18. Continuous energy Monte Carlo calculations for randomly distributed spherical fuels based on statistical geometry model

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Isao [Osaka Univ., Suita (Japan); Mori, Takamasa; Nakagawa, Masayuki; Itakura, Hirofumi

    1996-03-01

    The method to calculate neutronics parameters of a core composed of randomly distributed spherical fuels has been developed based on a statistical geometry model with a continuous energy Monte Carlo method. This method was implemented in a general purpose Monte Carlo code MCNP, and a new code MCNP-CFP had been developed. This paper describes the model and method how to use it and the validation results. In the Monte Carlo calculation, the location of a spherical fuel is sampled probabilistically along the particle flight path from the spatial probability distribution of spherical fuels, called nearest neighbor distribution (NND). This sampling method was validated through the following two comparisons: (1) Calculations of inventory of coated fuel particles (CFPs) in a fuel compact by both track length estimator and direct evaluation method, and (2) Criticality calculations for ordered packed geometries. This method was also confined by applying to an analysis of the critical assembly experiment at VHTRC. The method established in the present study is quite unique so as to a probabilistic model of the geometry with a great number of spherical fuels distributed randomly. Realizing the speed-up by vector or parallel computations in future, it is expected to be widely used in calculation of a nuclear reactor core, especially HTGR cores. (author).

  19. CALIPSO - a computer code for the calculation of fluiddynamics, thermohydraulics and changes of geometry in failing fuel elements of a fast breeder reactor

    International Nuclear Information System (INIS)

    Kedziur, F.

    1982-07-01

    The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.) [de

  20. Model of the final borehole geometry for helical laser drilling

    Science.gov (United States)

    Kroschel, Alexander; Michalowski, Andreas; Graf, Thomas

    2018-05-01

    A model for predicting the borehole geometry for laser drilling is presented based on the calculation of a surface of constant absorbed fluence. It is applicable to helical drilling of through-holes with ultrashort laser pulses. The threshold fluence describing the borehole surface is fitted for best agreement with experimental data in the form of cross-sections of through-holes of different shapes and sizes in stainless steel samples. The fitted value is similar to ablation threshold fluence values reported for laser ablation models.

  1. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

    The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

  2. Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched (<20% {sup 235}U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

  3. Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

    International Nuclear Information System (INIS)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S.

    2011-01-01

    A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched ( 235 U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

  4. Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

    International Nuclear Information System (INIS)

    Koch, R.; Findeisen, A.

    1986-04-01

    The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

  5. Bond graph modeling of nuclear reactor dynamics

    International Nuclear Information System (INIS)

    Tylee, J.L.

    1981-01-01

    A tenth-order linear model of a pressurized water reactor (PWR) is developed using bond graph techniques. The model describes the nuclear heat generation process and the transfer of this heat to the reactor coolant. Comparisons between the calculated model response and test data from a small-scale PWR show the model to be an adequate representation of the actual plant dynamics. Possible application of the model in an advanced plant diagnostic system is discussed

  6. COSTANZA-AX, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Axial Geometry. COSTANZA-CYL, 1-D Neutronics and Thermodynamics of Liquid Cooled Reactor in Cylindrical Geometry

    International Nuclear Information System (INIS)

    Agazzi, A.; Forti, G.; Vincenti, E.

    1984-01-01

    1 - Nature of physical problem solved: Purpose of the programmes is to study reactor dynamics, considering the variation of the spatial flux distribution. The two programmes COSTANZA-CYL and COSTANZA-AX, solve the kinetics diffusion equations in two groups and one dimension (plane geometry for COSTANZA-AX, radial geometry for COSTANZA-CYL). The neutronic calculation is coupled with the calculation of the heat transmission from the fuel to the cladding and to the coolant, and with the thermo-hydraulics of channels with forced circulation of liquid coolant. The geometry of fuel element and channel may be cylindrical or slab. Up to ten groups of delayed neutrons are allowed. Temperature feedback of fuel (Doppler) and coolant are considered independently and affect the nuclear constants. Control rod movement or diffused poison concentrations are simulated by externally imposed variations of the thermal absorption cross section in the different regions of the reactors. Inlet temperatures and mass flow in the coolant channels may be varied according to any externally given time table. 2 - Method of solution: The kinetic diffusion equations in two groups are solved by finite-difference method. 3 - Restrictions on the complexity of the problem: 10 concentric regions; 10 coolant channels; 10 groups of delayed neutrons

  7. Reactor Thermal Hydraulic Numerical Calculation And Modeling

    International Nuclear Information System (INIS)

    Duong Ngoc Hai; Dang The Ba

    2008-01-01

    In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

  8. Modeling moisture ingress through simplified concrete crack geometries

    DEFF Research Database (Denmark)

    Pease, Bradley Justin; Michel, Alexander; Geiker, Mette Rica

    2011-01-01

    , considered to have two parts; 1) a coalesced crack length which behaves as a free-surface for moisture ingress, and 2) an isolated microcracking length which resists ingress similarly to the bulk material. Transport model results are compared to experimental results from steel fibre reinforced concrete wedge......This paper introduces a numerical model for ingress in cracked steel fibre reinforced concrete. Details of a simplified crack are preset in the model’s geometry using the cracked hinge model (CHM). The total crack length estimated using the CHM was, based on earlier work on conventional concrete...... on moisture ingress. Results from the transport model indicate the length of the isolated microcracks was approximately 19 mm for the investigated concrete composition....

  9. Cosmological evolution in a two-brane warped geometry model

    Directory of Open Access Journals (Sweden)

    Sumit Kumar

    2015-07-01

    Full Text Available We study an effective 4-dimensional scalar–tensor field theory, originated from an underlying brane–bulk warped geometry, to explore the scenario of inflation. It is shown that the inflaton potential naturally emerges from the radion energy–momentum tensor which in turn results in an inflationary model of the Universe on the visible brane that is consistent with the recent results from the Planck's experiment. The dynamics of modulus stabilization from the inflaton rolling condition is demonstrated. The implications of our results in the context of recent BICEP2 results are also discussed.

  10. Chemical reactor modeling multiphase reactive flows

    CERN Document Server

    Jakobsen, Hugo A

    2014-01-01

    Chemical Reactor Modeling closes the gap between Chemical Reaction Engineering and Fluid Mechanics.  The second edition consists of two volumes: Volume 1: Fundamentals. Volume 2: Chemical Engineering Applications In volume 1 most of the fundamental theory is presented. A few numerical model simulation application examples are given to elucidate the link between theory and applications. In volume 2 the chemical reactor equipment to be modeled are described. Several engineering models are introduced and discussed. A survey of the frequently used numerical methods, algorithms and schemes is provided. A few practical engineering applications of the modeling tools are presented and discussed. The working principles of several experimental techniques employed in order to get data for model validation are outlined. The monograph is based on lectures regularly taught in the fourth and fifth years graduate courses in transport phenomena and chemical reactor modeling, and in a post graduate course in modern reactor m...

  11. Conductive solar wind models in rapidly diverging flow geometries

    International Nuclear Information System (INIS)

    Holzer, T.E.; Leer, E.

    1980-01-01

    A detailed parameter study of conductive models of the solar wind has been carried out, extending the previous similar studies of Durney (1972) and Durney and Hundhausen (1974) by considering collisionless inhibition of thermal conduction, rapidly diverging flow geometries, and the structure of solutions for the entire n 0 -T 0 plane (n 0 and T 0 are the coronal base density and temperature). Primary emphasis is placed on understanding the complex effects of the physical processes operative in conductive solar wind models. There are five points of particular interest that have arisen from the study: (1) neither collisionless inhibition of thermal conduction nor rapidly diverging flow geometries can significantly increase the solar wind speed at 1 AU; (2) there exists a firm upper limit on the coronal base temperature consistent with observed values of the coronal base pressure and solar wind mass flux density; (3) the principal effect of rapidly diverging flow geometries is a decrease in the solar wind mass flux density at 1 AU and an increase in the mass flux density at the coronal base; (4) collisionless inhibition of thermal conduction can lead to a solar wind flow speed that either increases or decreases with increasing coronal base density (n 0 ) and temperature (T 0 , depending on the region of the n 0 -T 0 plane considered; (5) there is a region of the n 0 -T/sub o/ plane at high coronal base densities where low-speed, high-mass-flux, transonic solar wind flows exist: a region not previously considered

  12. A Geometry Deformation Model for Braided Continuum Manipulators

    Directory of Open Access Journals (Sweden)

    S. M. Hadi Sadati

    2017-06-01

    Full Text Available Continuum manipulators have gained significant attention in the robotic community due to their high dexterity, deformability, and reachability. Modeling of such manipulators has been shown to be very complex and challenging. Despite many research attempts, a general and comprehensive modeling method is yet to be established. In this paper, for the first time, we introduce the bending effect in the model of a braided extensile pneumatic actuator with both stiff and bendable threads. Then, the effect of the manipulator cross-section deformation on the constant curvature and variable curvature models is investigated using simple analytical results from a novel geometry deformation method and is compared to experimental results. We achieve 38% mean reference error simulation accuracy using our constant curvature model for a braided continuum manipulator in presence of body load and 10% using our variable curvature model in presence of extensive external loads. With proper model assumptions and taking to account the cross-section deformation, a 7–13% increase in the simulation mean error accuracy is achieved compared to a fixed cross-section model. The presented models can be used for the exact modeling and design optimization of compound continuum manipulators by providing an analytical tool for the sensitivity analysis of the manipulator performance. Our main aim is the application in minimal invasive manipulation with limited workspaces and manipulators with regional tunable stiffness in their cross section.

  13. A self-regulating model of bedrock river channel geometry

    Science.gov (United States)

    Stark, C. P.

    2006-02-01

    The evolution of many mountain landscapes is controlled by the incision of bedrock river channels. While the rate of incision is set by channel shape through its mediation of flow, the channel shape is itself set by the history of bedrock erosion. This feedback between channel geometry and incision determines the speed of landscape response to tectonic or climatic forcing. Here, a model for the dynamics of bedrock channel shape is derived from geometric arguments, a normal flow approximation for channel flow, and a threshold bed shear stress assumption for bedrock abrasion. The model dynamics describe the competing effects of channel widening, tilting, bending, and variable flow depth. Transient solutions suggest that channels may take ~1-10 ky to adapt to changes in discharge, implying that channel disequilibrium is commonplace. If so, landscape evolution models will need to include bedrock channel dynamics if they are to probe the effects of climate change.

  14. Simulation of hydrogen release and combustion in large scale geometries: models and methods

    International Nuclear Information System (INIS)

    Beccantini, A.; Dabbene, F.; Kudriakov, S.; Magnaud, J.P.; Paillere, H.; Studer, E.

    2003-01-01

    The simulation of H2 distribution and combustion in confined geometries such as nuclear reactor containments is a challenging task from the point of view of numerical simulation, as it involves quite disparate length and time scales, which need to resolved appropriately and efficiently. Cea is involved in the development and validation of codes to model such problems, for external clients such as IRSN (TONUS code), Technicatome (NAUTILUS code) or for its own safety studies. This paper provides an overview of the physical and numerical models developed for such applications, as well as some insight into the current research topics which are being pursued. Examples of H2 mixing and combustion simulations are given. (authors)

  15. Noncommutative geometry and its application to the standard model

    Energy Technology Data Exchange (ETDEWEB)

    Martinetti, Pierre [Georg-August Universitaet, Goettingen (Germany)

    2009-07-01

    We give an overview of the description of the standard model of particle physics minimally coupled to gravity within the framework of noncommutative geometry. Especially we study in detail the metric structure of spacetime that emerges from the spectral triple recently proposed by Chamseddine, Connes and Marcolli. Within this framework points of spacetime acquire an internal structure inherited from the gauge group of the standard model. A distance is defined on this generalized spacetime which is fully encoded by the Yang-Mills gauge fields together with the Higgs field. We focus on some explicit examples, underlying the link between this distance and other distances well known by physicists and mathematicians, such has the Carnot-Caratheodory horizontal distance or the Monge-Kantorovitch transport distance.

  16. A review on research activities using the SANS spectrometer in transmission geometry at ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Adib, M.

    1999-01-01

    The phased double rotor facility operating at ET-RR-1 reactor (2MW) was rearranged to operate as SANS spectrometer in transmission geometry. The rotors are suspended in magnetic fields and are spinning up to 16,000 rpm producing bursts of polyenergetic neutrons with wavelengths from 0.2 nm to 6.5 nm and beam divergence of 17' on the sample. The review on research activities using the SANS spectrometer and its applications for powder particle size determination and the long wavelength fluctuation of magnetization of the Fe-Ni alloys are discussed. (author)

  17. Material and geometry options and performance characteristics for a test reactor

    International Nuclear Information System (INIS)

    Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

    1993-01-01

    For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

  18. Spherical model for superfluidity in a restricted geometry

    International Nuclear Information System (INIS)

    Fishman, S.; Ziman, T.A.L.

    1982-01-01

    The spherical model is solved on a hypercubic lattice in d dimensions, each bond of which is decorated with l spins. The thermodynamic functions and the helicity modulus, analogous to a superfluid density, are calculated. We find that at least two spherical fields are required for the model to exhibit low-temperature properties that can approximate reasonably those of O(n) models. The heuristic prediction that the critical temperature behaves as T/sub c/(l)approx.(l+1) -1 is checked for the model and found to hold quite accurately even for small l(> or approx. =2). The helicity modulus and magnetization of the two-constraint spherical model are found to scale approximately with the critical temperature, but the relation between them is more complex than in the undecorated model. This relation is used to check heuristic arguments concerning the helicity modulus at low temperatures. We comment on the relevance to physical systems, in particular, the problem of boson condensation in a restricted geometry

  19. Stochastic model of energetic nuclear reactor

    International Nuclear Information System (INIS)

    Bojko, R.V.; Ryazanov, V.V.

    2002-01-01

    Behaviour of nuclear reactor was treated using the theory of branching processes. As mathematical model descriptive the neutron number in time the Markov occasional process is proposed. Application of branching occasional processes with variable regime to the description of neutron behaviour in the reactor makes possible conducting strong description of critical operation regime and demonstrates the severity of the process. Three regimes of the critical behaviour depending on the sign of manipulated variables and feedbacks were discovered. Probability regularities peculiar to the behaviour of the reactor are embodied to the suggested stochastic model [ru

  20. A model for nuclear research reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Barati, Ramin, E-mail: Barati.ramin@aut.ac.ir; Setayeshi, Saeed, E-mail: setayesh@aut.ac.ir

    2013-09-15

    Highlights: • A thirty-fourth order model is used to simulate the dynamics of a research reactor. • We consider delayed neutrons fraction as a function of time. • Variable fuel and temperature reactivity coefficients are used. • WIMS, BORGES and CITVAP codes are used for initial condition calculations. • Results are in agreement with experimental data rather than common codes. -- Abstract: In this paper, a useful thirty-fourth order model is presented to simulate the kinetics and dynamics of a research reactor core. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant and fuel temperatures (Doppler effects) with variable reactivity coefficients, xenon, samarium, boron concentration, fuel burn up and thermal hydraulics. WIMS and CITVAP codes are used to extract neutron cross sections and calculate the initial neuron flux respectively. The purpose is to present a model with results similar to reality as much as possible with reducing common simplifications in reactor modeling to be used in different analyses such as reactor control, functional reliability and safety. The model predictions are qualified by comparing with experimental data, detailed simulations of reactivity insertion transients, and steady state for Tehran research reactor reported in the literature and satisfactory results have been obtained.

  1. Modelling aerosol behavior in reactor cooling systems

    International Nuclear Information System (INIS)

    McDonald, B.H.

    1990-01-01

    This paper presents an overview of some of the areas of concern in using computer codes to model fission-product aerosol behavior in the reactor cooling system (RCS) of a water-cooled nuclear reactor during a loss-of-coolant accident. The basic physical processes that require modelling include: fission product release and aerosol formation in the reactor core, aerosol transport and deposition in the reactor core and throughout the rest of the RCS, and the interaction between aerosol transport processes and the thermalhydraulics. In addition to these basic physical processes, chemical reactions can have a large influence on the nature of the aerosol and its behavior in the RCS. The focus is on the physics and the implications of numerical methods used in the computer codes to model aerosol behavior in the RCS

  2. Flow model study of 'Monju' reactor vessel

    International Nuclear Information System (INIS)

    Miyaguchi, Kimihide

    1980-01-01

    In the case of designing the structures in nuclear reactors, various problems to be considered regarding thermo-hydrodynamics exist, such as the distribution of flow quantity and the pressure loss in reactors and the thermal shock to inlet and outlet nozzles. In order to grasp the flow characteristics of coolant in reactors, the 1/2 scale model of the reactor structure of ''Monju'' was attached to the water flow testing facility in the Oarai Engineering Center, and the simulation experiment has been carried out. The flow characteristics in reactors clarified by experiment and analysis so far are the distribution of flow quantity between high and low pressure regions in reactors, the distribution of flow quantity among flow zones in respective regions of high and low pressure, the pressure loss in respective parts in reactors, the flow pattern and the mixing effect of coolant in upper and lower plenums, the effect of the twisting angle of inlet nozzles on the flow characteristics in lower plenums, the effect of internal cylinders on the flow characteristics in upper plenums and so on. On the basis of these test results, the improvement of the design of structures in reactors was made, and the confirmation test on the improved structures was carried out. The testing method, the calculation method, the test results and the reflection to the design of actual machines are described. (Kako, I.)

  3. Modeling of the reactor core

    International Nuclear Information System (INIS)

    1999-01-01

    In order to improve technical - economical parameters fuel with 2.4% enrichment and burnable absorber is started to be used at Ignalina NPP. Using code QUABOX/CUBBOX the main neutronic - physical characteristics were calculated for selected reactor core conditions

  4. Differential geometry based solvation model II: Lagrangian formulation.

    Science.gov (United States)

    Chen, Zhan; Baker, Nathan A; Wei, G W

    2011-12-01

    Solvation is an elementary process in nature and is of paramount importance to more sophisticated chemical, biological and biomolecular processes. The understanding of solvation is an essential prerequisite for the quantitative description and analysis of biomolecular systems. This work presents a Lagrangian formulation of our differential geometry based solvation models. The Lagrangian representation of biomolecular surfaces has a few utilities/advantages. First, it provides an essential basis for biomolecular visualization, surface electrostatic potential map and visual perception of biomolecules. Additionally, it is consistent with the conventional setting of implicit solvent theories and thus, many existing theoretical algorithms and computational software packages can be directly employed. Finally, the Lagrangian representation does not need to resort to artificially enlarged van der Waals radii as often required by the Eulerian representation in solvation analysis. The main goal of the present work is to analyze the connection, similarity and difference between the Eulerian and Lagrangian formalisms of the solvation model. Such analysis is important to the understanding of the differential geometry based solvation model. The present model extends the scaled particle theory of nonpolar solvation model with a solvent-solute interaction potential. The nonpolar solvation model is completed with a Poisson-Boltzmann (PB) theory based polar solvation model. The differential geometry theory of surfaces is employed to provide a natural description of solvent-solute interfaces. The optimization of the total free energy functional, which encompasses the polar and nonpolar contributions, leads to coupled potential driven geometric flow and PB equations. Due to the development of singularities and nonsmooth manifolds in the Lagrangian representation, the resulting potential-driven geometric flow equation is embedded into the Eulerian representation for the purpose of

  5. Indoor Modelling Benchmark for 3D Geometry Extraction

    Science.gov (United States)

    Thomson, C.; Boehm, J.

    2014-06-01

    A combination of faster, cheaper and more accurate hardware, more sophisticated software, and greater industry acceptance have all laid the foundations for an increased desire for accurate 3D parametric models of buildings. Pointclouds are the data source of choice currently with static terrestrial laser scanning the predominant tool for large, dense volume measurement. The current importance of pointclouds as the primary source of real world representation is endorsed by CAD software vendor acquisitions of pointcloud engines in 2011. Both the capture and modelling of indoor environments require great effort in time by the operator (and therefore cost). Automation is seen as a way to aid this by reducing the workload of the user and some commercial packages have appeared that provide automation to some degree. In the data capture phase, advances in indoor mobile mapping systems are speeding up the process, albeit currently with a reduction in accuracy. As a result this paper presents freely accessible pointcloud datasets of two typical areas of a building each captured with two different capture methods and each with an accurate wholly manually created model. These datasets are provided as a benchmark for the research community to gauge the performance and improvements of various techniques for indoor geometry extraction. With this in mind, non-proprietary, interoperable formats are provided such as E57 for the scans and IFC for the reference model. The datasets can be found at: http://indoor-bench.github.io/indoor-bench.

  6. Some simulation aspects, from molecular systems to stochastic geometries of pebble bed reactors

    International Nuclear Information System (INIS)

    Mazzolo, A.

    2009-06-01

    After a brief presentation of his teaching and supervising activities, the author gives an overview of his research activities: investigation of atoms under high intensity magnetic field (investigation of the electronic structure under these fields), studies of theoretical and numerical electrochemistry (simulation coupling molecular dynamics and quantum calculations, comprehensive simulations of molecular dynamics), and studies relating stochastic geometry and neutron science

  7. Solving the neutron diffusion equation on combinatorial geometry computational cells for reactor physics calculations

    International Nuclear Information System (INIS)

    Azmy, Y. Y.

    2004-01-01

    An approach is developed for solving the neutron diffusion equation on combinatorial geometry computational cells, that is computational cells composed by combinatorial operations involving simple-shaped component cells. The only constraint on the component cells from which the combinatorial cells are assembled is that they possess a legitimate discretization of the underlying diffusion equation. We use the Finite Difference (FD) approximation of the x, y-geometry diffusion equation in this work. Performing the same combinatorial operations involved in composing the combinatorial cell on these discrete-variable equations yields equations that employ new discrete variables defined only on the combinatorial cell's volume and faces. The only approximation involved in this process, beyond the truncation error committed in discretizing the diffusion equation over each component cell, is a consistent-order Legendre series expansion. Preliminary results for simple configurations establish the accuracy of the solution to the combinatorial geometry solution compared to straight FD as the system dimensions decrease. Furthermore numerical results validate the consistent Legendre-series expansion order by illustrating the second order accuracy of the combinatorial geometry solution, the same as standard FD. Nevertheless the magnitude of the error for the new approach is larger than FD's since it incorporates the additional truncated series approximation. (authors)

  8. Optimizing multi-pinhole SPECT geometries using an analytical model

    International Nuclear Information System (INIS)

    Rentmeester, M C M; Have, F van der; Beekman, F J

    2007-01-01

    State-of-the-art multi-pinhole SPECT devices allow for sub-mm resolution imaging of radio-molecule distributions in small laboratory animals. The optimization of multi-pinhole and detector geometries using simulations based on ray-tracing or Monte Carlo algorithms is time-consuming, particularly because many system parameters need to be varied. As an efficient alternative we develop a continuous analytical model of a pinhole SPECT system with a stationary detector set-up, which we apply to focused imaging of a mouse. The model assumes that the multi-pinhole collimator and the detector both have the shape of a spherical layer, and uses analytical expressions for effective pinhole diameters, sensitivity and spatial resolution. For fixed fields-of-view, a pinhole-diameter adapting feedback loop allows for the comparison of the system resolution of different systems at equal system sensitivity, and vice versa. The model predicts that (i) for optimal resolution or sensitivity the collimator layer with pinholes should be placed as closely as possible around the animal given a fixed detector layer, (ii) with high-resolution detectors a resolution improvement up to 31% can be achieved compared to optimized systems, (iii) high-resolution detectors can be placed close to the collimator without significant resolution losses, (iv) interestingly, systems with a physical pinhole diameter of 0 mm can have an excellent resolution when high-resolution detectors are used

  9. Unified Stochastic Geometry Model for MIMO Cellular Networks with Retransmissions

    KAUST Repository

    Afify, Laila H.

    2016-10-11

    This paper presents a unified mathematical paradigm, based on stochastic geometry, for downlink cellular networks with multiple-input-multiple-output (MIMO) base stations (BSs). The developed paradigm accounts for signal retransmission upon decoding errors, in which the temporal correlation among the signal-to-interference-plus-noise-ratio (SINR) of the original and retransmitted signals is captured. In addition to modeling the effect of retransmission on the network performance, the developed mathematical model presents twofold analysis unification for MIMO cellular networks literature. First, it integrates the tangible decoding error probability and the abstracted (i.e., modulation scheme and receiver type agnostic) outage probability analysis, which are largely disjoint in the literature. Second, it unifies the analysis for different MIMO configurations. The unified MIMO analysis is achieved by abstracting unnecessary information conveyed within the interfering signals by Gaussian signaling approximation along with an equivalent SISO representation for the per-data stream SINR in MIMO cellular networks. We show that the proposed unification simplifies the analysis without sacrificing the model accuracy. To this end, we discuss the diversity-multiplexing tradeoff imposed by different MIMO schemes and shed light on the diversity loss due to the temporal correlation among the SINRs of the original and retransmitted signals. Finally, several design insights are highlighted.

  10. Unified Stochastic Geometry Model for MIMO Cellular Networks with Retransmissions

    KAUST Repository

    Afify, Laila H.; Elsawy, Hesham; Al-Naffouri, Tareq Y.; Alouini, Mohamed-Slim

    2016-01-01

    This paper presents a unified mathematical paradigm, based on stochastic geometry, for downlink cellular networks with multiple-input-multiple-output (MIMO) base stations (BSs). The developed paradigm accounts for signal retransmission upon decoding errors, in which the temporal correlation among the signal-to-interference-plus-noise-ratio (SINR) of the original and retransmitted signals is captured. In addition to modeling the effect of retransmission on the network performance, the developed mathematical model presents twofold analysis unification for MIMO cellular networks literature. First, it integrates the tangible decoding error probability and the abstracted (i.e., modulation scheme and receiver type agnostic) outage probability analysis, which are largely disjoint in the literature. Second, it unifies the analysis for different MIMO configurations. The unified MIMO analysis is achieved by abstracting unnecessary information conveyed within the interfering signals by Gaussian signaling approximation along with an equivalent SISO representation for the per-data stream SINR in MIMO cellular networks. We show that the proposed unification simplifies the analysis without sacrificing the model accuracy. To this end, we discuss the diversity-multiplexing tradeoff imposed by different MIMO schemes and shed light on the diversity loss due to the temporal correlation among the SINRs of the original and retransmitted signals. Finally, several design insights are highlighted.

  11. Reactor systems modeling for ICF hybrids

    International Nuclear Information System (INIS)

    Berwald, D.H.; Meier, W.R.

    1980-10-01

    The computational models of ICF reactor subsystems developed by LLNL and TRW are described and a computer program was incorporated for use in the EPRI-sponsored Feasibility Assessment of Fusion-Fission Hybrids. Representative parametric variations have been examined. Many of the ICF subsystem models are very preliminary and more quantitative models need to be developed and included in the code

  12. Using 3D Geometric Models to Teach Spatial Geometry Concepts.

    Science.gov (United States)

    Bertoline, Gary R.

    1991-01-01

    An explanation of 3-D Computer Aided Design (CAD) usage to teach spatial geometry concepts using nontraditional techniques is presented. The software packages CADKEY and AutoCAD are described as well as their usefulness in solving space geometry problems. (KR)

  13. Hydrodynamic models for slurry bubble column reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gidaspow, D. [IIT Center, Chicago, IL (United States)

    1995-12-31

    The objective of this investigation is to convert a {open_quotes}learning gas-solid-liquid{close_quotes} fluidization model into a predictive design model. This model is capable of predicting local gas, liquid and solids hold-ups and the basic flow regimes: the uniform bubbling, the industrially practical churn-turbulent (bubble coalescence) and the slugging regimes. Current reactor models incorrectly assume that the gas and the particle hold-ups (volume fractions) are uniform in the reactor. They must be given in terms of empirical correlations determined under conditions that radically differ from reactor operation. In the proposed hydrodynamic approach these hold-ups are computed from separate phase momentum balances. Furthermore, the kinetic theory approach computes the high slurry viscosities from collisions of the catalyst particles. Thus particle rheology is not an input into the model.

  14. Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Oezdemir, Erdal; Moon, Kang Hoon; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Kim, Yongdeog [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES.

  15. ALGOL geometrical module for reactor and reactor cell calculations in the R-Z geometry with the Monte Carlo method

    International Nuclear Information System (INIS)

    Usikov, D.A.

    1975-01-01

    A description of a geometrical module used in a program of the ARMONT complex of the Monte Carlo calculations is given. The geometrical module is designed to simulate the particle trajectory in the R-Z geometry. The geometrical module follows the particle trajectory from the start point to the next collision or flight-out points. The flight direction at the scattering point is assumed isotropic in the laboratory coordinate system. In the module the angle between the flight direction before and after collision is not determined. The principles for the module construction are presented alongside with the text-module in the ALGOL language. The module is optimumized as to the counting rate and it is rather compact not to cause difficulties due to the translator limitations in common translation with other program blocks based on the use of the Monte Carlo calculations

  16. Hydrodynamic model of hydrogen-flame propagation in reactor vessels

    International Nuclear Information System (INIS)

    Baer, M.R.; Ratzel, A.C.

    1982-01-01

    A hydrodynamic model for hydrogen flame propagation in reactor geometries is presented. This model is consistent with the theory of slow combustion in which the gasdynamic field equations are treated in the limit of small Mach numbers. To the lowest order, pressure is spatially uniform. The flame is treated as a density and entropy discontinuity which propagates at prescribed burning velocities, corresponding to laminar or turbulent flames. Radiation cooling of the burned combustion gases and possible preheating of the unburned gases during propagation of the flame is included using a molecular gas-band thermal radiation model. Application of this model has been developed for 1-D variable area flame propagation. Multidimensional effects induced by hydrodynamics and buoyancy are introduced as a correction to the burn velocity (which reflects a modification of planar flame surface to a distorted surface) using experimentally measured pressure-rise time data for hydrogen/air deflagrations in cylindrical vessels

  17. A reduced fidelity model for the rotary chemical looping combustion reactor

    KAUST Repository

    Iloeje, Chukwunwike O.

    2017-01-11

    The rotary chemical looping combustion reactor has great potential for efficient integration with CO capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel fraction for the selected oxygen carrier materials.

  18. Statistical modelling of railway track geometry degradation using Hierarchical Bayesian models

    International Nuclear Information System (INIS)

    Andrade, A.R.; Teixeira, P.F.

    2015-01-01

    Railway maintenance planners require a predictive model that can assess the railway track geometry degradation. The present paper uses a Hierarchical Bayesian model as a tool to model the main two quality indicators related to railway track geometry degradation: the standard deviation of longitudinal level defects and the standard deviation of horizontal alignment defects. Hierarchical Bayesian Models (HBM) are flexible statistical models that allow specifying different spatially correlated components between consecutive track sections, namely for the deterioration rates and the initial qualities parameters. HBM are developed for both quality indicators, conducting an extensive comparison between candidate models and a sensitivity analysis on prior distributions. HBM is applied to provide an overall assessment of the degradation of railway track geometry, for the main Portuguese railway line Lisbon–Oporto. - Highlights: • Rail track geometry degradation is analysed using Hierarchical Bayesian models. • A Gibbs sampling strategy is put forward to estimate the HBM. • Model comparison and sensitivity analysis find the most suitable model. • We applied the most suitable model to all the segments of the main Portuguese line. • Tackling spatial correlations using CAR structures lead to a better model fit

  19. Observations of the behaviour of gas in the wake behind a corner blockage in fast breeder reactor subassembly geometry

    International Nuclear Information System (INIS)

    Fukuzawa, Y.

    1979-07-01

    Observations were made of gas behaviour in the wake behind a 21% corner blockage in the subassembly geometry of a liquid metal fast breeder reactor. The test section used represented one half of the reactor fuel subassembly, divided along the vertical plane of symmetry through the blockage. A glass wall occupied the position of this plane. Water was allowed to flow between glass rods simulating fuel pins, the velocity being changed from 1.2 to 4.5 m/s. Argon was injected into the wake or into the flow upstream of the blockage, the injection rate being changed from 1 to 230 Ncm 3 /s (standard temperature and pressure). From the present experiment, the following is evident: The gas is accumulated in the wake behind the blockage, forming a gas cavity. The flow patterns of the two-phase mixture in the wake are classified into three types, depending on the liquid velocity. In the lower velocity range, a gas cavity cannot be present at rest, rising up through the wake as a single bubble due to buoyancy. In the higher velocity range, the gas cavity is broken up by the liquid flow forces, only small gas bubbles circulating in the wake. In the velocity range in between, the gas cavity is present in the wake. The cavity size depends on the gas injection rate and on the liquid velocity. From the results, the possibility of fuel failure caused by fission gas release at a blockage in the fast breeder reactor can be considered to depend on the operating conditions of the reactor, specially on the coolant velocity. (orig.) [de

  20. Computational modelling for diffusion of neutrons problems inside nuclear multiplying medium on bidimensional cartesian rectangular geometry

    International Nuclear Information System (INIS)

    Couto, Nozimar do

    2003-01-01

    Diffusion theory is traditionally applied to nuclear reactor global calculations. Based on the good results generated by the one-dimensional spectral nodal diffusion (SND) method for benchmark problems, we offer the SND method for nuclear reactor global calculations in X,Y geometry. In this method, the continuity equation and Flick law are transverse integrated in each spatial direction leading to a system of two 'one-dimensional' equations coupled by the transverse leakage terms. We then apply the SND method to numerically solve this system with constant approximations for the transverse leakage terms. We perform a spectral analysis to determine the local general solution of each 'one-dimensional' nodal equation with flat approximation for the transverse leakages. We used special auxiliary equations with parameters that are to be determined in order to preserve the analytical general solutions in the numerical algorithm. By considering continuity conditions at the node interfaces and appropriate boundary conditions, we obtain a solvable system of discretized equations involving the node-edge average scalar fluxes at each estimate of the dominant eigenvalue (k eff ) in the outer power iterations. As we considered approximations to the transverse leakages, the SND method is not free of spatial truncation errors. Nevertheless, it generated good results for the typical model problems that we considered. (author)

  1. Investigation of fluid flow in various geometries related to nuclear reactor using PIV system

    International Nuclear Information System (INIS)

    Kansal, A.K.; Maheshwari, N.K.; Singh, R.K.; Vijayan, P.K.; Saha, D.; Singh, R.K.; Joshi, V.M.

    2011-01-01

    Particle Image Velocimetry (PIV) is a non-intrusive technique for simultaneously measuring the velocities at many points in a fluid flow. The PIV system used is comprised of Nd:YAG laser source, CCD (Charged Coupled Device) camera, timing controller (to control the laser and camera) and software used for analyzing the flow velocities. Several case studies related to nuclear reactor were performed with the PIV system. Some of the cases like flow in circular tube, submerged jet, natural convection in a water pool, flow field of moderator inlet diffuser of 500 MWe Pressurised Heavy Water Reactor (PHWR) and fluidic flow control device (FFCD) used in advanced accumulator of Emergency Core Cooling System (ECCS) have been studied using PIV system. Theoretical studies have been performed and comparisons with PIV results are also given in the present studies. (author)

  2. Contribution to analytical theory of neutron resonance absorption in heterogeneous reactor systems with cylindrical geometry

    International Nuclear Information System (INIS)

    Slipicevic, K.

    1968-12-01

    Following a review of the existing theories od resonance absorption this thesis includes a new approach for calculating the effective resonance integral of absorbed neutrons, new approximate formula for the penetration factor, an analysis of the effective resonance integral and the correction of the resonance integral taking into account the interference of potential and resonance dissipation. A separate chapter is devoted to calculation of the effective resonance integral for the regular reactor lattice with cylindrical fuel elements

  3. Influence of Irradiance, Flow Rate, Reactor Geometry, and Photopromoter Concentration in Mineralization Kinetics of Methane in Air and in Aqueous Solutions by Photocatalytic Membranes Immobilizing Titanium Dioxide

    Directory of Open Access Journals (Sweden)

    Ignazio Renato Bellobono

    2008-01-01

    Full Text Available Photomineralization of methane in air (10.0–1000 ppm (mass/volume of C at 100% relative humidity (dioxygen as oxygen donor was systematically studied at 318±3 K in an annular laboratory-scale reactor by photocatalytic membranes immobilizing titanium dioxide as a function of substrate concentration, absorbed power per unit length of membrane, reactor geometry, and concentration of a proprietary vanadium alkoxide as photopromoter. Kinetics of both substrate disappearance, to yield intermediates, and total organic carbon (TOC disappearance, to yield carbon dioxide, were followed. At a fixed value of irradiance (0.30 W⋅cm-1, the mineralization experiments in gaseous phase were repeated as a function of flow rate (4–400 m3⋅h−1. Moreover, at a standard flow rate of 300 m3⋅h−1, the ratio between the overall reaction volume and the length of the membrane was varied, substantially by varying the volume of reservoir, from and to which circulation of gaseous stream took place. Photomineralization of methane in aqueous solutions was also studied, in the same annular reactor and in the same conditions, but in a concentration range of 0.8–2.0 ppm of C, and by using stoichiometric hydrogen peroxide as an oxygen donor. A kinetic model was employed, from which, by a set of differential equations, four final optimised parameters, k1 and K1, k2 and K2, were calculated, which is able to fit the whole kinetic profile adequately. The influence of irradiance on k1 and k2, as well as of flow rate on K1 and K2, is rationalized. The influence of reactor geometry on k values is discussed in view of standardization procedures of photocatalytic experiments. Modeling of quantum yields, as a function of substrate concentration and irradiance, as well as of concentration of photopromoter, was carried out very satisfactorily. Kinetics of hydroxyl radicals reacting between themselves, leading to hydrogen peroxide, other than with substrate or

  4. A reduced fidelity model for the rotary chemical looping combustion reactor

    International Nuclear Information System (INIS)

    Iloeje, Chukwunwike O.; Zhao, Zhenlong; Ghoniem, Ahmed F.

    2017-01-01

    Highlights: • Methodology for developing a reduced fidelity rotary CLC reactor model is presented. • The reduced model determines optimal reactor configuration that meets design and operating requirements. • A 4-order of magnitude reduction in computational cost is achieved with good prediction accuracy. • Sensitivity studies demonstrate importance of accurate kinetic parameters for reactor optimization. - Abstract: The rotary chemical looping combustion reactor has great potential for efficient integration with CO_2 capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel

  5. Model predictive controller design of hydrocracker reactors

    OpenAIRE

    GÖKÇE, Dila

    2011-01-01

    This study summarizes the design of a Model Predictive Controller (MPC) in Tüpraş, İzmit Refinery Hydrocracker Unit Reactors. Hydrocracking process, in which heavy vacuum gasoil is converted into lighter and valuable products at high temperature and pressure is described briefly. Controller design description, identification and modeling studies are examined and the model variables are presented. WABT (Weighted Average Bed Temperature) equalization and conversion increase are simulate...

  6. BR2 reactor core steady state transient modeling

    International Nuclear Information System (INIS)

    Makarenko, A.; Petrova, T.

    2000-01-01

    A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)

  7. The Hitchin model, Poisson-quasi-Nijenhuis, geometry and symmetry reduction

    International Nuclear Information System (INIS)

    Zucchini, Roberto

    2007-01-01

    We revisit our earlier work on the AKSZ-like formulation of topological sigma model on generalized complex manifolds, or Hitchin model, [20]. We show that the target space geometry geometry implied by the BV master equations is Poisson-quasi-Nijenhuis geometry recently introduced and studied by Stienon and Xu (in the untwisted case) in [44]. Poisson-quasi-Nijenhuis geometry is more general than generalized complex geometry and comprises it as a particular case. Next, we show how gauging and reduction can be implemented in the Hitchin model. We find that the geometry resulting form the BV master equation is closely related to but more general than that recently described by Lin and Tolman in [40, 41], suggesting a natural framework for the study of reduction of Poisson-quasi-Nijenhuis manifolds

  8. CAD-based automatic modeling method for Geant4 geometry model through MCAM

    International Nuclear Information System (INIS)

    Wang, D.; Nie, F.; Wang, G.; Long, P.; LV, Z.

    2013-01-01

    The full text of publication follows. Geant4 is a widely used Monte Carlo transport simulation package. Before calculating using Geant4, the calculation model need be established which could be described by using Geometry Description Markup Language (GDML) or C++ language. However, it is time-consuming and error-prone to manually describe the models by GDML. Automatic modeling methods have been developed recently, but there are some problems that exist in most present modeling programs, specially some of them were not accurate or adapted to specifically CAD format. To convert the GDML format models to CAD format accurately, a Geant4 Computer Aided Design (CAD) based modeling method was developed for automatically converting complex CAD geometry model into GDML geometry model. The essence of this method was dealing with CAD model represented with boundary representation (B-REP) and GDML model represented with constructive solid geometry (CSG). At first, CAD model was decomposed to several simple solids which had only one close shell. And then the simple solid was decomposed to convex shell set. Then corresponding GDML convex basic solids were generated by the boundary surfaces getting from the topological characteristic of a convex shell. After the generation of these solids, GDML model was accomplished with series boolean operations. This method was adopted in CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport (MCAM), and tested with several models including the examples in Geant4 install package. The results showed that this method could convert standard CAD model accurately, and can be used for Geant4 automatic modeling. (authors)

  9. Computer Modeling of Platinum Reforming Reactors | Momoh ...

    African Journals Online (AJOL)

    This paper, instead of using a theoretical approach has considered a computer model as means of assessing the reformate composition for three-stage fixed bed reactors in platforming unit. This is done by identifying many possible hydrocarbon transformation reactions that are peculiar to the process unit, identify the ...

  10. Finsler Geometry Modeling of an Orientation-Asymmetric Surface Model for Membranes

    Science.gov (United States)

    Proutorov, Evgenii; Koibuchi, Hiroshi

    2017-12-01

    In this paper, a triangulated surface model is studied in the context of Finsler geometry (FG) modeling. This FG model is an extended version of a recently reported model for two-component membranes, and it is asymmetric under surface inversion. We show that the definition of the model is independent of how the Finsler length of a bond is defined. This leads us to understand that the canonical (or Euclidean) surface model is obtained from the FG model such that it is uniquely determined as a trivial model from the viewpoint of well definedness.

  11. Polymer mixtures in confined geometries: Model systems to explore ...

    Indian Academy of Sciences (India)

    to mean field behavior for very long chains, the critical behavior of mixtures confined into thin film geometry falls in the 2d Ising class irrespective of chain length. ..... AB interface does not approach the wall; (b) corresponds to a temperature .... Very recently, these theoretical studies have been extended to polymer mixtures.

  12. Bianchi-IX string cosmological model in Lyra geometry

    Indian Academy of Sciences (India)

    field in Lyra's geometry will either include a creation field and be equal to Hoyle's cre- ation field cosmology or contain a special vacuum field which together with the gauge vector term may be considered as a cosmological term. Subsequent investigations were done by several authors in scalar–tensor theory and cos-.

  13. Modeling a nuclear reactor for experimental purposes

    International Nuclear Information System (INIS)

    Berta, V.T.

    1980-01-01

    The Loss-of-Fluid Test (LOFT) Facility is a scale model of a commercial PWR and is as fully functional and operational as the generic commercial counterpart. LOFT was designed and built for experimental purposes as part of the overall NRC reactor safety research program. The purpose of LOFT is to assess the capability of reactor safety systems to perform their intended functions during occurrences of off-normal conditions in a commercial nuclear reactor. Off-normal conditions arising from large and small break loss-of-coolant accidents (LOCA), operational transients, and anticipated transients without scram (ATWS) were to be investigated. This paper describes the LOFT model of the generic PWR and summarizes the experiments that have been conducted in the context of the significant findings involving the complex transient thermal-hydraulics and the consequent effects on the commercial reactor analytical licensing techniques. Through these techniques the validity of the LOFT model as a scaled counterpart of the generic PWR is shown

  14. Accelerating navigation in the VecGeom geometry modeller

    Science.gov (United States)

    Wenzel, Sandro; Zhang, Yang; pre="for the"> VecGeom Developers,

    2017-10-01

    The VecGeom geometry library is a relatively recent effort aiming to provide a modern and high performance geometry service for particle detector simulation in hierarchical detector geometries common to HEP experiments. One of its principal targets is the efficient use of vector SIMD hardware instructions to accelerate geometry calculations for single track as well as multi-track queries. Previously, excellent performance improvements compared to Geant4/ROOT could be reported for elementary geometry algorithms at the level of single shape queries. In this contribution, we will focus on the higher level navigation algorithms in VecGeom, which are the most important components as seen from the simulation engines. We will first report on our R&D effort and developments to implement SIMD enhanced data structures to speed up the well-known “voxelised” navigation algorithms, ubiquitously used for particle tracing in complex detector modules consisting of many daughter parts. Second, we will discuss complementary new approaches to improve navigation algorithms in HEP. These ideas are based on a systematic exploitation of static properties of the detector layout as well as automatic code generation and specialisation of the C++ navigator classes. Such specialisations reduce the overhead of generic- or virtual function based algorithms and enhance the effectiveness of the SIMD vector units. These novel approaches go well beyond the existing solutions available in Geant4 or TGeo/ROOT, achieve a significantly superior performance, and might be of interest for a wide range of simulation backends (GeantV, Geant4). We exemplify this with concrete benchmarks for the CMS and ALICE detectors.

  15. The geometry of terrestrial laser scanning; identification of errors, modeling and mitigation of scanning geometry

    NARCIS (Netherlands)

    Soudarissanane, S.S.

    2016-01-01

    Over the past few decades, Terrestrial Laser Scanners are increasingly being used in a broad spectrum of applications, from surveying to civil engineering, medical modeling and forensics. Especially surveying applications require on one hand a quickly obtainable, high resolution point cloud but also

  16. Models of iodine behavior in reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1992-10-01

    Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents.

  17. Models of iodine behavior in reactor containments

    International Nuclear Information System (INIS)

    Weber, C.F.; Beahm, E.C.; Kress, T.S.

    1992-10-01

    Models are developed for many phenomena of interest concerning iodine behavior in reactor containments during severe accidents. Processes include speciation in both gas and liquid phases, reactions with surfaces, airborne aerosols, and other materials, and gas-liquid interface behavior. Although some models are largely empirical formulations, every effort has been made to construct mechanistic and rigorous descriptions of relevant chemical processes. All are based on actual experimental data generated at the Oak Ridge National Laboratory (ORNL) or elsewhere, and, hence, considerable data evaluation and parameter estimation are contained in this study. No application or encoding is attempted, but each model is stated in terms of rate processes, with the intention of allowing mechanistic simulation. Taken together, this collection of models represents a best estimate iodine behavior and transport in reactor accidents

  18. Regression models of reactor diagnostic signals

    International Nuclear Information System (INIS)

    Vavrin, J.

    1989-01-01

    The application is described of an autoregression model as the simplest regression model of diagnostic signals in experimental analysis of diagnostic systems, in in-service monitoring of normal and anomalous conditions and their diagnostics. The method of diagnostics is described using a regression type diagnostic data base and regression spectral diagnostics. The diagnostics is described of neutron noise signals from anomalous modes in the experimental fuel assembly of a reactor. (author)

  19. Modeling photonic crystal waveguides with noncircular geometry using green function method

    International Nuclear Information System (INIS)

    Uvarovaa, I.; Tsyganok, B.; Bashkatov, Y.; Khomenko, V.

    2012-01-01

    Currently in the field of photonics is an acute problem fast and accurate simulation photonic crystal waveguides with complex geometry. This paper describes an improved method of Green's functions for non-circular geometries. Based on comparison of selected efficient numerical method for finding the eigenvalues for the Green's function method for non-circular holes chosen effective method for our purposes. Simulation is realized in Maple environment. The simulation results confirmed experimentally. Key words: photonic crystal, waveguide, modeling, Green function, complex geometry

  20. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

  1. Pebble Bed Reactor Dust Production Model

    Energy Technology Data Exchange (ETDEWEB)

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

  2. Pebble Bed Reactor Dust Production Model

    International Nuclear Information System (INIS)

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-01-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production

  3. Numerical simulations and mathematical models of flows in complex geometries

    DEFF Research Database (Denmark)

    Hernandez Garcia, Anier

    The research work of the present thesis was mainly aimed at exploiting one of the strengths of the Lattice Boltzmann methods, namely, the ability to handle complicated geometries to accurately simulate flows in complex geometries. In this thesis, we perform a very detailed theoretical analysis...... and through the Chapman-Enskog multi-scale expansion technique the dependence of the kinetic viscosity on each scheme is investigated. Seeking for optimal numerical schemes to eciently simulate a wide range of complex flows a variant of the finite element, off-lattice Boltzmann method [5], which uses...... the characteristic based integration is also implemented. Using the latter scheme, numerical simulations are conducted in flows of different complexities: flow in a (real) porous network and turbulent flows in ducts with wall irregularities. From the simulations of flows in porous media driven by pressure gradients...

  4. An immersed boundary method for modeling a dirty geometry data

    Science.gov (United States)

    Onishi, Keiji; Tsubokura, Makoto

    2017-11-01

    We present a robust, fast, and low preparation cost immersed boundary method (IBM) for simulating an incompressible high Re flow around highly complex geometries. The method is achieved by the dispersion of the momentum by the axial linear projection and the approximate domain assumption satisfying the mass conservation around the wall including cells. This methodology has been verified against an analytical theory and wind tunnel experiment data. Next, we simulate the problem of flow around a rotating object and demonstrate the ability of this methodology to the moving geometry problem. This methodology provides the possibility as a method for obtaining a quick solution at a next large scale supercomputer. This research was supported by MEXT as ``Priority Issue on Post-K computer'' (Development of innovative design and production processes) and used computational resources of the K computer provided by the RIKEN Advanced Institute for Computational Science.

  5. Comparison of low enriched uranium (UAlx-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

    International Nuclear Information System (INIS)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G.; Nishiyama, Pedro J.B. de O.

    2011-01-01

    The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium ( 235 U) UAl x dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ( 99 Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of 99 Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

  6. Fischer-Tropsch Slurry Reactor modeling

    Energy Technology Data Exchange (ETDEWEB)

    Soong, Y.; Gamwo, I.K.; Harke, F.W. [Pittsburgh Energy Technology Center, PA (United States)] [and others

    1995-12-31

    This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas, solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.

  7. A methodology for modeling photocatalytic reactors for indoor pollution control using previously estimated kinetic parameters

    Energy Technology Data Exchange (ETDEWEB)

    Passalia, Claudio; Alfano, Orlando M. [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina); Brandi, Rodolfo J., E-mail: rbrandi@santafe-conicet.gov.ar [INTEC - Instituto de Desarrollo Tecnologico para la Industria Quimica, CONICET - UNL, Gueemes 3450, 3000 Santa Fe (Argentina); FICH - Departamento de Medio Ambiente, Facultad de Ingenieria y Ciencias Hidricas, Universidad Nacional del Litoral, Ciudad Universitaria, 3000 Santa Fe (Argentina)

    2012-04-15

    Highlights: Black-Right-Pointing-Pointer Indoor pollution control via photocatalytic reactors. Black-Right-Pointing-Pointer Scaling-up methodology based on previously determined mechanistic kinetics. Black-Right-Pointing-Pointer Radiation interchange model between catalytic walls using configuration factors. Black-Right-Pointing-Pointer Modeling and experimental validation of a complex geometry photocatalytic reactor. - Abstract: A methodology for modeling photocatalytic reactors for their application in indoor air pollution control is carried out. The methodology implies, firstly, the determination of intrinsic reaction kinetics for the removal of formaldehyde. This is achieved by means of a simple geometry, continuous reactor operating under kinetic control regime and steady state. The kinetic parameters were estimated from experimental data by means of a nonlinear optimization algorithm. The second step was the application of the obtained kinetic parameters to a very different photoreactor configuration. In this case, the reactor is a corrugated wall type using nanosize TiO{sub 2} as catalyst irradiated by UV lamps that provided a spatially uniform radiation field. The radiative transfer within the reactor was modeled through a superficial emission model for the lamps, the ray tracing method and the computation of view factors. The velocity and concentration fields were evaluated by means of a commercial CFD tool (Fluent 12) where the radiation model was introduced externally. The results of the model were compared experimentally in a corrugated wall, bench scale reactor constructed in the laboratory. The overall pollutant conversion showed good agreement between model predictions and experiments, with a root mean square error less than 4%.

  8. Modeling of hydrogenation reactor of soya oil

    International Nuclear Information System (INIS)

    Sotudeh-Gharebagh, R.; Niknam, L.; Mostoufi, N.

    2008-01-01

    In this paper, a batch hydrogenation reactor performance was modeled using a hydrodynamic and reaction sub-models. The reaction expressions were obtained from the information reported in literature. Experimental studies were conducted in order to generate the experimental data needed to validate the model. The comparison between the experimental data and model predictions seems quite satisfactory considering the hydrodynamic limitations and simplifications made on the reaction scheme. The results of this study could be considered as framework in developing new process equipment and also soya oil product design for new applications

  9. Scaling laws for modeling nuclear reactor systems

    International Nuclear Information System (INIS)

    Nahavandi, A.N.; Castellana, F.S.; Moradkhanian, E.N.

    1979-01-01

    Scale models are used to predict the behavior of nuclear reactor systems during normal and abnormal operation as well as under accident conditions. Three types of scaling procedures are considered: time-reducing, time-preserving volumetric, and time-preserving idealized model/prototype. The necessary relations between the model and the full-scale unit are developed for each scaling type. Based on these relationships, it is shown that scaling procedures can lead to distortion in certain areas that are discussed. It is advised that, depending on the specific unit to be scaled, a suitable procedure be chosen to minimize model-prototype distortion

  10. Planning for Evolution in a Production Environment: Migration from a Legacy Geometry Code to an Abstract Geometry Modeling Language in STAR

    Science.gov (United States)

    Webb, Jason C.; Lauret, Jerome; Perevoztchikov, Victor

    2012-12-01

    Increasingly detailed descriptions of complex detector geometries are required for the simulation and analysis of today's high-energy and nuclear physics experiments. As new tools for the representation of geometry models become available during the course of an experiment, a fundamental challenge arises: how best to migrate from legacy geometry codes developed over many runs to the new technologies, such as the ROOT/TGeo [1] framework, without losing touch with years of development, tuning and validation. One approach, which has been discussed within the community for a number of years, is to represent the geometry model in a higher-level language independent of the concrete implementation of the geometry. The STAR experiment has used this approach to successfully migrate its legacy GEANT 3-era geometry to an Abstract geometry Modelling Language (AgML), which allows us to create both native GEANT 3 and ROOT/TGeo implementations. The language is supported by parsers and a C++ class library which enables the automated conversion of the original source code to AgML, supports export back to the original AgSTAR[5] representation, and creates the concrete ROOT/TGeo geometry implementation used by our track reconstruction software. In this paper we present our approach, design and experience and will demonstrate physical consistency between the original AgSTAR and new AgML geometry representations.

  11. CFD Modeling of Flow and Ion Exchange Kinetics in a Rotating Bed Reactor System

    DEFF Research Database (Denmark)

    Larsson, Hilde Kristina; Schjøtt Andersen, Patrick Alexander; Byström, Emil

    2017-01-01

    A rotating bed reactor (RBR) has been modeled using computational fluid dynamics (CFD). The flow pattern in the RBR was investigated and the flow through the porous material in it was quantified. A simplified geometry representing the more complex RBR geometry was introduced and the simplified...... model was able to reproduce the main characteristics of the flow. Alternating reactor shapes were investigated, and it was concluded that the use of baffles has a very large impact on the flows through the porous material. The simulations suggested, therefore, that even faster reaction rates could...... be achieved by making the baffles deeper. Two-phase simulations were performed, which managed to reproduce the deflection of the gas–liquid interface in an unbaffled system. A chemical reaction was implemented in the model, describing the ion-exchange phenomena in the porous material using four different...

  12. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  13. Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

    2012-07-01

    Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

  14. Transient Changes in Molecular Geometries and How to Model Them

    DEFF Research Database (Denmark)

    Dohn, Asmus Ougaard

    Light-induced chemical processes are accompanied by molecular motion on the femtosecond time scale. Uncovering this dynamical motion is central to understanding the chemical reaction on a fundamental level. This thesis focuses on the aspects of excess excitation energy dissipation via dynamic...... observe how the wide distribution of ground state geometries is responsible for decoherence, and that the solvent cage actually facilitates coherent motion, by blocking the newly discovered vibrational mode. We furthermore observe a non-specific, rotational solvent response to the excitation. The second...

  15. Micro-tomography based Geometry Modeling of Three-Dimensional Braided Composites

    Science.gov (United States)

    Fang, Guodong; Chen, Chenghua; Yuan, Shenggang; Meng, Songhe; Liang, Jun

    2018-06-01

    A tracking and recognizing algorithm is proposed to automatically generate irregular cross-sections and central path of braid yarn within the 3D braided composites by using sets of high resolution tomography images. Only the initial cross-sections of braid yarns in a tomography image after treatment are required to be calibrated manually as searching cross-section template. The virtual geometry of 3D braided composites including some detailed geometry information, such as the braid yarn squeezing deformation, braid yarn distortion and braid yarn path deviation etc., can be reconstructed. The reconstructed geometry model can reflect the change of braid configurations during solidification process. The geometry configurations and mechanical properties of the braided composites are analyzed by using the reconstructed geometry model.

  16. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  17. Investigation on the influence of electrode geometry on characteristics of coaxial dielectric barrier discharge reactor driven by an oscillating microsecond pulsed power supply

    Science.gov (United States)

    Miao, Chuanrun; Liu, Feng; Wang, Qian; Cai, Meiling; Fang, Zhi

    2018-03-01

    In this paper, an oscillating microsecond pulsed power supply with rise time of several tens of nanosecond (ns) is used to excite a coaxial DBD with double layer dielectric barriers. The effects of various electrode geometries by changing the size of inner quartz tube (different electrode gaps) on the discharge uniformity, power deposition, energy efficiency, and operation temperature are investigated by electrical, optical, and temperature diagnostics. The electrical parameters of the coaxial DBD are obtained from the measured applied voltage and current using an equivalent electrical model. The energy efficiency and the power deposition in air gap of coaxial DBD with various electrode geometries are also obtained with the obtained electrical parameters, and the heat loss and operation temperature are analyzed by a heat conduction model. It is found that at the same applied voltage, with the increasing of the air gap, the discharge uniformity becomes worse and the discharge power deposition and the energy efficiency decrease. At 2.5 mm air gap and 24 kV applied voltage, the energy efficiency of the coaxial DBD reaches the maximum value of 68.4%, and the power deposition in air gap is 23.6 W and the discharge uniformity is the best at this case. The corresponding operation temperature of the coaxial DBD reaches 64.3 °C after 900 s operation and the temperature of the inner dielectric barrier is 114.4 °C under thermal balance. The experimental results provide important experimental references and are important to optimize the design and the performance of coaxial DBD reactor.

  18. Modelling of a falling sludge bed reactor using AQUASIM | Ristow ...

    African Journals Online (AJOL)

    Modelling of a falling sludge bed reactor using AQUASIM. ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search · USING AJOL · RESOURCES ... a system of mixed reactors connected by water flow and mass flux streams.

  19. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    International Nuclear Information System (INIS)

    Vierow, Karen

    2008-01-01

    This project is investigating countercurrent flow and 'flooding' phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the 'surge line' and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008

  20. Entropy Error Model of Planar Geometry Features in GIS

    Institute of Scientific and Technical Information of China (English)

    LI Dajun; GUAN Yunlan; GONG Jianya; DU Daosheng

    2003-01-01

    Positional error of line segments is usually described by using "g-band", however, its band width is in relation to the confidence level choice. In fact, given different confidence levels, a series of concentric bands can be obtained. To overcome the effect of confidence level on the error indicator, by introducing the union entropy theory, we propose an entropy error ellipse index of point, then extend it to line segment and polygon,and establish an entropy error band of line segment and an entropy error donut of polygon. The research shows that the entropy error index can be determined uniquely and is not influenced by confidence level, and that they are suitable for positional uncertainty of planar geometry features.

  1. On stochastic geometry modeling of cellular uplink transmission with truncated channel inversion power control

    KAUST Repository

    Elsawy, Hesham; Hossain, Ekram

    2014-01-01

    Using stochastic geometry, we develop a tractable uplink modeling paradigm for outage probability and spectral efficiency in both single and multi-tier cellular wireless networks. The analysis accounts for per user equipment (UE) power control

  2. Hillslope hydrological modeling : the role of bedrock geometry and hillslope-stream interaction

    NARCIS (Netherlands)

    Shahedi, K.

    2008-01-01

    Keywords: Hillslope hydrology, hydrological modeling, bedrock geometry, boundary condition, numerical solution.

    This thesis focuses on hillslope subsurface flow as a dominant control on the hydrological processes defining the catchment response to rainfall. Due to the difficulties

  3. A Stochastic Geometry Model for Multi-hop Highway Vehicular Communication

    KAUST Repository

    Farooq, Muhammad Junaid; Elsawy, Hesham; Alouini, Mohamed-Slim

    2015-01-01

    dissemination. This paper exploits stochastic geometry to develop a tractable and accurate modeling framework to characterize the multi-hop transmissions for vehicular networks in a multi-lane highway setup. In particular, we study the tradeoffs between per

  4. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  5. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  6. Neutronic modeling of pebble bed reactors in APOLLO2

    International Nuclear Information System (INIS)

    Grimod, M.

    2010-01-01

    In this thesis we develop a new iterative homogenization technique for pebble bed reactors, based on a 'macro-stochastic' transport approximation in the collision probability method. A model has been developed to deal with the stochastic distribution of pebbles with different burnup in the core, considering spectral differences in homogenization and depletion calculations. This is generally not done in the codes presently used for pebble bed analyses, where a pebble with average isotopic composition is considered to perform the cell calculation. Also an iterative core calculation scheme has been set up, where the low-order RZ S N full-core calculation computes the entering currents in the spectrum zones subdividing the core. These currents, together with the core k eff , are then used as surface source in the fine-group heterogeneous calculation of the multi-pebble geometries. The developed method has been verified using reference Monte Carlo simulations of a simplified PBMR- 400 model. The pebbles in this model are individually positioned and have different randomly assigned burnup values. The APOLLO2 developed method matches the reference core k eff within ± 100 pcm, with relative differences on the production shape factors within ± 4%, and maximum discrepancy of 3% at the hotspot. Moreover, the first criticality experiment of the HTR-10 reactor was used to perform a first validation of the developed model. The computed critical number of pebbles to be loaded in the core is very close to the experimental value of 16890, only 77 pebbles less. A method to calculate the equilibrium reactor state was also developed and applied to analyze the simplified PBMR-400 model loaded with different fuel types (UO 2 , Pu, Pu + MA). The potential of the APOLLO2 method to compute different fluxes for the different pebble types of a multi-pebble geometry was used to evaluate the bias committed by the average composition pebble approximation. Thanks to a 'compensation of error

  7. Morphing the feature-based multi-blocks of normative/healthy vertebral geometries to scoliosis vertebral geometries: development of personalized finite element models.

    Science.gov (United States)

    Hadagali, Prasannaah; Peters, James R; Balasubramanian, Sriram

    2018-03-12

    Personalized Finite Element (FE) models and hexahedral elements are preferred for biomechanical investigations. Feature-based multi-block methods are used to develop anatomically accurate personalized FE models with hexahedral mesh. It is tedious to manually construct multi-blocks for large number of geometries on an individual basis to develop personalized FE models. Mesh-morphing method mitigates the aforementioned tediousness in meshing personalized geometries every time, but leads to element warping and loss of geometrical data. Such issues increase in magnitude when normative spine FE model is morphed to scoliosis-affected spinal geometry. The only way to bypass the issue of hex-mesh distortion or loss of geometry as a result of morphing is to rely on manually constructing the multi-blocks for scoliosis-affected spine geometry of each individual, which is time intensive. A method to semi-automate the construction of multi-blocks on the geometry of scoliosis vertebrae from the existing multi-blocks of normative vertebrae is demonstrated in this paper. High-quality hexahedral elements were generated on the scoliosis vertebrae from the morphed multi-blocks of normative vertebrae. Time taken was 3 months to construct the multi-blocks for normative spine and less than a day for scoliosis. Efforts taken to construct multi-blocks on personalized scoliosis spinal geometries are significantly reduced by morphing existing multi-blocks.

  8. Modelling chemical behavior of water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ball, R G.J.; Hanshaw, J; Mason, P K; Mignanelli, M A [AEA Technology, Harwell (United Kingdom)

    1997-08-01

    For many applications, large computer codes have been developed which use correlation`s, simplifications and approximations in order to describe the complex situations which may occur during the operation of nuclear power plant or during fault scenarios. However, it is important to have a firm physical basis for simplifications and approximations in such codes and, therefore, there has been an emphasis on modelling the behaviour of materials and processes on a more detailed or fundamental basis. The application of fundamental modelling techniques to simulated various chemical phenomena in thermal reactor fuel systems are described in this paper. These methods include thermochemical modelling, kinetic and mass transfer modelling and atomistic simulation and examples of each approach are presented. In each of these applications a summary of the methods are discussed together with the assessment process adopted to provide the fundamental parameters which form the basis of the calculation. (author). 25 refs, 9 figs, 2 tabs.

  9. Nuclear reactor core modelling in multifunctional simulators

    International Nuclear Information System (INIS)

    Puska, E.K.

    1999-01-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  10. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  11. Modeling of Neoclassical Tearing Mode Stability for Generalized Toroidal Geometry

    International Nuclear Information System (INIS)

    A.L. Rosenberg; D.A. Gates; A. Pletzer; J.E. Menard; S.E. Kruger; C.C. Hegna; F. Paoletti; S. Sabbagh

    2002-01-01

    Neoclassical tearing modes (NTMs) can lead to disruption and loss of confinement. Previous analysis of these modes used large aspect ratio, low beta (plasma pressure/magnetic pressure) approximations to determine the effect of NTMs on tokamak plasmas. A more accurate tool is needed to predict the onset of these instabilities. As a follow-up to recent theoretical work, a code has been written which computes the tearing mode island growth rate for arbitrary tokamak geometry. It calls PEST-3 [A. Pletzer et al., J. Comput. Phys. 115, 530 (1994)] to compute delta prime, the resistive magnetohydrodynamic (MHD) matching parameter. The code also calls the FLUXGRID routines in NIMROD [A.H. Glasser et al., Plasma Phys. Controlled Fusion 41, A747 (1999)] for Dnc, DI and DR [C.C. Hegna, Phys. Plasmas 6, 3980 (1999); A.H. Glasser et al., Phys. Fluids 18, 875 (1975)], which are the bootstrap current driven term and the ideal and resistive interchange mode criterion, respectively. In addition to these components, the NIMROD routines calculate alphas-H, a new correction to the Pfirsch-Schlter term. Finite parallel transport effects were added and a National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)] equilibrium was analyzed. Another program takes the output of PEST-3 and allows the user to specify the rational surface, island width, and amount of detail near the perturbed surface to visualize the total helical flux. The results of this work will determine the stability of NTMs in an spherical torus (ST) [Y.-K.M. Peng et al., Nucl. Fusion 26, 769 (1986)] plasma with greater accuracy than previously achieved

  12. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

    El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

    2010-10-01

    The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

  13. MONTE CARLO ANALYSES OF THE YALINA THERMAL FACILITY WITH SERPENT STEREOLITHOGRAPHY GEOMETRY MODEL

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, Y.

    2015-01-01

    This paper analyzes the YALINA Thermal subcritical assembly of Belarus using two different Monte Carlo transport programs, SERPENT and MCNP. The MCNP model is based on combinatorial geometry and universes hierarchy, while the SERPENT model is based on Stereolithography geometry. The latter consists of unstructured triangulated surfaces defined by the normal and vertices. This geometry format is used by 3D printers and it has been created by: the CUBIT software, MATLAB scripts, and C coding. All the Monte Carlo simulations have been performed using the ENDF/B-VII.0 nuclear data library. Both MCNP and SERPENT share the same geometry specifications, which describe the facility details without using any material homogenization. Three different configurations have been studied with different number of fuel rods. The three fuel configurations use 216, 245, or 280 fuel rods, respectively. The numerical simulations show that the agreement between SERPENT and MCNP results is within few tens of pcms.

  14. COGEDIF - automatic TORT and DORT input generation from MORSE combinatorial geometry models

    International Nuclear Information System (INIS)

    Castelli, R.A.; Barnett, D.A.

    1992-01-01

    COGEDIF is an interactive utility which was developed to automate the preparation of two and three dimensional geometrical inputs for the ORNL-TORT and DORT discrete ordinates programs from complex three dimensional models described using the MORSE combinatorial geometry input description. The program creates either continuous or disjoint mesh input based upon the intersections of user defined meshing planes and the MORSE body definitions. The composition overlay of the combinatorial geometry is used to create the composition mapping of the discretized geometry based upon the composition found at the centroid of each of the mesh cells. This program simplifies the process of using discrete orthogonal mesh cells to represent non-orthogonal geometries in large models which require mesh sizes of the order of a million cells or more. The program was specifically written to take advantage of the new TORT disjoint mesh option which was developed at ORNL

  15. A Numerical Model for Trickle Bed Reactors

    Science.gov (United States)

    Propp, Richard M.; Colella, Phillip; Crutchfield, William Y.; Day, Marcus S.

    2000-12-01

    Trickle bed reactors are governed by equations of flow in porous media such as Darcy's law and the conservation of mass. Our numerical method for solving these equations is based on a total-velocity splitting, sequential formulation which leads to an implicit pressure equation and a semi-implicit mass conservation equation. We use high-resolution finite-difference methods to discretize these equations. Our solution scheme extends previous work in modeling porous media flows in two ways. First, we incorporate physical effects due to capillary pressure, a nonlinear inlet boundary condition, spatial porosity variations, and inertial effects on phase mobilities. In particular, capillary forces introduce a parabolic component into the recast evolution equation, and the inertial effects give rise to hyperbolic nonconvexity. Second, we introduce a modification of the slope-limiting algorithm to prevent our numerical method from producing spurious shocks. We present a numerical algorithm for accommodating these difficulties, show the algorithm is second-order accurate, and demonstrate its performance on a number of simplified problems relevant to trickle bed reactor modeling.

  16. Geometry parameters for musculoskeletal modelling of the shoulder system

    NARCIS (Netherlands)

    Van der Helm, F C; Veeger, DirkJan (H. E. J.); Pronk, G M; Van der Woude, L H; Rozendal, R H

    A dynamical finite-element model of the shoulder mechanism consisting of thorax, clavicula, scapula and humerus is outlined. The parameters needed for the model are obtained in a cadaver experiment consisting of both shoulders of seven cadavers. In this paper, in particular, the derivation of

  17. RSMASS: A simple model for estimating reactor and shield masses

    International Nuclear Information System (INIS)

    Marshall, A.C.; Aragon, J.; Gallup, D.

    1987-01-01

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, thermal/hydraulic limits, or fuel damage limits, whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should be applicable to a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations

  18. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Limaiem, I.

    2006-12-01

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  19. Geometry of the Universe

    International Nuclear Information System (INIS)

    Gurevich, L.Eh.; Gliner, Eh.B.

    1978-01-01

    Problems of investigating the Universe space-time geometry are described on a popular level. Immediate space-time geometries, corresponding to three cosmologic models are considered. Space-time geometry of a closed model is the spherical Riemann geonetry, of an open model - is the Lobachevskij geometry; and of a plane model - is the Euclidean geometry. The Universe real geometry in the contemporary epoch of development is based on the data testifying to the fact that the Universe is infinitely expanding

  20. Coalescing colony model: Mean-field, scaling, and geometry

    Science.gov (United States)

    Carra, Giulia; Mallick, Kirone; Barthelemy, Marc

    2017-12-01

    We analyze the coalescing model where a `primary' colony grows and randomly emits secondary colonies that spread and eventually coalesce with it. This model describes population proliferation in theoretical ecology, tumor growth, and is also of great interest for modeling urban sprawl. Assuming the primary colony to be always circular of radius r (t ) and the emission rate proportional to r (t) θ , where θ >0 , we derive the mean-field equations governing the dynamics of the primary colony, calculate the scaling exponents versus θ , and compare our results with numerical simulations. We then critically test the validity of the circular approximation for the colony shape and show that it is sound for a constant emission rate (θ =0 ). However, when the emission rate is proportional to the perimeter, the circular approximation breaks down and the roughness of the primary colony cannot be discarded, thus modifying the scaling exponents.

  1. Optimized numerical annular flow dryout model using the drift-flux model in tube geometry

    International Nuclear Information System (INIS)

    Chun, Ji Han; Lee, Un Chul

    2008-01-01

    Many experimental analyses for annular film dryouts, which is one of the Critical Heat Flux (CHF) mechanisms, have been performed because of their importance. Numerical approaches must also be developed in order to assess the results from experiments and to perform pre-tests before experiments. Various thermal-hydraulic codes, such as RELAP, COBRATF, MARS, etc., have been used in the assessment of the results of dryout experiments and in experimental pre-tests. These thermal-hydraulic codes are general tools intended for the analysis of various phenomena that could appear in nuclear power plants, and many models applying these codes are unnecessarily complex for the focused analysis of dryout phenomena alone. In this study, a numerical model was developed for annular film dryout using the drift-flux model from uniform heated tube geometry. Several candidates of models that strongly affect dryout, such as the entrainment model, deposition model, and the criterion for the dryout point model, were tested as candidates for inclusion in an optimized annular film dryout model. The optimized model was developed by adopting the best combination of these candidate models, as determined through comparison with experimental data. This optimized model showed reasonable results, which were better than those of MARS code

  2. Dynamic hyperbolic geometry: building intuition and understanding mediated by a Euclidean model

    Science.gov (United States)

    Moreno-Armella, Luis; Brady, Corey; Elizondo-Ramirez, Rubén

    2018-05-01

    This paper explores a deep transformation in mathematical epistemology and its consequences for teaching and learning. With the advent of non-Euclidean geometries, direct, iconic correspondences between physical space and the deductive structures of mathematical inquiry were broken. For non-Euclidean ideas even to become thinkable the mathematical community needed to accumulate over twenty centuries of reflection and effort: a precious instance of distributed intelligence at the cultural level. In geometry education after this crisis, relations between intuitions and geometrical reasoning must be established philosophically, rather than taken for granted. One approach seeks intuitive supports only for Euclidean explorations, viewing non-Euclidean inquiry as fundamentally non-intuitive in nature. We argue for moving beyond such an impoverished approach, using dynamic geometry environments to develop new intuitions even in the extremely challenging setting of hyperbolic geometry. Our efforts reverse the typical direction, using formal structures as a source for a new family of intuitions that emerge from exploring a digital model of hyperbolic geometry. This digital model is elaborated within a Euclidean dynamic geometry environment, enabling a conceptual dance that re-configures Euclidean knowledge as a support for building intuitions in hyperbolic space-intuitions based not directly on physical experience but on analogies extending Euclidean concepts.

  3. Conservation laws and geometry of perturbed coset models

    CERN Document Server

    Bakas, Ioannis

    1994-01-01

    We present a Lagrangian description of the $SU(2)/U(1)$ coset model perturbed by its first thermal operator. This is the simplest perturbation that changes sign under Krammers--Wannier duality. The resulting theory, which is a 2--component generalization of the sine--Gordon model, is then taken in Minkowski space. For negative values of the coupling constant $g$, it is classically equivalent to the $O(4)$ non--linear $\\s$--model reduced in a certain frame. For $g > 0$, it describes the relativistic motion of vortices in a constant external field. Viewing the classical equations of motion as a zero curvature condition, we obtain recursive relations for the infinitely many conservation laws by the abelianization method of gauge connections. The higher spin currents are constructed entirely using an off--critical generalization of the $W_{\\infty}$ generators. We give a geometric interpretation to the corresponding charges in terms of embeddings. Applications to the chirally invariant $U(2)$ Gross--Neveu model ar...

  4. Geometry of coexistence in the interacting boson model

    International Nuclear Information System (INIS)

    Van Isacker, P.; Frank, A.; Vargas, C.E.

    2004-01-01

    The Interacting Boson Model (IBM) with configuration mixing is applied to describe the phenomenon of coexistence in nuclei. The analysis suggests that the IBM with configuration mixing, used in conjunction with a (matrix) coherent-state method, may be a reliable tool for the study of geometric aspects of shape coexistence in nuclei

  5. Geometry of neural networks and models with singularities

    International Nuclear Information System (INIS)

    Fukumizu, Kenji

    2001-01-01

    This paper discusses maximum likelihood estimation with unidentifiability of parameters. Unidentifiability is formulated as a conic singularity of the model. It is known that the likelihood ratio may have unusually large order in unidentifiable cases. A sufficient condition for such large order is given and applied to neural networks

  6. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    International Nuclear Information System (INIS)

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m 2 ) (1.1E+6 BTU/(ft 2 hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient

  7. Dynamic Modeling Accuracy Dependence on Errors in Sensor Measurements, Mass Properties, and Aircraft Geometry

    Science.gov (United States)

    Grauer, Jared A.; Morelli, Eugene A.

    2013-01-01

    A nonlinear simulation of the NASA Generic Transport Model was used to investigate the effects of errors in sensor measurements, mass properties, and aircraft geometry on the accuracy of dynamic models identified from flight data. Measurements from a typical system identification maneuver were systematically and progressively deteriorated and then used to estimate stability and control derivatives within a Monte Carlo analysis. Based on the results, recommendations were provided for maximum allowable errors in sensor measurements, mass properties, and aircraft geometry to achieve desired levels of dynamic modeling accuracy. Results using other flight conditions, parameter estimation methods, and a full-scale F-16 nonlinear aircraft simulation were compared with these recommendations.

  8. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

  9. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given. (author)

  10. Geometry of surfaces associated to Grassmannian sigma models

    International Nuclear Information System (INIS)

    Delisle, L; Hussin, V; Zakrzewski, W J

    2015-01-01

    We investigate the geometric characteristics of constant Gaussian curvature surfaces obtained from solutions of the G(m, n) sigma model. Most of these solutions are related to the Veronese sequence. We show that we can distinguish surfaces with the same Gaussian curvature using additional quantities like the topological charge and the mean curvature. The cases of G(1,n) = CP n-1 and G(2,n) are used to illustrate these characteristics. (paper)

  11. Geometry Based Design Automation : Applied to Aircraft Modelling and Optimization

    OpenAIRE

    Amadori, Kristian

    2012-01-01

    Product development processes are continuously challenged by demands for increased efficiency. As engineering products become more and more complex, efficient tools and methods for integrated and automated design are needed throughout the development process. Multidisciplinary Design Optimization (MDO) is one promising technique that has the potential to drastically improve concurrent design. MDO frameworks combine several disciplinary models with the aim of gaining a holistic perspective of ...

  12. Modelling of vapour explosion in a stratified geometry

    International Nuclear Information System (INIS)

    Brayer, Claude

    1994-01-01

    A vapour explosion is the explosive vaporisation of a volatile liquid in contact with another hotter liquid. Such a violent vaporisation requires an intimate mixing and a fine fragmentation of both liquids. Based on a synthesis of published experimental results, the author of this research thesis reports the development of a new physical model which describes the explosion. In this model, the explosion propagation is due to the propagation of the pressure wave associated with this this explosion, all along the vapour film which initially separates both liquids. The author takes the presence of water in the liquid initially located over the film into account. This presence of vapour explains experimental propagation rates. Another consequence, when the pressure wave passes, is an acceleration of liquids at different rates below and above the film. The author considers that a mixture layer then forms from the point of disappearance of the film, between both liquids, and that fragmentation is due to the turbulence in this mixture layer. This fragmentation model is then introduced into an Euler thermodynamic, three-dimensional and multi-constituents code of calculation, MC3D, to study the influence of fragmentation on thermal exchanges between the various constituents on the volatile liquid vaporisation [fr

  13. [Crop geometry identification based on inversion of semiempirical BRDF models].

    Science.gov (United States)

    Zhao, Chun-jiang; Huang, Wen-jiang; Mu, Xu-han; Wang, Jin-diz; Wang, Ji-hua

    2009-09-01

    With the rapid development of remote sensing technology, the application of remote sensing has extended from single view angle to multi-view angles. It was studied for the qualitative and quantitative effect of average leaf angle (ALA) on crop canopy reflected spectrum. Effect of ALA on canopy reflected spectrum can not be ignored with inversion of leaf area index (LAI) and monitoring of crop growth condition by remote sensing technology. Investigations of the effect of erective and horizontal varieties were conducted by bidirectional canopy reflected spectrum and semiempirical bidirectional reflectance distribution function (BRDF) models. The sensitive analysis was done based on the weight for the volumetric kernel (fvol), the weight for the geometric kernel (fgeo), and the weight for constant corresponding to isotropic reflectance (fiso) at red band (680 nm) and near infrared band (800 nm). By combining the weights of the red and near-infrared bands, the semiempirical models can obtain structural information by retrieving biophysical parameters from the physical BRDF model and a number of bidirectional observations. So, it will allow an on-site and non-sampling mode of crop ALA identification, which is useful for using remote sensing for crop growth monitoring and for improving the LAI inversion accuracy, and it will help the farmers in guiding the fertilizer and irrigation management in the farmland without a priori knowledge.

  14. Perbandingan antara Keefektifan Model Guided Discovery Learning dan Project-Based Learning pada Matakuliah Geometri

    Directory of Open Access Journals (Sweden)

    Okky Riswandha Imawan

    2015-12-01

    Abstract This research aims to describe the effectiveness and effectiveness differences of the Guided Discovery Learning (GDL Model and the Project Based Learning (PjBL Model in terms of achievement, self-confidence, and critical thinking skills of students on the Solid Geometry subjects. This research was quasi experimental. The research subjects were two undergraduate classes of Mathematics Education Program, Ahmad Dahlan University, in their second semester, established at random. The data analysis to test the effectiveness of the GDL and PjBL Models in terms of each of the dependent variables used the t-test. The data analysis to test differences between effectiveness of the GDL and that of the PjBL Model used the MANOVA test. The results of this research show that viewed from achievement, self confidence, and critical thinking skills of the students are the application of the GDL Model on Solid Geometry subject is effective, the application of the PjBL Model on Solid Geometry subject is effective, and there is no difference in the effectiveness of GDL and PjBL Models on Solid Geometry subject in terms of achievement, self confidence, and critical thinking skills of the students. Keywords: guided discovery learning model, project-based learning model, achievement, self-confidence, critical thinking skills

  15. Two-Dimensional Model for Reactive-Sorption Columns of Cylindrical Geometry: Analytical Solutions and Moment Analysis.

    Science.gov (United States)

    Khan, Farman U; Qamar, Shamsul

    2017-05-01

    A set of analytical solutions are presented for a model describing the transport of a solute in a fixed-bed reactor of cylindrical geometry subjected to the first (Dirichlet) and third (Danckwerts) type inlet boundary conditions. Linear sorption kinetic process and first-order decay are considered. Cylindrical geometry allows the use of large columns to investigate dispersion, adsorption/desorption and reaction kinetic mechanisms. The finite Hankel and Laplace transform techniques are adopted to solve the model equations. For further analysis, statistical temporal moments are derived from the Laplace-transformed solutions. The developed analytical solutions are compared with the numerical solutions of high-resolution finite volume scheme. Different case studies are presented and discussed for a series of numerical values corresponding to a wide range of mass transfer and reaction kinetics. A good agreement was observed in the analytical and numerical concentration profiles and moments. The developed solutions are efficient tools for analyzing numerical algorithms, sensitivity analysis and simultaneous determination of the longitudinal and transverse dispersion coefficients from a laboratory-scale radial column experiment. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  16. Emergence of geometry: A two-dimensional toy model

    International Nuclear Information System (INIS)

    Alfaro, Jorge; Espriu, Domene; Puigdomenech, Daniel

    2010-01-01

    We review the similarities between the effective chiral Lagrangrian, relevant for low-energy strong interactions, and the Einstein-Hilbert action. We use these analogies to suggest a specific mechanism whereby gravitons would emerge as Goldstone bosons of a global SO(D)xGL(D) symmetry broken down to SO(D) by fermion condensation. We propose a two-dimensional toy model where a dynamical zweibein is generated from a topological theory without any preexisting metric structure, the space being endowed only with an affine connection. A metric appears only after the symmetry breaking; thus the notion of distance is an induced effective one. In spite of several nonstandard features this simple toy model appears to be renormalizable and at long distances is described by an effective Lagrangian that corresponds to that of two-dimensional gravity (Liouville theory). The induced cosmological constant is related to the dynamical mass M acquired by the fermion fields in the breaking, which also acts as an infrared regulator. The low-energy expansion is valid for momenta k>M, i.e. for supra-horizon scales. We briefly discuss a possible implementation of a similar mechanism in four dimensions.

  17. The emergence of geometry: a two-dimensional toy model

    CERN Document Server

    Alfaro, Jorge; Puigdomenech, Daniel

    2010-01-01

    We review the similarities between the effective chiral lagrangrian, relevant for low-energy strong interactions, and the Einstein-Hilbert action. We use these analogies to suggest a specific mechanism whereby gravitons would emerge as Goldstone bosons of a global SO(D) X GL(D) symmetry broken down to SO(D) by fermion condensation. We propose a two-dimensional toy model where a dynamical zwei-bein is generated from a topological theory without any pre-existing metric structure, the space being endowed only with an affine connection. A metric appears only after the symmetry breaking; thus the notion of distance is an induced effective one. In spite of several non-standard features this simple toy model appears to be renormalizable and at long distances is described by an effective lagrangian that corresponds to that of two-dimensional gravity (Liouville theory). The induced cosmological constant is related to the dynamical mass M acquired by the fermion fields in the breaking, which also acts as an infrared re...

  18. 3D Printing of Molecular Models with Calculated Geometries and p Orbital Isosurfaces

    Science.gov (United States)

    Carroll, Felix A.; Blauch, David N.

    2017-01-01

    3D printing was used to prepare models of the calculated geometries of unsaturated organic structures. Incorporation of p orbital isosurfaces into the models enables students in introductory organic chemistry courses to have hands-on experience with the concept of orbital alignment in strained and unstrained p systems.

  19. Mathematical modeling for prediction and optimization of TIG welding pool geometry

    Directory of Open Access Journals (Sweden)

    U. Esme

    2009-04-01

    Full Text Available In this work, nonlinear and multi-objective mathematical models were developed to determine the process parameters corresponding to optimum weld pool geometry. The objectives of the developed mathematical models are to maximize tensile load (TL, penetration (P, area of penetration (AP and/or minimize heat affected zone (HAZ, upper width (UW and upper height (UH depending upon the requirements.

  20. Beyond the Standard Model with noncommutative geometry, strolling towards quantum gravity

    International Nuclear Information System (INIS)

    Martinetti, Pierre

    2015-01-01

    Noncommutative geometry in its many incarnations appears at the crossroad of many researches in theoretical and mathematical physics: from models of quantum spacetime(with or without breaking of Lorentz symmetry) to loop gravity and string theory, from early considerations on UV-divergenciesin quantum field theory to recent models of gauge theories on noncommutatives pacetime, from Connes description of the standard model of elementary particles to recent Pati-Salam like extensions. We list several of these applications, emphasizing also the original point of view brought by noncommutative geometry on the nature of time. This text serves as an introduction to the volume of proceedings of the parallel session “Noncommutative geometry and quantum gravity”, as a part of the conference “Conceptual and technical challenges in quantum gravity” organized at the University of Rome La Sapienza sin September 2014. (paper)

  1. Theoretical models of Kapton heating in solar array geometries

    Science.gov (United States)

    Morton, Thomas L.

    1992-01-01

    In an effort to understand pyrolysis of Kapton in solar arrays, a computational heat transfer program was developed. This model allows for the different materials and widely divergent length scales of the problem. The present status of the calculation indicates that thin copper traces surrounded by Kapton and carrying large currents can show large temperature increases, but the other configurations seen on solar arrays have adequate heat sinks to prevent substantial heating of the Kapton. Electron currents from the ambient plasma can also contribute to heating of thin traces. Since Kapton is stable at temperatures as high as 600 C, this indicates that it should be suitable for solar array applications. There are indications that the adhesive sued in solar arrays may be a strong contributor to the pyrolysis problem seen in solar array vacuum chamber tests.

  2. Advanced Small Modular Reactor Economics Model Development

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J [ORNL

    2014-10-01

    The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

  3. Information geometry

    CERN Document Server

    Ay, Nihat; Lê, Hông Vân; Schwachhöfer, Lorenz

    2017-01-01

    The book provides a comprehensive introduction and a novel mathematical foundation of the field of information geometry with complete proofs and detailed background material on measure theory, Riemannian geometry and Banach space theory. Parametrised measure models are defined as fundamental geometric objects, which can be both finite or infinite dimensional. Based on these models, canonical tensor fields are introduced and further studied, including the Fisher metric and the Amari-Chentsov tensor, and embeddings of statistical manifolds are investigated. This novel foundation then leads to application highlights, such as generalizations and extensions of the classical uniqueness result of Chentsov or the Cramér-Rao inequality. Additionally, several new application fields of information geometry are highlighted, for instance hierarchical and graphical models, complexity theory, population genetics, or Markov Chain Monte Carlo. The book will be of interest to mathematicians who are interested in geometry, inf...

  4. Modeling the PUSPATI TRIGA Reactor using MCNP code

    International Nuclear Information System (INIS)

    Mohamad Hairie Rabir; Mark Dennis Usang; Naim Syauqi Hamzah; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

    2012-01-01

    The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core and shielding with literally no physical approximation. (author)

  5. Computerized cost model for pressurized water reactors

    International Nuclear Information System (INIS)

    Meneely, T.K.; Tabata, Hiroaki; Labourey, P.

    1999-01-01

    A computerized cost model has been developed in order to allow utility users to improve their familiarity with pressurized water reactor overnight capital costs and the various factors which influence them. This model organizes its cost data in the standard format of the Energy Economic Data Base (EEDB), and encapsulates simplified relationships between physical plant design information and capital cost information in a computer code. Model calculations are initiated from a base case, which was established using traditional cost calculation techniques. The user enters a set of plant design parameters, selected to allow consideration of plant models throughout the typical three- and four-loop PWR power range, and for plant sites in Japan, Europe, and the United States. Calculation of the new capital cost is then performed in a very brief time. The presentation of the program's output allows comparison of various cases with each other or with separately calculated baseline data. The user can start at a high level summary, and by selecting values of interest on a display grid show progressively more and more detailed information, including links to background information such as individual cost driver accounts and physical plant variables for each case. Graphical presentation of the comparison summaries is provided, and the numerical results may be exported to a spreadsheet for further processing. (author)

  6. Morphology and stratal geometry of the Antarctic continental shelf: Insights from models

    Science.gov (United States)

    Cooper, Alan K.; Barker, Peter F.; Brancolini, Giuliano

    1997-01-01

    Reconstruction of past ice-sheet fluctuations from the stratigraphy of glaciated continental shelves requires understanding of the relationships among the stratal geometry, glacial and marine sedimentary processes, and ice dynamics. We investigate the formation of the morphology and the broad stratal geometry of topsets on the Antarctic continental shelf with numerical models. Our models assume that the stratal geometry and morphology are principally the results of time-integrated effects of glacial erosion and sedimentation related to the location of the seaward edge of the grounded ice. The location of the grounding line varies with time almost randomly across the shelf. With these simple assumptions, the models can successfully mimic salient features of the morphology and the stratal geometry. The models suggest that the current shelf has gradually evolved to its present geometry by many glacial advances and retreats of the grounding line to different locations across the shelf. The locations of the grounding line do not appear to be linearly correlated with either fluctuations in the 5 l s O record (which presumably represents changes in the global ice volume) or with the global sea-level curve, suggesting that either a more complex relationship exists or local effects dominate. The models suggest that erosion of preglacial sediments is confined to the inner shelf, and erosion decreases and deposition increases toward the shelf edge. Some of the deposited glacial sediments must be derived from continental erosion. The sediments probably undergo extensive transport and reworking obliterating much of the evidence for their original depositional environment. The flexural rigidity and the tectonic subsidence of the underlying lithosphere modify the bathymetry of the shelf, but probably have little effect on the stratal geometry. Our models provide several guidelines for the interpretation of unconformities, the nature of preserved topset deposits, and the

  7. Adapting Dynamic Mathematical Models to a Pilot Anaerobic Digestion Reactor

    Directory of Open Access Journals (Sweden)

    F. Haugen, R. Bakke, and B. Lie

    2013-04-01

    Full Text Available A dynamic model has been adapted to a pilot anaerobic reactor fed diarymanure. Both steady-state data from online sensors and laboratory analysis anddynamic operational data from online sensors are used in the model adaptation.The model is based on material balances, and comprises four state variables,namely biodegradable volatile solids, volatile fatty acids, acid generatingmicrobes (acidogens, and methane generating microbes (methanogens. The modelcan predict the methane gas flow produced in the reactor. The model may beused for optimal reactor design and operation, state-estimation and control.Also, a dynamic model for the reactor temperature based on energy balance ofthe liquid in the reactor is adapted. This model may be used for optimizationand control when energy and economy are taken into account.

  8. Geometry optimization method versus predictive ability in QSPR modeling for ionic liquids

    Science.gov (United States)

    Rybinska, Anna; Sosnowska, Anita; Barycki, Maciej; Puzyn, Tomasz

    2016-02-01

    Computational techniques, such as Quantitative Structure-Property Relationship (QSPR) modeling, are very useful in predicting physicochemical properties of various chemicals. Building QSPR models requires calculating molecular descriptors and the proper choice of the geometry optimization method, which will be dedicated to specific structure of tested compounds. Herein, we examine the influence of the ionic liquids' (ILs) geometry optimization methods on the predictive ability of QSPR models by comparing three models. The models were developed based on the same experimental data on density collected for 66 ionic liquids, but with employing molecular descriptors calculated from molecular geometries optimized at three different levels of the theory, namely: (1) semi-empirical (PM7), (2) ab initio (HF/6-311+G*) and (3) density functional theory (B3LYP/6-311+G*). The model in which the descriptors were calculated by using ab initio HF/6-311+G* method indicated the best predictivity capabilities ({{Q}}_{{EXT}}2 = 0.87). However, PM7-based model has comparable values of quality parameters ({{Q}}_{{EXT}}2 = 0.84). Obtained results indicate that semi-empirical methods (faster and less expensive regarding CPU time) can be successfully employed to geometry optimization in QSPR studies for ionic liquids.

  9. Modelling of turbulence and combustion for simulation of gas explosions in complex geometries

    Energy Technology Data Exchange (ETDEWEB)

    Arntzen, Bjoern Johan

    1998-12-31

    This thesis analyses and presents new models for turbulent reactive flows for CFD (Computational Fluid Dynamics) simulation of gas explosions in complex geometries like offshore modules. The course of a gas explosion in a complex geometry is largely determined by the development of turbulence and the accompanying increased combustion rate. To be able to model the process it is necessary to use a CFD code as a starting point, provided with a suitable turbulence and combustion model. The modelling and calculations are done in a three-dimensional finite volume CFD code, where complex geometries are represented by a porosity concept, which gives porosity on the grid cell faces, depending on what is inside the cell. The turbulent flow field is modelled with a k-{epsilon} turbulence model. Subgrid models are used for production of turbulence from geometry not fully resolved on the grid. Results from laser doppler anemometry measurements around obstructions in steady and transient flows have been analysed and the turbulence models have been improved to handle transient, subgrid and reactive flows. The combustion is modelled with a burning velocity model and a flame model which incorporates the burning velocity into the code. Two different flame models have been developed: SIF (Simple Interface Flame model), which treats the flame as an interface between reactants and products, and the {beta}-model where the reaction zone is resolved with about three grid cells. The flame normally starts with a quasi laminar burning velocity, due to flame instabilities, modelled as a function of flame radius and laminar burning velocity. As the flow field becomes turbulent, the flame uses a turbulent burning velocity model based on experimental data and dependent on turbulence parameters and laminar burning velocity. The laminar burning velocity is modelled as a function of gas mixture, equivalence ratio, pressure and temperature in reactant. Simulations agree well with experiments. 139

  10. Application of autoregressive moving average model in reactor noise analysis

    International Nuclear Information System (INIS)

    Tran Dinh Tri

    1993-01-01

    The application of an autoregressive (AR) model to estimating noise measurements has achieved many successes in reactor noise analysis in the last ten years. The physical processes that take place in the nuclear reactor, however, are described by an autoregressive moving average (ARMA) model rather than by an AR model. Consequently more correct results could be obtained by applying the ARMA model instead of the AR model to reactor noise analysis. In this paper the system of the generalised Yule-Walker equations is derived from the equation of an ARMA model, then a method for its solution is given. Numerical results show the applications of the method proposed. (author)

  11. The development of model generators for specific reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chow, J.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    Authoring reactor models is a routine task for practitioners in nuclear engineering for reactor design, safety analysis, and code validation. The conventional approach is to use a text-editor to either manually manipulate an existing model or to assemble a new model by copying and pasting or direct typing. This approach is error-prone and substantial effort is required for verification. Alternatively, models can be generated programmatically for a specific system via a centralized data source and with rigid algorithms to generate models consistently and efficiently. This approach is demonstrated here for model generators for MCNP and KENO for the ZED-2 reactor. (author)

  12. Stochastic processes analysis in nuclear reactor using ARMA models

    International Nuclear Information System (INIS)

    Zavaljevski, N.

    1990-01-01

    The analysis of ARMA model derived from general stochastic state equations of nuclear reactor is given. The dependence of ARMA model parameters on the main physical characteristics of RB nuclear reactor in Vinca is presented. Preliminary identification results are presented, observed discrepancies between theory and experiment are explained and the possibilities of identification improvement are anticipated. (author)

  13. Neutron field control cybernetics model of RBMK reactor operator

    International Nuclear Information System (INIS)

    Polyakov, V.V.; Postnikov, V.V.; Sviridenkov, A.N.

    1992-01-01

    Results on parameter optimization for cybernetics model of RBMK reactor operator by power release control function are presented. Convolutions of various criteria applied previously in algorithms of the program 'Adviser to reactor operator' formed the basis of the model. 7 refs.; 4 figs

  14. Modeling atmospheric dispersion for reactor accident consequence evaluation

    International Nuclear Information System (INIS)

    Alpert, D.J.; Gudiksen, P.H.; Woodard, K.

    1982-01-01

    Atmospheric dispersion models are a central part of computer codes for the evaluation of potential reactor accident consequences. A variety of ways of treating to varying degrees the many physical processes that can have an impact on the predicted consequences exists. The currently available models are reviewed and their capabilities and limitations, as applied to reactor accident consequence analyses, are discussed

  15. Effects of homogeneous geometry models in simulating the fuel balls in HTR-10

    International Nuclear Information System (INIS)

    Wang Mengjen; Liang Jenqhorng; Peir Jinnjer; Chao Dersheng

    2012-01-01

    In this study, the core geometry of HTR-10 was simulated using four different models including: (1) model 1 - an explicit double heterogeneous geometry, (2) model 2 - a mixing of UO 2 kernel and four layers in each TRISO particle into one, (3) model 3 - a mixing of 8,335 TRISO particles and the inner graphite matrix in each fuel ball into one, and (4) model 4 - a mixing of the outer graphite shell, 8,335 TRISO particles, and the inner graphite matrix in each fuel ball into one. The associated initial core computations were performed using the MCNP version 1.51 computer code. The experimental fuel loading height of 123 cm was employed for each model. The results revealed that the multiplication factors ranged from largest to smallest with model 1, model 2, model 3, and model 4. The neutron spectrum in the fuel region of each models varied from the hardest to the softest are model 1, model 2, model 3, and model 4 while the averaged neutron spectrum in fuel ball from hardest to softest are model 4, model 3, model 2, and model 1. In addition, the CPU execution times extended from longest to shortest with model 1, model 2, model 3, and model 4. (author)

  16. A model to describe the performance of the UASB reactor.

    Science.gov (United States)

    Rodríguez-Gómez, Raúl; Renman, Gunno; Moreno, Luis; Liu, Longcheng

    2014-04-01

    A dynamic model to describe the performance of the Upflow Anaerobic Sludge Blanket (UASB) reactor was developed. It includes dispersion, advection, and reaction terms, as well as the resistances through which the substrate passes before its biotransformation. The UASB reactor is viewed as several continuous stirred tank reactors connected in series. The good agreement between experimental and simulated results shows that the model is able to predict the performance of the UASB reactor (i.e. substrate concentration, biomass concentration, granule size, and height of the sludge bed).

  17. Nitritation performance and biofilm development of co- and counter-diffusion biofilm reactors: Modeling and experimental comparison

    DEFF Research Database (Denmark)

    Wang, Rongchang; Terada, Akihiko; Lackner, Susanne

    2009-01-01

    A comparative study was conducted on the start-up performance and biofilm development in two different biofilm reactors with aim of obtaining partial nitritation. The reactors were both operated under oxygen limited conditions, but differed in geometry. While substrates (O-2, NH3) co......-diffused in one geometry, they counter-diffused in the other. Mathematical simulations of these two geometries were implemented in two 1-D multispecies biofilm models using the AQUASIM software. Sensitivity analysis results showed that the oxygen mass transfer coefficient (K-i) and maximum specific growth rate...... results showed that the counter-diffusion biofilms developed faster and attained a larger maximum biofilm thickness than the co-diffusion biofilms. Under oxygen limited condition (DO

  18. Modeling and fabrication of an RF MEMS variable capacitor with a fractal geometry

    KAUST Repository

    Elshurafa, Amro M.

    2013-08-16

    In this paper, we model, fabricate, and measure an electrostatically actuated MEMS variable capacitor that utilizes a fractal geometry and serpentine-like suspension arms. Explicitly, a variable capacitor that possesses a top suspended plate with a specific fractal geometry and also possesses a bottom fixed plate complementary in shape to the top plate has been fabricated in the PolyMUMPS process. An important benefit that was achieved from using the fractal geometry in designing the MEMS variable capacitor is increasing the tuning range of the variable capacitor beyond the typical ratio of 1.5. The modeling was carried out using the commercially available finite element software COMSOL to predict both the tuning range and pull-in voltage. Measurement results show that the tuning range is 2.5 at a maximum actuation voltage of 10V.

  19. Room acoustics modeling using a point-cloud representation of the room geometry

    DEFF Research Database (Denmark)

    Markovic, Milos; Olesen, Søren Krarup; Hammershøi, Dorte

    2013-01-01

    Room acoustics modeling is usually based on the room geometry that is parametrically described prior to a sound transmission calculation. This is a highly room-specific task and rather time consuming if a complex geometry is to be described. Here, a run time generic method for an arbitrary room...... geometry acquisition is presented. The method exploits a depth sensor of the Kinect device that provides a point based information of a scanned room interior. After post-processing of the Kinect output data, a 3D point-cloud model of the room is obtained. Sound transmission between two selected points...... level of user immersion by a real time acoustical simulation of a dynamic scenes....

  20. MATLAB/SIMULINK model of CANDU reactor for control studies

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2006-01-01

    In this paper a MATLAB/SIMULINK model is developed for a CANDU type reactor. The data for the reactor are taken from an Indian PHWR, which is very similar to CANDU in its design. Among the different feedback mechanisms in the core of the reactor, only xenon has been considered which plays an important role in spatial oscillations. The model is verified under closed loop scenarios with simple PI controller. The results of the simulation show that this model can be used for controller design and simulation of the reactor systems. Adding models of the other components of a CANDU reactor would ultimately result in a complete model of CANDU plant in MATLAB/SIMULINK. (author)

  1. Neutronics modeling of TRIGA reactor at the University of Utah using agent, KENO6 and MCNP5 codes

    International Nuclear Information System (INIS)

    Yang, X.; Xiao, S.; Choe, D.; Jevremovic, T.

    2010-01-01

    The TRIGA reactor at the University of Utah is modelled in 2D using the AGENT state-of-the-art methodology based on the Method of Characteristics (MOC) and R-function theory supporting detailed reactor analysis of reactor geometries of any type. The TRIGA reactor is also modelled using KENO6 and MCNP5 for comparison. The spatial flux and reaction rates distribution are visualized by AGENT graphics support. All methodologies are in use in to study the effect of different fuel configurations in developing practical educational exercises for students studying reactor physics. At the University of Utah we train graduate and undergraduate students in obtaining the Nuclear Regulatory Commission license in operating the TRIGA reactor. The computational models as developed are in support of these extensive training classes and in helping students visualize the reactor core characteristics in regard to neutron transport under various operational conditions. Additionally, the TRIGA reactor is under the consideration for power uprate; this fleet of computational tools once benchmarked against real measurements will provide us with validated 3D simulation models for simulating operating conditions of TRIGA. (author)

  2. Temperature profile in a fix-bed reactor and with cylindrical geometry by the method of orthogonal collocation

    International Nuclear Information System (INIS)

    Basta, C.

    1982-01-01

    Using the method of orthogonal colocation the boundary problem for a fix bed with cylindrical geometry was solved. The axial disposal term was despicable and the results were compared with those the explicit finite difference method. (E.G.) [pt

  3. Spinning geometry = Twisted geometry

    International Nuclear Information System (INIS)

    Freidel, Laurent; Ziprick, Jonathan

    2014-01-01

    It is well known that the SU(2)-gauge invariant phase space of loop gravity can be represented in terms of twisted geometries. These are piecewise-linear-flat geometries obtained by gluing together polyhedra, but the resulting geometries are not continuous across the faces. Here we show that this phase space can also be represented by continuous, piecewise-flat three-geometries called spinning geometries. These are composed of metric-flat three-cells glued together consistently. The geometry of each cell and the manner in which they are glued is compatible with the choice of fluxes and holonomies. We first remark that the fluxes provide each edge with an angular momentum. By studying the piecewise-flat geometries which minimize edge lengths, we show that these angular momenta can be literally interpreted as the spin of the edges: the geometries of all edges are necessarily helices. We also show that the compatibility of the gluing maps with the holonomy data results in the same conclusion. This shows that a spinning geometry represents a way to glue together the three-cells of a twisted geometry to form a continuous geometry which represents a point in the loop gravity phase space. (paper)

  4. The abstract geometry modeling language (AgML): experience and road map toward eRHIC

    International Nuclear Information System (INIS)

    Webb, Jason; Lauret, Jerome; Perevoztchikov, Victor

    2014-01-01

    The STAR experiment has adopted an Abstract Geometry Modeling Language (AgML) as the primary description of our geometry model. AgML establishes a level of abstraction, decoupling the definition of the detector from the software libraries used to create the concrete geometry model. Thus, AgML allows us to support both our legacy GEANT 3 simulation application and our ROOT/TGeo based reconstruction software from a single source, which is demonstrably self- consistent. While AgML was developed primarily as a tool to migrate away from our legacy FORTRAN-era geometry codes, it also provides a rich syntax geared towards the rapid development of detector models. AgML has been successfully employed by users to quickly develop and integrate the descriptions of several new detectors in the RHIC/STAR experiment including the Forward GEM Tracker (FGT) and Heavy Flavor Tracker (HFT) upgrades installed in STAR for the 2012 and 2013 runs. AgML has furthermore been heavily utilized to study future upgrades to the STAR detector as it prepares for the eRHIC era. With its track record of practical use in a live experiment in mind, we present the status, lessons learned and future of the AgML language as well as our experience in bringing the code into our production and development environments. We will discuss the path toward eRHIC and pushing the current model to accommodate for detector miss-alignment and high precision physics.

  5. Monte Carlo modelling of the Belgian materials testing reactor BR2: present status

    International Nuclear Information System (INIS)

    Verboomen, B.; Aoust, Th.; Raedt, Ch. de; Beeckmans de West-Meerbeeck, A.

    2001-01-01

    A very detailed 3-D MCNP-4B model of the BR2 reactor was developed to perform all neutron and gamma calculations needed for the design of new experimental irradiation rigs. The Monte Carlo model of BR2 includes the nearly exact geometrical representation of fuel elements (now with their axially varying burn-up), of partially inserted control and regulating rods, of experimental devices and of radioisotope production rigs. The multiple level-geometry possibilities of MCNP-4B are fully exploited to obtain sufficiently flexible tools to cope with the very changing core loading. (orig.)

  6. Application of adobe flash media to optimize jigsaw learning model on geometry material

    Science.gov (United States)

    Imam, P.; Imam, S.; Ikrar, P.

    2018-05-01

    This study aims to determine and describe the effectiveness of the application of adobe flash media for jigsaw learning model on geometry material. In this study, the modified jigsaw learning with adobe flash media is called jigsaw-flash model. This research was conducted in Surakarta. The research method used is mix method research with exploratory sequential strategy. The results of this study indicate that students feel more comfortable and interested in studying geometry material taught by jigsaw-flash model. In addition, students taught using the jigsaw-flash model are more active and motivated than the students who were taught using ordinary jigsaw models. This shows that the use of the jigsaw-flash model can increase student participation and motivation. It can be concluded that the adobe flash media can be used as a solution to reduce the level of student abstraction in learning mathematics.

  7. KENO3D Visualization Tool for KENO V.a and KENO-VI Geometry Models

    International Nuclear Information System (INIS)

    Horwedel, J.E.; Bowman, S.M.

    2000-01-01

    Criticality safety analyses often require detailed modeling of complex geometries. Effective visualization tools can enhance checking the accuracy of these models. This report describes the KENO3D visualization tool developed at the Oak Ridge National Laboratory (ORNL) to provide visualization of KENO V.a and KENO-VI criticality safety models. The development of KENO3D is part of the current efforts to enhance the SCALE (Standardized Computer Analyses for Licensing Evaluations) computer software system

  8. Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

    International Nuclear Information System (INIS)

    Pan, Dongqing; Chien Jen, Tien; Li, Tao; Yuan, Chris

    2014-01-01

    This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired

  9. Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Dongqing; Chien Jen, Tien [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, Milwaukee, Wisconsin 53201 (United States); Li, Tao [School of Mechanical Engineering, Dalian University of Technology, Dalian 116024 (China); Yuan, Chris, E-mail: cyuan@uwm.edu [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, 3200 North Cramer Street, Milwaukee, Wisconsin 53211 (United States)

    2014-01-15

    This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired.

  10. Advanced DPSM approach for modeling ultrasonic wave scattering in an arbitrary geometry

    Science.gov (United States)

    Yadav, Susheel K.; Banerjee, Sourav; Kundu, Tribikram

    2011-04-01

    Several techniques are used to diagnose structural damages. In the ultrasonic technique structures are tested by analyzing ultrasonic signals scattered by damages. The interpretation of these signals requires a good understanding of the interaction between ultrasonic waves and structures. Therefore, researchers need analytical or numerical techniques to have a clear understanding of the interaction between ultrasonic waves and structural damage. However, modeling of wave scattering phenomenon by conventional numerical techniques such as finite element method requires very fine mesh at high frequencies necessitating heavy computational power. Distributed point source method (DPSM) is a newly developed robust mesh free technique to simulate ultrasonic, electrostatic and electromagnetic fields. In most of the previous studies the DPSM technique has been applied to model two dimensional surface geometries and simple three dimensional scatterer geometries. It was difficult to perform the analysis for complex three dimensional geometries. This technique has been extended to model wave scattering in an arbitrary geometry. In this paper a channel section idealized as a thin solid plate with several rivet holes is formulated. The simulation has been carried out with and without cracks near the rivet holes. Further, a comparison study has been also carried out to characterize the crack. A computer code has been developed in C for modeling the ultrasonic field in a solid plate with and without cracks near the rivet holes.

  11. Geometry and transport in a model of two coupled quadratic nonlinear waveguides

    DEFF Research Database (Denmark)

    Stirling, James R.; Bang, Ole; Christiansen, Peter Leth

    2008-01-01

    This paper applies geometric methods developed to understand chaos and transport in Hamiltonian systems to the study of power distribution in nonlinear waveguide arrays. The specific case of two linearly coupled X(2) waveguides is modeled and analyzed in terms of transport and geometry in the pha...

  12. Geometry Laboratory (GEOLAB) surface modeling and grid generation technology and services

    Science.gov (United States)

    Kerr, Patricia A.; Smith, Robert E.; Posenau, Mary-Anne K.

    1995-01-01

    The facilities and services of the GEOmetry LABoratory (GEOLAB) at the NASA Langley Research Center are described. Included in this description are the laboratory functions, the surface modeling and grid generation technologies used in the laboratory, and examples of the tasks performed in the laboratory.

  13. Finite element modeling of the neuron-electrode interface: stimulus transfer and geometry

    NARCIS (Netherlands)

    Buitenweg, Jan R.; Rutten, Wim; Marani, Enrico

    1999-01-01

    The relation between stimulus transfer and the geometry of the neuron-electrode interface can not be determined properly using electrical equivalent circuits, since current that flows from the sealing gap through the neuronal membrane is difficult to model in these circuits. Therefore, finite

  14. Importance of the energy-dependent geometry in the 16O+ 16O optical model potential

    International Nuclear Information System (INIS)

    Pantis, G.; Ioannidis, K.; Poirier, P.

    1985-01-01

    Optical model potentials with various forms of energy-dependent geometry have been considered for the description of 16 O+ 16 O elastic scattering. It is shown that the variation with energy of the imaginary radius leads to a reasonable fit of the cross-section data, throughout the energy range

  15. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  16. Verification of thermo-fluidic CVD reactor model

    International Nuclear Information System (INIS)

    Lisik, Z; Turczynski, M; Ruta, L; Raj, E

    2014-01-01

    Presented paper describes the numerical model of CVD (Chemical Vapour Deposition) reactor created in ANSYS CFX, whose main purpose is the evaluation of numerical approaches used to modelling of heat and mass transfer inside the reactor chamber. Verification of the worked out CVD model has been conducted with measurements under various thermal, pressure and gas flow rate conditions. Good agreement between experimental and numerical results confirms correctness of the elaborated model.

  17. KENO3D, Visualisation Tool for KENO V.A and KENO-VI Geometry Models

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: The KENO3D Visualization Tool for KENO Geometry Models is a powerful state-of-the-art visualization tool that enables KENO V.a users and KENO-VI to interactively display their three-dimensional geometry models. The KENO3D interactive options include: - Shaded or wire-frame images ; - Standard views such as top view, side view, front view, and isometric(3-D) view; - Rotating the model ; - Zooming in on selected locations ; - Selecting parts of the model to display ; - Editing colors and displaying legends ; - Displaying properties of any unit in the model ; - Creating cut-away views ; - Removing units from the model; - Printing image or saving image to a common graphics formats. KENO3D was developed for use by criticality safety specialists that use the KENO three-dimensional Monte Carlo criticality computer code. KENO V.a and KENO-VI are part of the SCALE (Standardized Computer Analyses for Licensing Evaluations) computer software system developed at Ridge National Laboratory (ORNL) that is widely used and accepted around the world for criticality safety analyses. 2 - Methods: KENO3D reads CSAS, KENO V.a, or KENO-VI input files. It attempts to verify that the KENO geometry input is 'legal', i.e., it conforms to the code input guidelines. KENO3D prints a warning message for illegal geometry input, and if possible, it displays the illegal KENO geometry to facilitate debugging of the input. Problems with more than 300,000 KENO V.a bodies have been successfully tested and displayed. KENO3D has the look and feel of a typical PC Windows application. Toolbar buttons are included for all major menu options. There is a setup dialog that allows the user to specify toolbars that should be displayed

  18. Modeling cavities exhibiting strong lateral confinement using open geometry Fourier modal method

    DEFF Research Database (Denmark)

    Häyrynen, Teppo; Gregersen, Niels

    2016-01-01

    We have developed a computationally efficient Fourier-Bessel expansion based open geometry formalism for modeling the optical properties of rotationally symmetric photonic nanostructures. The lateral computation domain is assumed infinite so that no artificial boundary conditions are needed. Instead,...... around a geometry specific dominant transverse wavenumber region. We will use the developed approach to investigate the Q factor and mode confinement in cavities where top DBR mirror has small rectangular defect confining the modes laterally on the defect region....

  19. Study of skin model and geometry effects on thermal performance of thermal protective fabrics

    Science.gov (United States)

    Zhu, Fanglong; Ma, Suqin; Zhang, Weiyuan

    2008-05-01

    Thermal protective clothing has steadily improved over the years as new materials and improved designs have reached the market. A significant method that has brought these improvements to the fire service is the NFPA 1971 standard on structural fire fighters’ protective clothing. However, this testing often neglects the effects of cylindrical geometry on heat transmission in flame resistant fabrics. This paper deals with methods to develop cylindrical geometry testing apparatus incorporating novel skin bioheat transfer model to test flame resistant fabrics used in firefighting. Results show that fabrics which shrink during the test can have reduced thermal protective performance compared with the qualities measured with a planar geometry tester. Results of temperature differences between skin simulant sensors of planar and cylindrical tester are also compared. This test method provides a new technique to accurately and precisely characterize the thermal performance of thermal protective fabrics.

  20. An improved algorithm to convert CAD model to MCNP geometry model based on STEP file

    International Nuclear Information System (INIS)

    Zhou, Qingguo; Yang, Jiaming; Wu, Jiong; Tian, Yanshan; Wang, Junqiong; Jiang, Hai; Li, Kuan-Ching

    2015-01-01

    Highlights: • Fully exploits common features of cells, making the processing efficient. • Accurately provide the cell position. • Flexible to add new parameters in the structure. • Application of novel structure in INP file processing, conveniently evaluate cell location. - Abstract: MCNP (Monte Carlo N-Particle Transport Code) is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. Its input file, the INP file, has the characteristics of complicated form and is error-prone when describing geometric models. Due to this, a conversion algorithm that can solve the problem by converting general geometric model to MCNP model during MCNP aided modeling is highly needed. In this paper, we revised and incorporated a number of improvements over our previous work (Yang et al., 2013), which was proposed and targeted after STEP file and INP file were analyzed. Results of experiments show that the revised algorithm is more applicable and efficient than previous work, with the optimized extraction of geometry and topology information of the STEP file, as well as the production efficiency of output INP file. This proposed research is promising, and serves as valuable reference for the majority of researchers involved with MCNP-related researches

  1. Feasibility Study of Ex Ovo Chick Chorioallantoic Artery Model for Investigating Pulsatile Variation of Arterial Geometry.

    Directory of Open Access Journals (Sweden)

    Kweon-Ho Nam

    Full Text Available Despite considerable research efforts on the relationship between arterial geometry and cardiovascular pathology, information is lacking on the pulsatile geometrical variation caused by arterial distensibility and cardiomotility because of the lack of suitable in vivo experimental models and the methodological difficulties in examining the arterial dynamics. We aimed to investigate the feasibility of using a chick embryo system as an experimental model for basic research on the pulsatile variation of arterial geometry. Optical microscope video images of various arterial shapes in chick chorioallantoic circulation were recorded from different locations and different embryo samples. The high optical transparency of the chorioallantoic membrane (CAM allowed clear observation of tiny vessels and their movements. Systolic and diastolic changes in arterial geometry were visualized by detecting the wall boundaries from binary images. Several to hundreds of microns of wall displacement variations were recognized during a pulsatile cycle. The spatial maps of the wall motion harmonics and magnitude ratio of harmonic components were obtained by analyzing the temporal brightness variation at each pixel in sequential grayscale images using spectral analysis techniques. The local variations in the spectral characteristics of the arterial wall motion were reflected well in the analysis results. In addition, mapping the phase angle of the fundamental frequency identified the regional variations in the wall motion directivity and phase shift. Regional variations in wall motion phase angle and fundamental-to-second harmonic ratio were remarkable near the bifurcation area. In summary, wall motion in various arterial geometry including straight, curved and bifurcated shapes was well observed in the CAM artery model, and their local and cyclic variations could be characterized by Fourier and wavelet transforms of the acquired video images. The CAM artery model with

  2. A Physically—Based Geometry Model for Transport Distance Estimation of Rainfall-Eroded Soil Sediment

    Directory of Open Access Journals (Sweden)

    Qian-Gui Zhang

    2016-01-01

    Full Text Available Estimations of rainfall-induced soil erosion are mostly derived from the weight of sediment measured in natural runoff. The transport distance of eroded soil is important for evaluating landscape evolution but is difficult to estimate, mainly because it cannot be linked directly to the eroded sediment weight. The volume of eroded soil is easier to calculate visually using popular imaging tools, which can aid in estimating the transport distance of eroded soil through geometry relationships. In this study, we present a straightforward geometry model to predict the maximum sediment transport distance incurred by rainfall events of various intensity and duration. In order to verify our geometry prediction model, a series of experiments are reported in the form of a sediment volume. The results show that cumulative rainfall has a linear relationship with the total volume of eroded soil. The geometry model can accurately estimate the maximum transport distance of eroded soil by cumulative rainfall, with a low root-mean-square error (4.7–4.8 and a strong linear correlation (0.74–0.86.

  3. A CAD based geometry model for simulation and analysis of particle detector data

    Energy Technology Data Exchange (ETDEWEB)

    Milde, Michael; Losekamm, Martin; Poeschl, Thomas; Greenwald, Daniel; Paul, Stephan [Technische Universitaet Muenchen, 85748 Garching (Germany)

    2016-07-01

    The development of a new particle detector requires a good understanding of its setup. A detailed model of the detector's geometry is not only needed during construction, but also for simulation and data analysis. To arrive at a consistent description of the detector geometry a representation is needed that can be easily implemented in different software tools used during data analysis. We developed a geometry representation based on CAD files that can be easily used within the Geant4 simulation framework and analysis tools based on the ROOT framework. This talk presents the structure of the geometry model and show its implementation using the example of the event reconstruction developed for the Multi-purpose Active-target Particle Telescope (MAPT). The detector consists of scintillating plastic fibers and can be used as a tracking detector and calorimeter with omnidirectional acceptance. To optimize the angular resolution and the energy reconstruction of measured particles, a detailed detector model is needed at all stages of the reconstruction.

  4. INTERVAL OBSERVER FOR A BIOLOGICAL REACTOR MODEL

    Directory of Open Access Journals (Sweden)

    T. A. Kharkovskaia

    2014-05-01

    Full Text Available The method of an interval observer design for nonlinear systems with parametric uncertainties is considered. The interval observer synthesis problem for systems with varying parameters consists in the following. If there is the uncertainty restraint for the state values of the system, limiting the initial conditions of the system and the set of admissible values for the vector of unknown parameters and inputs, the interval existence condition for the estimations of the system state variables, containing the actual state at a given time, needs to be held valid over the whole considered time segment as well. Conditions of the interval observers design for the considered class of systems are shown. They are: limitation of the input and state, the existence of a majorizing function defining the uncertainty vector for the system, Lipschitz continuity or finiteness of this function, the existence of an observer gain with the suitable Lyapunov matrix. The main condition for design of such a device is cooperativity of the interval estimation error dynamics. An individual observer gain matrix selection problem is considered. In order to ensure the property of cooperativity for interval estimation error dynamics, a static transformation of coordinates is proposed. The proposed algorithm is demonstrated by computer modeling of the biological reactor. Possible applications of these interval estimation systems are the spheres of robust control, where the presence of various types of uncertainties in the system dynamics is assumed, biotechnology and environmental systems and processes, mechatronics and robotics, etc.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  6. MODELING OF TUBULAR ELECTROCHEMICAL REACTOR FOR DYE REMOVAL

    Directory of Open Access Journals (Sweden)

    V. VIJAYAKUMAR

    2017-06-01

    Full Text Available The aim of the present investigation is to model a tubular electrochemical reactor for the treatment of synthetic dye wastewater. The tubular reactor was modeled and solved by finite difference method. For the model solution, the column was divided into 11 nodes in the axial direction and the variation in the radial direction has been neglected. An initial dye concentration of 200 mg L-1was taken in the reservoir. The reactor was operated in a batch with recirculation operation. Based on preliminary experiments all parameters have been optimized. The model simulation is compared with the experimental value and it is observed that the model fairly matches well with the experiment. The modeling of tubular electrochemical reactors for dye waste water treatment could be useful in the design and scale up of electrochemical process.

  7. Effect of the kinetics of ammonium and nitrite oxidation on nitritation success or failure for different biofilm reactor geometries

    DEFF Research Database (Denmark)

    Lackner, Susanne; Smets, Barth F.

    2012-01-01

    was on the influence of key biokinetic parameters (maximum specific growth rates, oxygen and nitrogen affinity constants of AOB (ammonium oxidizing bacteria) and NOB (nitrite oxidizing bacteria)) and their ratios on nitritation efficiency in these geometries. This exhaustive simulation study revealed that nitritation...... strongly depends on the chosen kinetic parameters of AOB and NOB. The maximum specific growth rates (μmax,AOB and μmax,NOB) had the strongest impact on nitritation efficiency (NE). In comparison, the counter-diffusion geometry yielded more parameter combinations (27.5%) that resulted in high NE than the co...

  8. Almost-commutative geometries beyond the standard model II: new colours

    International Nuclear Information System (INIS)

    Stephan, Christoph A

    2007-01-01

    We will present an extension of the standard model of particle physics in its almost-commutative formulation. This extension is guided by the minimal approach to almost-commutative geometries employed by Iochum et al (2004 J. Math. Phys. 45 5003 (Preprint hep-th/0312276)), Jureit and Stephan (2005 J. Math. Phys. 46 043512 (Preprint hep-th/0501134)), Schuecker (2005 Preprint hep-th/0501181), Jureit et al (2005 J. Math. Phys. 46 072303 (Preprint hep-th/0503190)) and Jureit and Stephan (2006 Preprint hep-th/0610040), although the model presented here is not minimal itself. The corresponding almost-commutative geometry leads to a Yang-Mills-Higgs model which consists of the standard model and two new fermions of opposite electromagnetic charge which may possess a new colour-like gauge group. As a new phenomenon, grand unification is no longer required by the spectral action

  9. Slab1.0: A three-dimensional model of global subduction zone geometries

    Science.gov (United States)

    Hayes, Gavin P.; Wald, David J.; Johnson, Rebecca L.

    2012-01-01

    We describe and present a new model of global subduction zone geometries, called Slab1.0. An extension of previous efforts to constrain the two-dimensional non-planar geometry of subduction zones around the focus of large earthquakes, Slab1.0 describes the detailed, non-planar, three-dimensional geometry of approximately 85% of subduction zones worldwide. While the model focuses on the detailed form of each slab from their trenches through the seismogenic zone, where it combines data sets from active source and passive seismology, it also continues to the limits of their seismic extent in the upper-mid mantle, providing a uniform approach to the definition of the entire seismically active slab geometry. Examples are shown for two well-constrained global locations; models for many other regions are available and can be freely downloaded in several formats from our new Slab1.0 website, http://on.doi.gov/d9ARbS. We describe improvements in our two-dimensional geometry constraint inversion, including the use of ‘average’ active source seismic data profiles in the shallow trench regions where data are otherwise lacking, derived from the interpolation between other active source seismic data along-strike in the same subduction zone. We include several analyses of the uncertainty and robustness of our three-dimensional interpolation methods. In addition, we use the filtered, subduction-related earthquake data sets compiled to build Slab1.0 in a reassessment of previous analyses of the deep limit of the thrust interface seismogenic zone for all subduction zones included in our global model thus far, concluding that the width of these seismogenic zones is on average 30% larger than previous studies have suggested.

  10. Thermal modeling of the ceramic composite fuel for light water reactors

    International Nuclear Information System (INIS)

    Revankar, S.T.; Latta, R.; Solomon, A.A.

    2005-01-01

    Full text of publication follows: Composite fuel designs capable of providing improved thermal performance are of great interest in advanced reactor designs where high efficiency and long fuel cycles are desired. Thermal modeling of the composite fuel consisting of continuous second phase in a ceramic (uranium oxide) matrix has been carried out with detailed examination of the microstructure of the composite and the interface. Assuming that constituent phases are arranged as slabs, upper and lower bounds for the thermal conductivity of the composite are derived analytically. Bounding calculations on the thermal conductivity of the composite were performed for SiC dispersed in the UO 2 matrix. It is found that with 10% SiC, the thermal conductivity increases from 5.8 to 9.8 W/m.deg. K at 500 K, or an increase of 69% was observed in UO 2 matrix. The finite element analysis computer program ANSYS was used to create composite fuel geometries with set boundary conditions to produce accurate thermal conductivity predictions. A model developed also accounts for SiC-matrix interface resistance and the addition of coatings or interaction barriers. The first set of calculations using the code was to model simple series and parallel fuel slab geometries, and then advance to inter-connected parallel pathways. The analytical calculations were compared with the ANSYS results. The geometry of the model was set up as a 1 cm long by 400 micron wide rectangle. This rectangle was then divided into one hundred sections with the first ninety percent of a single section being UO 2 and the remaining ten percent consisting of SiC. The model was then meshed using triangular type elements. The boundary conditions were set with the sides of the rectangle being adiabatic and having an assigned temperature at the end of the rectangle. A heat flux was then applied to one end of the model producing a temperature gradient. The effective thermal conductivity was then calculated using the geometry

  11. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Borum, Robert C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chaleff, Ethan S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rogerson, Doug W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J. [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation, Canton, MI (United States)

    2014-08-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  12. Dynamic modeling of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    March-Leuba, J.; Ibn-Khayat, M.

    1990-01-01

    The purpose of this paper is to provide a summary description and some applications of a computer model that has been developed to simulate the dynamic behavior of the advanced neutron source (ANS) reactor. The ANS dynamic model is coded in the advanced continuous simulation language (ACSL), and it represents the reactor core, vessel, primary cooling system, and secondary cooling systems. The use of a simple dynamic model in the early stages of the reactor design has proven very valuable not only in the development of the control and plant protection system but also of components such as pumps and heat exchangers that are usually sized based on steady-state calculations

  13. Self consistent MHD modeling of the solar wind from coronal holes with distinct geometries

    Science.gov (United States)

    Stewart, G. A.; Bravo, S.

    1995-01-01

    Utilizing an iterative scheme, a self-consistent axisymmetric MHD model for the solar wind has been developed. We use this model to evaluate the properties of the solar wind issuing from the open polar coronal hole regions of the Sun, during solar minimum. We explore the variation of solar wind parameters across the extent of the hole and we investigate how these variations are affected by the geometry of the hole and the strength of the field at the coronal base.

  14. (U) Influence of Compaction Model Form on Planar and Cylindrical Compaction Geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fredenburg, David A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carney, Theodore Clayton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fichtl, Christopher Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ramsey, Scott D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2018-01-05

    The dynamic compaction response of CeO2 is examined within the frameworks of the Ramp and P-a compaction models. Hydrocode calculations simulating the dynamic response of CeO2 at several distinct pressures within the compaction region are investigated in both planar and cylindrically convergent geometries. Findings suggest additional validation of the compaction models is warranted under complex loading configurations.

  15. Free-energy analysis of spin models on hyperbolic lattice geometries.

    Science.gov (United States)

    Serina, Marcel; Genzor, Jozef; Lee, Yoju; Gendiar, Andrej

    2016-04-01

    We investigate relations between spatial properties of the free energy and the radius of Gaussian curvature of the underlying curved lattice geometries. For this purpose we derive recurrence relations for the analysis of the free energy normalized per lattice site of various multistate spin models in the thermal equilibrium on distinct non-Euclidean surface lattices of the infinite sizes. Whereas the free energy is calculated numerically by means of the corner transfer matrix renormalization group algorithm, the radius of curvature has an analytic expression. Two tasks are considered in this work. First, we search for such a lattice geometry, which minimizes the free energy per site. We conjecture that the only Euclidean flat geometry results in the minimal free energy per site regardless of the spin model. Second, the relations among the free energy, the radius of curvature, and the phase transition temperatures are analyzed. We found out that both the free energy and the phase transition temperature inherit the structure of the lattice geometry and asymptotically approach the profile of the Gaussian radius of curvature. This achievement opens new perspectives in the AdS-CFT correspondence theories.

  16. Geometry modeling of single track cladding deposited by high power diode laser with rectangular beam spot

    Science.gov (United States)

    Liu, Huaming; Qin, Xunpeng; Huang, Song; Hu, Zeqi; Ni, Mao

    2018-01-01

    This paper presents an investigation on the relationship between the process parameters and geometrical characteristics of the sectional profile for the single track cladding (STC) deposited by High Power Diode Laser (HPDL) with rectangle beam spot (RBS). To obtain the geometry parameters, namely cladding width Wc and height Hc of the sectional profile, a full factorial design (FFD) of experiment was used to conduct the experiments with a total of 27. The pre-placed powder technique has been employed during laser cladding. The influence of the process parameters including laser power, powder thickness and scanning speed on the Wc and Hc was analyzed in detail. A nonlinear fitting model was used to fit the relationship between the process parameters and geometry parameters. And a circular arc was adopted to describe the geometry profile of the cross-section of STC. The above models were confirmed by all the experiments. The results indicated that the geometrical characteristics of the sectional profile of STC can be described as the circular arc, and the other geometry parameters of the sectional profile can be calculated only using Wc and Hc. Meanwhile, the Wc and Hc can be predicted through the process parameters.

  17. Safe design and operation of fluidized-bed reactors: Choice between reactor models

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1990-01-01

    For three different catalytic fluidized bed reactor models, two models presented by Werther and a model presented by van Deemter, the region of safe and unique operation for a chosen reaction system was investigated. Three reaction systems were used: the oxidation of benzene to maleic anhydride, the

  18. Geometry characteristics modeling and process optimization in coaxial laser inside wire cladding

    Science.gov (United States)

    Shi, Jianjun; Zhu, Ping; Fu, Geyan; Shi, Shihong

    2018-05-01

    Coaxial laser inside wire cladding method is very promising as it has a very high efficiency and a consistent interaction between the laser and wire. In this paper, the energy and mass conservation law, and the regression algorithm are used together for establishing the mathematical models to study the relationship between the layer geometry characteristics (width, height and cross section area) and process parameters (laser power, scanning velocity and wire feeding speed). At the selected parameter ranges, the predicted values from the models are compared with the experimental measured results, and there is minor error existing, but they reflect the same regularity. From the models, it is seen the width of the cladding layer is proportional to both the laser power and wire feeding speed, while it firstly increases and then decreases with the increasing of the scanning velocity. The height of the cladding layer is proportional to the scanning velocity and feeding speed and inversely proportional to the laser power. The cross section area increases with the increasing of feeding speed and decreasing of scanning velocity. By using the mathematical models, the geometry characteristics of the cladding layer can be predicted by the known process parameters. Conversely, the process parameters can be calculated by the targeted geometry characteristics. The models are also suitable for multi-layer forming process. By using the optimized process parameters calculated from the models, a 45 mm-high thin-wall part is formed with smooth side surfaces.

  19. Numerical modeling of a nuclear production reactor cooling lake

    International Nuclear Information System (INIS)

    Hamm, L.L.; Pepper, D.W.

    1987-01-01

    A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

  20. A simplified model of aerosol removal by natural processes in reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A.; Washington, K.E.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States); Burson, S.B. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-07-01

    Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.

  1. Babcock and Wilcox model for predicting in-reactor densification

    International Nuclear Information System (INIS)

    Buescher, B.J.; Pegram, J.W.

    1975-06-01

    The B and W fuel densification model is used to describe the extent and kinetics of in-reactor densification in B and W production fuel. The model and approach are qualified against an extensive data base available through B and W's participation in the EEI Fuel Densification Program. Out-of-reactor resintering tests on representative pellets from each batch of fuel are used to provide input parameters to the B and W densification model. The B and W densification model predicts in-reactor densification very accurately for pellets operated at heat rates above 5 kW/ft and with considerable conservation for pellets operated at heat rates less than 5 kW/ft. This model represents a technically rigorous and conservative basis for predicting the extent and kinetics of in-reactor densification. 9 references. (U.S.)

  2. Representing Misalignments of the STAR Geometry Model using AgML

    Science.gov (United States)

    Webb, Jason C.; Lauret, Jérôme; Perevotchikov, Victor; Smirnov, Dmitri; Van Buren, Gene

    2017-10-01

    The STAR Heavy Flavor Tracker (HFT) was designed to provide high-precision tracking for the identification of charmed hadron decays in heavy-ion collisions at RHIC. It consists of three independently mounted subsystems, providing four precision measurements along the track trajectory, with the goal of pointing decay daughters back to vertices displaced by less than 100 microns from the primary event vertex. The ultimate efficiency and resolution of the physics analysis will be driven by the quality of the simulation and reconstruction of events in heavy-ion collisions. In particular, it is important that the geometry model properly accounts for the relative misalignments of the HFT subsystems, along with the alignment of the HFT relative to STARs primary tracking detector, the Time Projection Chamber (TPC). The Abstract Geometry Modeling Language (AgML) provides a single description of the STAR geometry, generating both our simulation (GEANT 3) and reconstruction geometries (ROOT). AgML implements an ideal detector model, while misalignments are stored separately in database tables. These have historically been applied at the hit level. Simulated detector hits are projected from their ideal position along the track’s trajectory, until they intersect the misaligned detector volume, where the struck detector element is calculated for hit digitization. This scheme has worked well as hit errors have been negligible compared with the size of sensitive volumes. The precision and complexity of the HFT detector require us to apply misalignments to the detector volumes themselves. In this paper we summarize the extension of the AgML language and support libraries to enable the static misalignment of our reconstruction and simulation geometries, discussing the design goals, limitations and path to full misalignment support in ROOT/VMC-based simulation.

  3. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  4. Dependence of Dynamic Modeling Accuracy on Sensor Measurements, Mass Properties, and Aircraft Geometry

    Science.gov (United States)

    Grauer, Jared A.; Morelli, Eugene A.

    2013-01-01

    The NASA Generic Transport Model (GTM) nonlinear simulation was used to investigate the effects of errors in sensor measurements, mass properties, and aircraft geometry on the accuracy of identified parameters in mathematical models describing the flight dynamics and determined from flight data. Measurements from a typical flight condition and system identification maneuver were systematically and progressively deteriorated by introducing noise, resolution errors, and bias errors. The data were then used to estimate nondimensional stability and control derivatives within a Monte Carlo simulation. Based on these results, recommendations are provided for maximum allowable errors in sensor measurements, mass properties, and aircraft geometry to achieve desired levels of dynamic modeling accuracy. Results using additional flight conditions and parameter estimation methods, as well as a nonlinear flight simulation of the General Dynamics F-16 aircraft, were compared with these recommendations

  5. Nuclear reactor power control system based on flexibility model

    International Nuclear Information System (INIS)

    Li Gang; Zhao Fuyu; Li Chong; Tai Yun

    2011-01-01

    Design the nuclear reactor power control system in this paper to cater to a nonlinear nuclear reactor. First, calculate linear power models at five power levels of the reactor as five local models and design controllers of the local models as local controllers. Every local controller consists of an optimal controller contrived by the toolbox of Optimal Controller Designer (OCD) and a proportion-integration-differentiation (PID) controller devised via Genetic Algorithm (GA) to set parameters of the PID controller. According to the local models and controllers, apply the principle of flexibility model developed in the paper to obtain the flexibility model and the flexibility controller at every power level. Second, the flexibility model and the flexibility controller at a level structure the power control system of this level. The set of the whole power control systems corresponding to global power levels is to approximately carry out the power control of the reactor. Finally, the nuclear reactor power control system is simulated. The simulation result shows that the idea of flexibility model is feasible and the nuclear reactor power control system is effective. (author)

  6. COPS model estimates of LLEA availability near selected reactor sites

    International Nuclear Information System (INIS)

    Berkbigler, K.P.

    1979-11-01

    The COPS computer model has been used to estimate local law enforcement agency (LLEA) officer availability in the neighborhood of selected nuclear reactor sites. The results of these analyses are presented both in graphic and tabular form in this report

  7. Modeling issues associated with production reactor safety assessment

    International Nuclear Information System (INIS)

    Stack, D.W.; Thomas, W.R.

    1990-01-01

    This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs

  8. Simulation of MILD combustion using Perfectly Stirred Reactor model

    KAUST Repository

    Chen, Z.; Vanteru, Mahendra Reddy; Ruan, S.; Doan, N. A K; Roberts, William L.; Swaminathan, N.

    2016-01-01

    A simple model based on a Perfectly Stirred Reactor (PSR) is proposed for moderate or intense low-oxygen dilution (MILD) combustion. The PSR calculation is performed covering the entire flammability range and the tabulated chemistry approach is used

  9. LRS Bianchi Type II Massive String Cosmological Models with Magnetic Field in Lyra's Geometry

    Directory of Open Access Journals (Sweden)

    Raj Bali

    2013-01-01

    Full Text Available Bianchi type II massive string cosmological models with magnetic field and time dependent gauge function ( in the frame work of Lyra's geometry are investigated. The magnetic field is in -plane. To get the deterministic solution, we have assumed that the shear ( is proportional to the expansion (. This leads to , where and are metric potentials and is a constant. We find that the models start with a big bang at initial singularity and expansion decreases due to lapse of time. The anisotropy is maintained throughout but the model isotropizes when . The physical and geometrical aspects of the model in the presence and absence of magnetic field are also discussed.

  10. Random detailed model for probabilistic neutronic calculation in pebble bed Very High Temperature Reactors

    International Nuclear Information System (INIS)

    Perez Curbelo, J.; Rosales, J.; Garcia, L.; Garcia, C.; Brayner, C.

    2013-01-01

    The pebble bed nuclear reactor is one of the main candidates for the next generation of nuclear power plants. In pebble bed type HTRs, the fuel is contained within graphite pebbles in the form of TRISO particles, which form a randomly packed bed inside a graphite-walled cylindrical cavity. Pebble bed reactors (PBR) offer the opportunity to meet the sustainability requirements, such as nuclear safety, economic competitiveness, proliferation resistance and a minimal production of radioactive waste. In order to simulate PBRs correctly, the double heterogeneity of the system must be considered. It consists on randomly located pebbles into the core and TRISO particles into the fuel pebbles. These features are often neglected due to the difficulty to model with MCPN code. The main reason is that there is a limited number of cells and surfaces to be defined. In this study, a computational tool which allows getting a new geometrical model of fuel pebbles for neutronic calculations with MCNPX code, was developed. The heterogeneity of system is considered, and also the randomly located TRISO particles inside the pebble. Four proposed fuel pebble models were compared regarding their effective multiplication factor and energy liberation profiles. Such models are: Homogeneous Pebble, Five Zone Homogeneous Pebble, Detailed Geometry, and Randomly Detailed Geometry. (Author)

  11. 2d forward modelling of marine CSEM survey geometry for seabed logging

    International Nuclear Information System (INIS)

    Hussain, N.; Noh, M.; Yahya, N.B.

    2011-01-01

    Hydrocarbon reserve exploration in deep water is done by geophysical surveys. Previously seismic geophysical surveys were explicitly used but it has indistinct results for both water and hydrocarbon saturated reservoir. Recent development for the detection of hydrocarbon reservoir in deeper water is Marine Controlled Source Electromagnetic (MCSEM) geophysical survey. MCSEM is sensitive to electrical conductivity of rocks by which it can differentiate between hydrocarbon reservoir and water saturated reservoir. MCSEM survey geometry put vital role and may causes for anomalies in synthetic data. Consequentially MCSEM is sensitive to survey geometry (e.g. source dipping, rotation and speed, receivers' orientation etc) which causes anomalies. The interpretation for delineating subsurface structure from survey data need to well understand the effects of survey geometry anomalies. Forward modelling is an alternative rather real time survey to study the aforementioned anomalies. In this paper finite difference method (FDM) is implemented for 2D forward modelling in the sense of qualitative understanding to how induced Electromagnetic (EM) signal changes its overall pattern while interact with physical earth properties. A stratified earth structure is developed and modelled in MatLabTM software to study the behaviour of EM field with physical earth properties. Obtained results of 2D geological models are also discussed in this paper. (author)

  12. Babcock and Wilcox model for predicting in-reactor densification

    International Nuclear Information System (INIS)

    Buescher, B.J.; Pegram, J.W.

    1977-07-01

    The B and W densification model is based on a correlation between in-reactor densification and a thermal resintering test. The densification model has been found to predict in-reactor densification with a remarkable degree of accuracy for fuel pellets operated at heat rates above 5 kW/ft and with considerable conservatism for pellelts operating at heat rates below 5 kW/ft

  13. Modeling of reflood of severely damaged reactor core

    International Nuclear Information System (INIS)

    Bachrata, A.

    2012-01-01

    The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

  14. FUZZY REGRESSION MODEL TO PREDICT THE BEAD GEOMETRY IN THE ROBOTIC WELDING PROCESS

    Institute of Scientific and Technical Information of China (English)

    B.S. Sung; I.S. Kim; Y. Xue; H.H. Kim; Y.H. Cha

    2007-01-01

    Recently, there has been a rapid development in computer technology, which has in turn led todevelop the fully robotic welding system using artificial intelligence (AI) technology. However, therobotic welding system has not been achieved due to difficulties of the mathematical model andsensor technologies. The possibilities of the fuzzy regression method to predict the bead geometry,such as bead width, bead height, bead penetration and bead area in the robotic GMA (gas metalarc) welding process is presented. The approach, a well-known method to deal with the problemswith a high degree of fuzziness, is used to build the relationship between four process variablesand the four quality characteristics, respectively. Using these models, the proper prediction of theprocess variables for obtaining the optimal bead geometry can be determined.

  15. Numerical modeling of optical coherent transient processes with complex configurations - I. Angled beam geometry

    International Nuclear Information System (INIS)

    Chang Tiejun; Tian Mingzhen; Randall Babbitt, Wm.

    2004-01-01

    We present a theoretical model for optical coherent transient (OCT) processes based on Maxwell-Bloch equations for angled beam geometry. This geometry is critical in various OCT applications where the desired coherence outputs need to be spatially separated from the rest of the field. The model takes into account both the local interactions between inhomogeneously broadened two-level atoms and the laser fields, and the field propagation in optically thick media. Under the small-angle condition, the spatial dimensions transversing to the main propagation direction were treated with spatial Fourier transform to make the numerical computations for the practical settings confined within a reasonable time frame. The simulations for analog correlators and continuous processing based on stimulated photon echo have been performed using the simulator developed using the theory

  16. Metabolic modeling of synthesis gas fermentation in bubble column reactors.

    Science.gov (United States)

    Chen, Jin; Gomez, Jose A; Höffner, Kai; Barton, Paul I; Henson, Michael A

    2015-01-01

    A promising route to renewable liquid fuels and chemicals is the fermentation of synthesis gas (syngas) streams to synthesize desired products such as ethanol and 2,3-butanediol. While commercial development of syngas fermentation technology is underway, an unmet need is the development of integrated metabolic and transport models for industrially relevant syngas bubble column reactors. We developed and evaluated a spatiotemporal metabolic model for bubble column reactors with the syngas fermenting bacterium Clostridium ljungdahlii as the microbial catalyst. Our modeling approach involved combining a genome-scale reconstruction of C. ljungdahlii metabolism with multiphase transport equations that govern convective and dispersive processes within the spatially varying column. The reactor model was spatially discretized to yield a large set of ordinary differential equations (ODEs) in time with embedded linear programs (LPs) and solved using the MATLAB based code DFBAlab. Simulations were performed to analyze the effects of important process and cellular parameters on key measures of reactor performance including ethanol titer, ethanol-to-acetate ratio, and CO and H2 conversions. Our computational study demonstrated that mathematical modeling provides a complementary tool to experimentation for understanding, predicting, and optimizing syngas fermentation reactors. These model predictions could guide future cellular and process engineering efforts aimed at alleviating bottlenecks to biochemical production in syngas bubble column reactors.

  17. Mathematical Modeling of Resonant Processes in Confined Geometry of Atomic and Atom-Ion Traps

    Science.gov (United States)

    Melezhik, Vladimir S.

    2018-02-01

    We discuss computational aspects of the developed mathematical models for resonant processes in confined geometry of atomic and atom-ion traps. The main attention is paid to formulation in the nondirect product discrete-variable representation (npDVR) of the multichannel scattering problem with nonseparable angular part in confining traps as the boundary-value problem. Computational efficiency of this approach is demonstrated in application to atomic and atom-ion confinement-induced resonances we predicted recently.

  18. Boiling water reactor modeling capabilities of MMS-02

    International Nuclear Information System (INIS)

    May, R.S.; Abdollahian, D.A.; Elias, E.; Shak, D.P.

    1987-01-01

    During the development period for the Modular Modeling System (MMS) library modules, the Boiling Water Reactor (BWR) has been the last major component to be addressed. The BWRX module includes models of the reactor core, reactor vessel, and recirculation loop. A pre-release version was made available for utility use in September 1983. Since that time a number of changes have been incorporated in BWRX to (1) improve running time for most transient events of interest, (2) extend its capability to include certain events of interest in reactor safety analysis, and (3) incorporate a variety of improvements to the module interfaces and user input formats. The purposes of this study were to briefly review the module structure and physical models, to point the differences between the MMS-02 BWRX module and the BWRX version previously available in the TESTREV1 library, to provide guidelines for choosing among the various user options, and to present some representative results

  19. Three dimensional modeling of depositional geometries. A case study from Tofane Group (Dolomites, Italy).

    Science.gov (United States)

    Gattolin, G.; Franceschi, M.; Breda, A.; Teza, G.; Preto, N.

    2012-04-01

    At the end of the Early Carnian, the Carnian Pluvial Event (CPE) resulted in a major crisis of carbonate factories. The sharp change in carbonate production lead to a dramatic modifications in depositional geometries. Steep clinoforms of the high-relief pre-crisis carbonate platforms were replaced by low-angle ramps. Spatial characters of depositional geometries can be decisive in identifying the genesis of geological bodies. We here show how 3D modeling techniques can be applied to help in quantifying and highlighting their variations. As case study we considered two outcrops in the Tofane Group (Dolomites, Italy). The first outcrop (bottom of southern walls of Tofana di Rozes) exposes a platform-to-basin transect of pre- and post-crisis platforms, the second (Dibona hut) a clinostratified carbonate body deposited during the Carnian crisis. Outcrop conditions at both sites, with vertical and hardly accessible walls, make the field tracing of depositional geometries particularly challenging. Line drawing on high resolution pictures can help (e.g. for clinoforms), but its use for quantification is hampered by perspective deformation. Three dimensional acquisition and modeling allow to retrieve the true spatial characters of sedimentary bodies in these outcrops. The geometry of the carbonate body at Dibona (~ 15000 sqm) was acquired with terrestrial LiDAR, while for Tofana photogrammetric techniques were applied because of the extension of the outcrop itself (~ 240000 sqm) and the lack of suitable points of view for terrestrial laser scanning. At Tofana, field observations reveal the presence of tens-hundreds m large carbonate mounds grown on a pre-existing inclined surface, intercalated with skeletal carbonates and siltites-arenites. This system rapidly evolves into a carbonate-clastic ramp. Photogrammetric topography acquisition permitted to place and visualize geological features in a three dimensional frame, thus obtaining a conceptual sedimentological model. A 3

  20. A statistical skull geometry model for children 0-3 years old.

    Directory of Open Access Journals (Sweden)

    Zhigang Li

    Full Text Available Head injury is the leading cause of fatality and long-term disability for children. Pediatric heads change rapidly in both size and shape during growth, especially for children under 3 years old (YO. To accurately assess the head injury risks for children, it is necessary to understand the geometry of the pediatric head and how morphologic features influence injury causation within the 0-3 YO population. In this study, head CT scans from fifty-six 0-3 YO children were used to develop a statistical model of pediatric skull geometry. Geometric features important for injury prediction, including skull size and shape, skull thickness and suture width, along with their variations among the sample population, were quantified through a series of image and statistical analyses. The size and shape of the pediatric skull change significantly with age and head circumference. The skull thickness and suture width vary with age, head circumference and location, which will have important effects on skull stiffness and injury prediction. The statistical geometry model developed in this study can provide a geometrical basis for future development of child anthropomorphic test devices and pediatric head finite element models.

  1. A computational approach to modeling cellular-scale blood flow in complex geometry

    Science.gov (United States)

    Balogh, Peter; Bagchi, Prosenjit

    2017-04-01

    We present a computational methodology for modeling cellular-scale blood flow in arbitrary and highly complex geometry. Our approach is based on immersed-boundary methods, which allow modeling flows in arbitrary geometry while resolving the large deformation and dynamics of every blood cell with high fidelity. The present methodology seamlessly integrates different modeling components dealing with stationary rigid boundaries of complex shape, moving rigid bodies, and highly deformable interfaces governed by nonlinear elasticity. Thus it enables us to simulate 'whole' blood suspensions flowing through physiologically realistic microvascular networks that are characterized by multiple bifurcating and merging vessels, as well as geometrically complex lab-on-chip devices. The focus of the present work is on the development of a versatile numerical technique that is able to consider deformable cells and rigid bodies flowing in three-dimensional arbitrarily complex geometries over a diverse range of scenarios. After describing the methodology, a series of validation studies are presented against analytical theory, experimental data, and previous numerical results. Then, the capability of the methodology is demonstrated by simulating flows of deformable blood cells and heterogeneous cell suspensions in both physiologically realistic microvascular networks and geometrically intricate microfluidic devices. It is shown that the methodology can predict several complex microhemodynamic phenomena observed in vascular networks and microfluidic devices. The present methodology is robust and versatile, and has the potential to scale up to very large microvascular networks at organ levels.

  2. A statistical skull geometry model for children 0-3 years old.

    Science.gov (United States)

    Li, Zhigang; Park, Byoung-Keon; Liu, Weiguo; Zhang, Jinhuan; Reed, Matthew P; Rupp, Jonathan D; Hoff, Carrie N; Hu, Jingwen

    2015-01-01

    Head injury is the leading cause of fatality and long-term disability for children. Pediatric heads change rapidly in both size and shape during growth, especially for children under 3 years old (YO). To accurately assess the head injury risks for children, it is necessary to understand the geometry of the pediatric head and how morphologic features influence injury causation within the 0-3 YO population. In this study, head CT scans from fifty-six 0-3 YO children were used to develop a statistical model of pediatric skull geometry. Geometric features important for injury prediction, including skull size and shape, skull thickness and suture width, along with their variations among the sample population, were quantified through a series of image and statistical analyses. The size and shape of the pediatric skull change significantly with age and head circumference. The skull thickness and suture width vary with age, head circumference and location, which will have important effects on skull stiffness and injury prediction. The statistical geometry model developed in this study can provide a geometrical basis for future development of child anthropomorphic test devices and pediatric head finite element models.

  3. Validation of a numerical release model (REPCOM) for the Finnish reactor waste disposal systems: Pt.1

    International Nuclear Information System (INIS)

    Nykyri, Mikko

    1987-05-01

    The aim of the work is to model experimentally the inner structures and materials of two reactor waste repositories and to use the results for the validation work of a numerical near field release model, REPCOM. The experimental modelling of the multibarrier systems is conducted on a laboratory scale by using the same principal materials as are employed in the Finnish reactor waste disposal concepts. The migration of radionuclides is studied in two or more consecutive material layers. The laboratory arrangements include the following test materials: bituminized resin, cemented resin, concrete, crushed rock, and water. The materials correspond to the local materials in the planned disposal systems. Cs-137, Co-60, Sr-85, and Sr-90 are used as tracers, with which the resin, water, and crushed rock are labeled depending on the specimen type. The basic specimen geometries are cylindrical and cubic. In the cylindrical geometry the test materials were placed into PVC-tubes. The corresponding numerical model is one-dimensional. In the cubic geometry the materials were placed inside each other. The boundaries form cubes, and the numerical model is three-dimensional. Altogether 12 test system types were produced. The gamma active nuclides in the cylindrical samples were measured nondestructively with a scanner in order to determine the activity profiles in the specimens. The gamma active nuclides in the cubic samples and the beta emeitting Sr-90 in separate samples will be measured after splitting the samples. One to five activity profiles were determined for each cylindrical gamma-active sample. There are already clear diffusion profiles to be had for strontium in crushed rock, and for cesium in crushed rock and in concrete. Cobalt indicated no diffusion. No activity profiles were measured for the cubic samples or for the beta active, Sr-90-doped samples

  4. Visualizing Three-dimensional Slab Geometries with ShowEarthModel

    Science.gov (United States)

    Chang, B.; Jadamec, M. A.; Fischer, K. M.; Kreylos, O.; Yikilmaz, M. B.

    2017-12-01

    Seismic data that characterize the morphology of modern subducted slabs on Earth suggest that a two-dimensional paradigm is no longer adequate to describe the subduction process. Here we demonstrate the effect of data exploration of three-dimensional (3D) global slab geometries with the open source program ShowEarthModel. ShowEarthModel was designed specifically to support data exploration, by focusing on interactivity and real-time response using the Vrui toolkit. Sixteen movies are presented that explore the 3D complexity of modern subduction zones on Earth. The first movie provides a guided tour through the Earth's major subduction zones, comparing the global slab geometry data sets of Gudmundsson and Sambridge (1998), Syracuse and Abers (2006), and Hayes et al. (2012). Fifteen regional movies explore the individual subduction zones and regions intersecting slabs, using the Hayes et al. (2012) slab geometry models where available and the Engdahl and Villasenor (2002) global earthquake data set. Viewing the subduction zones in this way provides an improved conceptualization of the 3D morphology within a given subduction zone as well as the 3D spatial relations between the intersecting slabs. This approach provides a powerful tool for rendering earth properties and broadening capabilities in both Earth Science research and education by allowing for whole earth visualization. The 3D characterization of global slab geometries is placed in the context of 3D slab-driven mantle flow and observations of shear wave splitting in subduction zones. These visualizations contribute to the paradigm shift from a 2D to 3D subduction framework by facilitating the conceptualization of the modern subduction system on Earth in 3D space.

  5. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  6. Modelling and control design for SHARON/Anammox reactor sequence

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    metabolism against fast chemical reaction and mass transfer. Likewise, the analysis of the dynamics contributed to establish qualitatively the requirements for control of the reactors, both for regulation and for optimal operation. Work in progress on quantitatively analysing different control structure......With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial...

  7. A friendly Maple module for one and two group reactor model

    International Nuclear Information System (INIS)

    Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O.

    2015-01-01

    The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

  8. A friendly Maple module for one and two group reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O., E-mail: camila.oliv.baptista@gmail.com, E-mail: pavanguilherme@gmail.com, E-mail: kelmo.lins@gmail.com, E-mail: marcelovilelasilva@gmail.com, E-mail: rodrigowerner@hotmail.com, E-mail: neutron201566@yahoo.com, E-mail: vellozo@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

  9. Path integral representation of Lorentzian spinfoam model, asymptotics and simplicial geometries

    International Nuclear Information System (INIS)

    Han, Muxin; Krajewski, Thomas

    2014-01-01

    A new path integral representation of Lorentzian Engle–Pereira–Rovelli–Livine spinfoam model is derived by employing the theory of unitary representation of SL(2,C). The path integral representation is taken as a starting point of semiclassical analysis. The relation between the spinfoam model and classical simplicial geometry is studied via the large-spin asymptotic expansion of the spinfoam amplitude with all spins uniformly large. More precisely, in the large-spin regime, there is an equivalence between the spinfoam critical configuration (with certain nondegeneracy assumption) and a classical Lorentzian simplicial geometry. Such an equivalence relation allows us to classify the spinfoam critical configurations by their geometrical interpretations, via two types of solution-generating maps. The equivalence between spinfoam critical configuration and simplical geometry also allows us to define the notion of globally oriented and time-oriented spinfoam critical configuration. It is shown that only at the globally oriented and time-oriented spinfoam critical configuration, the leading-order contribution of spinfoam large-spin asymptotics gives precisely an exponential of Lorentzian Regge action of General Relativity. At all other (unphysical) critical configurations, spinfoam large-spin asymptotics modifies the Regge action at the leading-order approximation. (paper)

  10. KENO3D visualization tool for KENO V.a geometry models

    International Nuclear Information System (INIS)

    Bowman, S.M.; Horwedel, J.E.

    1999-01-01

    The standardized computer analyses for licensing evaluations (SCALE) computer software system developed at Oak Ridge National Laboratory (ORNL) is widely used and accepted around the world for criticality safety analyses. SCALE includes the well-known KENO V.a three-dimensional Monte Carlo criticality computer code. Criticality safety analysis often require detailed modeling of complex geometries. Checking the accuracy of these models can be enhanced by effective visualization tools. To address this need, ORNL has recently developed a powerful state-of-the-art visualization tool called KENO3D that enables KENO V.a users to interactively display their three-dimensional geometry models. The interactive options include the following: (1) having shaded or wireframe images; (2) showing standard views, such as top view, side view, front view, and isometric three-dimensional view; (3) rotating the model; (4) zooming in on selected locations; (5) selecting parts of the model to display; (6) editing colors and displaying legends; (7) displaying properties of any unit in the model; (8) creating cutaway views; (9) removing units from the model; and (10) printing image or saving image to common graphics formats

  11. Unified tractable model for downlink MIMO cellular networks using stochastic geometry

    KAUST Repository

    Afify, Laila H.

    2016-07-26

    Several research efforts are invested to develop stochastic geometry models for cellular networks with multiple antenna transmission and reception (MIMO). On one hand, there are models that target abstract outage probability and ergodic rate for simplicity. On the other hand, there are models that sacrifice simplicity to target more tangible performance metrics such as the error probability. Both types of models are completely disjoint in terms of the analytic steps to obtain the performance measures, which makes it challenging to conduct studies that account for different performance metrics. This paper unifies both techniques and proposes a unified stochastic-geometry based mathematical paradigm to account for error probability, outage probability, and ergodic rates in MIMO cellular networks. The proposed model is also unified in terms of the antenna configurations and leads to simpler error probability analysis compared to existing state-of-the-art models. The core part of the analysis is based on abstracting unnecessary information conveyed within the interfering signals by assuming Gaussian signaling. To this end, the accuracy of the proposed framework is verified against state-of-the-art models as well as system level simulations. We provide via this unified study insights on network design by reflecting system parameters effect on different performance metrics. © 2016 IEEE.

  12. Modeling cavities exhibiting strong lateral confinement using open geometry Fourier modal method

    Science.gov (United States)

    Häyrynen, Teppo; Gregersen, Niels

    2016-04-01

    We have developed a computationally efficient Fourier-Bessel expansion based open geometry formalism for modeling the optical properties of rotationally symmetric photonic nanostructures. The lateral computation domain is assumed infinite so that no artificial boundary conditions are needed. Instead, the leakage of the modes due to an imperfect field confinement is taken into account by using a basis functions that expand the whole infinite space. The computational efficiency is obtained by using a non-uniform discretization in the frequency space in which the lateral expansion modes are more densely sampled around a geometry specific dominant transverse wavenumber region. We will use the developed approach to investigate the Q factor and mode confinement in cavities where top DBR mirror has small rectangular defect confining the modes laterally on the defect region.

  13. Overview of the reactor safety study consequence model

    International Nuclear Information System (INIS)

    Wall, I.B.; Yaniv, S.S.; Blond, R.M.; McGrath, P.E.; Church, H.W.; Wayland, J.R.

    1977-01-01

    The Reactor Safety Study (WASH-1400) is a comprehensive assessment of the potential risk to the public from accidents in light water power reactors. The engineering analysis of the plants is described in detail in the Reactor Safety Study: it provides an estimate of the probability versus magnitude of the release of radioactive material. The consequence model, which is the subject of this paper, describes the progression of the postulated accident after the release of the radioactive material from the containment. A brief discussion of the manner in which the consequence calculations are performed is presented. The emphasis in the description is on the models and data that differ significantly from those previously used for these types of assessments. The results of the risk calculations for 100 light water power reactors are summarized

  14. Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

    International Nuclear Information System (INIS)

    Chen Yiping; Liu Songlin

    2010-01-01

    Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

  15. RELAP5 kinetics model development for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Judd, J.L.; Terry, W.K.

    1990-01-01

    A point-kinetics model of the Advanced Test Reactor has been developed for the RELAP5 code. Reactivity feedback parameters were calculated by a three-dimensional analysis with the PDQ neutron diffusion code. Analyses of several hypothetical reactivity insertion events by the new model and two earlier models are discussed. 3 refs., 10 figs., 6 tabs

  16. Modeling of hydrodynamic cavitation reactors: a unified approach

    NARCIS (Netherlands)

    Moholkar, V.S.; Pandit, A.B.

    2001-01-01

    An attempt has been made to present a unified theoretical model for the cavitating flow in a hydrodynamic cavitation reactor using the nonlinear continuum mixture model for two-phase flow as the basis. This model has been used to describe the radial motion of bubble in the cavitating flow in two

  17. A Monte Carlo modeling on charging effect for structures with arbitrary geometries

    Science.gov (United States)

    Li, C.; Mao, S. F.; Zou, Y. B.; Li, Yong Gang; Zhang, P.; Li, H. M.; Ding, Z. J.

    2018-04-01

    Insulating materials usually suffer charging effects when irradiated by charged particles. In this paper, we present a Monte Carlo study on the charging effect caused by electron beam irradiation for sample structures with any complex geometry. When transporting in an insulating solid, electrons encounter elastic and inelastic scattering events; the Mott cross section and a Lorentz-type dielectric function are respectively employed to describe such scatterings. In addition, the band gap and the electron–long optical phonon interaction are taken into account. The electronic excitation in inelastic scattering causes generation of electron–hole pairs; these negative and positive charges establish an inner electric field, which in turn induces the drift of charges to be trapped by impurities, defects, vacancies etc in the solid, where the distributions of trapping sites are assumed to have uniform density. Under charging conditions, the inner electric field distorts electron trajectories, and the surface electric potential dynamically alters secondary electron emission. We present, in this work, an iterative modeling method for a self-consistent calculation of electric potential; the method has advantages in treating any structure with arbitrary complex geometry, in comparison with the image charge method—which is limited to a quite simple boundary geometry. Our modeling is based on: the combination of the finite triangle mesh method for an arbitrary geometry construction; a self-consistent method for the spatial potential calculation; and a full dynamic description for the motion of deposited charges. Example calculations have been done to simulate secondary electron yield of SiO2 for a semi-infinite solid, the charging for a heterostructure of SiO2 film grown on an Au substrate, and SEM imaging of a SiO2 line structure with rough surfaces and SiO2 nanoparticles with irregular shapes. The simulations have explored interesting interlaced charge layer distribution

  18. A new control-oriented transient model of variable geometry turbocharger

    International Nuclear Information System (INIS)

    Bahiuddin, Irfan; Mazlan, Saiful Amri; Imaduddin, Fitrian; Ubaidillah

    2017-01-01

    The flow input of a variable geometry turbocharger turbine is highly unsteady due to rapid and periodic pressure dynamics in engine combustion chambers. Several VGT control methods have been developed to recover more energy from the highly pulsating exhaust gas flow. To develop a control system for the highly pulsating flow condition, an accurate and valid unsteady model is required. This study focuses on the derivation of governing the unsteady control-oriented model (COM) for a turbine of an actively controlled turbocharger (ACT). The COM has the capability to predict the turbocharger behaviour regarding the instantaneous turbine actual and isentropic powers in different effective throat areas. The COM is a modified version of a conventional mean value model (MVM) with an additional feature to calculate the turbine angular velocity and torque for determining the actual power. The simulation results were further compared with experimental data in two general scenarios. The first scenario was simulations on fixed geometry positions. The second simulation scenario considered the nozzle movement after receiving a signal from the controller in different cases. The comparison between simulation and experimental results showed similarities in the recovered power behaviours the turbine inlet area increases or vice versa. The model also has proved its reliability to replicate general behaviour as in the example of ACT cases presented in this paper. However, the model is incapable to replicate the detailed and complicated phenomena, such as choking effect and hysteresis effect. - Highlights: • A control-oriented model of a variable geometry turbocharger turbine is proposed. • Isentropic and actual power behaviour estimations on turbocharger turbine. • A simulation tool for developing active control systems of turbocharger turbines.

  19. Identification of Chemical Reactor Plant’s Mathematical Model

    OpenAIRE

    Pyakullya, Boris Ivanovich; Kladiev, Sergey Nikolaevich

    2015-01-01

    This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  20. Identification of Chemical Reactor Plant’s Mathematical Model

    Directory of Open Access Journals (Sweden)

    Pyakillya Boris

    2015-01-01

    Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  1. Improvements in the model of neutron calculations for research reactors

    International Nuclear Information System (INIS)

    Calzetta, Osvaldo; Leszczynski, Francisco

    1987-01-01

    Within the research program in the field of neutron physics calculations being carried out in the Nuclear Engineering Division at the Centro Atomico Bariloche, the errors which due to some typical approximations appear in the final results are researched. For research MTR type reactors, two approximations, for high and low enrichment are investigated: the treatment of the geometry and the method of few-group cell cross-sections calculation, particularly in the resonance energy region. Commonly, the cell constants used for the entire reactor calculation are obtained making an homogenization of the full fuel elements, by one-dimensional calculations. An improvement is made that explicitly includes the fuel element frames in the core calculation geometry. Besides, a detailed treatment-in energy and space- is used to find the resonance few-group cross sections, and a comparison of the results with detailed and approximated calculations is made. The least number and the best mesh of energy groups needed for cell calculations is fixed too. (Author) [es

  2. Almost-commutative geometries beyond the standard model: III. Vector doublets

    International Nuclear Information System (INIS)

    Squellari, Romain; Stephan, Christoph A

    2007-01-01

    We will present a new extension of the standard model of particle physics in its almost-commutative formulation. This extension has as its basis the algebra of the standard model with four summands (Iochum et al 2004 J. Math. Phys. 45 5003 (Preprint hep-th/0312276), Jureit J-H and Stephan C 2005 J. Math. Phys. 46 043512 (Preprint hep-th/0501134), Schuecker T 2005 Krajewski diagrams and spin lifts Preprint hep-th/0501181, Jureit et al 2005 J. Math. Phys. 46 072303 (Preprint hep-th/0503190), Jureit J-H and Stephan C 2006 On a classification of irreducible almost commutative geometries: IV (Preprint hep-th/0610040)), and enlarges only the particle content by an arbitrary number of generations of left-right symmetric doublets which couple vectorially to the U(1) Y x SU(2) w subgroup of the standard model. As in the model presented in Stephan (2007 Almost-commutative geometries beyond the standard model: II. New Colours Preprint hep-th/0706.0595), which introduced particles with a new colour, grand unification is no longer required by the spectral action. The new model may also possess a candidate for dark matter in the hundred TeV mass range with neutrino-like cross section

  3. Digital Tomosynthesis System Geometry Analysis Using Convolution-Based Blur-and-Add (BAA) Model.

    Science.gov (United States)

    Wu, Meng; Yoon, Sungwon; Solomon, Edward G; Star-Lack, Josh; Pelc, Norbert; Fahrig, Rebecca

    2016-01-01

    Digital tomosynthesis is a three-dimensional imaging technique with a lower radiation dose than computed tomography (CT). Due to the missing data in tomosynthesis systems, out-of-plane structures in the depth direction cannot be completely removed by the reconstruction algorithms. In this work, we analyzed the impulse responses of common tomosynthesis systems on a plane-to-plane basis and proposed a fast and accurate convolution-based blur-and-add (BAA) model to simulate the backprojected images. In addition, the analysis formalism describing the impulse response of out-of-plane structures can be generalized to both rotating and parallel gantries. We implemented a ray tracing forward projection and backprojection (ray-based model) algorithm and the convolution-based BAA model to simulate the shift-and-add (backproject) tomosynthesis reconstructions. The convolution-based BAA model with proper geometry distortion correction provides reasonably accurate estimates of the tomosynthesis reconstruction. A numerical comparison indicates that the simulated images using the two models differ by less than 6% in terms of the root-mean-squared error. This convolution-based BAA model can be used in efficient system geometry analysis, reconstruction algorithm design, out-of-plane artifacts suppression, and CT-tomosynthesis registration.

  4. Improvements of the reactivity devices modeling for the advanced CANDU reactor

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Dahmani, M.; Hebert, A.

    2008-01-01

    In the context of the ACR TM (Advanced CANDU Reactor), 3D transport calculations are required in order to simulate the reactivity devices located perpendicularly to the fuel channels. The computational scheme that is usually used for CANDU-6 and ACR reactors is based on a simplified supercell geometry in which the fuel clusters and devices are replaced by annuli. Recently, an exact modeling of 3D supercell configurations was introduced within the framework of the ACR calculations. However, with such a model, fine meshing requirements lead to problems that are very demanding in terms of computational resources. In this paper, we present improvements introduced in the ACR context to reduce the cost of the 3D supercell calculations. Two avenues of investigations are reported. First, the introduction of an accelerated characteristics method permits to reduce the computational burden of such calculations involving a large number of regions. In addition, contrarily to CANDU-6 supercell configurations, the ACR 3D geometry is prismatic and consequently a special tracking procedure can be used. This approach introduces no approximation and is significantly faster than the general 3D tracking technique. Thanks to these modifications in the computational procedure, 3D supercell calculations with a level of mesh discretization comparable to 2D cell configurations become affordable for industrial applications

  5. Analytical evaluation of neutron diffusion equation for the geometry of very intense continuous high flux pulsed reactor

    International Nuclear Information System (INIS)

    Narain, Rajendra

    1995-01-01

    Using the concept of Very Intense Continuous High Flux Pulsed Reactor to obtain a rotating high flux pulse in an annular core an analytical treatment for the quasi-static solution with a moving reflector is presented. Under quasi-static situation, time averaged values for important parameters like multiplication factor, flux, leakage do not change with time. As a result the instantaneous solution can be considered to be separable in time and space after correcting for the coordinates for the motion of the pulser. The space behaviour of the pulser is considered as exp(-αx 2 ). Movement of delayed neutron precursors is also taken into account. (author). 4 refs

  6. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  7. Development of modeling tools for pin-by-pin precise reactor simulation

    International Nuclear Information System (INIS)

    Ma Yan; Li Shu; Li Gang; Zhang Baoyin; Deng Li; Fu Yuanguang

    2013-01-01

    In order to develop large-scale transport simulation and calculation method (such as simulation of whole reactor core pin-by-pin problem), the Institute of Applied Physics and Computational Mathematics developed the neutron-photon coupled transport code JMCT and the toolkit JCOGIN. Creating physical calculation model easily and efficiently can essentially reduce problem solving time. Currently, lots of visual modeling programs have been developed based on different CAD systems. In this article, the developing idea of a visual modeling tool based on field oriented development was introduced. Considering the feature of physical modeling, fast and convenient operation modules were developed. In order to solve the storage and conversion problems of large scale models, the data structure and conversional algorithm based on the hierarchical geometry tree were designed. The automatic conversion and generation of physical model input file for JMCT were realized. By using this modeling tool, the Dayawan reactor whole core physical model was created, and the transformed file was delivered to JMCT for transport calculation. The results validate the correctness of the visual modeling tool. (authors)

  8. Parametric study of the Incompletely Stirred Reactor modeling

    Energy Technology Data Exchange (ETDEWEB)

    Mobini, K. [Department of Mechanical Engineering, Shahid Rajaee University, Lavizan, Tehran (Iran); Bilger, R.W. [School of Aerospace, Mechanical and Mechatronic Engineering, University of Sydney, Sydney (Australia)

    2009-09-15

    The Incompletely Stirred Reactor (ISR) is a generalization of the widely-used Perfectly Stirred Reactor (PSR) model and allows for incomplete mixing within the reactor. Its formulation is based on the Conditional Moment Closure (CMC) method. This model is applicable to nonpremixed combustion with strong recirculation such as in a gas turbine combustor primary zone. The model uses the simplifying assumptions that the conditionally-averaged reactive-scalar concentrations are independent of position in the reactor: this results in ordinary differential equations in mixture fraction space. The simplicity of the model permits the use of very complex chemical mechanisms. The effects of the detailed chemistry can be found while still including the effects of micromixing. A parametric study is performed here on an ISR for combustion of methane at overall stoichiometric conditions to investigate the sensitivity of the model to different parameters. The focus here is on emissions of nitric oxide and carbon monoxide. It is shown that the most important parameters in the ISR model are reactor residence time, the chemical mechanism and the core-averaged Probability Density Function (PDF). Using several different shapes for the core-averaged PDF, it is shown that use of a bimodal PDF with a low minimum at stoichiometric mixture fraction and a large variance leads to lower nitric oxide formation. The 'rich-plus-lean' mixing or staged combustion strategy for combustion is thus supported. (author)

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

    International Nuclear Information System (INIS)

    Bayoumi, M.A.A.

    1976-10-01

    A bibliographic study shows that the experimental studies examined, have been developed to understand the phenomenon acting on the mixing between the sub-channels of which geometries are such these of rod bundles used in some nuclear reactors. Experimental devices and tests have been developed to study the influence of the following parameters, operating conditions, pressure, flow rate, power brought to the bundle and inlet temperature on the distribution of flow rates and vapor content among the different sub-channels. By means of non isokinetic sampling, one has determined the enthalpy of the fluid participating to the mixing between the communicating sub-channels and it has been shown that the value of this enthalpy depends strongly on the type of fluid flow and that this enthalpy cannot be either the enthalpy of one of the two sub-channels, nor (always) an average of these two enthalpies. The experimental results have been compared with calculations developed with the code FLICA, concerning the mass velocity distribution, the exchange term of linear momentum, and the variation of the transversal enthalpy with regard to the type of fluid flow. A study of local void ratio measurement, by means of optical probes, has been proposed. The present study has been carried out with a smooth geometry [fr

  10. Human eye analytical and mesh-geometry models for ophthalmic dosimetry using MCNP6

    International Nuclear Information System (INIS)

    Angelocci, Lucas V.; Fonseca, Gabriel P.; Yoriyaz, Helio

    2015-01-01

    Eye tumors can be treated with brachytherapy using Co-60 plaques, I-125 seeds, among others materials. The human eye has regions particularly vulnerable to ionizing radiation (e.g. crystalline) and dosimetry for this region must be taken carefully. A mathematical model was proposed in the past [1] for the eye anatomy to be used in Monte Carlo simulations to account for dose distribution in ophthalmic brachytherapy. The model includes the description for internal structures of the eye that were not treated in previous works. The aim of this present work was to develop a new eye model based on the Mesh geometries of the MCNP6 code. The methodology utilized the ABAQUS/CAE (Simulia 3DS) software to build the Mesh geometry. For this work, an ophthalmic applicator containing up to 24 model Amersham 6711 I-125 seeds (Oncoseed) was used, positioned in contact with a generic tumor defined analytically inside the eye. The absorbed dose in eye structures like cornea, sclera, choroid, retina, vitreous body, lens, optical nerve and optical nerve wall were calculated using both models: analytical and MESH. (author)

  11. Human eye analytical and mesh-geometry models for ophthalmic dosimetry using MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Angelocci, Lucas V.; Fonseca, Gabriel P.; Yoriyaz, Helio, E-mail: hyoriyaz@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Eye tumors can be treated with brachytherapy using Co-60 plaques, I-125 seeds, among others materials. The human eye has regions particularly vulnerable to ionizing radiation (e.g. crystalline) and dosimetry for this region must be taken carefully. A mathematical model was proposed in the past [1] for the eye anatomy to be used in Monte Carlo simulations to account for dose distribution in ophthalmic brachytherapy. The model includes the description for internal structures of the eye that were not treated in previous works. The aim of this present work was to develop a new eye model based on the Mesh geometries of the MCNP6 code. The methodology utilized the ABAQUS/CAE (Simulia 3DS) software to build the Mesh geometry. For this work, an ophthalmic applicator containing up to 24 model Amersham 6711 I-125 seeds (Oncoseed) was used, positioned in contact with a generic tumor defined analytically inside the eye. The absorbed dose in eye structures like cornea, sclera, choroid, retina, vitreous body, lens, optical nerve and optical nerve wall were calculated using both models: analytical and MESH. (author)

  12. Modelling of crack chemistry in sensitized stainless steel in boiling water reactor environments

    International Nuclear Information System (INIS)

    Turnbull, A.

    1997-01-01

    An advanced model has been used to predict the chemistry and potential in a stress corrosion crack in sensitized stainless steel in a boiling water reactor (BWR) environment. The model assumes trapezoidal crack geometry, incorporates anodic reaction and cathodic reduction within the crack, and takes into account the limited solubility of cations in high temperature water. The results indicate that the crack tip potential is not independent of the external potential, and that the reactions on the walls of the crack must be included for reliable prediction. Accordingly, both the modelling assumptions of Ford and Andresen and of Macdonald and Urquidi-Macdonald, whilst having merit, are not fully satisfactory. Extended application of the model for improved prediction of stress corrosion crack growth rate is constrained by limitations in electrochemical data which are currently inadequate. (author)

  13. Fuel requirements for experimental devices in MTR reactors. A perturbation model for reactor core analysis

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1991-01-01

    Irradiation in neutron absorbing devices, requiring high fast neutron fluxes in the core or high thermal fluxes in the reflector and flux traps, lead to higher density fuel and larger core dimensions. A perturbation model of the reactor core helps to estimate the fuel requirements. (orig.)

  14. Development of a sensitivity analysis systems in nuclear reactors through generalized perturbation theory at first order in 2 D geometries

    International Nuclear Information System (INIS)

    Garcia, Juan Matias

    2005-01-01

    Perturbation Methods represent a powerful tool to do sensitivity analysis, and they found many aplications in nuclear engineering.As an introduction to this kind of analysis, we develope a program that apply the Generalized Perturbation Theory or GPT Method to bidimensional system of rectangular geometry.We first consider an homogeneous system of non-multiplying material and then an heterogeneous system with region of multiplying material, with the intention of make concret aplications of perturbation method to nuclear engineering problems.The program, that we called Pert, determines neutron fluxes and importance functions applying the Multigroup Diffusion Theory; and also solves the integrals required to calculate sensitivity coefficients.Using this perturbation methods we could verify the low computational cost required to make this kind of analysis and the simplicity of the equations systems involved, allowing us to make elaborates sensitivity analysis for the responses of our interest

  15. Numerical algebraic geometry for model selection and its application to the life sciences

    KAUST Repository

    Gross, Elizabeth

    2016-10-12

    Researchers working with mathematical models are often confronted by the related problems of parameter estimation, model validation and model selection. These are all optimization problems, well known to be challenging due to nonlinearity, non-convexity and multiple local optima. Furthermore, the challenges are compounded when only partial data are available. Here, we consider polynomial models (e.g. mass-action chemical reaction networks at steady state) and describe a framework for their analysis based on optimization using numerical algebraic geometry. Specifically, we use probability-one polynomial homotopy continuation methods to compute all critical points of the objective function, then filter to recover the global optima. Our approach exploits the geometrical structures relating models and data, and we demonstrate its utility on examples from cell signalling, synthetic biology and epidemiology.

  16. Modelling of MOCVD reactor: new 3D approach

    International Nuclear Information System (INIS)

    Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

    2014-01-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  17. A two-point kinetic model for the PROTEUS reactor

    International Nuclear Information System (INIS)

    Dam, H. van.

    1995-03-01

    A two-point reactor kinetic model for the PROTEUS-reactor is developed and the results are described in terms of frequency dependent reactivity transfer functions for the core and the reflector. It is shown that at higher frequencies space-dependent effects occur which imply failure of the one-point kinetic model. In the modulus of the transfer functions these effects become apparent above a radian frequency of about 100 s -1 , whereas for the phase behaviour the deviation from a point model already starts at a radian frequency of 10 s -1 . (orig.)

  18. Fission-product burnup chain model for research reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

    1990-12-01

    A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

  19. Modelling of MOCVD Reactor: New 3D Approach

    Science.gov (United States)

    Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

    2014-04-01

    The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

  20. Reliability modeling of Clinch River breeder reactor electrical shutdown systems

    International Nuclear Information System (INIS)

    Schatz, R.A.; Duetsch, K.L.

    1974-01-01

    The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

  1. Modelling of temperature distribution and pulsations in fast reactor units

    International Nuclear Information System (INIS)

    Ushakov, P.A.; Sorokin, A.P.

    1994-01-01

    Reasons for the occurrence of thermal stresses in reactor units have been analyzed. The main reasons for this analysis are: temperature non-uniformity at the output of reactor core and breeder and the ensuing temperature pulsation; temperature pulsations due to mixing of sodium jets of a different temperature; temperature nonuniformity and pulsations resulting from the part of loops (circuits) un-plug; temperature nonuniformity and fluctuations in transient and accidental shut down of reactor or transfer to cooling by natural circulation. The results of investigating the thermal hydraulic characteristics are obtained by modelling the processes mentioned above. Analysis carried out allows the main lines of investigation to be defined and conclusions can be drawn regarding the problem of temperature distribution and fluctuation in fast reactor units

  2. Comparative study of cost models for tokamak DEMO fusion reactors

    International Nuclear Information System (INIS)

    Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Ban, Kanae; Kondo, Takuya; Tobita, Kenji; Goto, Takuya

    2012-01-01

    Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study. (author)

  3. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    Ahmmed Saadi Ibrehem

    2011-05-01

    Full Text Available A modified model for the neutralization process of Stirred Tank Reactors (CSTR reactor is presented in this study. The model accounts for the effect of strong acid [HCL] flowrate and strong base [NaOH] flowrate with the ionic concentrations of [Cl-] and [Na+] on the Ph of the system. In this work, the effect of important reactor parameters such as ionic concentrations and acid and base flowrates on the dynamic behavior of the CSTR is investigated and the behavior of mathematical model is compared with the reported models for the McAvoy model and Jutila model. Moreover, the results of the model are compared with the experimental data in terms of pH dynamic study. A good agreement is observed between our model prediction and the actual plant data. © 2011 BCREC UNDIP. All rights reserved(Received: 1st March 2011, Revised: 28th March 2011; Accepted: 7th April 2011[How to Cite: A.S. Ibrehem. (2011. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor. Bulletin of Chemical Reaction Engineering & Catalysis, 6(1: 47-52. doi:10.9767/bcrec.6.1.825.47-52][How to Link / DOI: http://dx.doi.org/10.9767/bcrec.6.1.825.47-52 || or local:  http://ejournal.undip.ac.id/index.php/bcrec/article/view/825 ] | View in 

  4. Towards an efficient multiphysics model for nuclear reactor dynamics

    Directory of Open Access Journals (Sweden)

    Obaidurrahman K.

    2015-01-01

    Full Text Available Availability of fast computer resources nowadays has facilitated more in-depth modeling of complex engineering systems which involve strong multiphysics interactions. This multiphysics modeling is an important necessity in nuclear reactor safety studies where efforts are being made worldwide to combine the knowledge from all associated disciplines at one place to accomplish the most realistic simulation of involved phenomenon. On these lines coupled modeling of nuclear reactor neutron kinetics, fuel heat transfer and coolant transport is a regular practice nowadays for transient analysis of reactor core. However optimization between modeling accuracy and computational economy has always been a challenging task to ensure the adequate degree of reliability in such extensive numerical exercises. Complex reactor core modeling involves estimation of evolving 3-D core thermal state, which in turn demands an expensive multichannel based detailed core thermal hydraulics model. A novel approach of power weighted coupling between core neutronics and thermal hydraulics presented in this work aims to reduce the bulk of core thermal calculations in core dynamics modeling to a significant extent without compromising accuracy of computation. Coupled core model has been validated against a series of international benchmarks. Accuracy and computational efficiency of the proposed multiphysics model has been demonstrated by analyzing a reactivity initiated transient.

  5. Modelization of physical phenomena in research reactors with the help of new developments in transport methods, and methodology validation with experimental data

    International Nuclear Information System (INIS)

    Rauck, St.

    2000-10-01

    The aim of this work is to develop a scheme for experimental reactors, based on transport equations. This type of reactors is characterized by a small core, a complex, very heterogeneous geometry and a large leakage. The possible insertion of neutron beams in the reflector and the presence of absorbers in the core increase the difficulty of the 3D-geometrical description and the physical modeling of the component parameters of the reactor. The Orphee reactor has been chosen for our study. Physical models (homogenization, collapsing cross section in few groups, albedo multigroup condition) have been developed in the APOLLO2 and CRONOS2 codes to calculate flux and power maps in a 3D-geometry, with different burnup and through transport equations. Comparisons with experimental measurements have shown the interest of taking into account anisotropy, steep flux gradients by using Sn methods, and on the other hand using a 12-group cross section library. The modeling of neutron beams has been done outside the core modeling through Monte Carlo calculations and with the total geometry, including a large thickness of heavy water. Thanks to this calculations, one can evaluate the neutron beams anti-reactivity and determinate the core cycle. We assure these methods more accurate than usual transport-diffusion calculations will be used for the conception of new research reactors. (author)

  6. Determination of contaminants in nuclear materials by measuring the capture gamma rays of thermal neutrons in a reactor internal geometry

    International Nuclear Information System (INIS)

    Suarez, A.A.

    1980-01-01

    A new method for analysis of impurities in nuclear fuel material was developed. Prompt gamma rays following thermal neutron capture, from a sample placed inside the research reactor were analyzed with a solid state high resolution detector. A number of improvements were introduced to improve the background-to-signal ratio, and the sensitivity of the method: use of collimeters for gamma rays and 6 Li 2 CO 3 filters to eliminate thermal neutrons from the beam were supplemented with the application of a pair spectrometer. Using a 42.5 cm 3 true coaxial Ge(Li) detector, and two optically separated NaI (Tl) scintillation detector, the sensitivity of the method for quantitative determination of impurities reached 30 p.p.m. The reproducibility of the results was better than 2%

  7. Morphing methods to parameterize specimen-specific finite element model geometries.

    Science.gov (United States)

    Sigal, Ian A; Yang, Hongli; Roberts, Michael D; Downs, J Crawford

    2010-01-19

    Shape plays an important role in determining the biomechanical response of a structure. Specimen-specific finite element (FE) models have been developed to capture the details of the shape of biological structures and predict their biomechanics. Shape, however, can vary considerably across individuals or change due to aging or disease, and analysis of the sensitivity of specimen-specific models to these variations has proven challenging. An alternative to specimen-specific representation has been to develop generic models with simplified geometries whose shape is relatively easy to parameterize, and can therefore be readily used in sensitivity studies. Despite many successful applications, generic models are limited in that they cannot make predictions for individual specimens. We propose that it is possible to harness the detail available in specimen-specific models while leveraging the power of the parameterization techniques common in generic models. In this work we show that this can be accomplished by using morphing techniques to parameterize the geometry of specimen-specific FE models such that the model shape can be varied in a controlled and systematic way suitable for sensitivity analysis. We demonstrate three morphing techniques by using them on a model of the load-bearing tissues of the posterior pole of the eye. We show that using relatively straightforward procedures these morphing techniques can be combined, which allows the study of factor interactions. Finally, we illustrate that the techniques can be used in other systems by applying them to morph a femur. Morphing techniques provide an exciting new possibility for the analysis of the biomechanical role of shape, independently or in interaction with loading and material properties. Copyright 2009 Elsevier Ltd. All rights reserved.

  8. An empirical model of diagnostic x-ray attenuation under narrow-beam geometry

    International Nuclear Information System (INIS)

    Mathieu, Kelsey B.; Kappadath, S. Cheenu; White, R. Allen; Atkinson, E. Neely; Cody, Dianna D.

    2011-01-01

    Purpose: The purpose of this study was to develop and validate a mathematical model to describe narrow-beam attenuation of kilovoltage x-ray beams for the intended applications of half-value layer (HVL) and quarter-value layer (QVL) estimations, patient organ shielding, and computer modeling. Methods: An empirical model, which uses the Lambert W function and represents a generalized Lambert-Beer law, was developed. To validate this model, transmission of diagnostic energy x-ray beams was measured over a wide range of attenuator thicknesses [0.49-33.03 mm Al on a computed tomography (CT) scanner, 0.09-1.93 mm Al on two mammography systems, and 0.1-0.45 mm Cu and 0.49-14.87 mm Al using general radiography]. Exposure measurements were acquired under narrow-beam geometry using standard methods, including the appropriate ionization chamber, for each radiographic system. Nonlinear regression was used to find the best-fit curve of the proposed Lambert W model to each measured transmission versus attenuator thickness data set. In addition to validating the Lambert W model, we also assessed the performance of two-point Lambert W interpolation compared to traditional methods for estimating the HVL and QVL [i.e., semilogarithmic (exponential) and linear interpolation]. Results: The Lambert W model was validated for modeling attenuation versus attenuator thickness with respect to the data collected in this study (R 2 > 0.99). Furthermore, Lambert W interpolation was more accurate and less sensitive to the choice of interpolation points used to estimate the HVL and/or QVL than the traditional methods of semilogarithmic and linear interpolation. Conclusions: The proposed Lambert W model accurately describes attenuation of both monoenergetic radiation and (kilovoltage) polyenergetic beams (under narrow-beam geometry).

  9. An empirical model of diagnostic x-ray attenuation under narrow-beam geometry.

    Science.gov (United States)

    Mathieu, Kelsey B; Kappadath, S Cheenu; White, R Allen; Atkinson, E Neely; Cody, Dianna D

    2011-08-01

    The purpose of this study was to develop and validate a mathematical model to describe narrow-beam attenuation of kilovoltage x-ray beams for the intended applications of half-value layer (HVL) and quarter-value layer (QVL) estimations, patient organ shielding, and computer modeling. An empirical model, which uses the Lambert W function and represents a generalized Lambert-Beer law, was developed. To validate this model, transmission of diagnostic energy x-ray beams was measured over a wide range of attenuator thicknesses [0.49-33.03 mm Al on a computed tomography (CT) scanner, 0.09-1.93 mm Al on two mammography systems, and 0.1-0.45 mm Cu and 0.49-14.87 mm Al using general radiography]. Exposure measurements were acquired under narrow-beam geometry using standard methods, including the appropriate ionization chamber, for each radiographic system. Nonlinear regression was used to find the best-fit curve of the proposed Lambert W model to each measured transmission versus attenuator thickness data set. In addition to validating the Lambert W model, we also assessed the performance of two-point Lambert W interpolation compared to traditional methods for estimating the HVL and QVL [i.e., semi-logarithmic (exponential) and linear interpolation]. The Lambert W model was validated for modeling attenuation versus attenuator thickness with respect to the data collected in this study (R2 > 0.99). Furthermore, Lambert W interpolation was more accurate and less sensitive to the choice of interpolation points used to estimate the HVL and/or QVL than the traditional methods of semilogarithmic and linear interpolation. The proposed Lambert W model accurately describes attenuation of both monoenergetic radiation and (kilovoltage) polyenergetic beams (under narrow-beam geometry).

  10. LPV model development and control of a solution copolymerization reactor

    NARCIS (Netherlands)

    Rahme, S.; Abbas, H.M.S.; Meskin, N.; Tóth, R.; Mohammadpour, J.

    2016-01-01

    In this paper, linear parameter-varying (LPV) control is considered for a solution copolymerization reactor, which takes into account the time-varying nature of the parameters of the process. The nonlinear model of the process is first converted to an exact LPV model representation in the

  11. Neutrino Mass Models: impact of non-zero reactor angle

    International Nuclear Information System (INIS)

    King, Stephen F.

    2011-01-01

    In this talk neutrino mass models are reviewed and the impact of a non-zero reactor angle and other deviations from tri-bi maximal mixing are discussed. We propose some benchmark models, where the only way to discriminate between them is by high precision neutrino oscillation experiments.

  12. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.P.

    1994-01-01

    The model comprises the whole primary circuit, including steam generators, loops, coolant pumps, main isolating valves and certainly the reactor pressure vessel and its internals. It was developed using the finite-element-code ANSYS. The model has a modular structure, so that various operational and assembling states can easily be considered. (orig./DG)

  13. Computational fluid dynamic modeling of fluidized-bed polymerization reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rokkam, Ram [Iowa State Univ., Ames, IA (United States)

    2012-01-01

    Polyethylene is one of the most widely used plastics, and over 60 million tons are produced worldwide every year. Polyethylene is obtained by the catalytic polymerization of ethylene in gas and liquid phase reactors. The gas phase processes are more advantageous, and use fluidized-bed reactors for production of polyethylene. Since they operate so close to the melting point of the polymer, agglomeration is an operational concern in all slurry and gas polymerization processes. Electrostatics and hot spot formation are the main factors that contribute to agglomeration in gas-phase processes. Electrostatic charges in gas phase polymerization fluidized bed reactors are known to influence the bed hydrodynamics, particle elutriation, bubble size, bubble shape etc. Accumulation of electrostatic charges in the fluidized-bed can lead to operational issues. In this work a first-principles electrostatic model is developed and coupled with a multi-fluid computational fluid dynamic (CFD) model to understand the effect of electrostatics on the dynamics of a fluidized-bed. The multi-fluid CFD model for gas-particle flow is based on the kinetic theory of granular flows closures. The electrostatic model is developed based on a fixed, size-dependent charge for each type of particle (catalyst, polymer, polymer fines) phase. The combined CFD model is first verified using simple test cases, validated with experiments and applied to a pilot-scale polymerization fluidized-bed reactor. The CFD model reproduced qualitative trends in particle segregation and entrainment due to electrostatic charges observed in experiments. For the scale up of fluidized bed reactor, filtered models are developed and implemented on pilot scale reactor.

  14. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    plate may have been underestimated and thus the heat flux had been underestimated. The MELCOR model predicts a film thickness on the order of 100 microns, which agrees very well with film flow model developed in this study for scaling analysis. However, the expected differences in film thicknesses for near vacuum and near atmospheric test conditions are not significant. Further study on the behavior of condensate film is expected to refine the simulation results. Possible refinements include but are not limited to, the followings: CFD simulation focusing on the liquid film behavior and benchmarking with experimental analyses for simpler geometries. 16 1 INTRODUCTION This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). The experimental results are employed to validate the containment condensation model in reactor containment system safety analysis code for integral SMRs. Such a containment condensation model is important to demonstrate the adequate cooling. In the three years of investigation, following the original proposal, the following planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental

  15. Modeling and simulation of CANDU reactor and its regulating system

    Science.gov (United States)

    Javidnia, Hooman

    Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

  16. Interactive Modeling of Architectural Freeform Structures - Combining Geometry with Fabrication and Statics

    KAUST Repository

    Jiang, Caigui

    2014-09-01

    This paper builds on recent progress in computing with geometric constraints, which is particularly relevant to architectural geometry. Not only do various kinds of meshes with additional properties (like planar faces, or with equilibrium forces in their edges) become available for interactive geometric modeling, but so do other arrangements of geometric primitives, like honeycomb structures. The latter constitute an important class of geometric objects, with relations to “Lobel” meshes, and to freeform polyhedral patterns. Such patterns are particularly interesting and pose research problems which go beyond what is known for meshes, e.g. with regard to their computing, their flexibility, and the assessment of their fairness.

  17. Modelling Plane Geometry: the connection between Geometrical Visualization and Algebraic Demonstration

    Science.gov (United States)

    Pereira, L. R.; Jardim, D. F.; da Silva, J. M.

    2017-12-01

    The teaching and learning of Mathematics contents have been challenging along the history of the education, both for the teacher, in his dedicated task of teaching, as for the student, in his arduous and constant task of learning. One of the topics that are most discussed in these contents is the difference between the concepts of proof and demonstration. This work presents an interesting discussion about such concepts considering the use of the mathematical modeling approach for teaching, applied to some examples developed in the classroom with a group of students enrolled in the discipline of Geometry of the Mathematics curse of UFVJM.

  18. Literature Reviews on Modeling Internal Geometry of Textile Composites and Rate-Independent Continuum Damage

    Science.gov (United States)

    Su-Yuen, Hsu

    2011-01-01

    Textile composite materials have good potential for constructing composite structures where the effects of three-dimensional stresses are critical or geometric complexity is a manufacturing concern. There is a recent interest in advancing competence within Langley Research Center for modeling the degradation of mechanical properties of textile composites. In an initial effort, two critical areas are identified to pursue: (1) Construction of internal geometry of textile composites, and (2) Rate-independent continuum damage mechanics. This report documents reviews on the two subjects. Various reviewed approaches are categorized, their assumptions, methods, and progress are briefed, and then critiques are presented. Each review ends with recommended research.

  19. Evaluation of the use of color-set geometry during lattice physics constants generation for boiling water reactor simulation

    International Nuclear Information System (INIS)

    Evans, S.; Ivanov, K.

    2013-01-01

    Current methods for BWR nuclear design and analysis consist of using lattice physics neutron transport methods to generate the two-group homogenized cross-sections that are then used in a nodal diffusion theory code. The lattice transport solutions are performed for a single assembly with reflective boundary conditions, which is a practical approximation. A method is developed to account for assembly exposure distributions (environment) in the core within the lattice transport calculations with the use of color-sets (2x2) geometry. The loading pattern is examined and an appropriate number of characteristic color-set cells are selected for analysis. Treatment of the co-resident exposed fuel within this method is also presented. The calculation process was followed for a recent BWR cycle design with comparisons being performed on both a lattice and core-wide basis to evaluate the proposed method. The lattice based comparisons show noticeable differences in the pin power distribution predictions, which require further investigation to see how this translates into core performance calculations. The core-wide comparisons show minor differences and are generally in a good agreement, which is expected with this small perturbation. A slight improvement was noticed in the reduction of the power distribution uncertainty. However, given the additional amount of work and computer run time increase, further evaluation, especially of core pin power predictions, is needed to consider this method for production level design and safety analysis calculations. (authors)

  20. Monte Carlo Modeling Electronuclear Processes in Cascade Subcritical Reactor

    CERN Document Server

    Bznuni, S A; Zhamkochyan, V M; Polyanskii, A A; Sosnin, A N; Khudaverdian, A G

    2000-01-01

    Accelerator driven subcritical cascade reactor composed of the main thermal neutron reactor constructed analogous to the core of the VVER-1000 reactor and a booster-reactor, which is constructed similar to the core of the BN-350 fast breeder reactor, is taken as a model example. It is shown by means of Monte Carlo calculations that such system is a safe energy source (k_{eff}=0.94-0.98) and it is capable of transmuting produced radioactive wastes (neutron flux density in the thermal zone is PHI^{max} (r,z)=10^{14} n/(cm^{-2} s^{-1}), neutron flux in the fast zone is respectively equal PHI^{max} (r,z)=2.25 cdot 10^{15} n/(cm^{-2} s^{-1}) if the beam current of the proton accelerator is k_{eff}=0.98 and I=5.3 mA). Suggested configuration of the "cascade" reactor system essentially reduces the requirements on the proton accelerator current.

  1. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  2. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  3. A modular approach to lead-cooled reactors modelling

    Energy Technology Data Exchange (ETDEWEB)

    Casamassima, V. [CESI RICERCA, via Rubattino 54, I-20134 Milano (Italy)], E-mail: casamassima@cesiricerca.it; Guagliardi, A. [CESI RICERCA, via Rubattino 54, I-20134 Milano (Italy)], E-mail: guagliardi@cesiricerca.it

    2008-06-15

    After an overview of the lego plant simulation tools (LegoPST), the paper gives some details about the ongoing LegoPST extension for modelling lead fast reactor plants. It refers to a simple mathematical model of the liquid lead channel dynamic process and shows the preliminary results of its application in dynamic simulation of the BREST 300 liquid lead steam generator. Steady state results agree with reference data [IAEA-TECDOC 1531, Fast Reactor Database, 2006 Update] both for water and lead.

  4. A modular approach to lead-cooled reactors modelling

    International Nuclear Information System (INIS)

    Casamassima, V.; Guagliardi, A.

    2008-01-01

    After an overview of the lego plant simulation tools (LegoPST), the paper gives some details about the ongoing LegoPST extension for modelling lead fast reactor plants. It refers to a simple mathematical model of the liquid lead channel dynamic process and shows the preliminary results of its application in dynamic simulation of the BREST 300 liquid lead steam generator. Steady state results agree with reference data [IAEA-TECDOC 1531, Fast Reactor Database, 2006 Update] both for water and lead

  5. System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

    International Nuclear Information System (INIS)

    Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

    1996-01-01

    In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

  6. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  7. Experimental Validation and Model Verification for a Novel Geometry ICPC Solar Collector

    DEFF Research Database (Denmark)

    Perers, Bengt; Duff, William S.; Daosukho, Jirachote

    A novel geometry ICPC solar collector was developed at the University of Chicago and Colorado State University. A ray tracing model has been designed to investigate the optical performance of both the horizontal and vertical fin versions of this collector. Solar radiation is modeled as discrete...... to the desired incident angle of the sun’s rays, performance of the novel ICPC solar collector at various specified angles along the transverse and longitudinal evacuated tube directions were experimentally determined. To validate the ray tracing model, transverse and longitudinal performance predictions...... at the corresponding specified incident angles are compared to the Sandia results. A 100 m2 336 Novel ICPC evacuated tube solar collector array has been in continuous operation at a demonstration project in Sacramento California since 1998. Data from the initial operation of the array are used to further validate...

  8. Information geometry and population genetics the mathematical structure of the Wright-Fisher model

    CERN Document Server

    Hofrichter, Julian; Tran, Tat Dat

    2017-01-01

    The present monograph develops a versatile and profound mathematical perspective of the Wright--Fisher model of population genetics. This well-known and intensively studied model carries a rich and beautiful mathematical structure, which is uncovered here in a systematic manner. In addition to approaches by means of analysis, combinatorics and PDE, a geometric perspective is brought in through Amari's and Chentsov's information geometry. This concept allows us to calculate many quantities of interest systematically; likewise, the employed global perspective elucidates the stratification of the model in an unprecedented manner. Furthermore, the links to statistical mechanics and large deviation theory are explored and developed into powerful tools. Altogether, the manuscript provides a solid and broad working basis for graduate students and researchers interested in this field.

  9. 3D Digital Surveying and Modelling of Cave Geometry: Application to Paleolithic Rock Art.

    Science.gov (United States)

    González-Aguilera, Diego; Muñoz-Nieto, Angel; Gómez-Lahoz, Javier; Herrero-Pascual, Jesus; Gutierrez-Alonso, Gabriel

    2009-01-01

    3D digital surveying and modelling of cave geometry represents a relevant approach for research, management and preservation of our cultural and geological legacy. In this paper, a multi-sensor approach based on a terrestrial laser scanner, a high-resolution digital camera and a total station is presented. Two emblematic caves of Paleolithic human occupation and situated in northern Spain, "Las Caldas" and "Peña de Candamo", have been chosen to put in practise this approach. As a result, an integral and multi-scalable 3D model is generated which may allow other scientists, pre-historians, geologists…, to work on two different levels, integrating different Paleolithic Art datasets: (1) a basic level based on the accurate and metric support provided by the laser scanner; and (2) a advanced level using the range and image-based modelling.

  10. Programming While Construction of Engineering 3D Models of Complex Geometry

    Science.gov (United States)

    Kheyfets, A. L.

    2017-11-01

    The capabilities of geometrically accurate computational 3D models construction with the use of programming are presented. The construction of models of an architectural arch and a glo-boid worm gear is considered as an example. The models are designed in the AutoCAD pack-age. Three programs of construction are given. The first program is for designing a multi-section architectural arch. The control of the arch’s geometry by impacting its main parameters is shown. The second program is for designing and studying the working surface of a globoid gear’s worm. The article shows how to make the animation for this surface’s formation. The third program is for formation of a worm gear cavity surface. The cavity formation dynamics is studied. The programs are written in the AutoLisp programming language. The program texts are provided.

  11. Monte Carlo modeling of fiber-scintillator flow-cell radiation detector geometry

    International Nuclear Information System (INIS)

    Rucker, T.L.; Ross, H.H.; Tennessee Univ., Knoxville; Schweitzer, G.K.

    1988-01-01

    A Monte Carlo computer calculation is described which models the geometric efficiency of a fiber-scintillator flow-cell radiation detector designed to detect radiolabeled compounds in liquid chromatography eluates. By using special mathematical techniques, an efficiency prediction with a precision of 1% is obtained after generating only 1000 random events. Good agreement is seen between predicted and experimental efficiency except for very low energy beta emission where the geometric limitation on efficiency is overcome by pulse height limitations which the model does not consider. The modeling results show that in the test system, the detection efficiency for low energy beta emitters is limited primarily by light generation and collection rather than geometry. (orig.)

  12. Validation of new 3-D neutronics model in APROS for hexagonal geometry

    International Nuclear Information System (INIS)

    Rintala, J.

    2010-01-01

    APROS - Advanced PROcess Simulation environment-is a widely used simulation tool for nuclear power plant modelling. Earlier the three-dimensional neutronics calculation has been based on model using the difference method. The original three-dimensional core model is mainly used in power plant simulator applications, where it fits well because of its speed. For safety analysis purposes, however, a new model was considered to be an important improvement to meet the accuracy requirements. A sophisticated nodal model used already in HEXTRAN and TRAB-3D was decided to be implemented into APROS. The hexagonal part of the model has now been implemented and tested. For practical reasons, the model was programmed from scratch into APROS and also some small improvements were added and thus, an extensive validation program was necessary to prove the correct behaviour of the model. In this paper, the most important results from AER kinetic benchmarks 2 and 3 calculations are shown as well as the calculation results against data achieved LR-0 test reactor space-time kinetic experiments. Since the model is similar to the one in HEXTRAN, the results in benchmarks are compared to the results by it. In LR-0 calculations, results by both, original and new model are presented and compared to the measurements. The results shows that the implementation of the model has been successful and the new model improves the accuracy of three-dimensional neutronics calculation in APROS into the level required in safety analyses. (Author)

  13. Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

    International Nuclear Information System (INIS)

    Matijević, Mario; Pevec, Dubravko; Trontl, Krešimir

    2015-01-01

    Highlights: • Shielding analysis of the typical PWR primary loop components was performed. • FW-CADIS methodology was thoroughly investigated using SCALE6.0 code package. • Versatile ability of SCALE6.0/FW-CADIS for deep penetration models was proved. • The adjoint source with focus on specific material can improve MC modeling. - Abstract: The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the S N part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high S N /P N parameters. Detailed voxelization (homogenization) process in Denovo together with high S N /P N parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR

  14. Comparison of low enriched uranium (UAl{sub x}-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)

    2011-07-01

    The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium (<20% {sup 235}U) UAl{sub x} dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

  15. 3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

    International Nuclear Information System (INIS)

    Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

    1974-01-01

    1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

  16. General empirical model for 60Co generation in pressurized water reactors with continuous refueling

    International Nuclear Information System (INIS)

    Urrutia, G.A.; Blesa, M.A.; Fernandez-Prini, R.; Maroto, A.J.G.

    1984-01-01

    A simplified model is presented that permits one to calculate the average activity on the fuel elements of a reactor that operates under continuous refueling, based on the assumption of crud interchange between fuel element surface and coolant in the form of particulate material only and using the crud specific activity as an empirical parameter determined in plant. The net activity flux from core to out-of-core components is then calculated in the form of parametric curves depending on crud specific activity and rate of particulate release from fuel surface. In pressure vessel reactors, contribution to out-ofcore radionuclide inventory arising in the release of activated materials from core components must be taken into account. The contribution from in situ activation of core components is calculated from the rates of release and the specific activities corresponding to the exposed surface of the component (calculated in a straightforward way on the basis of core geometry and neutron fluxes). The rates of release can be taken from the literature, or in the case of cobalt-rich alloys, can be calculated from experimentally determined cobalt contents of structural components and crud. For pressure vessel reactors operating under continuous refueling, activation of deposited crud and release of activated materials are compared; the latter, in certain cases, may represent a sizable (and even the largest) fraction of the total cobalt activity. It is proposed that the ratio of activities of 59 Fe to 54 Mn may be used as a diagnostic tool for in situ activation of structural materials; available data indicate ratios close to unity for pressure tube heavy water reactors (no in situ activation) and ratios around 4 to 10 for pressure vessel, heavy water reactors

  17. Properties of autoregressive model in reactor noise analysis, 1

    International Nuclear Information System (INIS)

    Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

    1987-01-01

    Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

  18. Modeling and simulation of pressurized water reactor power plant

    International Nuclear Information System (INIS)

    Wang, S.J.

    1983-01-01

    Two kinds of balance of plant (BOP) models of a pressurized water reactor (PWR) system are developed in this work - the detailed BOP model and the simple BOP model. The detailed model is used to simulate the normal operational performance of a whole BOP system. The simple model is used to combine with the NSSS model for a whole plant simulation. The trends of the steady state values of the detailed model are correct and the dynamic responses are reasonable. The simple BOP model approach starts the modelling work from the overall point of view. The response of the normalized turbine power and the feedwater inlet temperature to the steam generator of the simple model are compared with those of the detailed model. Both the steady state values and the dynamic responses are close to those of the detailed model. The simple BOP model is found adequate to represent the main performance of the BOP system. The simple balance of plant model was coupled with a NSSS model for a whole plant simulation. The NSSS model consists of the reactor core model, the steam generator model, and the coolant temperature control system. A closed loop whole plant simulation for an electric load perturbation was performed. The results are plausible. The coupling effect between the NSSS system and the BOP system was analyzed. The feedback of the BOP system has little effect on the steam generator performance, while the performance of the BOP system is strongly affected by the steam flow rate from the NSSS

  19. One-dimensional reactor kinetics model for RETRAN

    International Nuclear Information System (INIS)

    Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.

    1981-01-01

    This paper describes a one-dimensional spatial neutron kinetics model that was developed for the RETRAN code. The RETRAN -01 code has a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. A one-dimensional neutronics model has been developed for RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects for many operational transients. 19 refs

  20. Fuzzy model-based control of a nuclear reactor

    International Nuclear Information System (INIS)

    Van Den Durpel, L.; Ruan, D.

    1994-01-01

    The fuzzy model-based control of a nuclear power reactor is an emerging research topic world-wide. SCK-CEN is dealing with this research in a preliminary stage, including two aspects, namely fuzzy control and fuzzy modelling. The aim is to combine both methodologies in contrast to conventional model-based PID control techniques, and to state advantages of including fuzzy parameters as safety and operator feedback. This paper summarizes the general scheme of this new research project

  1. Component and system simulation models for High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs

  2. A spectral nodal method for eigenvalue S{sub N} transport problems in two-dimensional rectangular geometry for energy multigroup nuclear reactor global calculations

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: halves@iprj.uerj.br, E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pos-Graduacao em Modelagem Computacional

    2015-07-01

    A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S{sub N}) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S{sub N} discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S{sub N} transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S{sub N} eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)

  3. A spectral nodal method for eigenvalue SN transport problems in two-dimensional rectangular geometry for energy multigroup nuclear reactor global calculations

    International Nuclear Information System (INIS)

    Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C.

    2015-01-01

    A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S N ) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S N discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S N transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S N eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)

  4. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

    International Nuclear Information System (INIS)

    Ojeda S, J.; Morales S, J.; Chavez M, C.

    2009-10-01

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  5. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

    International Nuclear Information System (INIS)

    Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

    2010-01-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

  6. Reactor noise diagnostics based on multivariate autoregressive modeling: Application to LOFT [Loss-of-Fluid-Test] reactor process noise

    International Nuclear Information System (INIS)

    Gloeckler, O.; Upadhyaya, B.R.

    1987-01-01

    Multivariate noise analysis of power reactor operating signals is useful for plant diagnostics, for isolating process and sensor anomalies, and for automated plant monitoring. In order to develop a reliable procedure, the previously established techniques for empirical modeling of fluctuation signals in power reactors have been improved. Application of the complete algorithm to operational data from the Loss-of-Fluid-Test (LOFT) Reactor showed that earlier conjectures (based on physical modeling) regarding the perturbation sources in a Pressurized Water Reactor (PWR) affecting coolant temperature and neutron power fluctuations can be systematically explained. This advanced methodology has important implication regarding plant diagnostics, and system or sensor anomaly isolation. 6 refs., 24 figs

  7. Identification of the reduced order models of a BWR reactor

    International Nuclear Information System (INIS)

    Hernandez S, A.

    2004-01-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  8. Approach to decision modeling for an ignition test reactor

    International Nuclear Information System (INIS)

    Howland, H.R.; Varljen, T.C.

    1977-01-01

    A comparison matrix decision model is applied to candidates for a D-T ignition tokamak (TNS), including assessment of semi-quantifiable or judgemental factors as well as quantitative ones. The results show that TNS is mission-sensitive with a choice implied between near-term achievability and reactor technology

  9. Modeling of heat transfer in wall-cooled tubular reactors

    NARCIS (Netherlands)

    Koning, G.W.; Westerterp, K.R.

    1999-01-01

    In a pilot scale wall-cooled tubular reactor, temperature profiles have been measured with and without reaction. As a model reaction oxidation of carbon monoxide in air over a copper chromite catalyst has been used. The kinetics of this reaction have been determined separately in two kinetic

  10. Reactor accident calculation models in use in the Nordic countries

    International Nuclear Information System (INIS)

    Tveten, U.

    1984-01-01

    The report relates to a subproject under a Nordic project called ''Large reactor accidents - consequences and mitigating actions''. In the first part of the report short descriptions of the various models are given. A systematic list by subject is then given. In the main body of the report chapter and subchapter headings are by subject. (Auth.)

  11. Improved Nuclear Reactor and Shield Mass Model for Space Applications

    Science.gov (United States)

    Robb, Kevin

    2004-01-01

    New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

  12. Consequence model of the German reactor safety study

    International Nuclear Information System (INIS)

    Bayer, A.; Aldrich, D.; Burkart, K.; Horsch, F.; Hubschmann, W.; Schueckler, M.; Vogt, S.

    1979-01-01

    The consequency model developed for phase A of the German Reactor Safety Study (RSS) is similar in many respects to its counterpart in WASH-1400. As in that previous study, the model describes the atmosphere dispersion and transport of radioactive material released from the containment during a postulated reactor accident, and predicts its interaction with and influence on man. Differences do exist between the two models however, for the following reasons: (1) to more adequately reflect central European conditions, (2) to include improved submodels, and (3) to apply additional data and knowledge that have become available since publication of WASH-1400. The consequence model as used in phase A of the German RSS is described, highlighting differences between it and the U.S. model

  13. Vibration tests on some models of PEC reactor core elements

    International Nuclear Information System (INIS)

    Bonacina, G.; Castoldi, A.; Zola, M.; Cecchini, F.; Martelli, A.; Vincenzi, D.

    1982-01-01

    This paper describes the aims of the experimental tests carried out at ISMES, within an agreement with the Department of Fast Reactors of ENEA, on some models of the elements of PEC Fast Nuclear Reactor Core in the frame of the activities for the seismic verification of the PEC core. The seismic verification is briefly described with particular attention to the problems arising from the shocks among the various elements during an earthquake, as well as the computer code used, the purpose and the techniques used to perform tests, some results and the first comparison between the theory and the experimental data

  14. Modeling earthquake sequences along the Manila subduction zone: Effects of three-dimensional fault geometry

    Science.gov (United States)

    Yu, Hongyu; Liu, Yajing; Yang, Hongfeng; Ning, Jieyuan

    2018-05-01

    To assess the potential of catastrophic megathrust earthquakes (MW > 8) along the Manila Trench, the eastern boundary of the South China Sea, we incorporate a 3D non-planar fault geometry in the framework of rate-state friction to simulate earthquake rupture sequences along the fault segment between 15°N-19°N of northern Luzon. Our simulation results demonstrate that the first-order fault geometry heterogeneity, the transitional-segment (possibly related to the subducting Scarborough seamount chain) connecting the steeper south segment and the flatter north segment, controls earthquake rupture behaviors. The strong along-strike curvature at the transitional-segment typically leads to partial ruptures of MW 8.3 and MW 7.8 along the southern and northern segments respectively. The entire fault occasionally ruptures in MW 8.8 events when the cumulative stress in the transitional-segment is sufficiently high to overcome the geometrical inhibition. Fault shear stress evolution, represented by the S-ratio, is clearly modulated by the width of seismogenic zone (W). At a constant plate convergence rate, a larger W indicates on average lower interseismic stress loading rate and longer rupture recurrence period, and could slow down or sometimes stop ruptures that initiated from a narrower portion. Moreover, the modeled interseismic slip rate before whole-fault rupture events is comparable with the coupling state that was inferred from the interplate seismicity distribution, suggesting the Manila trench could potentially rupture in a M8+ earthquake.

  15. Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Thanigaiyarasu, G.; Chellapandi, P.

    2014-01-01

    Highlights: • Mathematical modelling of dynamic behaviour of safety rod during scram action in fast reactor. • Effects of hydraulics, structural interaction and geometry on drop time of safety rod are understood. • Using simplified model, drop time can be assessed replacing detailed CFD analysis. • Sensitivities of the related parameters on drop time are understood. • Experimental validation qualifies the modelling and computer software developed. - Abstract: Performance of safety rod and its drive mechanism which are parts of shutdown systems in sodium cooled fast reactor (SFR) plays a major role in ensuring safe operation of the plant during all the design basis events. The safety rods are to be inserted into the core within a stipulated time during off-normal conditions of the reactor. Mathematical modelling of dynamic behaviour of a safety rod and its drive mechanism in a typical 500 MWe SFR during scram action is considered in the present study. A full-scale prototype system has undergone qualification tests in air, water and in sodium simulating the operating conditions in the reactor. In this paper, the salient features of the safety rod and its mechanism, details related to mathematical modelling and sensitivity of the parameters having influence on drop time are presented. The outcomes of the numerical analysis are compared with the experimental results. In this process, the mathematical model and the computer software developed are validated

  16. Dispersive versus constant-geometry models of the neutron-208Pb mean field

    International Nuclear Information System (INIS)

    Mahaux, C.; Sartor, R.

    1990-01-01

    Phenomenological optical-model analyses of differential elastic scattering cross sections of neutrons by 208 Pb indicate that the radius of the real part of the potential decreases with increasing energy in the domain 4< E<40 MeV. On the other hand, the experimental total cross section is compatible with a real potential whose radial shape is energy independent. In order to clarify this situation, we compare a 'constant geometry' model whose real part has an energy-independent radial shape with a 'dispersive model' whose real part has an energy-dependent radial shape calculated from the dispersion relation which connects the real and imaginary parts of the field. The following three main features are considered. (i) The junction of the optical-model potential with the shell-model potential at negative energy. (ii) The agreement between the calculated total and differential cross sections and their experimental values. (iii) The extent to which the real part of the optical-model potential can be accurately determined by analyzing the total cross section only. It is concluded that the presently available experimental data support the existence of an energy dependence of the radial shape of the real potential, in keeping with the dispersion relation. A new parametrization of a 'dispersive' mean field is also presented. It does not involve more parameters than the previously published one but takes better account of the physical properties of the spectral functions; it is shown to improve the agreement between predicted and experimental scattering data. (orig.)

  17. Neural network modeling of chaotic dynamics in nuclear reactor flows

    International Nuclear Information System (INIS)

    Welstead, S.T.

    1992-01-01

    Neural networks have many scientific applications in areas such as pattern classification and time series prediction. The universal approximation property of these networks, however, can also be exploited to provide researchers with tool for modeling observed nonlinear phenomena. It has been shown that multilayer feed forward networks can capture important global nonlinear properties, such as chaotic dynamics, merely by training the network on a finite set of observed data. The network itself then provides a model of the process that generated the data. Characterizations such as the existence and general shape of a strange attractor and the sign of the largest Lyapunov exponent can then be extracted from the neural network model. In this paper, the author applies this idea to data generated from a nonlinear process that is representative of convective flows that can arise in nuclear reactor applications. Such flows play a role in forced convection heat removal from pressurized water reactors and boiling water reactors, and decay heat removal from liquid-metal-cooled reactors, either by natural convection or by thermosyphons

  18. The relation between geometry, hydrology and stability of complex hillslopes examined using low-dimensional hydrological models

    NARCIS (Netherlands)

    Talebi, A.

    2008-01-01

    Key words: Hillslope geometry, Hillslope hydrology, Hillslope stability, Complex hillslopes, Modeling shallow landslides, HSB model, HSB-SM model.

    The hydrologic response of a hillslope to rainfall involves a complex, transient saturated-unsaturated interaction that usually leads to a

  19. Simulation of MILD combustion using Perfectly Stirred Reactor model

    KAUST Repository

    Chen, Z.

    2016-07-06

    A simple model based on a Perfectly Stirred Reactor (PSR) is proposed for moderate or intense low-oxygen dilution (MILD) combustion. The PSR calculation is performed covering the entire flammability range and the tabulated chemistry approach is used with a presumed joint probability density function (PDF). The jet, in hot and diluted coflow experimental set-up under MILD conditions, is simulated using this reactor model for two oxygen dilution levels. The computed results for mean temperature, major and minor species mass fractions are compared with the experimental data and simulation results obtained recently using a multi-environment transported PDF approach. Overall, a good agreement is observed at three different axial locations for these comparisons despite the over-predicted peak value of CO formation. This suggests that MILD combustion can be effectively modelled by the proposed PSR model with lower computational cost.

  20. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  1. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    International Nuclear Information System (INIS)

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-01-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  2. Thermohydraulic modeling and simulation of breeder reactors

    International Nuclear Information System (INIS)

    Agrawal, A.K.; Khatib-Rahbar, M.; Curtis, R.T.; Hetrick, D.L.; Girijashankar, P.V.

    1982-01-01

    This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is discussed

  3. Fracture network modelling: an integrated approach for realisation of complex fracture network geometries

    International Nuclear Information System (INIS)

    Srivastava, R.M.

    2007-01-01

    In its efforts to improve geological support of the safety case, Ontario Power Generation's Deep Geologic Repository Technology Programme (DGRTP) has developed a procedure (Srivastava, 2002) for creating realistic 3-D fracture network models (FNMs) that honor information typically available at the time of preliminary site characterisation: By accommodating all of the these various pieces of 'hard' and 'soft' data, these FNMs provide a single, coherent and consistent model that can serve the needs of many preliminary site characterisation studies. The detailed, complex and realistic models of 3-D fracture geometry produced by this method can serve as the basis for developing rock property models to be used in flow and transport studies. They can also be used for exploring the suitability of a proposed site by providing quantitative assessments of the probability that a proposed repository with a specified geometry will be intersected by fractures. When integrated with state-of-the-art scientific visualisation, these models can also help in the planning of additional data gathering activities by identifying critical fractures that merit further detailed investigation. Finally, these FNMs can serve as one of the central elements of the presentation and explanation of the Descriptive Conceptual Geosphere Model (DCM) to other interested parties, including non-technical audiences. In addition to being ideally suited to preliminary site characterisation, the approach also readily incorporates field data that may become available during subsequent site investigations, including ground reconnaissance, borehole programmes and other subsurface studies. A single approach can therefore serve the needs of the site characterisation from its inception through several years of data collection and more detailed site-specific investigations, accommodating new data as they become available and updating the FNMs accordingly. The FNMs from this method are probabilistic in the sense that

  4. Flooding Experiments and Modeling for Improved Reactor Safety

    International Nuclear Information System (INIS)

    Solmos, M.; Hogan, K.J.; VIerow, K.

    2008-01-01

    Countercurrent two-phase flow and 'flooding' phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing

  5. Linear Dynamics Model for Steam Cooled Fast Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollmer, H

    1968-04-15

    A linear analytical dynamic model is developed for steam cooled fast power reactors. All main components of such a plant are investigated on a general though relatively simple basis. The model is distributed in those parts concerning the core but lumped as to the external plant components. Coolant is considered as compressible and treated by the actual steam law. Combined use of analogue and digital computer seems most attractive.

  6. Computer modelling of radioactive source terms at a tokamak reactor

    International Nuclear Information System (INIS)

    Meide, A.

    1984-12-01

    The Monte Carlo code MCNP has been used to create a simple three-dimensional mathematical model representing 1/12 of a tokamak fusion reactor for studies of the exposure rate level from neutrons as well as gamma rays from the activated materials, and for later estimates of the consequences to the environment, public, and operating personnel. The model is based on the recommendations from the NET/INTOR workshops. (author)

  7. Validation of the dynamic model for a pressurized water reactor

    International Nuclear Information System (INIS)

    Zwingelstein, Gilles.

    1979-01-01

    Dynamic model validation is a necessary procedure to assure that the developed empirical or physical models are satisfactorily representing the dynamic behavior of the actual plant during normal or abnormal transients. For small transients, physical models which represent isolated core, isolated steam generator and the overall pressurized water reactor are described. Using data collected during the step power changes that occured during the startup procedures, comparisons of experimental and actual transients are given at 30% and 100% of full power. The agreement between the transients derived from the model and those recorded on the plant indicates that the developed models are well suited for use for functional or control studies

  8. MODELING THE ELECTROLYTIC DECHLORINATION OF TRICHLOROETHYLENE IN A GRANULAR GRAPHITE-PACKED REACTOR

    Science.gov (United States)

    A comprehensive reactor model was developed for the electrolytic dechlorination of trichloroethylene (TCE) at a granular-graphite cathode. The reactor model describes the dynamic processes of TCE dechlorination and adsorption, and the formation and dechlorination of all the major...

  9. Integrated fuel-cycle models for fast breeder reactors

    International Nuclear Information System (INIS)

    Ott, K.O.; Maudlin, P.J.

    1981-01-01

    Breeder-reactor fuel-cycle analysis can be divided into four different areas or categories. The first category concerns questions about the spatial variation of the fuel composition for single loading intervals. Questions of the variations in the fuel composition over several cycles represent a second category. Third, there is a need for a determination of the breeding capability of the reactor. The fourth category concerns the investigation of breeding and long-term fuel logistics. Two fuel-cycle models used to answer questions in the third and fourth area are presented. The space- and time-dependent actinide balance, coupled with criticality and fuel-management constraints, is the basis for both the Discontinuous Integrated Fuel-Cycle Model and the Continuous Integrated Fuel-Cycle Model. The results of the continuous model are compared with results obtained from detailed two-dimensional space and multigroup depletion calculations. The continuous model yields nearly the same results as the detailed calculation, and this is with a comparatively insignificant fraction of the computational effort needed for the detailed calculation. Thus, the integrated model presented is an accurate tool for answering questions concerning reactor breeding capability and long-term fuel logistics. (author)

  10. Reactor kinetics revisited: a coefficient based model (CBM)

    International Nuclear Information System (INIS)

    Ratemi, W.M.

    2011-01-01

    In this paper, a nuclear reactor kinetics model based on Guelph expansion coefficients calculation ( Coefficients Based Model, CBM), for n groups of delayed neutrons is developed. The accompanying characteristic equation is a polynomial form of the Inhour equation with the same coefficients of the CBM- kinetics model. Those coefficients depend on Universal abc- values which are dependent on the type of the fuel fueling a nuclear reactor. Furthermore, such coefficients are linearly dependent on the inserted reactivity. In this paper, the Universal abc- values have been presented symbolically, for the first time, as well as with their numerical values for U-235 fueled reactors for one, two, three, and six groups of delayed neutrons. Simulation studies for constant and variable reactivity insertions are made for the CBM kinetics model, and a comparison of results, with numerical solutions of classical kinetics models for one, two, three, and six groups of delayed neutrons are presented. The results show good agreements, especially for single step insertion of reactivity, with the advantage of the CBM- solution of not encountering the stiffness problem accompanying the numerical solutions of the classical kinetics model. (author)

  11. Studies on modelling of bubble driven flows in chemical reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grevskott, Sverre

    1997-12-31

    Multiphase reactors are widely used in the process industry, especially in the petrochemical industry. They very often are characterized by very good thermal control and high heat transfer coefficients against heating and cooling surfaces. This thesis first reviews recent advances in bubble column modelling, focusing on the fundamental flow equations, drag forces, transversal forces and added mass forces. The mathematical equations for the bubble column reactor are developed, using an Eulerian description for the continuous and dispersed phase in tensor notation. Conservation equations for mass, momentum, energy and chemical species are given, and the k-{epsilon} and Rice-Geary models for turbulence are described. The different algebraic solvers used in the model are described, as are relaxation procedures. Simulation results are presented and compared with experimental values. Attention is focused on the modelling of void fractions and gas velocities in the column. The energy conservation equation has been included in the bubble column model in order to model temperature distributions in a heated reactor. The conservation equation of chemical species has been included to simulate absorption of CO{sub 2}. Simulated axial and radial mass fraction profiles for CO{sub 2} in the gas phase are compared with measured values. Simulations of the dynamic behaviour of the column are also presented. 189 refs., 124 figs., 1 tab.

  12. Fractal Geometry Enables Classification of Different Lung Morphologies in a Model of Experimental Asthma

    Science.gov (United States)

    Obert, Martin; Hagner, Stefanie; Krombach, Gabriele A.; Inan, Selcuk; Renz, Harald

    2015-06-01

    Animal models represent the basis of our current understanding of the pathophysiology of asthma and are of central importance in the preclinical development of drug therapies. The characterization of irregular lung shapes is a major issue in radiological imaging of mice in these models. The aim of this study was to find out whether differences in lung morphology can be described by fractal geometry. Healthy and asthmatic mouse groups, before and after an acute asthma attack induced by methacholine, were studied. In vivo flat-panel-based high-resolution Computed Tomography (CT) was used for mice's thorax imaging. The digital image data of the mice's lungs were segmented from the surrounding tissue. After that, the lungs were divided by image gray-level thresholds into two additional subsets. One subset contained basically the air transporting bronchial system. The other subset corresponds mainly to the blood vessel system. We estimated the fractal dimension of all sets of the different mouse groups using the mass radius relation (mrr). We found that the air transporting subset of the bronchial lung tissue enables a complete and significant differentiation between all four mouse groups (mean D of control mice before methacholine treatment: 2.64 ± 0.06; after treatment: 2.76 ± 0.03; asthma mice before methacholine treatment: 2.37 ± 0.16; after treatment: 2.71 ± 0.03; p < 0.05). We conclude that the concept of fractal geometry allows a well-defined, quantitative numerical and objective differentiation of lung shapes — applicable most likely also in human asthma diagnostics.

  13. Radionuclide transport in running waters, sensitivity analysis of bed-load, channel geometry and model discretisation

    International Nuclear Information System (INIS)

    Jonsson, Karin; Elert, Mark

    2006-08-01

    In this report, further investigations of the model concept for radionuclide transport in stream, developed in the SKB report TR-05-03 is presented. Especially three issues have been the focus of the model investigations. The first issue was to investigate the influence of assumed channel geometry on the simulation results. The second issue was to reconsider the applicability of the equation for the bed-load transport in the stream model, and finally the last issue was to investigate how the model discretisation will influence the simulation results. The simulations showed that there were relatively small differences in results when applying different cross-sections in the model. The inclusion of the exact shape of the cross-section in the model is therefore not crucial, however, if cross-sectional data exist, the overall shape of the cross-section should be used in the model formulation. This could e.g. be accomplished by using measured values of the stream width and depth in the middle of the stream and by assuming a triangular shape. The bed-load transport was in this study determined for different sediment characteristics which can be used as an order of magnitude estimation if no exact determinations of the bed-load are available. The difference in the calculated bed-load transport for the different materials was, however, found to be limited. The investigation of model discretisation showed that a fine model discretisation to account for numerical effects is probably not important for the performed simulations. However, it can be necessary for being able to account for different conditions along a stream. For example, the application of mean slopes instead of individual values in the different stream reaches can result in very different predicted concentrations

  14. Molecular dynamics modeling on the role of initial void geometry in a thin aluminum film under uniaxial tension

    International Nuclear Information System (INIS)

    Cui, Yi; Chen, Zengtao

    2015-01-01

    The effect of initial void geometry on damage progression in a thin aluminum film under uniaxial load is studied via molecular dynamics (MD) method. The embedded voids are with different initial geometries regarding shape, porosity and intervoid ligament distance (ILD). Major simulations are run upon twelve MD geometries with each containing 8–27 million atoms. The corresponding stress–strain relation is monitored during the microstructure evolution of the specimens. The critical stress to trigger the dislocation emission is found in line with the prediction of the Lubarda model. The simulation results reveal that the initial void geometry has substantial impact on the stress–strain relation especially for a specimen with larger initial porosity. (paper)

  15. One-velocity neutron diffusion calculations based on a two-group reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Bingulac, S; Radanovic, L; Lazarevic, B; Matausek, M; Pop-Jordanov, J [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1965-07-01

    Many processes in reactor physics are described by the energy dependent neutron diffusion equations which for many practical purposes can often be reduced to one-dimensional two-group equations. Though such two-group models are satisfactory from the standpoint of accuracy, they require rather extensive computations which are usually iterative and involve the use of digital computers. In many applications, however, and particularly in dynamic analyses, where the studies are performed on analogue computers, it is preferable to avoid iterative calculations. The usual practice in such situations is to resort to one group models, which allow the solution to be expressed analytically. However, the loss in accuracy is rather great particularly when several media of different properties are involved. This paper describes a procedure by which the solution of the two-group neutron diffusion. equations can be expressed analytically in the form which, from the computational standpoint, is as simple as the one-group model, but retains the accuracy of the two-group treatment. In describing the procedure, the case of a multi-region nuclear reactor of cylindrical geometry is treated, but the method applied and the results obtained are of more general application. Another approach in approximate solution of diffusion equations, suggested by Galanin is applicable only in special ideal cases.

  16. Stability of a two-volume MRxMHD model in slab geometry

    Science.gov (United States)

    Tuen, Li Huey

    Ideal MHD models are known to be inadequate to describe various physical attributes of a toroidal field with non-continuous symmetry, such as magnetic islands and stochastic regions. Motivated by this omission, a new variational principle MRXMHD was developed; rather than include an infinity of magnetic flux surfaces, MRxMHD has a finite number of flux surfaces, and thus supports partial plasma relaxation. The model comprises of relaxed plasma regions which are separated by nested ideal MHD interfaces (flux surfaces), and can be encased in a perfectly conducting wall. In each region the pressure is constant, but can jump across interfaces. The field and field pitch, or rotational transform, can also jump across the interfaces. Unlike ideal MHD, MRxMHD plasmas can support toroidally non-axisymmetric confined magnetic fields, magnetic islands and stochastic regions. In toroidally non-axisymmetric plasma, the existence of interfaces in MRxMHD is contingent on the irrationality of the rotational transform of flux surfaces. That is, the KAM theorem shows that invariant tori (flux surfaces) continue to exist for sufficiently small perturbations to an integrable system (which describes flux surfaces), provided that the rotational transform is sufficiently irrational. Building upon the MRxMHD stability model, we study the effects of irrationality of the rotational transform at interfaces in MRxMHD on plasma stability. We present an MRxMHD equilibrium model to investigate the effects of magnetic field pitch within the plasma and across the aforementioned flux surfaces within a chosen geometry. In this model, it is found that the 2D system stability conditions are dependent on the interface and resonant surface magnetic field pitch at minimised energy states, and the stability of a system as a function of magnetic field pitch destabilises at particular values of magnetic field pitch. We benchmark the treatment of a two-volume system, along with the calculations for

  17. Fault detection in IRIS reactor secondary loop using inferential models

    International Nuclear Information System (INIS)

    Perillo, Sergio R.P.; Upadhyaya, Belle R.; Hines, J. Wesley

    2013-01-01

    The development of fault detection algorithms is well-suited for remote deployment of small and medium reactors, such as the IRIS, and the development of new small modular reactors (SMR). However, an extensive number of tests are still to be performed for new engineering aspects and components that are not yet proven technology in the current PWRs, and present some technological challenges for its deployment since many of its features cannot be proven until a prototype plant is built. In this work, an IRIS plant simulation platform was developed using a Simulink® model. The dynamic simulation was utilized in obtaining inferential models that were used to detect faults artificially added to the secondary system simulations. The implementation of data-driven models and the results are discussed. (author)

  18. Theoretical modeling of electroosmotic flow in soft microchannels: A variational approach applied to the rectangular geometry

    Science.gov (United States)

    Sadeghi, Arman

    2018-03-01

    Modeling of fluid flow in polyelectrolyte layer (PEL)-grafted microchannels is challenging due to their two-layer nature. Hence, the pertinent studies are limited only to circular and slit geometries for which matching the solutions for inside and outside the PEL is simple. In this paper, a simple variational-based approach is presented for the modeling of fully developed electroosmotic flow in PEL-grafted microchannels by which the whole fluidic area is considered as a single porous medium of variable properties. The model is capable of being applied to microchannels of a complex cross-sectional area. As an application of the method, it is applied to a rectangular microchannel of uniform PEL properties. It is shown that modeling a rectangular channel as a slit may lead to considerable overestimation of the mean velocity especially when both the PEL and electric double layer (EDL) are thick. It is also demonstrated that the mean velocity is an increasing function of the fixed charge density and PEL thickness and a decreasing function of the EDL thickness and PEL friction coefficient. The influence of the PEL thickness on the mean velocity, however, vanishes when both the PEL thickness and friction coefficient are sufficiently high.

  19. Cooperative learning model with high order thinking skills questions: an understanding on geometry

    Science.gov (United States)

    Sari, P. P.; Budiyono; Slamet, I.

    2018-05-01

    Geometry, a branch of mathematics, has an important role in mathematics learning. This research aims to find out the effect of learning model, emotional intelligence, and the interaction between learning model and emotional intelligence toward students’ mathematics achievement. This research is quasi-experimental research with 2 × 3 factorial design. The sample in this research included 179 Senior High School students on 11th grade in Sukoharjo Regency, Central Java, Indonesia in academic year of 2016/2017. The sample was taken by using stratified cluster random sampling. The results showed that: the student are taught by Thinking Aloud Pairs Problem-Solving using HOTs questions provides better mathematics learning achievement than Make A Match using HOTs questions. High emotional intelligence students have better mathematics learning achievement than moderate and low emotional intelligence students, and moderate emotional intelligence students have better mathematics learning achievement than low emotional intelligence students. There is an interaction between learning model and emotional intelligence, and these affect mathematics learning achievement. We conclude that appropriate learning model can support learning activities become more meaningful and facilitate students to understand material. For further research, we suggest to explore the contribution of other aspects in cooperative learning modification to mathematics achievement.

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

    Science.gov (United States)

    Gruber, Mathias F.; Johnson, Carl J.; Tang, Chuyang; Jensen, Mogens H.; Yde, Lars; Hélix-Nielsen, Claus

    2012-01-01

    In forward osmosis (FO), an osmotic pressure gradient generated across a semi-permeable membrane is used to generate water transport from a dilute feed solution into a concentrated draw solution. This principle has shown great promise in the areas of water purification, wastewater treatment, seawater desalination and power generation. To ease optimization and increase understanding of membrane systems, it is desirable to have a comprehensive model that allows for easy investigation of all the major parameters in the separation process. Here we present experimental validation of a computational fluid dynamics (CFD) model developed to simulate FO experiments with asymmetric membranes. Simulations are compared with experimental results obtained from using two distinctly different complex three-dimensional membrane chambers. It is found that the CFD model accurately describes the solute separation process and water permeation through membranes under various flow conditions. It is furthermore demonstrated how the CFD model can be used to optimize membrane geometry in such as way as to promote the mass transfer. PMID:24958428

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

    Directory of Open Access Journals (Sweden)

    Lars Yde

    2012-11-01

    Full Text Available In forward osmosis (FO, an osmotic pressure gradient generated across a semi-permeable membrane is used to generate water transport from a dilute feed solution into a concentrated draw solution. This principle has shown great promise in the areas of water purification, wastewater treatment, seawater desalination and power generation. To ease optimization and increase understanding of membrane systems, it is desirable to have a comprehensive model that allows for easy investigation of all the major parameters in the separation process. Here we present experimental validation of a computational fluid dynamics (CFD model developed to simulate FO experiments with asymmetric membranes. Simulations are compared with experimental results obtained from using two distinctly different complex three-dimensional membrane chambers. It is found that the CFD model accurately describes the solute separation process and water permeation through membranes under various flow conditions. It is furthermore demonstrated how the CFD model can be used to optimize membrane geometry in such as way as to promote the mass transfer.

  2. Using geometry to improve model fitting and experiment design for glacial isostasy

    Science.gov (United States)

    Kachuck, S. B.; Cathles, L. M.

    2017-12-01

    As scientists we routinely deal with models, which are geometric objects at their core - the manifestation of a set of parameters as predictions for comparison with observations. When the number of observations exceeds the number of parameters, the model is a hypersurface (the model manifold) in the space of all possible predictions. The object of parameter fitting is to find the parameters corresponding to the point on the model manifold as close to the vector of observations as possible. But the geometry of the model manifold can make this difficult. By curving, ending abruptly (where, for instance, parameters go to zero or infinity), and by stretching and compressing the parameters together in unexpected directions, it can be difficult to design algorithms that efficiently adjust the parameters. Even at the optimal point on the model manifold, parameters might not be individually resolved well enough to be applied to new contexts. In our context of glacial isostatic adjustment, models of sparse surface observations have a broad spread of sensitivity to mixtures of the earth's viscous structure and the surface distribution of ice over the last glacial cycle. This impedes precise statements about crucial geophysical processes, such as the planet's thermal history or the climates that controlled the ice age. We employ geometric methods developed in the field of systems biology to improve the efficiency of fitting (geodesic accelerated Levenberg-Marquardt) and to identify the maximally informative sources of additional data to make better predictions of sea levels and ice configurations (optimal experiment design). We demonstrate this in particular in reconstructions of the Barents Sea Ice Sheet, where we show that only certain kinds of data from the central Barents have the power to distinguish between proposed models.

  3. A model to predict moisture conditions in concrete reactor containments

    International Nuclear Information System (INIS)

    Ahs, M.; Nilsson, L.O.; Poyet, S.; L'Hostis, V.

    2015-01-01

    Moisture has an impact in many of the degradation mechanisms that appear in the structures of a nuclear power plant. Moisture conditions in a reactor containment wall have been simulated by using a hygro-thermal model of drying concrete. Methods to estimate the temperature dependency of the sorption isotherms and moisture transport properties is suggested and applied in the model. This temperature dependency is included as there is a temperature gradient present through the containment wall. The hygro-thermal model was applied on a full scale 3D model of a real reactor containment building and the concrete relative humidity has been computed at 4 different moments: 1, 10, 20 and 30 years. The results show that the major part of the concrete is not dried at all even after 30 years of operation. It is also clear that the temperature distribution inside the whole concrete volume is affected by the variable boundary conditions. It was concluded that the suggested hygro-thermal model was appropriate to use as a method to estimate the existing conditions in a PWR reactor containment wall

  4. Modelling of infrared multiphoton absorption and dissociation for design of reactors for isotope separation by lasers

    International Nuclear Information System (INIS)

    Takeuchi, Kazuo; Nakane, Ryohei; Inoue, Cihiro

    1981-01-01

    A series of experiments were performed on infrared laser beam absorption (multiphoton absorption) and subsequent dissociation (multiphoton dissociation) of CF 3 Cl to propose models for the design of reactors for isotope separation by lasers. A parallel beam geometry was utilized in batch irradiation experiments to make direct compilation of lumped-parameter data possible. Multiphoton absorption is found to be expressed by a power-law extension of the law of Lambert and by an addition of a new term for buffer gas effect to the law of Beer. For reaction analysis, a method to evaluate the effect of incomplete mixing on apparent reaction rates is first presented. Secondly, multiphoton dissociation of Cf 3 Cl is found to occur in pseudo-first order fashion and the specific reaction rates for different beam fluence are shown to be correlated to the absorbed energy. (author)

  5. Modelling perspectives on radiation chemistry in BWR reactor core

    International Nuclear Information System (INIS)

    Ibe, Eishi

    1991-01-01

    Development of a full-system boiling water reactor core model started in 1982. The model included a two-region reactor core, one with and one without boiling. Key design parameters consider variable dose rates in a three-layer liquid downcomer. Dose rates in the core and downcomer include both generation and recombination reactions of species. Agreement is good between calculations and experimental data of oxygen concentration as a function of hydrogen concentration for different bubble sizes. Oxygen concentration is reduced in the reactor pressure vessel (RPV) by increasing bubble size. The multilayer model follows the oxygen data better than a single-layered model at high concentrations of hydrogen. Key reactions are reduced to five radiolysis reactions and four decomposition reactions for hydrogen peroxide. Calculations by the DOT 3 code showed dose rates from neutrons and gamma rays in various parts of the core. Concentrations of oxygen, hydrogen peroxide, and hydrogen were calculated by the model as a function of time from core inlet. Similar calculations for NWC and HWC were made as a function of height from core inlet both in the boiling channel an the bypass channel. Finally the model was applied to calculate the oxygen plus half the hydrogen peroxide concentrations as a function of hydrogen concentration to compare with data from five plants. Power density distribution with core height was given for an early stage and an end stage of a cycle. Increases of dose rates in the turbine for seven plants were shown as a function of increased hydrogen concentration in the reactor water

  6. Energy deposition in STARFIRE reactor components

    International Nuclear Information System (INIS)

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry

  7. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  8. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  9. Observations of Tin/Water Thermal Explosions in a Long-Tube Geometry. Their Interpretation and Consequences for the Detonation Model

    International Nuclear Information System (INIS)

    Hall, R.W.; Board, S.J.; Baines, M.

    1979-01-01

    This paper presents details of experiments designed to test the detonation model of thermal explosions (Board et al, 1975); on this theory large-scale explosions should propagate steadily at supersonic velocities through a fuel coolant mixture, giving a yield which has been shown to depend on details of the fragmentation and heat transfer behind the shock front. Observations of propagating explosions have been reported previously. In the present work, a long-tube geometry is used since in 1D, propagation measurements are particularly easy to interpret. Also, in 2D and 3D geometries radial flow can tend to extinguish shock waves and if a single-phase region of coolant is present, pressure pulses can propagate ahead of the two-phase shock in the intermixed region. This paper describes the six experiments that all use molten tin and water mixtures. In the first four, detailed pressure measurement was the main objective; the last two are attempts at flow visualization to aid the interpretation of these. The results obtained and the implications for the detonation model are discussed. A detailed interpretation in terms of fragmentation and heat transfer processes behind the shock is attempted. The implications of the work for reactor materials are then briefly outlined

  10. A porous medium model for predicting the duct wall temperature of sodium fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yiqi, E-mail: yyu@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Merzari, Elia; Obabko, Aleksandr [Mathematics and Computer Science Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Thomas, Justin [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States)

    2015-12-15

    Highlights: • The proposed models are 400 times less computationally expensive than CFD simulations. • The proposed models show good duct wall temperature agreement with CFD simulations. • The paper provides an efficient tool for coupled radial core expansion calculation. - Abstract: Porous medium models have been established for predicting duct wall temperature of sodium fast reactor rod bundle assembly, which is much less computationally expensive than conventional CFD simulations that explicitly represent the wire-wrap and fuel pin geometry. Three porous medium models are proposed in this paper. Porous medium model 1 takes the whole assembly as one porous medium of uniform characteristics in the conventional approach. Porous medium model 2 distinguishes the pins along the assembly's edge from those in the interior with two distinct regions, each with a distinct porosity, resistance, and volumetric heat source. This accounts for the different fuel-to-coolant volume ratio in the two regions, which is important for predicting the temperature of the assembly's exterior duct wall. In Porous medium model 3, a precise resistance distribution was employed to define the characteristic of the porous medium. The results show that both porous medium model 2 and 3 can capture the average duct wall temperature well. Furthermore, the local duct wall variations due to different sub-channel patterns in bare rod bundles are well captured by porous medium model 3, although the wire effect on the duct wall temperature in wire wrap rod bundle has not been fully reproduced yet.

  11. The estimation of geometry and motion of a surface from image sequences by means of linearisation of a paramatric model

    NARCIS (Netherlands)

    Korsten, Maarten J.; Houkes, Z.

    1990-01-01

    A method is given to estimate the geometry and motion of a moving body surface from image sequences. To this aim a parametric model of the surface is used, in order to reformulate the problem to one of parameter estimation. After linearization of the model standard linear estimation methods can be

  12. Mathematical modeling of water radiolysis in the Syrian MNSR reactor

    International Nuclear Information System (INIS)

    Soukieh, M.

    2009-11-01

    Because it is difficult to measure the concentration of the radiolytic species in reactors under operating conduction, they must be estimated by computer simulation techniques. This study discusses the mathematical modeling of water radiolysis modeling of the MNSR nuclear reactor cooling water. The mathematical model comprising of 13 differential equations describe 55 chemical reactions of radiolytic species e - a q H + , OH - , H, H 2 , OH, HO 2 , O 2 , HO - 2 , O - , O - 2 , O - 3 . The mathematical model have been tested and it shows a good agreement of the computed values in this work with the results cited in references [1,18] in case of only γray irradiation of pure water with dose rate of 1.18x10 19 eV/L s. The neutron fluxes and dose rates at the interface of cladding-water for the different fuel rings in the MNSR core are determined using MCNP-4C code. In addition, the time dependent of the radiolytic specie concentrations were estimated for max. and min. dose rates and at temperature of 20 degree centigrade in the MNSR. The radiolytic specie concentrations reach the steady sate after about 200-400 s. The radiolytic specie concentrations order of H 2 , O 2 , H 2 O 2 were about ppb. Also this study shows the possibility of suppressed the water radiolysis reactions by adding hydrogen to the MNSR reactor cooling water. (author)

  13. Modeling of Hybrid Growth Wastewater Bio-reactor

    International Nuclear Information System (INIS)

    EI Nashaei, S.; Garhyan, P.; Prasad, P.; Abdel Halim, H.S.; Ibrahim, G.

    2004-01-01

    The attached/suspended growth mixed reactors are considered one of the recently tried approaches to improve the performance of the biological treatment by increasing the volume of the accumulated biomass in terms of attached growth as well as suspended growth. Moreover, the domestic WW can be easily mixed with a high strength non-hazardous industrial wastewater and treated together in these bio-reactors if the need arises. Modeling of Hybrid hybrid growth wastewater reactor addresses the need of understanding the rational of such system in order to achieve better design and operation parameters. This paper aims at developing a heterogeneous mathematical model for hybrid growth system considering the effect of diffusion, external mass transfer, and power input to the system in a rational manner. The model will be based on distinguishing between liquid/solid phase (bio-film and bio-floc). This model would be a step ahead to the fine tuning the design of hybrid systems based on the experimental data of a pilot plant to be implemented in near future

  14. Tractable Stochastic Geometry Model for IoT Access in LTE Networks

    KAUST Repository

    Gharbieh, Mohammad; Elsawy, Hesham; Bader, Ahmed; Alouini, Mohamed-Slim

    2017-01-01

    The Internet of Things (IoT) is large-scale by nature. This is not only manifested by the large number of connected devices, but also by the high volumes of traffic that must be accommodated. Cellular networks are indeed a natural candidate for the data tsunami the IoT is expected to generate in conjunction with legacy human-type traffic. However, the random access process for scheduling request represents a major bottleneck to support IoT via LTE cellular networks. Accordingly, this paper develops a mathematical framework to model and study the random access channel (RACH) scalability to accommodate IoT traffic. The developed model is based on stochastic geometry and discrete time Markov chains (DTMC) to account for different access strategies and possible sources of inter-cell and intra-cell interferences. To this end, the developed model is utilized to assess and compare three different access strategies, which incorporate a combination of transmission persistency, back-off, and power ramping. The analysis and the results showcased herewith clearly illustrate the vulnerability of the random access procedure as the IoT intensity grows. Finally, the paper offers insights into effective scenarios for each transmission strategy in terms of IoT intensity and RACH detection thresholds.

  15. Tractable Stochastic Geometry Model for IoT Access in LTE Networks

    KAUST Repository

    Gharbieh, Mohammad

    2017-02-07

    The Internet of Things (IoT) is large-scale by nature. This is not only manifested by the large number of connected devices, but also by the high volumes of traffic that must be accommodated. Cellular networks are indeed a natural candidate for the data tsunami the IoT is expected to generate in conjunction with legacy human-type traffic. However, the random access process for scheduling request represents a major bottleneck to support IoT via LTE cellular networks. Accordingly, this paper develops a mathematical framework to model and study the random access channel (RACH) scalability to accommodate IoT traffic. The developed model is based on stochastic geometry and discrete time Markov chains (DTMC) to account for different access strategies and possible sources of inter-cell and intra-cell interferences. To this end, the developed model is utilized to assess and compare three different access strategies, which incorporate a combination of transmission persistency, back-off, and power ramping. The analysis and the results showcased herewith clearly illustrate the vulnerability of the random access procedure as the IoT intensity grows. Finally, the paper offers insights into effective scenarios for each transmission strategy in terms of IoT intensity and RACH detection thresholds.

  16. Development of neural network models for the prediction of solidification mode, weld bead geometry and sensitisation in austenitic stainless steels

    International Nuclear Information System (INIS)

    Vasudevan, M.; Raj, B.; Prasad Rao, K.

    2005-01-01

    Quantitative models describing the effect of weld composition on the solidification mode, ferrite content and process parameters on the weld bead geometry are necessary in order to design composition of the welding consumable to ensure primary ferritic solidification mode, proper ferrite content and to ensure right choice of process parameters to achieve good bead geometry. A quantitative model on sensitisation behaviour of austenitic stainless steels is also necessary to optimise the composition of the austenitic stainless steel and to limit the strain on the material in order to enhance the resistance to sensitisation. The present paper discuss the development of quantitative models using artificial neural networks to correlate weld metal composition with solidification mode, process parameter with weld bead geometry and time for sensitisation with composition, strain in the material before welding and the temperature of exposure in austenitic stainless steels. (author)

  17. Reactor scale modeling of multi-walled carbon nanotube growth

    International Nuclear Information System (INIS)

    Lombardo, Jeffrey J.; Chiu, Wilson K.S.

    2011-01-01

    As the mechanisms of carbon nanotube (CNT) growth becomes known, it becomes important to understand how to implement this knowledge into reactor scale models to optimize CNT growth. In past work, we have reported fundamental mechanisms and competing deposition regimes that dictate single wall carbon nanotube growth. In this study, we will further explore the growth of carbon nanotubes with multiple walls. A tube flow chemical vapor deposition reactor is simulated using the commercial software package COMSOL, and considered the growth of single- and multi-walled carbon nanotubes. It was found that the limiting reaction processes for multi-walled carbon nanotubes change at different temperatures than the single walled carbon nanotubes and it was shown that the reactions directly governing CNT growth are a limiting process over certain parameters. This work shows that the optimum conditions for CNT growth are dependent on temperature, chemical concentration, and the number of nanotube walls. Optimal reactor conditions have been identified as defined by (1) a critical inlet methane concentration that results in hydrogen abstraction limited versus hydrocarbon adsorption limited reaction kinetic regime, and (2) activation energy of reaction for a given reactor temperature and inlet methane concentration. Successful optimization of a CNT growth processes requires taking all of those variables into account.

  18. Development of an automated core model for nuclear reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

  19. SU-G-TeP1-08: LINAC Head Geometry Modeling for Cyber Knife System

    Energy Technology Data Exchange (ETDEWEB)

    Liang, B; Li, Y; Liu, B; Guo, B; Xu, X; Wei, R; Zhou, F [Beihang University, Beijing, Beijing (China); Xu, S [PLA General Hospital, Beijing, Beijing (China); Wu, Q [Duke University Medical Center, Durham, NC (United States)

    2016-06-15

    Purpose: Knowledge of the LINAC head information is critical for model based dose calculation algorithms. However, the geometries are difficult to measure precisely. The purpose of this study is to develop linac head models for Cyber Knife system (CKS). Methods: For CKS, the commissioning data were measured in water at 800mm SAD. The measured full width at half maximum (FWHM) for each cone was found greater than the nominal value, this was further confirmed by additional film measurement in air. Diameter correction, cone shift and source shift models (DCM, CSM and SSM) are proposed to account for the differences. In DCM, a cone-specific correction is applied. For CSM and SSM, a single shift is applied to the cone or source physical position. All three models were validated with an in-house developed pencil beam dose calculation algorithm, and further evaluated by the collimator scatter factor (Sc) correction. Results: The mean square error (MSE) between nominal diameter and the FWHM derived from commissioning data and in-air measurement are 0.54mm and 0.44mm, with the discrepancy increasing with cone size. Optimal shift for CSM and SSM is found to be 9mm upward and 18mm downward, respectively. The MSE in FWHM is reduced to 0.04mm and 0.14mm for DCM and CSM (SSM). Both DCM and CSM result in the same set of Sc values. Combining all cones at SAD 600–1000mm, the average deviation from 1 in Sc of DCM (CSM) and SSM is 2.6% and 2.2%, and reduced to 0.9% and 0.7% for the cones with diameter greater than 15mm. Conclusion: We developed three geometrical models for CKS. All models can handle the discrepancy between vendor specifications and commissioning data. And SSM has the best performance for Sc correction. The study also validated that a point source can be used in CKS dose calculation algorithms.

  20. Parameter identification in a nonlinear nuclear reactor model using quasilinearization

    International Nuclear Information System (INIS)

    Barreto, J.M.; Martins Neto, A.F.; Tanomaru, N.

    1980-09-01

    Parameter identification in a nonlinear, lumped parameter, nuclear reactor model is carried out using discrete output power measurements during the transient caused by an external reactivity change. In order to minimize the difference between the model and the reactor power responses, the parameter promt neutron generation time and a parameter in fuel temperature reactivity coefficient equation are adjusted using quasilinearization. The influences of the external reactivity disturbance, the number and frequency of measurements and the measurement noise level on the method accuracy and rate of convergence are analysed through simulation. Procedures for the design of the identification experiments are suggested. The method proved to be very effective for low level noise measurements. (Author) [pt

  1. Nuclear reactor fuel rod behavior modelling and current trends

    International Nuclear Information System (INIS)

    Colak, Ue.

    2001-01-01

    Safety assessment of nuclear reactors is carried out by simulating the events to taking place in nuclear reactors by realistic computer codes. Such codes are developed in a way that each event is represented by differential equations derived based on physical laws. Nuclear fuel is an important barrier against radioactive fission gas release. The release of radioactivity to environment is the main concern and this can be avoided by preserving the integrity of fuel rod. Therefore, safety analyses should cover an assessment of fuel rod behavior with certain extent. In this study, common approaches for fuel behavior modeling are discussed. Methods utilized by widely accepted computer codes are reviewed. Shortcomings of these methods are explained. Current research topics to improve code reliability and problems encountered in fuel rod behavior modeling are presented

  2. WWER reactor fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    Stefanova, S.; Chantoin, P.; Kolev, I.

    1994-01-01

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

  3. Scale modeling flow-induced vibrations of reactor components

    International Nuclear Information System (INIS)

    Mulcahy, T.M.

    1982-06-01

    Similitude relationships currently employed in the design of flow-induced vibration scale-model tests of nuclear reactor components are reviewed. Emphasis is given to understanding the origins of the similitude parameters as a basis for discussion of the inevitable distortions which occur in design verification testing of entire reactor systems and in feature testing of individual component designs for the existence of detrimental flow-induced vibration mechanisms. Distortions of similitude parameters made in current test practice are enumerated and selected example tests are described. Also, limitations in the use of specific distortions in model designs are evaluated based on the current understanding of flow-induced vibration mechanisms and structural response

  4. Computer modeling of flow induced in-reactor vibrations

    International Nuclear Information System (INIS)

    Turula, P.; Mulcahy, T.M.

    1977-01-01

    An assessment of the reliability of finite element method computer models, as applied to the computation of flow induced vibration response of components used in nuclear reactors, is presented. The prototype under consideration was the Fast Flux Test Facility reactor being constructed for US-ERDA. Data were available from an extensive test program which used a scale model simulating the hydraulic and structural characteristics of the prototype components, subjected to scaled prototypic flow conditions as well as to laboratory shaker excitations. Corresponding analytical solutions of the component vibration problems were obtained using the NASTRAN computer code. Modal analyses and response analyses were performed. The effect of the surrounding fluid was accounted for. Several possible forcing function definitions were considered. Results indicate that modal computations agree well with experimental data. Response amplitude comparisons are good only under conditions favorable to a clear definition of the structural and hydraulic properties affecting the component motion. 20 refs

  5. WWER reactor fuel performance, modelling and experimental support. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Chantoin, P; Kolev, I [eds.

    1994-12-31

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: (1) WWER Fuel Performance and Economics: Status and Improvement Prospects: (2) WWER Fuel Behaviour Modelling and Experimental Support; (3) Licensing of WWER Fuel and Fuel Analysis Codes; (4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items.

  6. Fully coupled modeling of burnup dependent light water reactor fuel performance using COMSOL Multiphysics

    International Nuclear Information System (INIS)

    Liu Rong; Zhou Wenzhong; Prudil, Andrew

    2015-01-01

    This paper presents the development of a light water reactor fuel performance code, which considers almost all the related physical models, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the equations are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. Comparisons are made for the simulation results between COMSOL and another simulation tool of BISON. The comparisons show the capability of our simulation tool to predict light water UO 2 fuel performances. In our modeling and simulation work, the performance of enhanced thermal conductivity UO 2 -BeO fuel and newly-adopted corrosion resistant SiC cladding material was also studied. UO 2 -BeO high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor would be improved. However, for SiC cladding, although due to its high thermal expansion, the gap closure time is delayed, irradiation induced point defects and defect-clusters in the SiC crystal will dramatically decrease SiC thermal conductivity, and cause significant increase in the fuel temperature. (author)

  7. Criticality assessment for prismatic high temperature reactors by fuel stochastic Monte Carlo modeling

    Energy Technology Data Exchange (ETDEWEB)

    Zakova, Jitka [Department of Nuclear and Reactor Physics, Royal Institute of Technology, KTH, Roslagstullsbacken 21, S-10691 Stockholm (Sweden)], E-mail: jitka.zakova@neutron.kth.se; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, ANL, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: alby@anl.gov

    2008-05-15

    Modeling of prismatic high temperature reactors requires a high precision description due to the triple heterogeneity of the core and also to the random distribution of fuel particles inside the fuel pins. On the latter issue, even with the most advanced Monte Carlo techniques, some approximation often arises while assessing the criticality level: first, a regular lattice of TRISO particles inside the fuel pins and, second, the cutting of TRISO particles by the fuel boundaries. We utilized two of the most accurate Monte Codes: MONK and MCNP, which are both used for licensing nuclear power plants in United Kingdom and in the USA, respectively, to evaluate the influence of the two previous approximations on estimating the criticality level of the Gas Turbine Modular Helium Reactor. The two codes exactly shared the same geometry and nuclear data library, ENDF/B, and only modeled different lattices of TRISO particles inside the fuel pins. More precisely, we investigated the difference between a regular lattice that cuts TRISO particles and a random lattice that axially repeats a region containing over 3000 non-cut particles. We have found that both Monte Carlo codes provide similar excesses of reactivity, provided that they share the same approximations.

  8. Hyperbolic geometry

    CERN Document Server

    Iversen, Birger

    1992-01-01

    Although it arose from purely theoretical considerations of the underlying axioms of geometry, the work of Einstein and Dirac has demonstrated that hyperbolic geometry is a fundamental aspect of modern physics

  9. System model for analysis of the mirror fusion-fission reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Carlson, G.A.

    1977-01-01

    This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters

  10. Estimation of exhaust gas aerodynamic force on the variable geometry turbocharger actuator: 1D flow model approach

    International Nuclear Information System (INIS)

    Ahmed, Fayez Shakil; Laghrouche, Salah; Mehmood, Adeel; El Bagdouri, Mohammed

    2014-01-01

    Highlights: • Estimation of aerodynamic force on variable turbine geometry vanes and actuator. • Method based on exhaust gas flow modeling. • Simulation tool for integration of aerodynamic force in automotive simulation software. - Abstract: This paper provides a reliable tool for simulating the effects of exhaust gas flow through the variable turbine geometry section of a variable geometry turbocharger (VGT), on flow control mechanism. The main objective is to estimate the resistive aerodynamic force exerted by the flow upon the variable geometry vanes and the controlling actuator, in order to improve the control of vane angles. To achieve this, a 1D model of the exhaust flow is developed using Navier–Stokes equations. As the flow characteristics depend upon the volute geometry, impeller blade force and the existing viscous friction, the related source terms (losses) are also included in the model. In order to guarantee stability, an implicit numerical solver has been developed for the resolution of the Navier–Stokes problem. The resulting simulation tool has been validated through comparison with experimentally obtained values of turbine inlet pressure and the aerodynamic force as measured at the actuator shaft. The simulator shows good compliance with experimental results

  11. Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

  12. Designing visual displays and system models for safe reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  13. Designing visual displays and system models for safe reactor operations

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.

    1995-01-01

    The material presented in this paper is based on two studies involving the design of visual displays and the user's prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator's perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors

  14. Numerical Modelling of Wood Gasification in Thermal Plasma Reactor

    Czech Academy of Sciences Publication Activity Database

    Hirka, Ivan; Živný, Oldřich; Hrabovský, Milan

    2017-01-01

    Roč. 37, č. 4 (2017), s. 947-965 ISSN 0272-4324 Institutional support: RVO:61389021 Keywords : Plasma modelling * CFD * Thermal plasma reactor * Biomass * Gasification * Syngas Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.355, year: 2016 https://link.springer.com/article/10.1007/s11090-017-9812-z

  15. Fast reactor fuel pin behaviour modelling in the UK

    International Nuclear Information System (INIS)

    Matthews, J.R.; Hughes, H.

    1979-01-01

    Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

  16. Fast reactor fuel pin behaviour modelling in the UK

    Energy Technology Data Exchange (ETDEWEB)

    Matthews, J R [UKAEA, Harwell, Didcot, Oxon (United Kingdom); Hughes, H [Springfields Nuclear Power Development Laboratories, Springfields, Salwick, Preston (United Kingdom)

    1979-12-01

    Two fuel behaviour codes have been applied extensively to fast reactor problems; SLEUTH developed at Sprlngfields Nuclear Laboratory and FRUMP at A.E.R.E. Harwell. The SLEUTH fuel pin endurance code was originally developed to define a programme of power cycling and power ramp experiments In Advanced Gas Cooled Reactors (AGRs) where, because of the very soft cladding, pellet clad interaction is severe. The code was required to define accelerated test conditions to generalise from the observed endurance to that under other power histories and to select for investigation the most significant design, material and operational variables. The weak clad and low coolant pressure combine to make fission gas swelling a major contributor to clad deformation while the high clad ductility renders the distribution of strain readily observable. This has led to a detailed study of strain concentrations using the SEER code. SLEUTH and SEER have subsequently been used to specify power cycling and power ramp 112 experiments in water cooled, fast and materials testing reactors with the aim of developing a unified quantitative model of pellet-clad interaction whatever the reactor system. The FRUMP fuel behaviour code was developed specifically for the interpretation of fast reactor fuel pin behaviour. Experience with earlier models was valuable In its development. Originally the model was developed to describe behaviour during normal operation, but subsequently the code has been used extensively in the field of accident studies. Much of the effort in FRUMP development has been devoted to the production of physical models of the various effects of irradiation and the temperature gradients on the structure of the fuel and clad. Each process is modelled as well as is permitted by current knowledge and the limitations of computing costs. Each sub-model has a form which reflects the underlying mechanisms, where quantities are unknown values are assigned semi-empirically, i.e. coefficients

  17. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  18. Twistor geometry

    NARCIS (Netherlands)

    van den Broek, P.M.

    1984-01-01

    The aim of this paper is to give a detailed exposition of the relation between the geometry of twistor space and the geometry of Minkowski space. The paper has a didactical purpose; no use has been made of differential geometry and cohomology.

  19. A Stochastic Geometry Model for Multi-hop Highway Vehicular Communication

    KAUST Repository

    Farooq, Muhammad Junaid

    2015-11-19

    Carrier sense multiple access (CSMA) protocol is standardized for vehicular communication to ensure a distributed and efficient communication between vehicles. However, several vehicular applications require efficient multi-hop information dissemination. This paper exploits stochastic geometry to develop a tractable and accurate modeling framework to characterize the multi-hop transmissions for vehicular networks in a multi-lane highway setup. In particular, we study the tradeoffs between per-hop packet forward progress, per-hop transmission success probability, and spatial frequency reuse (SFR) efficiency imposed by different packet forwarding schemes, namely, most forward with fixed radius (MFR), the nearest with forward progress (NFP), and the random with forward progress (RFP). We also define a new performance metric, denoted as the aggregate packet progress (APP), which is a dimensionless quantity that captures the aforementioned tradeoffs. To this end, the developed model reveals the interplay between the spectrum sensing threshold (th) of the CSMA protocol and the packet forwarding scheme. Our results show that, in contrary to ALOHA networks which always favor NFP, MFR may achieve the highest APP in CSMA networks if th is properly chosen.

  20. Probing emergent geometry through phase transitions in free vector and matrix models

    Energy Technology Data Exchange (ETDEWEB)

    Amado, Irene; Sundborg, Bo [The Oskar Klein Centre for Cosmoparticle Physics, Department of Physics, Stockholm University,AlbaNova, 106 91 Stockholm (Sweden); Thorlacius, Larus [The Oskar Klein Centre for Cosmoparticle Physics, Department of Physics, Stockholm University,AlbaNova, 106 91 Stockholm (Sweden); Science Institute, University of Iceland, Dunhaga 3, 107 Reykjavik (Iceland); Wintergerst, Nico [The Oskar Klein Centre for Cosmoparticle Physics, Department of Physics, Stockholm University,AlbaNova, 106 91 Stockholm (Sweden)

    2017-02-01

    Boundary correlation functions provide insight into the emergence of an effective geometry in higher spin gravity duals of O(N) or U(N) symmetric field theories. On a compact manifold, the singlet constraint leads to nontrivial dynamics at finite temperature and large N phase transitions even at vanishing ’t Hooft coupling. At low temperature, the leading behavior of boundary two-point functions is consistent with propagation through a bulk thermal anti de Sitter space. Above the phase transition, the two-point function shows significant departure from thermal AdS space and the emergence of localized black hole like objects in the bulk. In adjoint models, these objects appear at length scales of order of the AdS radius, consistent with a Hawking-Page transition, but in vector models they are parametrically larger than the AdS scale. In low dimensions, we find another crossover at large distances beyond which the correlation function again takes a thermal AdS form, albeit with a temperature dependent normalization factor.

  1. Thrust initiation and its control on tectonic wedge geometry: An insight from physical and numerical models

    Science.gov (United States)

    Bose, Santanu; Mandal, Nibir; Saha, Puspendu; Sarkar, Shamik; Lithgow-Bertelloni, Carolina

    2014-10-01

    We performed a series of sandbox experiments to investigate the initiation of thrust ramping in tectonic wedges on a mechanically continuous basal decollement. The experiments show that the decollement slope (β) is the key factor in controlling the location of thrust initiation with respect to the backstop (i.e. tectonic suture line). For β = 0, the ramping begins right at the backstop, followed by sequential thrusting in the frontal direction, leading to a typical mono-vergent wedge. In contrast, the ramp initiates away from the backstop as β > 0. Under this boundary condition an event of sequential back thrusting takes place prior to the onset of frontal thrust progression. These two-coupled processes eventually give rise to a bi-vergent geometry of the thrust wedge. Using the Drucker-Prager failure criterion in finite element (FE) models, we show the location of stress intensification to render a mechanical basis for the thrust initiation away from the backstop if β > 0. Our physical and FE model results explain why the Main Central Thrust (MCT) is located far away from the Indo-Tibetan plate contact (ITSZ) in the Himalayan fold-and-thrust belts.

  2. Effect of inlet geometry on macrosegregation during the direct chill casting of 7050 alloy billets: experiments and computer modelling

    International Nuclear Information System (INIS)

    Zhang, L; Miroux, A; Subroto, T; Katgerman, L; Eskin, D G

    2012-01-01

    Controlling macrosegregation is one of the major challenges in direct-chill (DC) casting of aluminium alloys. In this paper, the effect of the inlet geometry (which influences the melt distribution) on macrosegregation during the DC casting of 7050 alloy billets was studied experimentally and by using 2D computer modelling. The ALSIM model was used to determine the temperature and flow patterns during DC casting. The results from the computer simulations show that the sump profiles and flow patterns in the billet are strongly influenced by the melt flow distribution determined by the inlet geometry. These observations were correlated to the actual macrosegregation patterns found in the as-cast billets produced by having two different inlet geometries. The macrosegregation analysis presented here may assist in determining the critical parameters to consider for improving the casting of 7XXX aluminium alloys.

  3. Effect of inlet geometry on macrosegregation during the direct chill casting of 7050 alloy billets: experiments and computer modelling

    Science.gov (United States)

    Zhang, L.; Eskin, D. G.; Miroux, A.; Subroto, T.; Katgerman, L.

    2012-07-01

    Controlling macrosegregation is one of the major challenges in direct-chill (DC) casting of aluminium alloys. In this paper, the effect of the inlet geometry (which influences the melt distribution) on macrosegregation during the DC casting of 7050 alloy billets was studied experimentally and by using 2D computer modelling. The ALSIM model was used to determine the temperature and flow patterns during DC casting. The results from the computer simulations show that the sump profiles and flow patterns in the billet are strongly influenced by the melt flow distribution determined by the inlet geometry. These observations were correlated to the actual macrosegregation patterns found in the as-cast billets produced by having two different inlet geometries. The macrosegregation analysis presented here may assist in determining the critical parameters to consider for improving the casting of 7XXX aluminium alloys.

  4. Analytical modelling of hydrogen transport in reactor containments

    International Nuclear Information System (INIS)

    Manno, V.P.

    1983-09-01

    A versatile computational model of hydrogen transport in nuclear plant containment buildings is developed. The background and significance of hydrogen-related nuclear safety issues are discussed. A computer program is constructed that embodies the analytical models. The thermofluid dynamic formulation spans a wide applicability range from rapid two-phase blowdown transients to slow incompressible hydrogen injection. Detailed ancillary models of molecular and turbulent diffusion, mixture transport properties, multi-phase multicomponent thermodynamics and heat sink modelling are addressed. The numerical solution of the continuum equations emphasizes both accuracy and efficiency in the employment of relatively coarse discretization and long time steps. Reducing undesirable numerical diffusion is addressed. Problem geometry options include lumped parameter zones, one dimensional meshs, two dimensional Cartesian or axisymmetric coordinate systems and three dimensional Cartesian or cylindrical regions. An efficient lumped nodal model is included for simulation of events in which spatial resolution is not significant. Several validation calculations are reported

  5. Sensitivity of subject-specific models to errors in musculo-skeletal geometry.

    Science.gov (United States)

    Carbone, V; van der Krogt, M M; Koopman, H F J M; Verdonschot, N

    2012-09-21

    Subject-specific musculo-skeletal models of the lower extremity are an important tool for investigating various biomechanical problems, for instance the results of surgery such as joint replacements and tendon transfers. The aim of this study was to assess the potential effects of errors in musculo-skeletal geometry on subject-specific model results. We performed an extensive sensitivity analysis to quantify the effect of the perturbation of origin, insertion and via points of each of the 56 musculo-tendon parts contained in the model. We used two metrics, namely a Local Sensitivity Index (LSI) and an Overall Sensitivity Index (OSI), to distinguish the effect of the perturbation on the predicted force produced by only the perturbed musculo-tendon parts and by all the remaining musculo-tendon parts, respectively, during a simulated gait cycle. Results indicated that, for each musculo-tendon part, only two points show a significant sensitivity: its origin, or pseudo-origin, point and its insertion, or pseudo-insertion, point. The most sensitive points belong to those musculo-tendon parts that act as prime movers in the walking movement (insertion point of the Achilles Tendon: LSI=15.56%, OSI=7.17%; origin points of the Rectus Femoris: LSI=13.89%, OSI=2.44%) and as hip stabilizers (insertion points of the Gluteus Medius Anterior: LSI=17.92%, OSI=2.79%; insertion point of the Gluteus Minimus: LSI=21.71%, OSI=2.41%). The proposed priority list provides quantitative information to improve the predictive accuracy of subject-specific musculo-skeletal models. Copyright © 2012 Elsevier Ltd. All rights reserved.

  6. Geometry Modeling and Adaptive Control of Air-Breathing Hypersonic Vehicles

    Science.gov (United States)

    Vick, Tyler Joseph

    Air-breathing hypersonic vehicles have the potential to provide global reach and affordable access to space. Recent technological advancements have made scramjet-powered flight achievable, as evidenced by the successes of the X-43A and X-51A flight test programs over the last decade. Air-breathing hypersonic vehicles present unique modeling and control challenges in large part due to the fact that scramjet propulsion systems are highly integrated into the airframe, resulting in strongly coupled and often unstable dynamics. Additionally, the extreme flight conditions and inability to test fully integrated vehicle systems larger than X-51 before flight leads to inherent uncertainty in hypersonic flight. This thesis presents a means to design vehicle geometries, simulate vehicle dynamics, and develop and analyze control systems for hypersonic vehicles. First, a software tool for generating three-dimensional watertight vehicle surface meshes from simple design parameters is developed. These surface meshes are compatible with existing vehicle analysis tools, with which databases of aerodynamic and propulsive forces and moments can be constructed. A six-degree-of-freedom nonlinear dynamics simulation model which incorporates this data is presented. Inner-loop longitudinal and lateral control systems are designed and analyzed utilizing the simulation model. The first is an output feedback proportional-integral linear controller designed using linear quadratic regulator techniques. The second is a model reference adaptive controller (MRAC) which augments this baseline linear controller with an adaptive element. The performance and robustness of each controller are analyzed through simulated time responses to angle-of-attack and bank angle commands, while various uncertainties are introduced. The MRAC architecture enables the controller to adapt in a nonlinear fashion to deviations from the desired response, allowing for improved tracking performance, stability, and

  7. Comparison and analysis of 1D/2D/3D neutronics modeling for a fusion reactor

    International Nuclear Information System (INIS)

    Li, J.; Zeng, Q.; Chen, M.; Jiang, J.; Wu, Y.

    2007-01-01

    During the course of analyzing the characteristics for fusion reactors, the refined calculations are needed to confirm that the nuclear design requirements are met. Since the long computational time is consumed, the refined three-dimensional (3D) representation has been used primarily for establishing the baseline reference values, analyzing problems which cannot be reduced by symmetry considerations to lower dimensions, or where a high level of accuracy is desired locally. The two-dimensional (2D) or one-dimensional (1D) description leads itself readily to resolve many problems, such as the studies for the material fraction optimization, or for the blanket size optimization. The purpose of this paper is to find out the differences among different geometric descriptions, which can guide the way to approximate and simplify the computational model. The fusion power reactor named FDS-II was designed as an advanced fusion power reactor to demonstrate and validate the commercialization of fusion power by Institute of Plasma Physics, Chinese Academy of Science. In this contribution, the dual-cooled lithium lead (DLL) blanket of FDS-II was used as a reference for neutronics comparisons and analyses. The geometric descriptions include 1D concentric sphere model, 1D, 2D and 3D cylinder models. The home-developed multi-functional neutronics analysis code system VisualBUS, the Monte Carlo transport code MCNP and nuclear data library HENDL have been used for these analyses. The neutron wall loading distribution, tritium breeding ratio (TBR) and nuclear heat were calculated to evaluate the nuclear performance. The 3D calculation has been used as a comparison reference because it has the least errors in the treatment of geometry. It is suggested that the value of TBR calculated by the 1D approach should be greater than 1.3 to satisfy the practical need of tritium self-sufficiency. The distribution of nuclear heat based on the 2D and 3D models were similar since they all consider

  8. Tailoring weld geometry during keyhole mode laser welding using a genetic algorithm and a heat transfer model

    International Nuclear Information System (INIS)

    Rai, R; DebRoy, T

    2006-01-01

    Tailoring of weld attributes based on scientific principles remains an important goal in welding research. The current generation of unidirectional laser keyhole models cannot determine sets of welding variables that can lead to a particular weld attribute such as specific weld geometry. Here we show how a computational heat transfer model of keyhole mode laser welding can be restructured for systematic tailoring of weld attributes based on scientific principles. Furthermore, the model presented here can calculate multiple sets of laser welding variables, i.e. laser power, welding speed and beam defocus, with each set leading to the same weld pool geometry. Many sets of welding variables were obtained via a global search using a real number-based genetic algorithm, which was combined with a numerical heat transfer model of keyhole laser welding. The reliability of the numerical heat transfer calculations was significantly improved by optimizing values of the uncertain input parameters from a limited volume of experimental data. The computational procedure was applied to the keyhole mode laser welding of the 5182 Al-Mg alloy to calculate various sets of welding variables to achieve a specified weld geometry. The calculated welding parameter sets showed wide variations of the values of welding parameters, but each set resulted in a similar fusion zone geometry. The effectiveness of the computational procedure was examined by comparing the computed weld geometry for each set of welding parameters with the corresponding experimental geometry. The results provide hope that systematic tailoring of weld attributes via multiple pathways, each representing alternative welding parameter sets, is attainable based on scientific principles

  9. Photocatalytic mineralization of commercial herbicides in a pilot-scale solar CPC reactor: photoreactor modeling and reaction kinetics constants independent of radiation field.

    Science.gov (United States)

    Colina-Márquez, Jose; Machuca-Martínez, Fiderman; Li Puma, Gianluca

    2009-12-01

    The six-flux absorption-scattering model (SFM) of the radiation field in the photoreactor, combined with reaction kinetics and fluid-dynamic models, has proved to be suitable to describe the degradation of water pollutants in heterogeneous photocatalytic reactors, combining simplicity and accuracy. In this study, the above approach was extended to model the photocatalytic mineralization of a commercial herbicides mixture (2,4-D, diuron, and ametryne used in Colombian sugar cane crops) in a solar, pilot-scale, compound parabolic collector (CPC) photoreactor using a slurry suspension of TiO(2). The ray-tracing technique was used jointly with the SFM to determine the direction of both the direct and diffuse solar photon fluxes and the spatial profile of the local volumetric rate of photon absorption (LVRPA) in the CPC reactor. Herbicides mineralization kinetics with explicit photon absorption effects were utilized to remove the dependence of the observed rate constants from the reactor geometry and radiation field in the photoreactor. The results showed that the overall model fitted the experimental data of herbicides mineralization in the solar CPC reactor satisfactorily for both cloudy and sunny days. Using the above approach kinetic parameters independent of the radiation field in the reactor can be estimated directly from the results of experiments carried out in a solar CPC reactor. The SFM combined with reaction kinetics and fluid-dynamic models proved to be a simple, but reliable model, for solar photocatalytic applications.

  10. A reduced fidelity model for the rotary chemical looping combustion reactor

    KAUST Repository

    Iloeje, Chukwunwike O.; Zhao, Zhenlong; Ghoniem, Ahmed F.

    2017-01-01

    The rotary chemical looping combustion reactor has great potential for efficient integration with CO capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate

  11. Mathematical modeling of a three-phase trickle bed reactor

    Directory of Open Access Journals (Sweden)

    J. D. Silva

    2012-09-01

    Full Text Available The transient behavior in a three-phase trickle bed reactor system (N2/H2O-KCl/activated carbon, 298 K, 1.01 bar was evaluated using a dynamic tracer method. The system operated with liquid and gas phases flowing downward with constant gas flow Q G = 2.50 x 10-6 m³ s-1 and the liquid phase flow (Q L varying in the range from 4.25x10-6 m³ s-1 to 0.50x10-6 m³ s-1. The evolution of the KCl concentration in the aqueous liquid phase was measured at the outlet of the reactor in response to the concentration increase at reactor inlet. A mathematical model was formulated and the solutions of the equations fitted to the measured tracer concentrations. The order of magnitude of the axial dispersion, liquid-solid mass transfer and partial wetting efficiency coefficients were estimated based on a numerical optimization procedure where the initial values of these coefficients, obtained by empirical correlations, were modified by comparing experimental and calculated tracer concentrations. The final optimized values of the coefficients were calculated by the minimization of a quadratic objective function. Three correlations were proposed to estimate the parameters values under the conditions employed. By comparing experimental and predicted tracer concentration step evolutions under different operating conditions the model was validated.

  12. Plasma Reactors and Plasma Thrusters Modeling by Ar Complete Global Models

    Directory of Open Access Journals (Sweden)

    Chloe Berenguer

    2012-01-01

    Full Text Available A complete global model for argon was developed and adapted to plasma reactor and plasma thruster modeling. It takes into consideration ground level and excited Ar and Ar+ species and the reactor and thruster form factors. The electronic temperature, the species densities, and the ionization percentage, depending mainly on the pressure and the absorbed power, have been obtained and commented for various physical conditions.

  13. Thermal-hydraulic modeling of porous bed reactors

    International Nuclear Information System (INIS)

    Araj, K.J.; Nourbakhsh, H.P.

    1987-01-01

    Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

  14. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current US innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery

  15. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current U.S. innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery. (orig.)

  16. High Flux Isotope Reactor system RELAP5 input model

    International Nuclear Information System (INIS)

    Morris, D.G.; Wendel, M.W.

    1993-01-01

    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model

  17. Assessment for the Applicability of Effective Thermal Conductivity Models on the Prismatic Fuel Assembly of Very High Temperature Reactor

    International Nuclear Information System (INIS)

    Shin, Dong-ho; Cho, Hyoung-kyu; Tak, Nam-il; Park, Goon-cherl

    2014-01-01

    A prismatic gas-cooled reactor is promising reactor type in the Nuclear Hydrogen Development and Demonstration (NHDD) project which was launched at KAERI (Korea Atomic Energy Research Institute). One of the most favorable characteristics of a prismatic gas-cooled reactor is its inherent and passive safety. As one of its inherent safety features, the heat flows through the prismatic core radially during the High Pressure Conduction Cooling (HPCC) or Low Pressure Conduction Cooling (LPCC) event and the radial heat transfer cools down the reactor core passively under such conditions. To verify the inherent safety of its design, the GAMMA+ code that is used to analyze VHTR thermo-fluid transients has been developed by KAERI. The code adopts effective thermal conductivity (ETC) model to analyze radial heat transfer in the core as a lumped parameter model. It is because the fuel block has complex geometry with large number of coolant holes and fuel compacts and the detail heat transfer calculations on that geometry needs excessive computation resources. GAMMA+ is adopting the Maxwell-based ETC model, however, there are several ETC models that could be applied to the GAMMA+ code. In this study, several ETC models will be introduced. They will be compared to CFD calculations which have similar condition with the fuel block. And then the most appropriate ETC model will be suggested for calculating the ETC of the fuel block. For the CFD calculation, unit cell tests with simple geometries were conducted. With unit cell test, the applicability of the ETC models were investigated. And proper ETC models were used to calculate the ETC of the fuel block and the results were compared to that of CFD calculation on the fuel block. In this study, the ETC models are introduced and the applicability of the ETC models to VHTR fuel block was investigated. The results of the ETC models were compared to those of CFD calculation. The CFD calculations were conducted for square graphite block

  18. Reactor pressure vessel embrittlement: Insights from neural network modelling

    Science.gov (United States)

    Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

    2018-04-01

    Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

  19. Capability to model reactor regulating system in RFSP

    Energy Technology Data Exchange (ETDEWEB)

    Chow, H C; Rouben, B; Younis, M H; Jenkins, D A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Baudouin, A [Hydro-Quebec, Montreal, PQ (Canada); Thompson, P D [New Brunswick Electric Power Commission, Point Lepreau, NB (Canada). Point Lepreau Generating Station

    1996-12-31

    The Reactor Regulating System package extracted from SMOKIN-G2 was linked within RFSP to the spatial kinetics calculation. The objective is to use this new capability in safety analysis to model the actions of RRS in hypothetical events such as in-core LOCA or moderator drain scenarios. This paper describes the RRS modelling in RFSP and its coupling to the neutronics calculations, verification of the RRS control routine functions, sample applications and comparisons to SMOKIN-G2 results for the same transient simulations. (author). 7 refs., 6 figs.

  20. Vibration behavior of PWR reactor internals Model experiments and analysis

    International Nuclear Information System (INIS)

    Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

    1975-01-01

    In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

  1. Once-through CANDU reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    Croff, A.G.; Bjerke, M.A.

    1980-11-01

    Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

  2. Experimental and modeling analysis of fast ionization wave discharge propagation in a rectangular geometry

    International Nuclear Information System (INIS)

    Takashima, Keisuke; Adamovich, Igor V.; Xiong Zhongmin; Kushner, Mark J.; Starikovskaia, Svetlana; Czarnetzki, Uwe; Luggenhoelscher, Dirk

    2011-01-01

    Fast ionization wave (FIW), nanosecond pulse discharge propagation in nitrogen and helium in a rectangular geometry channel/waveguide is studied experimentally using calibrated capacitive probe measurements. The repetitive nanosecond pulse discharge in the channel was generated using a custom designed pulsed plasma generator (peak voltage 10-40 kV, pulse duration 30-100 ns, and voltage rise time ∼1 kV/ns), generating a sequence of alternating polarity high-voltage pulses at a pulse repetition rate of 20 Hz. Both negative polarity and positive polarity ionization waves have been studied. Ionization wave speed, as well as time-resolved potential distributions and axial electric field distributions in the propagating discharge are inferred from the capacitive probe data. ICCD images show that at the present conditions the FIW discharge in helium is diffuse and volume-filling, while in nitrogen the discharge propagates along the walls of the channel. FIW discharge propagation has been analyzed numerically using quasi-one-dimensional and two-dimensional kinetic models in a hydrodynamic (drift-diffusion), local ionization approximation. The wave speed and the electric field distribution in the wave front predicted by the model are in good agreement with the experimental results. A self-similar analytic solution of the fast ionization wave propagation equations has also been obtained. The analytic model of the FIW discharge predicts key ionization wave parameters, such as wave speed, peak electric field in the front, potential difference across the wave, and electron density as functions of the waveform on the high voltage electrode, in good agreement with the numerical calculations and the experimental results.

  3. Description of the RA-3 research reactor as a model facility

    International Nuclear Information System (INIS)

    Vicens, Hugo E.; Quintana, Jorge A.

    2001-01-01

    The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

  4. Modeling and Analysis of Cellular Networks using Stochastic Geometry: A Tutorial

    KAUST Repository

    Elsawy, Hesham; Salem, Ahmed Sultan; Alouini, Mohamed-Slim; Win, Moe Z.

    2016-01-01

    This paper presents a tutorial on stochastic geometry (SG) based analysis for cellular networks. This tutorial is distinguished by its depth with respect to wireless communication details and its focus on cellular networks. The paper starts by modeling and analyzing the baseband interference in a baseline single-tier downlink cellular network with single antenna base stations and universal frequency reuse. Then, it characterizes signal-to-interference-plus-noise-ratio (SINR) and its related performance metrics. In particular, a unified approach to conduct error probability, outage probability, and transmission rate analysis is presented. Although the main focus of the paper is on cellular networks, the presented unified approach applies for other types of wireless networks that impose interference protection around receivers. The paper then extends the unified approach to capture cellular network characteristics (e.g., frequency reuse, multiple antenna, power control, etc.). It also presents numerical examples associated with demonstrations and discussions. To this end, the paper highlights the state-of-the- art research and points out future research directions.

  5. Fractional cable equation for general geometry: A model of axons with swellings and anomalous diffusion

    Science.gov (United States)

    López-Sánchez, Erick J.; Romero, Juan M.; Yépez-Martínez, Huitzilin

    2017-09-01

    Different experimental studies have reported anomalous diffusion in brain tissues and notably this anomalous diffusion is expressed through fractional derivatives. Axons are important to understand neurodegenerative diseases such as multiple sclerosis, Alzheimer's disease, and Parkinson's disease. Indeed, abnormal accumulation of proteins and organelles in axons is a hallmark of these diseases. The diffusion in the axons can become anomalous as a result of this abnormality. In this case the voltage propagation in axons is affected. Another hallmark of different neurodegenerative diseases is given by discrete swellings along the axon. In order to model the voltage propagation in axons with anomalous diffusion and swellings, in this paper we propose a fractional cable equation for a general geometry. This generalized equation depends on fractional parameters and geometric quantities such as the curvature and torsion of the cable. For a cable with a constant radius we show that the voltage decreases when the fractional effect increases. In cables with swellings we find that when the fractional effect or the swelling radius increases, the voltage decreases. Similar behavior is obtained when the number of swellings and the fractional effect increase. Moreover, we find that when the radius swelling (or the number of swellings) and the fractional effect increase at the same time, the voltage dramatically decreases.

  6. Modeling and Analysis of Cellular Networks using Stochastic Geometry: A Tutorial

    KAUST Repository

    Elsawy, Hesham

    2016-11-03

    This paper presents a tutorial on stochastic geometry (SG) based analysis for cellular networks. This tutorial is distinguished by its depth with respect to wireless communication details and its focus on cellular networks. The paper starts by modeling and analyzing the baseband interference in a baseline single-tier downlink cellular network with single antenna base stations and universal frequency reuse. Then, it characterizes signal-to-interference-plus-noise-ratio (SINR) and its related performance metrics. In particular, a unified approach to conduct error probability, outage probability, and transmission rate analysis is presented. Although the main focus of the paper is on cellular networks, the presented unified approach applies for other types of wireless networks that impose interference protection around receivers. The paper then extends the unified approach to capture cellular network characteristics (e.g., frequency reuse, multiple antenna, power control, etc.). It also presents numerical examples associated with demonstrations and discussions. To this end, the paper highlights the state-of-the- art research and points out future research directions.

  7. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  8. Oklo reactors and implications for nuclear science

    OpenAIRE

    Davis, E. D.; Gould, C. R.; Sharapov, E. I.

    2014-01-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlie...

  9. Real-time advanced nuclear reactor core model

    International Nuclear Information System (INIS)

    Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

    1990-01-01

    The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

  10. Simulation of styrene polymerization reactors: kinetic and thermodynamic modeling

    Directory of Open Access Journals (Sweden)

    A. S. Almeida

    2008-06-01

    Full Text Available A mathematical model for the free radical polymerization of styrene is developed to predict the steady-state and dynamic behavior of a continuous process. Special emphasis is given for the kinetic and thermodynamic models, where the most sensitive parameters were estimated using data from an industrial plant. The thermodynamic model is based on a cubic equation of state and a mixing rule applied to the low-pressure vapor-liquid equilibrium of polymeric solutions, suitable for modeling the auto-refrigerated polymerization reactors, which use the vaporization rate to remove the reaction heat from the exothermic reactions. The simulation results show the high predictive capability of the proposed model when compared with plant data for conversion, average molecular weights, polydispersity, melt flow index, and thermal properties for different polymer grades.

  11. A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E.A. [Department of Mechanical Engineering, University of Texas, Austin, TX (United States); Deinert, M.R. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)]. E-mail: mrd6@cornell.edu; Cady, K.B. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)

    2006-10-15

    The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

  12. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  13. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  14. A simplified hydrodynamic model of hydrogen flame propagation in reactor vessels

    International Nuclear Information System (INIS)

    Baer, M.; Ratzel, A.

    1983-01-01

    A hydrodynamic model for hydrogen flame propagation in reactor geometries is presented. This model is consistent with the theory of slow combustion in which the gasdynamic field equations are treated in the limit of small Mach numbers. To the lowest order, pressure is spatially uniform. The flame is treated as a density and entropy discontinuity which propagates at prescribed burning velocities, corresponding to laminar or turbulent flames. Radiation cooling of the burned combustion gases and possible preheating of the unburned gases during propagation of the flame is included using a molecular gas-band thermal radiation model. Application of this model has been developed for 1-D variable area flame propagation. Multidimensional effects induced by hydrodynamics and buoyancy are introduced as a correction to the burn velocity (which reflects a modification of planar flame surface to a distorted surface) using experimentally measured pressure-rise time data for hydrogen/air deflagrations in cylindrical vessels. This semianalytical model of flame propagation reduces to a set of ordinary differential equations which describes the temporal variations of vessel pressure, burned volume and gas entropy. The thermodynamic state of the burned gas immediately following the flame is determined using an isobaric Hugoniot relationship. At other locations the burned gas thermodynamic states are determined using a Lagrangian particle tracking method. Results of a computer code using the method are presented

  15. Parameter estimation for LLDPE gas-phase reactor models

    Directory of Open Access Journals (Sweden)

    G. A. Neumann

    2007-06-01

    Full Text Available Product development and advanced control applications require models with good predictive capability. However, in some cases it is not possible to obtain good quality phenomenological models due to the lack of data or the presence of important unmeasured effects. The use of empirical models requires less investment in modeling, but implies the need for larger amounts of experimental data to generate models with good predictive capability. In this work, nonlinear phenomenological and empirical models were compared with respect to their capability to predict the melt index and polymer yield of a low-density polyethylene production process consisting of two fluidized bed reactors connected in series. To adjust the phenomenological model, the optimization algorithms based on the flexible polyhedron method of Nelder and Mead showed the best efficiency. To adjust the empirical model, the PLS model was more appropriate for polymer yield, and the melt index needed more nonlinearity like the QPLS models. In the comparison between these two types of models better results were obtained for the empirical models.

  16. Photocatalytic Oxidation of Azo Dyes and Oxalic Acid in Batch Reactors and CSTR: Introduction of Photon Absorption by Dyes to Kinetic Models

    Directory of Open Access Journals (Sweden)

    I. Grčić

    2018-04-01

    Full Text Available The possibilities of treating industrial effluents and water purification by advanced oxidation processes have been extensively studied; photocatalysis has emerged as a feasible alternative solution. In order to apply the photocatalytic treatment on a larger scale, relevant modeling approaches are necessary. The scope of this work was to investigate the applicability of recently published kinetic models in different reactor systems (batch and CSTR under UVA or UVC irradiation and in combination with two types of TiO2 catalyst, AEROXIDE® P25 and PC-500 for degradation of azo dyes (C.I. Reactive Violet 2, and C.I. Mordant Yellow 10, oxalic acid and their mixtures. The influences of reactor geometry and irradiation intensities on pollutant oxidation efficiency were examined. The effect of photon absorption by dyes in water matrix was thoroughly studied. Relevant kinetic models were introduced to the mass balance for particular reactor system. Resulting models were sufficient for description of pollutant degradation in batch reactors and CSTR. Experimental results showed 1.15 times higher mineralization extents achieved after 7 cycles in CSTR than in batch photoreactor of similar geometry within the equivalent time-span. The application of CSTR in-series could simplify the photocatalytic water treatment on a larger scale.

  17. Low-Rynolds number k-ε turbulence model for calculation of fast-reactor-channel flows

    International Nuclear Information System (INIS)

    Mikhin, V.I.

    2000-01-01

    For calculating the turbulent flows in the complex geometry channels typical for the nuclear reactor installation elements the low-Reynolds-number k-ε turbulence model with the model functions not containing the spatial coordinate like y + is proposed. Such spatial coordinate is usually used for modeling the turbulence near the wall correctly. The model completed on the developed flow of the non-viscous incompressible liquid in the plane channel correctly describes the transition from the laminar regime to the turbulent one. The calculated skin friction coefficients obey the well-known Dean and Zarbi - Reynolds laws. The mean velocity distributions are close to that obtained from the empirical three-layer Karman model. (author)

  18. Molecular geometry

    CERN Document Server

    Rodger, Alison

    1995-01-01

    Molecular Geometry discusses topics relevant to the arrangement of atoms. The book is comprised of seven chapters that tackle several areas of molecular geometry. Chapter 1 reviews the definition and determination of molecular geometry, while Chapter 2 discusses the unified view of stereochemistry and stereochemical changes. Chapter 3 covers the geometry of molecules of second row atoms, and Chapter 4 deals with the main group elements beyond the second row. The book also talks about the complexes of transition metals and f-block elements, and then covers the organometallic compounds and trans

  19. One-dimensional reactor kinetics model for RETRAN

    International Nuclear Information System (INIS)

    Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.

    1981-01-01

    Previous versions of RETRAN have had only a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. The principal assumption in deriving the point kinetics model is that the neutron flux may be separated into a time-dependent amplitude funtion and a time-independent shape function. Certain types of transients cannot be correctly analyzed under this assumption, since proper definitions for core average quantities such as reactivity or lifetime include the inner product of the adjoint flux with the perturbed flux. A one-dimensional neutronics model has been included in a preliminary version of RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects. This paper describes the neutronics model and discusses some of the analyses

  20. Stochastic models of edge turbulent transport in the thermonuclear reactors

    International Nuclear Information System (INIS)

    Volchenkov, Dima

    2005-01-01

    Two-dimensional stochastic model of turbulent transport in the scrape-off layer (SOL) of thermonuclear reactors is considered. Convective instability arisen in the system with respect to perturbations reveals itself in the strong outward bursts of particle density propagating ballistically across the SOL. The criterion of stability for the fluctuations of particle density is formulated. A possibility to stabilize the system depends upon the certain type of plasma waves interactions and the certain scenario of turbulence. A bias of limiter surface would provide a fairly good insulation of chamber walls excepting for the resonant cases. Pdf of the particle flux for the large magnitudes of flux events is modeled with a simple discrete time toy model of I-dimensional random walks concluding at the boundary. The spectra of wandering times feature the pdf of particle flux in the model and qualitatively reproduce the experimental statistics of transport events