WorldWideScience

Sample records for neutron transport phenomena

  1. Monte Carlo simulation of neutron transport phenomena

    International Nuclear Information System (INIS)

    Srinivasan, P.

    2009-01-01

    Neutron transport is one of the central problems in nuclear reactor related studies and other applied sciences. Some of the major applications of neutron transport include nuclear reactor design and safety, criticality safety of fissile material handling, neutron detector design and development, nuclear medicine, assessment of radiation damage to materials, nuclear well logging, forensic analysis etc. Most reactor and dosimetry studies assume that neutrons diffuse from regions of high to low density just like gaseous molecules diffuse to regions of low concentration or heat flow from high to low temperature regions. However while treatment of gaseous or heat diffusion is quite accurately modeled, treatment of neutron transport as simple diffusion is quite limited. In simple diffusion, the neutron trajectories are irregular, random and zigzag - where as in neutron transport low reaction cross sections (1 barn- 10 -24 cm 2 ) lead to long mean free paths which again depend on the nature and irregularities of the medium. Hence a more accurate representation of the neutron transport evolved based on the Boltzmann equation of kinetic gas theory. In fact the neutron transport equation is a linearized version of the Boltzmann gas equation based on neutron conservation with seven independent variables. The transport equation is difficult to solve except in simple cases amenable to numerical methods. The diffusion (equation) approximation follows from removing the angular dependence of the neutron flux

  2. Transport phenomena

    International Nuclear Information System (INIS)

    Kirczenow, G.; Marro, J.

    1974-01-01

    Some simple remarks on the basis of transport theory. - Entropy, dynamics and scattering theory. - Response, relaxation and fluctuation. - Fluctuating hydrodynamics and renormalization of susceptibilities and transport coefficients. - Irreversibility of the transport equations. - Ergodic theory and statistical mechanics. - Correlation functions in Heisenberg magnets. - On the Enskog hard-sphere kinetic eqquation and the transport phenomena of dense simple gases. - What can one learn from Lorentz models. - Conductivity in a magnetic field. - Transport properties in gases in presence of external fields. - Transport properties of dilute gases with internal structure. (orig.) [de

  3. Interfacial transport phenomena

    CERN Document Server

    Slattery, John C; Oh, Eun-Suok

    2007-01-01

    Revised and updated extensively from the previous editionDiscusses transport phenomena at common lines or three-phase lines of contactProvides a comprehensive summary about the extensions of continuum mechanics to the nanoscale.

  4. Modelling of transport phenomena

    International Nuclear Information System (INIS)

    Itoh, Kimitaka; Itoh, Sanae; Fukuyama, Atsushi.

    1993-09-01

    In this review article, we discuss key features of the transport phenomena and theoretical modelling to understand them. Experimental observations have revealed the nature of anomalous transport, i.e., the enhancement of the transport coefficients by the gradients of the plasma profiles, the pinch phenomena, the radial profile of the anomalous transport coefficients, the variation of the transport among the Bohm diffusion, Pseudo-classical confinement, L-mode and variety of improved confinement modes, and the sudden jumps such as L-H transition. Starting from the formalism of the transport matrix, the modelling based on the low frequency instabilities are reviewed. Theoretical results in the range of drift wave frequency are examined. Problems in theories based on the quasilinear and mixing-length estimates lead to the renewal of the turbulence theory, and the physics picture of the self-sustained turbulence is discussed. The theory of transport using the fluid equation of plasma is developed, showing that the new approach is very promising in explaining abovementioned characteristics of anomalous transport in both L-mode and improved confinement plasmas. The interference of the fluxes is the key to construct the physics basis of the bifurcation theory for the L-H transition. The present status of theories on the mechanisms of improved confinement is discussed. Modelling on the nonlocal nature of transport is briefly discussed. Finally, the impact of the anomalous transport on disruptive phenomena is also described. (author) 95 refs

  5. Neutron transport

    International Nuclear Information System (INIS)

    Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Coste-Delclaux, Mireille; M'Backe Diop, Cheikh; Nicolas, Anne; Andrieux, Catherine; Archier, Pascal; Baudron, Anne-Marie; Bernard, David; Biaise, Patrick; Blanc-Tranchant, Patrick; Bonin, Bernard; Bouland, Olivier; Bourganel, Stephane; Calvin, Christophe; Chiron, Maurice; Damian, Frederic; Dumonteil, Eric; Fausser, Clement; Fougeras, Philippe; Gabriel, Franck; Gagnier, Emmanuel; Gallo, Daniele; Hudelot, Jean-Pascal; Hugot, Francois-Xavier; Dat Huynh, Tan; Jouanne, Cedric; Lautard, Jean-Jacques; Laye, Frederic; Lee, Yi-Kang; Lenain, Richard; Leray, Sylvie; Litaize, Olivier; Magnaud, Christine; Malvagi, Fausto; Mijuin, Dominique; Mounier, Claude; Naury, Sylvie; Nicolas, Anne; Noguere, Gilles; Palau, Jean-Marc; Le Pallec, Jean-Charles; Peneliau, Yannick; Petit, Odile; Poinot-Salanon, Christine; Raepsaet, Xavier; Reuss, Paul; Richebois, Edwige; Roque, Benedicte; Royer, Eric; Saint-Jean, Cyrille de; Santamarina, Alain; Serot, Olivier; Soldevila, Michel; Tommasi, Jean; Trama, Jean-Christophe; Tsilanizara, Aime; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre

    2013-10-01

    This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality

  6. Transport phenomena II essentials

    CERN Document Server

    REA, The Editors of

    2012-01-01

    REA's Essentials provide quick and easy access to critical information in a variety of different fields, ranging from the most basic to the most advanced. As its name implies, these concise, comprehensive study guides summarize the essentials of the field covered. Essentials are helpful when preparing for exams, doing homework and will remain a lasting reference source for students, teachers, and professionals. Transport Phenomena II covers forced convention, temperature distribution, free convection, diffusitivity and the mechanism of mass transfer, convective mass transfer, concentration

  7. Transport phenomena I essentials

    CERN Document Server

    REA, The Editors of

    2012-01-01

    REA's Essentials provide quick and easy access to critical information in a variety of different fields, ranging from the most basic to the most advanced. As its name implies, these concise, comprehensive study guides summarize the essentials of the field covered. Essentials are helpful when preparing for exams, doing homework and will remain a lasting reference source for students, teachers, and professionals. Transport Phenomena I includes viscosity, flow of Newtonian fluids, velocity distribution in laminar flow, velocity distributions with more than one independent variable, thermal con

  8. Transport phenomena in environmental engineering

    Science.gov (United States)

    Sander, Aleksandra; Kardum, Jasna Prlić; Matijašić, Gordana; Žižek, Krunoslav

    2018-01-01

    A term transport phenomena arises as a second paradigm at the end of 1950s with high awareness that there was a strong need to improve the scoping of chemical engineering science. At that point, engineers became highly aware that it is extremely important to take step forward from pure empirical description and the concept of unit operations only to understand the specific process using phenomenological equations that rely on three elementary physical processes: momentum, energy and mass transport. This conceptual evolution of chemical engineering was first presented with a well-known book of R. Byron Bird, Warren E. Stewart and Edwin N. Lightfoot, Transport Phenomena, published in 1960 [1]. What transport phenomena are included in environmental engineering? It is hard to divide those phenomena through different engineering disciplines. The core is the same but the focus changes. Intention of the authors here is to present the transport phenomena that are omnipresent in treatment of various process streams. The focus in this chapter is made on the transport phenomena that permanently occur in mechanical macroprocesses of sedimentation and filtration for separation in solid-liquid particulate systems and on the phenomena of the flow through a fixed and a fluidized bed of particles that are immanent in separation processes in packed columns and in environmental catalysis. The fundamental phenomena for each thermal and equilibrium separation process technology are presented as well. Understanding and mathematical description of underlying transport phenomena result in scoping the separation processes in a way that ChEs should act worldwide.

  9. Micro transport phenomena during boiling

    CERN Document Server

    Peng, Xiaofeng

    2011-01-01

    "Micro Transport Phenomena During Boiling" reviews the new achievements and contributions in recent investigations at microscale. It presents some original research results and discusses topics at the frontier of thermal and fluid sciences.

  10. Transport Phenomena in Gel

    Directory of Open Access Journals (Sweden)

    Masayuki Tokita

    2016-05-01

    Full Text Available Gel becomes an important class of soft materials since it can be seen in a wide variety of the chemical and the biological systems. The unique properties of gel arise from the structure, namely, the three-dimensional polymer network that is swollen by a huge amount of solvent. Despite the small volume fraction of the polymer network, which is usually only a few percent or less, gel shows the typical properties that belong to solids such as the elasticity. Gel is, therefore, regarded as a dilute solid because its elasticity is much smaller than that of typical solids. Because of the diluted structure, small molecules can pass along the open space of the polymer network. In addition to the viscous resistance of gel fluid, however, the substance experiences resistance due to the polymer network of gel during the transport process. It is, therefore, of importance to study the diffusion of the small molecules in gel as well as the flow of gel fluid itself through the polymer network of gel. It may be natural to assume that the effects of the resistance due to the polymer network of gel depends strongly on the network structure. Therefore, detailed study on the transport processes in and through gel may open a new insight into the relationship between the structure and the transport properties of gel. The two typical transport processes in and through gel, that is, the diffusion of small molecules due to the thermal fluctuations and the flow of gel fluid that is caused by the mechanical pressure gradient will be reviewed.

  11. Lunar magma transport phenomena

    Science.gov (United States)

    Spera, Frank J.

    1992-01-01

    An outline of magma transport theory relevant to the evolution of a possible Lunar Magma Ocean and the origin and transport history of the later phase of mare basaltic volcanism is presented. A simple model is proposed to evaluate the extent of fractionation as magma traverses the cold lunar lithosphere. If Apollo green glasses are primitive and have not undergone significant fractionation en route to the surface, then mean ascent rates of 10 m/s and cracks of widths greater than 40 m are indicated. Lunar tephra and vesiculated basalts suggest that a volatile component plays a role in eruption dynamics. The predominant vapor species appear to be CO CO2, and COS. Near the lunar surface, the vapor fraction expands enormously and vapor internal energy is converted to mixture kinetic energy with the concomitant high-speed ejection of vapor and pyroclasts to form lunary fire fountain deposits such as the Apollo 17 orange and black glasses and Apollo 15 green glass.

  12. Transport phenomena in multiphase flows

    CERN Document Server

    Mauri, Roberto

    2015-01-01

    This textbook provides a thorough presentation of the phenomena related to the transport of mass, momentum and energy.  It lays all the basic physical principles, then for the more advanced readers, it offers an in-depth treatment with advanced mathematical derivations and ends with some useful applications of the models and equations in specific settings. The important idea behind the book is to unify all types of transport phenomena, describing them within a common framework in terms of cause and effect, respectively represented by the driving force and the flux of the transported quantity. The approach and presentation are original in that the book starts with a general description of transport processes, providing the macroscopic balance relations of fluid dynamics and heat and mass transfer, before diving into the mathematical realm of continuum mechanics to derive the microscopic governing equations at the microscopic level. The book is a modular teaching tool and can be used either for an introductory...

  13. Transport phenomena in particulate systems

    CERN Document Server

    Freire, José Teixeira; Ferreira, Maria do Carmo

    2012-01-01

    This volume spans 10 chapters covering different aspects of transport phenomena including fixed and fluidized systems, spouted beds, electrochemical and wastewater treatment reactors. This e-book will be valuable for students, engineers and researchers aiming to keep updated on the latest developments on particulate systems.

  14. Gravitational Anomaly and Transport Phenomena

    International Nuclear Information System (INIS)

    Landsteiner, Karl; Megias, Eugenio; Pena-Benitez, Francisco

    2011-01-01

    Quantum anomalies give rise to new transport phenomena. In particular, a magnetic field can induce an anomalous current via the chiral magnetic effect and a vortex in the relativistic fluid can also induce a current via the chiral vortical effect. The related transport coefficients can be calculated via Kubo formulas. We evaluate the Kubo formula for the anomalous vortical conductivity at weak coupling and show that it receives contributions proportional to the gravitational anomaly coefficient. The gravitational anomaly gives rise to an anomalous vortical effect even for an uncharged fluid.

  15. Transport phenomena in porous media

    CERN Document Server

    Ingham, Derek B

    1998-01-01

    Research into thermal convection in porous media has substantially increased during recent years due to its numerous practical applications. These problems have attracted the attention of industrialists, engineers and scientists from many very diversified disciplines, such as applied mathematics, chemical, civil, environmental, mechanical and nuclear engineering, geothermal physics and food science. Thus, there is a wealth of information now available on convective processes in porous media and it is therefore appropriate and timely to undertake a new critical evaluation of this contemporary information. Transport Phenomena in Porous Media contains 17 chapters and represents the collective work of 27 of the world's leading experts, from 12 countries, in heat transfer in porous media. The recent intensive research in this area has substantially raised the expectations for numerous new practical applications and this makes the book a most timely addition to the existing literature. It includes recent major deve...

  16. Transport phenomena an introduction to advanced topics

    CERN Document Server

    Glasgow, Larry A

    2010-01-01

    Enables readers to apply transport phenomena principles to solve advanced problems in all areas of engineering and science This book helps readers elevate their understanding of, and their ability to apply, transport phenomena by introducing a broad range of advanced topics as well as analytical and numerical solution techniques. Readers gain the ability to solve complex problems generally not addressed in undergraduate-level courses, including nonlinear, multidimensional transport, and transient molecular and convective transport scenarios. Avoiding rote memorization, the author em

  17. A Connection between Transport Phenomena and Thermodynamics

    Science.gov (United States)

    Swaney, Ross; Bird, R. Byron

    2017-01-01

    Although students take courses in transport phenomena and thermodynamics, they probably do not ask whether these two subjects are related. Here we give an answer to that question. Specifically we give relationships between the equations of change for total energy, internal energy, and entropy of transport phenomena and key equations of equilibrium…

  18. Neutron transportation simulator

    International Nuclear Information System (INIS)

    Uenohara, Yuzo.

    1995-01-01

    In the present invention, problems in an existent parallelized monte carlo method is solved, and behaviors of neutrons in a large scaled system are accurately simulated at a high speed. Namely, a neutron transportation simulator according to the monte carlo method simulates movement of each of neutrons by using a parallel computer. In this case, the system to be processed is divided based on a space region and an energy region to which neutrons belong. Simulation of neutrons in the divided regions is allotted to each of performing devices of the parallel computer. Tarry data and nuclear data of the neutrons in each of the regions are memorized dispersedly to memories of each of the performing devices. A transmission means for simulating the behaviors of the neutrons in the region by each of the performing devices, as well as transmitting the information of the neutrons, when the neutrons are moved to other region, to the performing device in a transported portion are disposed to each of the performing devices. With such procedures, simulation for the neutrons in the allotted region can be conducted with small capacity of memories. (I.S.)

  19. Computational transport phenomena for engineering analyses

    CERN Document Server

    Farmer, Richard C; Cheng, Gary C; Chen, Yen-Sen

    2009-01-01

    Computational Transport PhenomenaOverviewTransport PhenomenaAnalyzing Transport PhenomenaA Computational Tool: The CTP CodeVerification, Validation, and GeneralizationSummaryNomenclatureReferencesThe Equations of ChangeIntroductionDerivation of The Continuity EquationDerivation of The Species Continuity EquationDerivation of The Equation Of MotionDerivation of The General Energy EquationNon-Newtonian FluidsGeneral Property BalanceAnalytical and Approximate Solutions for the Equations of ChangeSummaryNomenclatureReferencesPhysical PropertiesOverviewReal-Fluid ThermodynamicsChemical Equilibrium

  20. Kinetic theory and transport phenomena

    CERN Document Server

    Soto, Rodrigo

    2016-01-01

    This textbook presents kinetic theory, which is a systematic approach to describing nonequilibrium systems. The text is balanced between the fundamental concepts of kinetic theory (irreversibility, transport processes, separation of time scales, conservations, coarse graining, distribution functions, etc.) and the results and predictions of the theory, where the relevant properties of different systems are computed. The book is organised in thematic chapters where different paradigmatic systems are studied. The specific features of these systems are described, building and analysing the appropriate kinetic equations. Specifically, the book considers the classical transport of charges, the dynamics of classical gases, Brownian motion, plasmas, and self-gravitating systems, quantum gases, the electronic transport in solids and, finally, semiconductors. Besides these systems that are studied in detail, concepts are applied to some modern examples including the quark–gluon plasma, the motion of bacterial suspen...

  1. Advances in transport phenomena 2011

    CERN Document Server

    2014-01-01

    This new volume of the annual review “Advances in Transport Phenomena” series contains three in-depth review articles on the microfluidic fabrication of vesicles, the dielectrophoresis field-flow fractionation for continuous-flow separation of particles and cells in microfluidic devices, and the thermodynamic analysis and optimization of heat exchangers, respectively.

  2. Visualization of frosting phenomena by using neutron radiography

    International Nuclear Information System (INIS)

    Yoshimura, Tomoya; Matsumoto, Ryosuke; Umekawa, Hisashi; Ami, Takeyuki; Saito, Yasushi

    2012-01-01

    This study focuses on the frost formation on the fin-tube heat exchanger using neutron radiography. The visualization of the frost formation was estimated by the attenuation of the neutron beam through the water. The visualization image of the neutron radiography shows clearly the frost formation phenomena on the fin-tube heat exchanger. The rapid frost formation was observed at the fin and tube edges. Local mass transfer coefficient can be calculated from the differential images of the neutron radiography. (author)

  3. Transport phenomena in strongly correlated Fermi liquids

    International Nuclear Information System (INIS)

    Kontani, Hiroshi

    2013-01-01

    Comprehensive overview. Written by an expert of this topic. Provides the reader with current developments in the field. In conventional metals, various transport coefficients are scaled according to the quasiparticle relaxation time, τ, which implies that the relaxation time approximation (RTA) holds well. However, such a simple scaling does not hold in many strongly correlated electron systems, reflecting their unique electronic states. The most famous example would be cuprate high-Tc superconductors (HTSCs), where almost all the transport coefficients exhibit a significant deviation from the RTA results. To better understand the origin of this discrepancy, we develop a method for calculating various transport coefficients beyond the RTA by employing field theoretical techniques. Near the magnetic quantum critical point, the current vertex correction (CVC), which describes the electron-electron scattering beyond the relaxation time approximation, gives rise to various anomalous transport phenomena. We explain anomalous transport phenomena in cuprate HTSCs and other metals near their magnetic or orbital quantum critical point using a uniform approach. We also discuss spin related transport phenomena in strongly correlated systems. In many d- and f-electron systems, the spin current induced by the spin Hall effect is considerably greater because of the orbital degrees of freedom. This fact attracts much attention due to its potential application in spintronics. We discuss various novel charge, spin and heat transport phenomena in strongly correlated metals.

  4. Transport phenomena in strongly correlated Fermi liquids

    CERN Document Server

    Kontani, Hiroshi

    2013-01-01

    In conventional metals, various transport coefficients are scaled according to the quasiparticle relaxation time, \\tau, which implies that the relaxation time approximation (RTA) holds well. However, such a simple scaling does not hold in many strongly correlated electron systems, reflecting their unique electronic states. The most famous example would be cuprate high-Tc superconductors (HTSCs), where almost all the transport coefficients exhibit a significant deviation from the RTA results. To better understand the origin of this discrepancy, we develop a method for calculating various transport coefficients beyond the RTA by employing field theoretical techniques. Near the magnetic quantum critical point, the current vertex correction (CVC), which describes the electron-electron scattering beyond the relaxation time approximation, gives rise to various anomalous transport phenomena. We explain anomalous transport phenomena in cuprate HTSCs and other metals near their magnetic or orbital quantum critical poi...

  5. Single event phenomena in atmospheric neutron environments

    International Nuclear Information System (INIS)

    Gossett, C.A.; Hughlock, B.W.; Katoozi, M.; LaRue, G.S.; Wender, S.A.

    1993-01-01

    As integrated circuit technology achieves higher density through smaller feature sizes and as the airplane manufacturing industry integrates more sophisticated electronic components into the design of new aircraft, it has become increasingly important to evaluate the contribution of single event effects, primarily Single Event Upset (SEU), to the safety and reliability of commercial aircraft. In contrast to the effects of radiation on electronic systems in space applications for which protons and heavy ions are of major concern, in commercial aircraft applications the interactions of high energy neutrons are the dominant cause of single event effects. These high energy neutrons are produced by the interaction of solar and galactic cosmic rays, principally protons and heavy ions, in the upper atmosphere. This paper will describe direct experimental measurements of neutron-induced Single Event Effect (SEE) rates in commercial high density static random access memories in a neutron environment characteristic of that at commercial airplane altitudes. The first experimental measurements testing current models for neutron-silicon burst generation rates will be presented, as well as measurements of charge collection in silicon test structures as a function of neutron energy. These are the first laboratory SEE and charge collection measurements using a particle beam having a continuum energy spectrum and with a shape nearly identical to that observed during flight

  6. Neutron stochastic transport theory with delayed neutrons

    International Nuclear Information System (INIS)

    Munoz-Cobo, J.L.; Verdu, G.

    1987-01-01

    From the stochastic transport theory with delayed neutrons, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation. From this theory a general expression is obtained for the Feynman Y-function when delayed neutrons are included. The single mode approximation for the particular case of a subcritical assembly is developed, and it is shown that Y-function reduces to the familiar expression quoted in many books, when delayed neutrons are not considered, and spatial and source effects are not included. (author)

  7. PREFACE: Transport phenomena in proton conducting media Transport phenomena in proton conducting media

    Science.gov (United States)

    Eikerling, Michael

    2011-06-01

    eminently important field of transport phenomena in proton conducting media. Complex dynamics of fluids in disordered and crowded environments contents Electrostatic models of electron-driven proton transfer across a lipid membrane Anatoly Yu Smirnov, Lev G Mourokh and Franco Nori Molecular basis of proton uptake in single and double mutants of cytochrome c oxidase Rowan M Henry, David Caplan, Elisa Fadda and Régis Pomès Proton diffusion along biological membranes E S Medvedev and A A Stuchebrukhov Ab initio molecular dynamics of proton networks in narrow polymer electrolyte pores Mehmet A Ilhan and Eckhard Spohr A simulation study of field-induced proton-conduction pathways in dry ionomers Elshad Allahyarov, Philip L Taylor and Hartmut Löwen Molecular structure and transport dynamics in perfluoro sulfonyl imide membranes Nagesh Idupulapati, Ram Devanathan and Michel Dupuis The kinetics of water sorption in Nafion membranes: a small-angle neutron scattering study Gérard Gebel, Sandrine Lyonnard, Hakima Mendil-Jakani and Arnaud Morin Using 2H labeling with neutron radiography for the study of solid polymer electrolyte water transport properties P Boillat, P Oberholzer, B C Seyfang, A Kästner, R Perego, G G Scherer, E H Lehmann and A Wokaun Spatial distribution and dynamics of proton conductivity in fuel cell membranes: potential and limitations of electrochemical atomic force microscopy measurements E Aleksandrova, S Hink, R Hiesgen and E Roduner A review on phosphate based, solid state, protonic conductors for intermediate temperature fuel cells O Paschos, J Kunze, U Stimming and F Maglia A structural study of the proton conducting B-site ordered perovskite Ba3Ca1.18Ta1.82O8.73 Maarten C Verbraeken, Hermenegildo A L Viana, Philip Wormald and John T S Irvine

  8. Basic transport phenomena in materials engineering

    CERN Document Server

    Iguchi, Manabu

    2014-01-01

    This book presents the basic theory and experimental techniques of transport phenomena in materials processing operations. Such fundamental knowledge is highly useful for researchers and engineers in the field to improve the efficiency of conventional processes or develop novel technology. Divided into four parts, the book comprises 11 chapters describing the principles of momentum transfer, heat transfer, and mass transfer in single phase and multiphase systems. Each chapter includes examples with solutions and exercises to facilitate students’ learning. Diagnostic problems are also provided at the end of each part to assess students’ comprehension of the material.  The book is aimed primarily at students in materials science and engineering. However, it can also serve as a useful reference text in chemical engineering as well as an introductory transport phenomena text in mechanical engineering. In addition, researchers and engineers engaged in materials processing operations will find the material use...

  9. Thermal transport phenomena in nanoparticle suspensions

    International Nuclear Information System (INIS)

    Cardellini, Annalisa; Fasano, Matteo; Bozorg Bigdeli, Masoud; Chiavazzo, Eliodoro; Asinari, Pietro

    2016-01-01

    Nanoparticle suspensions in liquids have received great attention, as they may offer an approach to enhance thermophysical properties of base fluids. A good variety of applications in engineering and biomedicine has been investigated with the aim of exploiting the above potential. However, the multiscale nature of nanosuspensions raises several issues in defining a comprehensive modelling framework, incorporating relevant molecular details and much larger scale phenomena, such as particle aggregation and their dynamics. The objectives of the present topical review is to report and discuss the main heat and mass transport phenomena ruling macroscopic behaviour of nanosuspensions, arising from molecular details. Relevant experimental results are included and properly put in the context of recent observations and theoretical studies, which solved long-standing debates about thermophysical properties enhancement. Major transport phenomena are discussed and in-depth analysis is carried out for highlighting the role of geometrical (nanoparticle shape, size, aggregation, concentration), chemical (pH, surfactants, functionalization) and physical parameters (temperature, density). We finally overview several computational techniques available at different scales with the aim of drawing the attention on the need for truly multiscale predictive models. This may help the development of next-generation nanoparticle suspensions and their rational use in thermal applications. (topical review)

  10. Modeling in transport phenomena a conceptual approach

    CERN Document Server

    Tosun, Ismail

    2007-01-01

    Modeling in Transport Phenomena, Second Edition presents and clearly explains with example problems the basic concepts and their applications to fluid flow, heat transfer, mass transfer, chemical reaction engineering and thermodynamics. A balanced approach is presented between analysis and synthesis, students will understand how to use the solution in engineering analysis. Systematic derivations of the equations and the physical significance of each term are given in detail, for students to easily understand and follow up the material. There is a strong incentive in science and engineering to

  11. An introduction to neutron transport

    International Nuclear Information System (INIS)

    Wiesenfeld, Bernard

    2015-01-01

    Neutron transport science is the study of neutron transport in a nuclear reactor and of associated nuclear reactions, notably fission reactions. Heat released by these reactions can be used for several purposes: electricity production, hydrogen production, sea water desalination, urban heating, naval propulsion, space propulsion, and so on. This publication contains the course proposed at Mines ParisTech and at the Arts et Metiers ParisTech. It is an introduction to neutron transport science and aims at presenting fundamental physical principles of this original branch of nuclear physics, a so called 'low energies' branch whereas 'high energy' nuclear physics focuses on elementary particles. It addresses complex computation methods which have been developed during the last decades with computation codes of always higher performance. The first part presents elements of atom physics: origin of matter, properties of nuclei and atoms, notion of quantum mechanics, interaction between radiation and matter (ray absorption, Compton Effect and scattering, photoelectric effect). The second part introduces neutron transport by addressing the following issues: nuclear structure, the various aspects of the interaction between neutrons and matter, the evolution of the reactivity of a reactor in normal operation, the chain fission reaction kinetics, and neutron slowing down. The third part addresses various aspects of neutron transport calculation: expression of neutron assessment, scattering approximation, critical condition of a nuclear reactor, introduction to transport theory, peculiarities of fast breeder reactors. The last chapter 'from theory to practice' addresses the approach of the neutron scientist, proposes an overview of the main calculation codes, and presents fields of application (within or without nuclear fission)

  12. 3-D neutron transport benchmarks

    International Nuclear Information System (INIS)

    Takeda, T.; Ikeda, H.

    1991-03-01

    A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes

  13. Investigations of fundamental phenomena in quantum mechanics with neutrons

    International Nuclear Information System (INIS)

    Hasegawa, Yuji

    2014-01-01

    Neutron interferometer and polarimeter are used for the experimental investigations of quantum mechanical phenomena. Interferometry exhibits clear evidence of quantum-contextuality and polarimetry demonstrates conflicts of a contextual model of quantum mechanics á la Leggett. In these experiments, entanglements are achieved between degrees of freedom in a single-particle: spin, path and energy degrees of freedom are manipulated coherently and entangled. Both experiments manifest the fact that quantum contextuality is valid for phenomena with matter waves with high precision. In addition, another experiment is described which deals with error-disturbance uncertainty relation: we have experimentally tested error-disturbance uncertainty relations, one is derived by Heisenberg and the other by Ozawa. Experimental results confirm the fact that the Heisenberg's uncertainty relation is often violated and that the new relation by Ozawa is always larger than the limit. At last, as an example of a counterfactual phenomenon of quantum mechanics, observation of so-called quantum Cheshire Cat is carried out by using neutron interferometer. Experimental results suggest that pre- and post-selected neutrons travel through one of the arms of the interferometer while their magnetic moment is located in the other arm.

  14. Neutron scattering studies of pretransitional phenomena in structural phase transformations

    International Nuclear Information System (INIS)

    Shapiro, S.M.

    1979-03-01

    Materials exhibiting structural phase transformations are well known to possess pretransitional phenomena. Below the transition temperature, T/sub c/, an order parameter appears and the pretransitional effects are associated with the fluctuations of the order parameter. Neutron scattering techniques have proved invaluable in studying the temporal and spatial dependence of these fluctuations. SrTiO 3 is the prototypical example of a structural phase transformation exhibiting features observable in other transformations such as martensitic and order-disorder. The experimental evolution of the understanding of the phase transformation in SrTiO 3 will be reviewed and the features observed will be shown to typify other systems

  15. Linear stochastic neutron transport theory

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A new and direct derivation of the Bell-Pal fundamental equation for (low power) neutron stochastic behaviour in the Boltzmann continuum model is given. The development includes correlation of particle emission direction in induced and spontaneous fission. This leads to generalizations of the backward and forward equations for the mean and variance of neutron behaviour. The stochastic importance for neutron transport theory is introduced and related to the conventional deterministic importance. Defining equations and moment equations are derived and shown to be related to the backward fundamental equation with the detector distribution of the operational definition of stochastic importance playing the role of an adjoint source. (author)

  16. Transport phenomena in materials processing---1990

    International Nuclear Information System (INIS)

    Bishop, B.J.; Lior, N.; Lavine, A.; Flik, M.; Karwe, M.V.; Bergman, T.L.; Beckermann, C.; Charmchi, M.

    1990-01-01

    The papers contained in this volume represent a wide range of current research interests in processes such as food and polymer processing, casting, welding, machining, laser cutting, and superconductor processing. This volume includes papers presented in four sessions: Heat Transfer in Materials Processing; Thermal Phenomena in Superconductor Processing; Heat Transfer in Food and Polymer Processing; Heat Transfer in CAsting and Welding

  17. Transport phenomena in Newtonian fluids a concise primer

    CERN Document Server

    Olsson, Per

    2013-01-01

    This short primer provides a concise and tutorial-style introduction to transport phenomena in Newtonian fluids , in particular the transport of mass, energy and momentum.  The reader will find detailed derivations of the transport equations for these phenomena, as well as selected analytical solutions to the transport equations in some simple geometries. After a brief introduction to the basic mathematics used in the text, Chapter 2, which deals with momentum transport, presents a derivation of the Navier-Stokes-Duhem equation describing the basic flow in a Newtonian fluid.  Also provided at

  18. Visualization and measurement of fluid phenomena using neutron radiography techniques

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Hibiki, Takashi; Fujine, Shigenori; Yoneda, Kenji; Kanda, Keiji; Nishihara, Hideaki; Tsuruno, Akira; Matsubayashi, Masahito; Sobajima, Makoto; Ohtomo, Shoichi.

    1993-01-01

    This paper presents some of the results from recent work performed on the application of neutron radiography to visualization and measurement of fluid phenomena at the Research Reactor Institute of Kyoto University. Experiments have been performed on the following subjects with use of the NR systems at the Japan Research Reactor 3 and the Nuclear Safety Research Reactor of the Japan Atomic Energy Research Institute as well as the Kyoto University Research Reactor: air-water flow in rectangular ducts with 1.0 and 2.4 mm gaps, air-water flow and steam-water flow in a round tube with 4.0 mm inner diameter. The void fraction was measured by processing the images taken by the neutron radiography. The effect of several corrections in image processing was also discussed previously. It was shown that the proposed method could be useful in observing the flow regimes and measuring the void fraction of gas-liquid two-phase flow in narrow channels. (author)

  19. Quantum contextual phenomena observed in single-neutron interferometer experiments

    International Nuclear Information System (INIS)

    Hasegawa, Yuji; Rauch, Helmut

    2006-01-01

    Neutron optical experiments are presented, which exhibit quantum contextual phenomena. Entanglement is achieved not between particles, but between degrees of freedom, in this case, for a single-particle. Appropriate combinations of the direction of spin analysis and the position of the phase shifter allow an experimental verification of the violation of a Bell-like inequality. Our experiments manifest the fact that manipulation of the wavefunction in one Hilbert space influences the result of the measurement in the other Hilbert space: manipulation without touch! Next, we report another experiment which exhibits other peculiarity of quantum contextuality, e.g., originally intended to show a Kochen-Specker-like phenomenon. We have introduced inequalities for quantitative analysis of the experiments. The value obtained in the experiments clearly showed violations of prediction by non-contextual theory. Finally, we have accomplished a tomographic determination of entangled quantum state in single-neutrons. There, characteristics of the Bell-sate are confirmed: four poles for the real part of the density matrix are clearly seen

  20. Neutron measurement by transportable spectrometer

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Two levels of neutron spectrometry are in regular use at nuclear power plants: some techniques used in the laboratory produce detailed spectra but require specialist operators, while simple instruments used by non-specialists to measure the neutron dose-rate to operators provide little spectral information. The standard portable instruments are therefore of no use when anomalous readings are obtained which require further investigation. AEA Technology at Winfrith has developed a Transportable Neutron Spectrometer (TNS) which is designed to produce reasonable spectra in routine use by staff with no specialist skill in spectroscopy, and high-quality spectra in the hands of skilled staff. The TNS provides a level of information intermediate between those currently available, and is also designed to solve the problem of imperfect dose response which is common in portable dosimeters. The TNS system consists of a power supply, a probe and a signal processing and data acquisition unit. (author)

  1. Phenomena in thermal transport in fuels

    International Nuclear Information System (INIS)

    Chernatynskiy, A.; Tulenko, J.S.; Phillpot, S.R.; El-Azab, A.

    2015-01-01

    Thermal transport in nuclear fuels is a key performance metric that affects not only the power output, but is also an important consideration in potential accident situations. While the fundamental theory of the thermal transport in crystalline solids was extensively developed in the 1950's and 1960's, the pertinent analytic approaches contained significant simplifications of the physical processes. While these approaches enabled estimates of the thermal conductivity in bulk materials with microstructure, they were not comprehensive enough to provide the detailed guidance needed for the in-pile fuel performance. Rather, this guidance has come from data painfully accumulated over 50 years of experiments on irradiated uranium dioxide, the most widely used nuclear fuel. At this point, a fundamental theoretical understanding of the interplay between the microstructure and thermal conductivity of irradiated uranium dioxide fuel is still lacking. In this chapter, recent advances are summarised in the modelling approaches for thermal transport of uranium dioxide fuel. Being computational in nature, these modelling approaches can, at least in principle, describe in detail virtually all mechanisms affecting thermal transport at the atomistic level, while permitting the coupling of the atomistic-level simulations to the mesoscale continuum theory and thus enable the capture of the impact of microstructural evolution in fuel on thermal transport. While the subject of current studies is uranium dioxide, potential applications of the methods described in this chapter extend to the thermal performance of other fuel forms. (authors)

  2. Imaging unsteady three-dimensional transport phenomena

    Indian Academy of Sciences (India)

    2014-01-05

    Jan 5, 2014 ... The image data can be jointly analysed with the physical laws governing transport and principles of image formation. Hence, with the experiment suitably carried out, three-dimensional physical domains with unsteady processes can be accommodated. Optical methods promise to breach the holy grail of ...

  3. Charge Transport Phenomena in Peptide Molecular Junctions

    International Nuclear Information System (INIS)

    Luchini, A.; Petricoin, E.F.; Geho, D.H.; Liotta, L.A.; Long, D.P.; Vaisman, I.I.

    2008-01-01

    Inelastic electron tunneling spectroscopy (IETS) is a valuable in situ spectroscopic analysis technique that provides a direct portrait of the electron transport properties of a molecular species. In the past, IETS has been applied to small molecules. Using self-assembled nano electronic junctions, IETS was performed for the first time on a large polypeptide protein peptide in the phosphorylated and native form, yielding interpretable spectra. A reproducible 10-fold shift of the I/V characteristics of the peptide was observed upon phosphorylation. Phosphorylation can be utilized as a site-specific modification to alter peptide structure and thereby influence electron transport in peptide molecular junctions. It is envisioned that kinases and phosphatases may be used to create tunable systems for molecular electronics applications, such as biosensors and memory devices.

  4. Individual and collective motions: transport phenomena

    International Nuclear Information System (INIS)

    Richert, J.

    The present lectures deal with the description of deep inelastic heavy ion reaction mechanism by means of transport concepts which were developed in statistical mechanics. The applicability of those concepts is analysed in terms of a hierarchy of characteristic times. The specificity of this time scale in the heavy ion framework is underlined. The derivation of transport coefficients out of the full dynamical formulation of the problem is shown through a random matrix model. The mechanism of mass and charge transfer between the ions as well as the generation of intrinsic angular momenta in the interacting fragments is formulated, quantitatively worked out and compared to some experimental results by using phenomenological approaches. Weak points of the theoretical approaches and disagreement between phenomenological models and experimental data are discussed. Open problems and unsolved questions are sketched in the final conclusions [fr

  5. Investigation on neutron/gamma discrimination phenomena in plastic scintillators

    International Nuclear Information System (INIS)

    Blanc, Pauline

    2014-01-01

    This PhD topic was born from misunderstandings and incomplete knowledge of the mechanism and relative effectiveness of neutron and gamma-ray (n/γ) discrimination between plastic scintillators compared to liquid scintillators. The shape of the light pulse these materials generate following interaction with an ionizing particle (predominantly recoil protons in the case of neutrons and electrons in the case of gamma-rays) is different in time in a way that depends on the detected particle (nature and energy). It is this fact that enables separation (PSD). The behavior in liquid scintillators has been extensively studied experimentally for practical applications. Only recently has it been shown that a weak separation can also be achieved using specially prepared plastics. The study of this system presents an open field and the understanding of both liquids and plastics with respect to their PSD properties is far from complete. This work is dedicated to exploring the fundamental photophysical phenomena at play in the generation of luminescence emission, following the interaction of ionizing radiation with organic scintillators. For this purpose, firstly a detailed literature review of the state-of-the-art has been conducted extending from 1960 to the present day. Secondly a complete characterization of the main scintillating materials has been conducted to define their fluorescence properties and the characteristics of their scintillation under irradiation. Thirdly a proton beam has been used to simulate recoil protons to quantify under controlled laboratory conditions their specific energy deposition in a plastic scintillator with PSD properties. The fourth part of this thesis is devoted to the study of PSD efficiency of scintillators as a function of their molecular structure. This investigation has led to a plastic scintillator prepared in our laboratory with good PSD properties and a patent submission. Finally, photophysical experiments were performed using a

  6. Coupled electric and transport phenomena in porous media

    NARCIS (Netherlands)

    Li, Shuai

    2014-01-01

    The coupled electrical and transport properties of clay-containing porous media are the topics of interest in this study. Both experimental and numerical (pore network modeling) techniques are employed to gain insight into the macro-scale interaction between electrical and solute transport phenomena

  7. Killing symmetries in neutron transport

    International Nuclear Information System (INIS)

    Lukacs, B.; Racz, A.

    1992-10-01

    Although inside the reactor zone there is no exact continuous spatial symmetry, in certain configurations neutron flux distribution is close to a symmetrical one. In such cases the symmetrical solution could provide a good starting point to determine the non-symmetrical power distribution. All possible symmetries are determined in the 3-dimensional Euclidean space, and the form of the transport equation is discussed in such a coordinate system which is adapted to the particular symmetry. Possible spontaneous symmetry breakings are pointed out. (author) 6 refs

  8. Phenomena of charged particles transport in variable magnetic fields

    International Nuclear Information System (INIS)

    Savane, Sy Y.; Faza Barry, M.; Vladmir, L.; Diaby, I.

    2002-11-01

    This present work is dedicated to the study of the dynamical phenomena for the transport of ions in the presence of variable magnetic fields in front of the Jupiter wave shock. We obtain the spectrum of the accelerated ions and we study the conditions of acceleration by solving the transport equation in the planetocentric system. We discuss the theoretical results obtained and make a comparison with the experimental parameters in the region of acceleration behind the Jupiter wave shock. (author)

  9. Visualization and measurement by image processing of thermal hydraulic phenomena by neutron radiography

    International Nuclear Information System (INIS)

    Takenaka, Nobuyuki

    1996-01-01

    Neutron Radiography was applied to visualization of thermal hydraulic phenomena and measurement was carried out by image processing the visualized images. Since attenuation of thermal neutron rays is high in ordinary liquids like water and organic fluid while it is low in most of metals, liquid flow behaviors can be visualized through a metallic wall by neutron radiography. Measurement of void fraction and flow vector field which is important to study thermal hydraulic phenomena can be carried out by image processing the images obtained by the visualization. Various two-phase and liquid metal flows were visualized by a JRR-3M thermal neutron radiography system in the present study. Multi-dimensional void fraction distributions in two-phase flows and flow vector fields in liquid metals, which are difficult to measure by the other methods, were successfully measured by image processing. It was shown that neutron radiography was efficiently applicable to study thermal hydraulic phenomena. (author)

  10. Scattering of neutrons and critical phenomena in antiferromagnetic fermi liquid

    International Nuclear Information System (INIS)

    Akhiezer, I.A.; Barannik, E.A.

    1980-01-01

    The scattering of slow neutrons in an antiferromagnetic with collectivized magnetic electrons is considered and it is shown to significantly differ from the neutron scattering in an antiferromagnetic with localized magnetic electrons. The behaviour of scattering cross sections and fluctuation correlators near the Neel point is studied. These magnitudes are shown to increase with the critical index r=-1 [ru

  11. Some improved methods in neutron transport theory

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J; Stefanovic, D; Kocic, A; Matausek, M; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1973-07-01

    The methods described in this paper are: analytical approach to neutron spectra in case of energy dependent anisotropy of elastic scattering; Monte Carlo estimations of neutron absorption reaction rate during slowing down process; spherical harmonics treatment of space-angle-lethargy dependent slowing down transport equation; integral transport theory based on point-wise representation of variables.

  12. Anomalous transport phenomena in Fermi liquids with strong magnetic fluctuations

    International Nuclear Information System (INIS)

    Kontani, Hiroshi

    2008-01-01

    In this paper, we present recent developments in the theory of transport phenomena based on the Fermi liquid theory. In conventional metals, various transport coefficients are scaled according to the quasiparticles relaxation time, τ, which implies that the relaxation time approximation (RTA) holds well. However, such a simple scaling does not hold in many strongly correlated electron systems. The most famous example would be high-T c superconductors (HTSCs), where almost all the transport coefficients exhibit a significant deviation from the RTA results. This issue has been one of the most significant unresolved problems in HTSCs for a long time. Similar anomalous transport phenomena have been observed in metals near their antiferromagnetic (AF) quantum critical point (QCP). The main goal of this study is to demonstrate whether the anomalous transport phenomena in HTSC is evidence of a non-Fermi liquid ground state, or just RTA violation in strongly correlated Fermi liquids. Another goal is to establish a unified theory of anomalous transport phenomena in metals with strong magnetic fluctuations. For these purposes, we develop a method for calculating various transport coefficients beyond the RTA by employing field theoretical techniques. In a Fermi liquid, an excited quasiparticle induces other excited quasiparticles by collision, and current due to these excitations is called a current vertex correction (CVC). Landau noticed the existence of CVC first, which is indispensable for calculating transport coefficients in accord with the conservation laws. Here, we develop a transport theory involving resistivity and the Hall coefficient on the basis of the microscopic Fermi liquid theory, by considering the CVC. In nearly AF Fermi liquids, we find that the strong backward scattering due to AF fluctuations induces the CVC with prominent momentum dependence. This feature of the CVC can account for the significant enhancement in the Hall coefficient, magnetoresistance

  13. Phase-shift and spin-rotation phenomena in neutron interferometry

    International Nuclear Information System (INIS)

    Badurek, G.; Rauch, H.; Zeilinger, A.; Bauspiess, W.; Bonse, U.

    1976-01-01

    The perfect-crystal neutron interferometer was used to study characteristic phenomena arising from simultaneous phase shift and spin rotation of neutron waves. In accordance with theoretical predictions, the beams leaving the interferometer became partially polarized, even with unpolarized incident neutrons. The intensity and the polarization as a function of phase shift and spin rotation have been found to oscillate with the same period, displaying a mutual beat pattern

  14. Neutron transport equation - indications on homogenization and neutron diffusion

    International Nuclear Information System (INIS)

    Argaud, J.P.

    1992-06-01

    In PWR nuclear reactor, the practical study of the neutrons in the core uses diffusion equation to describe the problem. On the other hand, the most correct method to describe these neutrons is to use the Boltzmann equation, or neutron transport equation. In this paper, we give some theoretical indications to obtain a diffusion equation from the general transport equation, with some simplifying hypothesis. The work is organised as follows: (a) the most general formulations of the transport equation are presented: integro-differential equation and integral equation; (b) the theoretical approximation of this Boltzmann equation by a diffusion equation is introduced, by the way of asymptotic developments; (c) practical homogenization methods of transport equation is then presented. In particular, the relationships with some general and useful methods in neutronic are shown, and some homogenization methods in energy and space are indicated. A lot of other points of view or complements are detailed in the text or the remarks

  15. Computational transport phenomena of fluid-particle systems

    CERN Document Server

    Arastoopour, Hamid; Abbasi, Emad

    2017-01-01

    This book concerns the most up-to-date advances in computational transport phenomena (CTP), an emerging tool for the design of gas-solid processes such as fluidized bed systems. The authors examine recent work in kinetic theory and CTP and illustrate gas-solid processes’ many applications in the energy, chemical, pharmaceutical, and food industries. They also discuss the kinetic theory approach in developing constitutive equations for gas-solid flow systems and how it has advanced over the last decade as well as the possibility of obtaining innovative designs for multiphase reactors, such as those needed to capture CO2 from flue gases. Suitable as a concise reference and a textbook supplement for graduate courses, Computational Transport Phenomena of Gas-Solid Systems is ideal for practitioners in industries involved with the design and operation of processes based on fluid/particle mixtures, such as the energy, chemicals, pharmaceuticals, and food processing. Explains how to couple the population balance e...

  16. 8th International symposium on transport phenomena in combustion

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    The 8th International Symposium on Transport Phenomena in Combustion will be held in San Francisco, California, U.S.A., July 16-20, 1995, under the auspices of the Pacific Center of Thermal-Fluids Engineering. The purpose of the Symposium is to provide a forum for researchers and practitioners from around the world to present new developments and discuss the state of the art and future directions and priorities in the areas of transport phenomena in combustion. The Symposium is the eighth in a series; previous venues were Honolulu 1985, Tokyo 1987, Taipei 1988, Sydney 1991, Beijing 1992, Seoul 1993 and Acapulco 1994, with emphasis on various aspects of transport phenomena. The current Symposium theme is combustion. The Symposium has assembled a balanced program with topics ranging from fundamental research to contemporary applications of combustion theory. Invited keynote lecturers will provide extensive reviews of topics of great interest in combustion. Colloquia will stress recent advances and innovations in fire spread and suppression, and in low NO{sub x} burners, furnaces, boilers, internal combustion engines, and other practical combustion systems. Finally, numerous papers will contribute to the fundamental understanding of complex processes in combustion. This document contains abstracts of papers to be presented at the Symposium.

  17. New methods for analyzing transport phenomena in supersonic ejectors

    International Nuclear Information System (INIS)

    Lamberts, Olivier; Chatelain, Philippe; Bartosiewicz, Yann

    2017-01-01

    Highlights: • Simulation of a supersonic ejector with the open source software for CFD OpenFOAM. • Validation of the numerical tool based on flow structures obtained by schlieren. • Application of the momentum and energy tube analysis tools to a supersonic ejector. • Extension of this framework to exergy to construct exergy transport tubes. • Quantification of local transfers and losses of exergy within the ejector. - Abstract: This work aims at providing novel insights into the quantification and the location of the transfers and the irreversibilities within supersonic ejectors, and their connection with the entrainment. In this study, we propose two different and complementary approaches. First of all, recent analysis tools based on momentum and energy tubes (Meyers and Meneveau (2013)) are extended to the present compressible flow context and applied to the mean-flow structure of turbulent flow within the ejector. Furthermore, the transport equation for the mean-flow total exergy is derived and exergy transport tubes are proposed as a tool for the investigation of transport phenomena within supersonic ejectors. In addition to this topological approach, an analysis based on classical stream tubes is performed in order to quantitatively investigate transfers between the primary and the secondary streams all along the ejector. Finally, the present work identifies the location of exergy losses and their origins. Throughout this analysis, new local and cumulative parameters related to transfers and irreversibilities are introduced. The proposed methodology sheds light on the complex phenomena at play and may serve as a basis for the analysis of transport phenomena within supersonic ejectors. For the ejector under consideration, although global transfers are more important in on-design conditions, it is shown that the net gain in exergy of the secondary stream is maximum for a value of the back pressure that is close to the critical back pressure, as

  18. Transport phenomena and drying of solids and particulate materials

    CERN Document Server

    Lima, AG

    2014-01-01

    The purpose of this book, Transport Phenomena and Drying of Solids and Particulate Materials, is to provide a collection of recent contributions in the field of heat and mass transfer, transport phenomena, drying and wetting of solids and particulate materials. The main benefit of the book is that it discusses some of the most important topics related to the heat and mass transfer in solids and particulate materials. It includes a set of new developments in the field of basic and applied research work on the physical and chemical aspects of heat and mass transfer phenomena, drying and wetting processes, namely, innovations and trends in drying science and technology, drying mechanism and theory, equipment, advanced modelling, complex simulation and experimentation. At the same time, these topics will be going to the encounter of a variety of scientific and engineering disciplines. The book is divided in several chapters that intend to be a resume of the current state of knowledge for benefit of professional c...

  19. Macroscopic Modeling of Transport Phenomena in Direct Methanol Fuel Cells

    DEFF Research Database (Denmark)

    Olesen, Anders Christian

    An increasing need for energy efficiency and high energy density has sparked a growing interest in direct methanol fuel cells for portable power applications. This type of fuel cell directly generates electricity from a fuel mixture consisting of methanol and water. Although this technology...... surpasses batteries in important areas, fundamental research is still required to improve durability and performance. Particularly the transport of methanol and water within the cell structure is difficult to study in-situ. A demand therefore exist for the fundamental development of mathematical models...... for studying their transport. In this PhD dissertation the macroscopic transport phenomena governing direct methanol fuel cell operation are analyzed, discussed and modeled using the two-fluid approach in the computational fluid dynamics framework of CFX 14. The overall objective of this work is to extend...

  20. Exact solution of the neutron transport equation in spherical geometry

    Energy Technology Data Exchange (ETDEWEB)

    Anli, Fikret; Akkurt, Abdullah; Yildirim, Hueseyin; Ates, Kemal [Kahramanmaras Suetcue Imam Univ. (Turkey). Faculty of Sciences and Letters

    2017-03-15

    Solution of the neutron transport equation in one dimensional slab geometry construct a basis for the solution of neutron transport equation in a curvilinear geometry. Therefore, in this work, we attempt to derive an exact analytical benchmark solution for both neutron transport equations in slab and spherical medium by using P{sub N} approximation which is widely used in neutron transport theory.

  1. Transport phenomena of nanoparticles in plants and animals/humans.

    Science.gov (United States)

    Anjum, Naser A; Rodrigo, Miguel Angel Merlos; Moulick, Amitava; Heger, Zbynek; Kopel, Pavel; Zítka, Ondřej; Adam, Vojtech; Lukatkin, Alexander S; Duarte, Armando C; Pereira, Eduarda; Kizek, Rene

    2016-11-01

    The interaction of a plethora nanoparticles with major biota such as plants and animals/humans has been the subject of various multidisciplinary studies with special emphasis on toxicity aspects. However, reports are meager on the transport phenomena of nanoparticles in the plant-animal/human system. Since plants and animals/humans are closely linked via food chain, discussion is imperative on the main processes and mechanisms underlying the transport phenomena of nanoparticles in the plant-animal/human system, which is the main objective of this paper. Based on the literature appraised herein, it is recommended to perform an exhaustive exploration of so far least explored aspects such as reproducibility, predictability, and compliance risks of nanoparticles, and insights into underlying mechanisms in context with their transport phenomenon in the plant-animal/human system. The outcomes of the suggested studies can provide important clues for fetching significant benefits of rapidly expanding nanotechnology to the plant-animal/human health-improvements and protection as well. Copyright © 2016 Elsevier Inc. All rights reserved.

  2. Cavitation phenomena in a fuel injection nozzle of a diesel engine by neutron radiography

    International Nuclear Information System (INIS)

    Takenaka, N.; Kawabata, Y.; Miyata, D.; Kawabata, Y.; Sim, C. M.; Lim, I. C.

    2005-01-01

    Visualization of cavitation phenomena in a Diesel engine fuel injection nozzle was carried out by using neutron radiography system in Research Reactor Institute in Kyoto University and HANARO in Korea Atomic Energy Research Institute. A neutron chopper was synchronized to the engine rotation for high shutter speed exposures. A multi exposure method was applied to obtain a clear image as an ensemble average of the synchronized images. Some images were successfully obtained and suggested new understanding of the cavitation phenomena in a Diesel engine fuel injection nozzle

  3. Neutron and x-ray scattering studies of premartensitic phenomena

    International Nuclear Information System (INIS)

    Shapiro, S.M.

    1987-01-01

    This paper discusses neutron and x-ray investigations of some metallic alloys which are known to exhibit martensitic transformations. It is shown that precursor effects are usually present in the diffuse scattering and in the phonon dispersion curves, but the transition cannot be described in terms of the soft mode picture used in the Landau and Devonshire theory to describe structural phase transitions. In addition, it is noted that it is inappropriate to look at these microstructures as incommensurate systems, but more correctly as a coherent coexistence of two phases

  4. Type II critical phenomena of neutron star collapse

    International Nuclear Information System (INIS)

    Noble, Scott C.; Choptuik, Matthew W.

    2008-01-01

    We investigate spherically symmetric, general relativistic systems of collapsing perfect fluid distributions. We consider neutron star models that are driven to collapse by the addition of an initially 'ingoing' velocity profile to the nominally static star solution. The neutron star models we use are Tolman-Oppenheimer-Volkoff solutions with an initially isentropic, gamma law equation of state. The initial values of (1) the amplitude of the velocity profile, and (2) the central density of the star, span a parameter space, and we focus only on that region that gives rise to type II critical behavior, wherein black holes of arbitrarily small mass can be formed. In contrast to previously published work, we find that--for a specific value of the adiabatic index (Γ=2)--the observed type II critical solution has approximately the same scaling exponent as that calculated for an ultrarelativistic fluid of the same index. Further, we find that the critical solution computed using the ideal-gas equations of state asymptotes to the ultrarelativistic critical solution.

  5. Influence of coupling phenomena on the transport through compacted clays

    Energy Technology Data Exchange (ETDEWEB)

    Rosanne, M.; Koudina, N.; Adler, P.M. [IPGP, Paris (France); Tevissen, E. [ANDRA, Dept. Etude-Experimentation et Calcul, Chatenay Malabry (France)

    2001-07-01

    Our principal motivation was to study the influence of the coupling phenomena on transport through compacted clays. Coupled transports may occur when a pressure gradient {nabla}P, and electrical field E and a concentration gradient {nabla}C interact. These three gradients induce three fluxes. A flow is generated characterized by the seepage velocity U; a solute flux J{sub L} and a current density I are generated. Close to equilibrium, when the gradients are not to large, the problem is linear and the fluxes are linear functions of the gradients. A first series of experiments was performed with argillite to determine the diagonal properties, i.e., permeability, conductivity, and diffusion coefficient. In a second series of experiments, the voltage resulting from an imposed concentration gradient between two reservoirs separated by a clay sample was systematically measured; this corresponds to the coefficient L{sub 13}. (orig.)

  6. Modelling transport phenomena in a multi-physics context

    Science.gov (United States)

    Marra, Francesco

    2015-01-01

    Innovative heating research on cooking, pasteurization/sterilization, defrosting, thawing and drying, often focuses on areas which include the assessment of processing time, evaluation of heating uniformity, studying the impact on quality attributes of the final product as well as considering the energy efficiency of these heating processes. During the last twenty years, so-called electro-heating-processes (radio-frequency - RF, microwaves - MW and ohmic - OH) gained a wide interest in industrial food processing and many applications using the above mentioned technologies have been developed with the aim of reducing processing time, improving process efficiency and, in many cases, the heating uniformity. In the area of innovative heating, electro-heating accounts for a considerable portion of both the scientific literature and commercial applications, which can be subdivided into either direct electro-heating (as in the case of OH heating) where electrical current is applied directly to the food or indirect electro-heating (e.g. MW and RF heating) where the electrical energy is firstly converted to electromagnetic radiation which subsequently generates heat within a product. New software packages, which make easier solution of PDEs based mathematical models, and new computers, capable of larger RAM and more efficient CPU performances, allowed an increasing interest about modelling transport phenomena in systems and processes - as the ones encountered in food processing - that can be complex in terms of geometry, composition, boundary conditions but also - as in the case of electro-heating assisted applications - in terms of interaction with other physical phenomena such as displacement of electric or magnetic field. This paper deals with the description of approaches used in modelling transport phenomena in a multi-physics context such as RF, MW and OH assisted heating.

  7. Modelling transport phenomena in a multi-physics context

    Energy Technology Data Exchange (ETDEWEB)

    Marra, Francesco [Dipartimento di Ingegneria Chimica e Alimentare - Università degli studi di Salerno Via Ponte Don Melillo - 84084 Fisciano SA (Italy)

    2015-01-22

    Innovative heating research on cooking, pasteurization/sterilization, defrosting, thawing and drying, often focuses on areas which include the assessment of processing time, evaluation of heating uniformity, studying the impact on quality attributes of the final product as well as considering the energy efficiency of these heating processes. During the last twenty years, so-called electro-heating-processes (radio-frequency - RF, microwaves - MW and ohmic - OH) gained a wide interest in industrial food processing and many applications using the above mentioned technologies have been developed with the aim of reducing processing time, improving process efficiency and, in many cases, the heating uniformity. In the area of innovative heating, electro-heating accounts for a considerable portion of both the scientific literature and commercial applications, which can be subdivided into either direct electro-heating (as in the case of OH heating) where electrical current is applied directly to the food or indirect electro-heating (e.g. MW and RF heating) where the electrical energy is firstly converted to electromagnetic radiation which subsequently generates heat within a product. New software packages, which make easier solution of PDEs based mathematical models, and new computers, capable of larger RAM and more efficient CPU performances, allowed an increasing interest about modelling transport phenomena in systems and processes - as the ones encountered in food processing - that can be complex in terms of geometry, composition, boundary conditions but also - as in the case of electro-heating assisted applications - in terms of interaction with other physical phenomena such as displacement of electric or magnetic field. This paper deals with the description of approaches used in modelling transport phenomena in a multi-physics context such as RF, MW and OH assisted heating.

  8. Modelling transport phenomena in a multi-physics context

    International Nuclear Information System (INIS)

    Marra, Francesco

    2015-01-01

    Innovative heating research on cooking, pasteurization/sterilization, defrosting, thawing and drying, often focuses on areas which include the assessment of processing time, evaluation of heating uniformity, studying the impact on quality attributes of the final product as well as considering the energy efficiency of these heating processes. During the last twenty years, so-called electro-heating-processes (radio-frequency - RF, microwaves - MW and ohmic - OH) gained a wide interest in industrial food processing and many applications using the above mentioned technologies have been developed with the aim of reducing processing time, improving process efficiency and, in many cases, the heating uniformity. In the area of innovative heating, electro-heating accounts for a considerable portion of both the scientific literature and commercial applications, which can be subdivided into either direct electro-heating (as in the case of OH heating) where electrical current is applied directly to the food or indirect electro-heating (e.g. MW and RF heating) where the electrical energy is firstly converted to electromagnetic radiation which subsequently generates heat within a product. New software packages, which make easier solution of PDEs based mathematical models, and new computers, capable of larger RAM and more efficient CPU performances, allowed an increasing interest about modelling transport phenomena in systems and processes - as the ones encountered in food processing - that can be complex in terms of geometry, composition, boundary conditions but also - as in the case of electro-heating assisted applications - in terms of interaction with other physical phenomena such as displacement of electric or magnetic field. This paper deals with the description of approaches used in modelling transport phenomena in a multi-physics context such as RF, MW and OH assisted heating

  9. Exotic geophysical phenomena observed in an environmental neutron flux study using EAS PRISMA detectors

    Directory of Open Access Journals (Sweden)

    Alekseenko Victor

    2017-01-01

    Full Text Available Some exotic geophysical events are observed by a global net of electron-neutron detectors (en-detectors developed in the framework of the PRISMA EAS project. Our en-detectors running both on the Earth's surface and underground are continuously measuring the environmental thermal neutron flux. Thermal neutrons are in equilibrium with media and are therefore sensitive to many geophysical phenomena, which are exotic for people studying ultra high-energy cosmic rays or carrying out low background experiments deep underground.

  10. Numerical investigations for insulation particle transport phenomena in water flow

    International Nuclear Information System (INIS)

    Krepper, E.; Grahn, A.; Alt, S.; Kaestner, W.; Kratzsch, A.; Seeliger, A.

    2005-01-01

    The investigation of insulation debris generation, transport and sedimentation gains importance regarding the reactor safety research for PWR and BWR considering the long term behaviour of emergency core coolant systems during all types of LOCA. The insulation debris released near the break during LOCA consists of a mixture of very different particles concerning size, shape, consistence and other properties. Some fraction of the released insulation debris will be transported into the reactor sump where it may affect emergency core cooling. Open questions of generic interest are e.g. the sedimentation of the insulation debris in a water pool, possible re-suspension, transport in the sump water flow, particle load on strainers and corresponding difference pressure. A joint research project in cooperation with Institute of Process Technology, Process Automation and Measuring Technology (IPM) Zittau deals with the experimental investigation and the development of CFD models for the description of particle transport phenomena in coolant flow. While experiments are performed at the IPM-Zittau, theoretical work is concentrated at Forschungszentrum Rossendorf. In the present paper the basic concepts for CFD modelling are described and first results including feasibility studies are shown. During the ongoing work further results are expected. (author)

  11. Kinetic theory of nonlinear transport phenomena in complex plasmas

    International Nuclear Information System (INIS)

    Mishra, S. K.; Sodha, M. S.

    2013-01-01

    In contrast to the prevalent use of the phenomenological theory of transport phenomena, a number of transport properties of complex plasmas have been evaluated by using appropriate expressions, available from the kinetic theory, which are based on Boltzmann's transfer equation; in particular, the energy dependence of the electron collision frequency has been taken into account. Following the recent trend, the number and energy balance of all the constituents of the complex plasma and the charge balance on the particles is accounted for; the Ohmic loss has also been included in the energy balance of the electrons. The charging kinetics for the complex plasma comprising of uniformly dispersed dust particles, characterized by (i) uniform size and (ii) the Mathis, Rumpl, and Nordsieck power law of size distribution has been developed. Using appropriate expressions for the transport parameters based on the kinetic theory, the system of equations has been solved to investigate the parametric dependence of the complex plasma transport properties on the applied electric field and other plasma parameters; the results are graphically illustrated.

  12. Radiation transport phenomena and modeling - part A: Codes

    International Nuclear Information System (INIS)

    Lorence, L.J.

    1997-01-01

    The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped

  13. Transport phenomena in fuel cells : from microscale to macroscale

    Energy Technology Data Exchange (ETDEWEB)

    Djilali, N. [Victoria Univ., BC (Canada). Dept. of Mechanical Engineering]|[Victoria Univ., BC (Canada). Inst. for Integrated Energy Systems

    2006-07-01

    Proton Exchange Membrane (PEM) fuel cells rely on an array of thermofluid transport processes for the regulated supply of reactant gases and the removal of by-product heat and water. Flows are characterized by a broad range of length and time scales that take place in conjunction with reaction kinetics in a variety of regimes and structures. This paper examined some of the challenges related to computational fluid dynamics (CFD) modelling of PEM fuel cell transport phenomena. An overview of the main features, components and operation of PEM fuel cells was followed by a discussion of the various strategies used for component modelling of the electrolyte membrane; the gas diffusion layer; microporous layer; and flow channels. A review of integrated CFD models for PEM fuel cells included the coupling of electrochemical thermal and fluid transport with 3-D unit cell simulations; air-breathing micro-structured fuel cells; and stack level modelling. Physical models for modelling of transport at the micro-scale were also discussed. Results of the review indicated that the treatment of electrochemical reactions in a PEM fuel cell currently combines classical reaction kinetics with solutions procedures to resolve charged species transport, which may lead to thermodynamically inconsistent solutions for more complex systems. Proper representation of the surface coverage of all the chemical species at all reaction sites is needed, and secondary reactions such as platinum (Pt) dissolution and oxidation must be accounted for in order to model and understand degradation mechanisms in fuel cells. While progress has been made in CFD-based modelling of fuel cells, functional and predictive capabilities remain a challenge because of fundamental modelling and material characterization deficiencies in ionic and water transport in polymer membranes; 2-phase transport in porous gas diffusion electrodes and gas flow channels; inadequate macroscopic modelling and resolution of catalyst

  14. A Green function of neutron transport equation

    International Nuclear Information System (INIS)

    Simovic, R.

    1993-01-01

    In this paper the angularly dependent Green function of the neutron transport equation is derived analytically and approximately. By applying the analytical FDPN approximation up to eighth order, numerical values of the Green functions are obtained with the accuracy of six significant figures in the whole range of parameter c, angle cosine μ and distances x up to the ten optical lengths from the neutron source. (author)

  15. Heterogeneity effects in neutron transport computations

    International Nuclear Information System (INIS)

    Gelbard, E.M.

    1975-01-01

    A nuclear reactor is, generally, an intricate heterogeneous structure whose adjacent components may differ radically in their neutronic properties. The heterogeneities in the structure of the reactor complicate the work of the reactor analyst and tend to degrade the efficiency of the numerical methods used in reactor computations. Two types of heterogeneity effects are considered. First, certain singularities in the solution of the neutron transport equation, induced by heterogeneities, are briefly described. Second, the effect of heterogeneities on neutron leakage rates, and consequently on effective diffusion coefficients, are discussed. (5 figures) (U.S.)

  16. Study of a transportable neutron radiography system

    International Nuclear Information System (INIS)

    Souza, S.N.A. de.

    1991-05-01

    This work presents a study a transportable neutron radiography system for a 185 GBq 241 Am-Be (α, η) source with a neutron yield roughly 1,25 x 10 7 n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio of 14, for neutron fluxes up to 4,09 x 10 2 n.cm -2 .s -1 . Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10 2 n.cm -2 .s -1 and 2,65 x 10 2 n.cm -2 .s -1 , respectively. (author). 30 refs, 39 figs, 9 tabs

  17. Transportable type neutron level indicators

    International Nuclear Information System (INIS)

    Khatskevich, M.V.; Kalinin, O.V.; Moskovkin, V.N.; Molchanov, A.V.; Bobkov, A.D.; Rabotnov, Yu.A.

    1979-01-01

    Some peculiarities of designing level neutron converters (LNC) for portable indicators or level neutron relays are considered. The effect of the LNC geometry and other factors on measurement errors has been studied. Calibration results of the LNC with a neutron reflector and without it are presented. It is shown that the problem of level monitoring with the help of portable indicators can be solved practically for any volume, provided two LNC modifications with reflectors are available: the NPU-G modification with horizontal location of a counter for large volumes and the NPU-V with vertical location of a counter for lesser volumes. A possibility of perfecting LNC performances by shielding the counter with thermal neutron absorbers has been studied. The design of the NPU-V modification for the NIUP-2 level indicator is described. It is intended for tubes and cylinders 30-100 mm in diameter. Measurements carried out on different steel and aluminium vessels with a diameter ranging from 300 to 100 mm and a wall thickness of up to 16 mm with the help of the NPU-V and NPU-G modifications proved the efficiency of the LNC to control a variety of products (kerosine, gasoline, oils, acids, alkalis) [ru

  18. Lectures on neutron transport theory

    International Nuclear Information System (INIS)

    Benoist, P.

    1986-02-01

    This note is divided in two parts. In the first one the basis of transport theory, that is, the principal forms of the transport equation and the resulting theorems, are presented. The second part is particularly devoted to the applications of integral transport theory to reactor lattice problems [fr

  19. Basic physical phenomena, neutron production and scaling of the dense plasma focus

    International Nuclear Information System (INIS)

    Kaeppeler, H.J.

    This paper presents an attempt at establishing a model theory for the dense plasma focus in order to present a consistent interpretation of the basic physical phenomena leading to neutron production from both acceleration and thermal processes. To achieve this, the temporal history of the focus is divided into the compression of the plasma sheath, a qiescent and very dense phase with ensuing expansion, and an instable phase where the focus plasma is disrupted by instabilities. Finally, the decay of density, velocity and thermal fields is considered. Under the assumption that Io 2 /sigmaoRo 2 = const and to/Tc = const, scaling laws for plasma focus devices are derived. It is shown that while generally the neutron yield scales with the fourth power of maximum current, neutron production from thermal processes becomes increasingly important for large devices, while in the small devices neutron production from acceleration processes is by far predominant. (orig.) [de

  20. Uncertainty analysis of neutron transport calculation

    International Nuclear Information System (INIS)

    Oka, Y.; Furuta, K.; Kondo, S.

    1987-01-01

    A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

  1. Neutron radiography using a transportable superconducting cyclotron

    Energy Technology Data Exchange (ETDEWEB)

    Allen, D.A. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Hawkesworth, M.R. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Beynon, T.D. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Green, S. (School of Physics and Space Research, University of Birmingham, Birmingham, B15 2TT (United Kingdom)); Rogers, J.D. (Rolls-Royce, Derby (United Kingdom)); Allen, M.J. (Rolls-Royce, Derby (United Kingdom)); Plummer, H.C. (Rolls-Royce, MatEval, Derby (United Kingdom)); Boulding, N.J. (Oxford Instruments (United Kingdom)); Cox, M. (Oxford Instruments (United Kingdom)); McDougall, I. (Oxford Instruments (United Kingdom))

    1994-12-30

    A thermal neutron radiography system based on a compact 12 MeV superconducting proton cyclotron is described. Neutrons are generated using a thick beryllium target and moderated in high density polyethylene. Monte Carlo computer simulations have been used to model the neutron and photon transport in order to optimise the performance of the system. With proton beam currents in excess of 100 [mu]A, it can provide high thermal neutron fluxes with L/D ratios of between 50 and 300 for various applications. Both film and electronic imaging are used to produce radiographs. The electronic imaging system consists of a [sup 6]Li-loaded ZnS intensifier screen, and a low light CCD or SIT camera. High resolution images can be recorded and computer-controlled data processing, analysis and display are possible. ((orig.))

  2. 3D neutron transport modelization

    International Nuclear Information System (INIS)

    Warin, X.

    1996-12-01

    Some nodal methods to solve the transport equation in 3D are presented. Two nodal methods presented at an OCDE congress are described: a first one is a low degree one called RTN0; a second one is a high degree one called BDM1. The two methods can be made faster with a totally consistent DSA. Some results of parallelization show that: 98% of the time is spent in sweeps; transport sweeps are easily parallelized. (K.A.)

  3. 3D neutron transport modelization

    Energy Technology Data Exchange (ETDEWEB)

    Warin, X.

    1996-12-01

    Some nodal methods to solve the transport equation in 3D are presented. Two nodal methods presented at an OCDE congress are described: a first one is a low degree one called RTN0; a second one is a high degree one called BDM1. The two methods can be made faster with a totally consistent DSA. Some results of parallelization show that: 98% of the time is spent in sweeps; transport sweeps are easily parallelized. (K.A.). 10 refs.

  4. Coupled Transport Phenomena in the Opalinus Clay: Implications for Radionuclide Transport

    International Nuclear Information System (INIS)

    Soler, J.M.

    1999-09-01

    Coupled phenomena (thermal and chemical osmosis, hyperfiltration, coupled diffusion, thermal diffusion, thermal filtration, Dufour effect) may play an important role in fluid, solute and heat transport in clay-rich formations, such as the Opalinus Clay (OPA), which are being considered as potential hosts for radioactive waste repositories. In this study, the potential effects of coupled phenomena on radionuclide transport in the vicinity of a repository for vitrified high-level radioactive waste (HLW) and spent nuclear fuel (SF) hosted by the Opalinus Clay, at times equal to or greater than the expected lifetime of the waste canisters (about 1000 years), have been addressed. Firstly, estimates of the solute fluxes associated with chemical osmosis, hyperfiltration, thermal diffusion and thermal osmosis have been calculated. Available experimental data concerning coupled transport phenomena in compacted clays, and the hydrogeological and geochemical conditions to which the Opalinus Clay is subject, have been used for these estimates. These estimates suggest that thermal osmosis is the only coupled transport mechanism that could have a strong impact on solute and fluid transport in the vicinity of the repository. Secondly, estimates of the heat fluxes associated with thermal filtration and the Dufour effect in the vicinity of the repository have been calculated. The calculated heat fluxes are absolutely negligible compared to the heat flux caused by thermal conduction. As a further step to obtain additional insight into the effects of coupled phenomena on solute transport, the solute fluxes associated with advection, chemical diffusion, thermal and chemical osmosis, hyperfiltration and thermal diffusion have been incorporated into a simple one-dimensional transport equation. The analytical solution of this equation, with appropriate parameters, shows again that thermal osmosis is the only coupled transport mechanism that could have a strong effect on repository

  5. Coupled Transport Phenomena in the Opalinus Clay: Implications for Radionuclide Transport

    Energy Technology Data Exchange (ETDEWEB)

    Soler, J.M.

    1999-09-01

    Coupled phenomena (thermal and chemical osmosis, hyperfiltration, coupled diffusion, thermal diffusion, thermal filtration, Dufour effect) may play an important role in fluid, solute and heat transport in clay-rich formations, such as the Opalinus Clay (OPA), which are being considered as potential hosts for radioactive waste repositories. In this study, the potential effects of coupled phenomena on radionuclide transport in the vicinity of a repository for vitrified high-level radioactive waste (HLW) and spent nuclear fuel (SF) hosted by the Opalinus Clay, at times equal to or greater than the expected lifetime of the waste canisters (about 1000 years), have been addressed. Firstly, estimates of the solute fluxes associated with chemical osmosis, hyperfiltration, thermal diffusion and thermal osmosis have been calculated. Available experimental data concerning coupled transport phenomena in compacted clays, and the hydrogeological and geochemical conditions to which the Opalinus Clay is subject, have been used for these estimates. These estimates suggest that thermal osmosis is the only coupled transport mechanism that could have a strong impact on solute and fluid transport in the vicinity of the repository. Secondly, estimates of the heat fluxes associated with thermal filtration and the Dufour effect in the vicinity of the repository have been calculated. The calculated heat fluxes are absolutely negligible compared to the heat flux caused by thermal conduction. As a further step to obtain additional insight into the effects of coupled phenomena on solute transport, the solute fluxes associated with advection, chemical diffusion, thermal and chemical osmosis, hyperfiltration and thermal diffusion have been incorporated into a simple one-dimensional transport equation. The analytical solution of this equation, with appropriate parameters, shows again that thermal osmosis is the only coupled transport mechanism that could have a strong effect on repository

  6. Modeling of transport phenomena in concrete porous media.

    Science.gov (United States)

    Plecas, Ilija

    2014-02-01

    Two fundamental concerns must be addressed when attempting to isolate low-level waste in a disposal facility on land. The first concern is isolating the waste from water, or hydrologic isolation. The second is preventing movement of the radionuclides out of the disposal facility, or radionuclide migration. Particularly, we have investigated here the latter modified scenario. To assess the safety for disposal of radioactive waste-concrete composition, the leakage of 60Co from a waste composite into a surrounding fluid has been studied. Leakage tests were carried out by the original method, developed at the Vinča Institute. Transport phenomena involved in the leaching of a radioactive material from a cement composite matrix are investigated using three methods based on theoretical equations. These are: the diffusion equation for a plane source: an equation for diffusion coupled to a first-order equation, and an empirical method employing a polynomial equation. The results presented in this paper are from a 25-y mortar and concrete testing project that will influence the design choices for radioactive waste packaging for a future Serbian radioactive waste disposal center.

  7. Solving the equation of neutron transport

    International Nuclear Information System (INIS)

    Nasfi, Rim

    2009-01-01

    This work is devoted to the study of some numerical methods of resolution of the problem of transport of the neutrons. We started by introducing the equation integro-differential transport of the neutrons. Then we applied the finite element method traditional for stationary and nonstationary linear problems in 2D. A great part is reserved for the presentation of the mixed numerical diagram and mixed hybrid with two types of uniform grids: triangular and rectangular. Thereafter we treated some numerical examples by implementations in Matlab in order to test the convergence of each method. To finish, we had results of simulation by the Monte Carlo method on a problem of two-dimensional transport with an aim of comparing them with the results resulting from the finite element method mixed hybrids. Some remarks and prospects conclude this work.

  8. Neutronics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, David J.

    2018-11-11

    This paper discusses liquid-fuel molten salt reactors, how they will operate under normal, transient, and accident conditions, and the results of an expert elicitation to determine the corresponding neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC) to develop or identify modeling functionalities and tools required to carry out confirmatory analyses that examine the validity and accuracy of applicants’ calculations and help determine the margin of safety in plant design. NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures-of-merit (FoMs) (e.g., power distribution, fluence, kinetics parameters and reactivity). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs.

  9. Monte Carlo method for neutron transport problems

    International Nuclear Information System (INIS)

    Asaoka, Takumi

    1977-01-01

    Some methods for decreasing variances in Monte Carlo neutron transport calculations are presented together with the results of sample calculations. A general purpose neutron transport Monte Carlo code ''MORSE'' was used for the purpose. The first method discussed in this report is the method of statistical estimation. As an example of this method, the application of the coarse-mesh rebalance acceleration method to the criticality calculation of a cylindrical fast reactor is presented. Effective multiplication factor and its standard deviation are presented as a function of the number of histories and comparisons are made between the coarse-mesh rebalance method and the standard method. Five-group neutron fluxes at core center are also compared with the result of S4 calculation. The second method is the method of correlated sampling. This method was applied to the perturbation calculation of control rod worths in a fast critical assembly (FCA-V-3) Two methods of sampling (similar flight paths and identical flight paths) are tested and compared with experimental results. For every cases the experimental value lies within the standard deviation of the Monte Carlo calculations. The third method is the importance sampling. In this report a biased selection of particle flight directions discussed. This method was applied to the flux calculation in a spherical fast neutron system surrounded by a 10.16 cm iron reflector. Result-direction biasing, path-length stretching, and no biasing are compared with S8 calculation. (Aoki, K.)

  10. Multi-group neutron transport theory

    International Nuclear Information System (INIS)

    Zelazny, R.; Kuszell, A.

    1962-01-01

    Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr

  11. Prediction of transport phenomena in near and far field: interaction solid phase/fluid phase

    International Nuclear Information System (INIS)

    Mingarro, E.

    1995-01-01

    The prediction of transport phenomena in near and far field is presented in the present report. The study begins with the analysis of solid phases stability: solubility of storage waste: UO 2 and solubility of radionuclides the redox and sorption-desorption conditions are the last aspects studied to predict the transport phenomena

  12. Impact of Disorder on Spin Dependent Transport Phenomena

    KAUST Repository

    Saidaoui, Hamed

    2016-07-03

    The impact of the spin degree of freedom on the transport properties of electrons traveling through magnetic materials has been known since the pioneer work of Mott [1]. Since then it has been demonstrated that the spin angular momentum plays a key role in the scattering process of electrons in magnetic multilayers. This role has been emphasized by the discovery of the Giant Magnetoresistance in 1988 by Fert and Grunberg [2, 3]. Among the numerous applications and effects that emerged in mesoscopic devices two mechanisms have attracted our attention during the course of this thesis: the spin transfer torque and the spin Hall effects. The former consists in the transfer of the spin angular momentum from itinerant carriers to local magnetic moments [4]. This mechanism results in the current-driven magnetization switching and excitations, which has potential application in terms of magnetic data storage and non-volatile memories. The latter, spin Hall effect, is considered as well to be one of the most fascinating mechanisms in condensed matter physics due to its ability of generating non-equilibrium spin currents without the need for any magnetic materials. In fact the spin Hall effect relies only on the presence of the spin-orbit interaction in order to create an imbalance between the majority and minority spins. The objective of this thesis is to investigate the impact of disorder on spin dependent transport phenomena. To do so, we identified three classes of systems on which such disorder may have a dramatic influence: (i) antiferromagnetic materials, (ii) impurity-driven spin-orbit coupled systems and (iii) two dimensional semiconducting electron gases with Rashba spin-orbit coupling. Antiferromagnetic materials - We showed that in antiferromagnetic spin-valves, spin transfer torque is highly sensitive to disorder, which prevents its experimental observation. To solve this issue, we proposed to use either a tunnel barrier as a spacer or a local spin torque using

  13. Cosmic-ray neutron transport at a forest field site

    DEFF Research Database (Denmark)

    Andreasen, Mie; Jensen, Karsten Høgh; Desilets, Darin

    2017-01-01

    -ray neutron intensity is essential (e.g., the effect of vegetation, litter layer and soil type). In this study the environmental effect is examined by performing a sensitivity analysis using neutron transport modeling. We use a neutron transport model with various representations of the forest and different...

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

    Directory of Open Access Journals (Sweden)

    Jorge Pérez Mañes

    2014-01-01

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

  15. Analysis of cosmic ray neutron-induced single-event phenomena

    International Nuclear Information System (INIS)

    Tukamoto, Yasuyuki; Watanabe, Yukinobu; Nakashima, Hideki

    2003-01-01

    We have developed a database of cross sections for the n+ 28 Si reaction in the energy range between 2 MeV and 3 GeV in order to analyze single-event upset (SEU) phenomena induced by cosmic-ray neutrons in semiconductor memory devices. The data are applied to calculations of SEU cross sections using the Burst Generation Rate (BGR) model including two parameters, critical charge and effective depth. The calculated results are compared with measured SEU cross-sections for energies up to 160 MeV, and the reaction products that provide important effects on SEU are mainly investigated. (author)

  16. Progress in multidimensional neutron transport computation

    International Nuclear Information System (INIS)

    Lewis, E.E.

    1977-01-01

    The methods available for solution of the time-independent neutron transport problems arising in the analysis of nuclear systems are examined. The merits of deterministic and Monte Carlo methods are briefly compared. The capabilities of deterministic computational methods derived from the first-order form of the transport equation, from the second-order even-parity form of this equation, and from integral transport formulations are discussed in some detail. Emphasis is placed on the approaches for dealing with the related problems of computer memory requirements, computational cost, and achievable accuracy. Attention is directed to some areas where problems exist currently and where the need for further work appears to be particularly warranted

  17. Discrete elements method of neutron transport

    International Nuclear Information System (INIS)

    Mathews, K.A.

    1988-01-01

    In this paper a new neutron transport method, called discrete elements (L N ) is derived and compared to discrete ordinates methods, theoretically and by numerical experimentation. The discrete elements method is based on discretizing the Boltzmann equation over a set of elements of angle. The discrete elements method is shown to be more cost-effective than discrete ordinates, in terms of accuracy versus execution time and storage, for the cases tested. In a two-dimensional test case, a vacuum duct in a shield, the L N method is more consistently convergent toward a Monte Carlo benchmark solution

  18. Antisymmetrized molecular dynamics studies for exotic clustering phenomena in neutron-rich nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, M. [Hokkaido University, Department of Physics, Sapporo (Japan); Hokkaido University, Nuclear Reaction Data Centre, Faculty of Science, Sapporo (Japan); Suhara, T. [Matsue College of Technology, Matsue (Japan); Kanada-En' yo, Y. [Kyoto University, Department of Physics, Kyoto (Japan)

    2016-12-15

    We present a review of recent works on clustering phenomena in unstable nuclei studied by antisymmetrized molecular dynamics (AMD). The AMD studies in these decades have uncovered novel types of clustering phenomena brought about by the excess neutrons. Among them, this review focuses on the molecule-like structure of unstable nuclei. One of the earliest discussions on the clustering in unstable nuclei was made for neutron-rich Be and B isotopes. AMD calculations predicted that the ground state clustering is enhanced or reduced depending on the number of excess neutrons. Today, the experiments are confirming this prediction as the change of the proton radii. Behind this enhancement and reduction of the clustering, there are underlying shell effects called molecular and atomic orbits. These orbits form covalent and ionic bonding of the clusters analogous to the atomic molecules. It was found that this ''molecular-orbit picture'' reasonably explains the low-lying spectra of Be isotopes. The molecular-orbit picture is extended to other systems having parity asymmetric cluster cores and to the three cluster systems. O and Ne isotopes are the candidates of the former, while the 3α linear chains in C isotopes are the latter. For both subjects, many intensive studies are now in progress. We also pay a special attention to the observables which are the fingerprint of the clustering. In particular, we focus on the monopole and dipole transitions which are recently regarded as good probe for the clustering. We discuss how they have and will reveal the exotic clustering. (orig.)

  19. Center for low-gravity fluid mechanics and transport phenomena

    Science.gov (United States)

    Kassoy, D. R.; Sani, R. L.

    1991-01-01

    Research projects in several areas are discussed. Mass transport in vapor phase systems, droplet collisions and coalescence in microgravity, and rapid solidification of undercooled melts are discussed.

  20. The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

    2005-03-15

    A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well.

  1. The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

    International Nuclear Information System (INIS)

    Kim, K. S.; Ju, H. G.; Jeon, T. H. and others

    2005-03-01

    A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well

  2. Transition phenomena and thermal transport property in LHD plasmas with an electron internal transport barrier

    International Nuclear Information System (INIS)

    Shimozuma, T.; Kubo, S.; Idei, H.

    2005-01-01

    Two kinds of improved core confinement were observed during centrally focused Electron Cyclotron Heating (ECH) into plasmas sustained by Counter (CNTR) and Co Neutral Beam Injections (NBI) in the Large Helical Device (LHD). One shows transition phenomena to the high-electron-temperature state and has a clear electron Internal Transport Barrier (eITB) in CNTR NBI plasma. Another has no clear transition and no ECH power threshold, but shows a broad high temperature profiles with moderate temperature gradient, which indicates the improved core confinement with additional ECH in Co NBI plasma. The electron heat transport characteristics of these plasmas were directly investigated by using the heat pulse propagation excited by Modulated ECH (MECH). The difference of the features could be caused by the existence of the m/n=2/1 rational surface or island determined by the direction of NBI beam-driven current. (author)

  3. Transition phenomena and thermal transport properties in LHD plasmas with an electron internal transport barrier

    International Nuclear Information System (INIS)

    Shimozuma, T.; Kubo, S.; Idei, H.; Inagaki, S.; Tamura, N.; Tokuzawa, T.; Morisaki, T.; Watanabe, K.Y.; Ida, K.; Yamada, I.; Narihara, K.; Muto, S.; Yokoyama, M.; Yoshimura, Y.; Notake, T.; Ohkubo, K.; Seki, T.; Saito, K.; Kumazawa, R.; Mutoh, T.; Watari, T.; Komori, A.

    2005-01-01

    Two types of improved core confinement were observed during centrally focused electron cyclotron heating (ECH) into plasmas sustained by counter (CNTR) and Co neutral beam injections (NBI) in the Large Helical Device. The CNTR NBI plasma displayed transition phenomena to the high-electron-temperature state and had a clear electron internal transport barrier, while the Co NBI plasma did not show a clear transition or an ECH power threshold but showed broad high temperature profiles with moderate temperature gradient. This indicated that the Co NBI plasma with additional ECH also had an improved core confinement. The electron heat transport characteristics of these plasmas were directly investigated using heat pulse propagation excited by modulated ECH. These effects appear to be related to the m/n = 2/1 rational surface or the island induced by NBI beam-driven current

  4. Enhanced transport phenomena in CO2 sequestration and CO2 EOR

    NARCIS (Netherlands)

    Farajzadeh, R.

    2009-01-01

    The results of this thesis give insight into the (mass)-transfer during flow of gases, especially CO2, in various gas-liquid systems. A number of experiments was performed to investigate the transport phenomena through interfaces with and without surfactant monolayers. The observed phenomena have

  5. Diffusion and transport phenomena in a collisional magnetoplasma ...

    Indian Academy of Sciences (India)

    Boltzmann-transport equation is analytically solved for two-component magnetoplasma using Chapman-Enskog analysis to include collisional diffusion transport having anisotropies in both streaming velocity and temperature components. The modified collisional integrals are analytically solved with flux integrals and ...

  6. Asymptotic time dependent neutron transport in multidimensional systems

    International Nuclear Information System (INIS)

    Nagy, M.E.; Sawan, M.E.; Wassef, W.A.; El-Gueraly, L.A.

    1983-01-01

    A model which predicts the asymptotic time behavior of the neutron distribution in multi-dimensional systems is presented. The model is based on the kernel factorization method used for stationary neutron transport in a rectangular parallelepiped. The accuracy of diffusion theory in predicting the asymptotic time dependence is assessed. The use of neutron pulse experiments for predicting the diffusion parameters is also investigated

  7. Lab-scale tests on ISV vapor transport phenomena

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Gardner, B.M.

    1996-01-01

    In situ vitrification (ISV) is a promising technology for remediating buried waste sites and contaminated soil sites. However, concerns exist that low soil permeabilities may limit vapor transport away from the advancing melt front and cause a melt expulsion that breaches ISV containment. As a result, two ISV lab tests were conducted at the Idaho National Engineering Laboratory (INEL) using INEL soil (permeability: 10 -6 cm/s) and a low permeability (10 -10 cm/s) clay material. The clay test also had a ceramic tube inserted vertically through the center of the area being melted to provide one-dimensional data on vapor transport. Results confirm that low soil permeabilities can limit vapor transport away from the advancing ISV melt front. In addition, peak pressures inside the ceramic tube were significantly greater than those outside the tube, indicating the importance of horizontal vapor transport around the advancing ISV melt front

  8. Transport phenomena in sharply contrasting media with a diffusion barrier

    International Nuclear Information System (INIS)

    Dvoretskaya, O A; Kondratenko, P S

    2011-01-01

    Using the advection–diffusion equation, we analytically study contaminant transport in a sharply contrasting medium with a diffusion barrier due to localization of a contaminant source in a low-permeability medium. Anomalous diffusion behavior and a crossover between different transport regimes are observed. The diffusion barrier results in exponential attenuation of the source power, retardation of the contaminant plume growth and modification of the concentration distribution at large distances. (paper)

  9. Electron heat transport studies using transient phenomena in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Jacchia, A.; Angioni, C.; Manini, A.; Ryter, F.; Apostoliceanu, M.; Conway, G.; Fahrbach, H.-U.; Kirov, K.K.; Leuterer, F.; Reich, M.; Sutttrop, W.; Cirant, S.; Mantica, P.; De Luca, F.; Weiland, J.

    2005-01-01

    Experiments in tokamaks suggest that a critical gradient length may cause the resilient behavior of T e profiles, in the absence of ITBs. This agrees in general with ITG/TEM turbulence physics. Experiments in ASDEX Upgrade using modulation techniques with ECH and/or cold pulses demonstrate the existence of a threshold in R/L Te when T e >T i and T e ≤T i . For T e >T i linear stability analyses indicate that electron heat transport is dominated by TEM modes. They agree in the value of the threshold (both T e and n e ) and for the electron heat transport increase above the threshold. The stabilization of TEM modes by collisions yielded by gyro-kinetic calculations, which suggests a transition from TEM to ITG dominated transport at high collisionality, is experimentally demonstrated by comparing heat pulse and steady-state diffusivities. For the T e ∼T i discharges above the threshold the resilience, normalized by T e 3/2 , is similar to that of the TEM dominated cases, despite very different conditions. The heat pinch predicted by fluid modeling of ITG/TEM turbulence is investigated by perturbative transport in off-axis ECH-heated discharges. (author)

  10. Neutron transport on the connection machine

    International Nuclear Information System (INIS)

    Robin, F.

    1991-12-01

    Monte Carlo methods are heavily used at CEA and account for a a large part of the total CPU time of industrial codes. In the present work (done in the frame of the Parallel Computing Project of the CEL-V Applied Mathematics Department) we study and implement on the Connection Machine an optimised Monte Carlo algorithm for solving the neutron transport equation. This allows us to investigate the suitability of such an architecture for this kind of problem. This report describes the chosen methodology, the algorithm and its performances. We found that programming the CM-2 in CM Fortran is relatively easy and we got interesting performances as, on a 16 k, CM-2 they are the same level as those obtained on one processor of a CRAY X-MP with a well optimized vector code

  11. Concentration polarization: Electrodeposition and transport phenomena at overlimiting current

    DEFF Research Database (Denmark)

    Nielsen, Christoffer Peder

    a numerical sharp-interface model describing the electrode growth. This model differs from the established phase-field models, in that it is applicable at overlimiting current and implements electrode reactions in a consistent way. Comparison of the sharp-interface model to the results of the stability...... methods. The initial study concerns a fundamental problem in the study of concentration polarization at overlimiting current, namely the emergence of an extended space-charge region near the ion-selective interface. Based on the so-called quasi-uniform charge density assumption (QCD), we develop...... this coupled chemical and transport effect using two simple models of the reaction kinetics. The principal investigations are performed using numerical simulations, but in addition we derive an analytical model for the transport in the system. The analytical model reveals an important link between the current...

  12. Experimental transport phenomena and optimization strategies for thermoelectrics

    Energy Technology Data Exchange (ETDEWEB)

    Ehrlich, A C; Gillespie, D J

    1997-07-01

    When a new and promising thermoelectric material is discovered, an effort is undertaken to improve its figure of merit. If the effort is to be more efficient than one of trial and error with perhaps some rule of thumb guidance then it is important to be able to make the connection between experimental data and the underlying material characteristics, electronic and phononic, that influence the figure of merit. Transport and fermiology experimental data can be used to evaluate these material characteristics and thus establish trends as a function of some controllable parameter, such as composition. In this paper some of the generic-materials characteristics, generally believed to be required for a high figure of merit, will be discussed in terms of the experimental approach to their evaluation and optimization. Transport and fermiology experiments will be emphasized and both will be outlined in what they can reveal and what can be obscured by the simplifying assumptions generally used in their interpretation.

  13. On nonequilibrium many-body systems V: ultrafast transport phenomena

    International Nuclear Information System (INIS)

    Freire, V.N.; Vasconcellos, A.R.; Luzzi, R.

    1989-01-01

    The monequilibrium statistical operator method and its accompanying nonlinear quantum transport theory, are used to perform an analytical study of the ultrafast mobility transient of central-valley photoinjected carriers in direct-gap polar semiconductors. Expressions for the time-resolved mobility of the hot carriers are derived. A brief discussion of the carriers' diffusion coefficient is done. (A.C.A.S.) [pt

  14. Barodiffusion phenomena at active transport of na+ and K+ ions through the cell membrane

    International Nuclear Information System (INIS)

    Khrapijchuk, G.V.; Chalyi, A.V.; Nurishchenko, N.Je.

    2010-01-01

    The influence of ultrasound as the significant motive force of barodiffusion phenomena at the processes of active transport of Na + and K + ions through the cell membrane is considered. The dependence of membrane potential is theoretically estimated at active transport of natrium and potassium ions on the ultrasound intensity and pressure overfall between external and internal medium of the cell.

  15. Spin-transport-phenomena in metals, semiconductors, and insulators

    Energy Technology Data Exchange (ETDEWEB)

    Althammer, Matthias Klaus

    2012-07-19

    Assuming that one could deterministically inject, transport, manipulate, store and detect spin information in solid state devices, the well-established concepts of charge-based electronics could be transferred to the spin realm. This thesis explores the injection, transport, manipulation and storage of spin information in metallic conductors, semiconductors, as well as electrical insulators. On the one hand, we explore the spin-dependent properties of semiconducting zinc oxide thin films deposited via laser-molecular beam epitaxy (laser-MBE). After demonstrating that the zinc oxide films fabricated during this thesis have excellent structural, electrical, and optical properties, we investigate the spin-related properties by optical pump/probe, electrical injection/optical detection, and all electrical spin valve-based experiments. The two key results from these experiments are: (i) Long-lived spin states with spin dephasing times of 10 ns at 10 K related to donor bound excitons can be optically addressed. (ii) The spin dephasing times relevant for electrical transport-based experiments are {<=} 2 ns at 10 K and are correlated with structural quality. On the other hand we focus on two topics of current scientific interest: the comparison of the magnetoresistance to the magnetothermopower of conducting ferromagnets, and the investigation of pure spin currents generated in ferromagnetic insulator/normal metal hybrid structures. We investigate the magnetoresistance and magnetothermopower of gallium manganese arsenide and Heusler thin films as a function of external magnetic field orientation. Using a series expansion of the resistivity and Seebeck tensors and the inherent symmetry of the sample's crystal structure, we show that a full quantitative extraction of the transport tensors from such experiments is possible. Regarding the spin currents in ferromagnetic insulator/normal metal hybrid structures we studied the spin mixing conductance in yttrium iron garnet

  16. Gas generation phenomena in radioactive waste transportation packaging

    International Nuclear Information System (INIS)

    Nigrey, P.J.

    1998-01-01

    The interaction of radiation from radioactive materials with the waste matrix can lead to the deterioration of the waste form resulting in the possible of gaseous species. Depending on the type and characteristics of the radiation source, the generation of hydrogen may predominate. Since the interaction of alpha particles with the waste form results in significant energy transfer, other gases such as carbon oxides, methane, nitrogen oxides, oxygen, water, and helium are possible. The type of gases produced from the waste forms is determined by the mechanisms involved in the waste degradation. For transuranic wastes, the identified degradation mechanisms are reported to be caused by radiolysis, thermal decomposition or dewatering, chemical corrosion, and bacterial action. While all these mechanisms may be responsible for the building of gases during the storage of wastes, radiolysis and thermal decomposition appear to be main contributors during waste transport operations. (authors)

  17. Generic programming for deterministic neutron transport codes

    International Nuclear Information System (INIS)

    Plagne, L.; Poncot, A.

    2005-01-01

    This paper discusses the implementation of neutron transport codes via generic programming techniques. Two different Boltzmann equation approximations have been implemented, namely the Sn and SPn methods. This implementation experiment shows that generic programming allows us to improve maintainability and readability of source codes with no performance penalties compared to classical approaches. In the present implementation, matrices and vectors as well as linear algebra algorithms are treated separately from the rest of source code and gathered in a tool library called 'Generic Linear Algebra Solver System' (GLASS). Such a code architecture, based on a linear algebra library, allows us to separate the three different scientific fields involved in transport codes design: numerical analysis, reactor physics and computer science. Our library handles matrices with optional storage policies and thus applies both to Sn code, where the matrix elements are computed on the fly, and to SPn code where stored matrices are used. Thus, using GLASS allows us to share a large fraction of source code between Sn and SPn implementations. Moreover, the GLASS high level of abstraction allows the writing of numerical algorithms in a form which is very close to their textbook descriptions. Hence the GLASS algorithms collection, disconnected from computer science considerations (e.g. storage policy), is very easy to read, to maintain and to extend. (authors)

  18. Electric Transport Phenomena of Nanocomposite Organic Polymer Thin Films

    Science.gov (United States)

    Jira, Nicholas C.; Sabirianov, Ildar; Ilie, Carolina C.

    We discuss herein the nanocomposite organic thin film diodes for the use of plasmonic solar cells. This experimental work follows the theoretical calculations done for plasmonic solar cells using the MNPBEM toolbox for MatLab. These calculations include dispersion curves and amount of light scattering cross sections for different metallic nanoparticles. This study gives us clear ideas on what to expect from different metals, allowing us to make the best choice on what to use to obtain the best results. One specific technique for light trapping in thin films solar cells utilizes metal nanoparticles on the surface of the semiconductor. The characteristics of the metal, semiconductor interface allows for light to be guided in between them causing it to be scattered, allowing for more chances of absorption. The samples were fabricated using organic thin films made from polymers and metallic nanoparticles, more specifically Poly(1-vinylpyrrolidone-co-2-dimethylaminoethyl methacrylate) copolymer and silver or gold nanoparticles. The two fabrication methods applied include spin coating and Langmuir-Blodgett technique. The transport properties are obtained by analyzing the I-V curves. We will also discuss the resistance, resistivity, conductance, density of charge carriers. SUNY Oswego SCAC Grant.

  19. Comparison of neutron transport calculations with NRC test results

    International Nuclear Information System (INIS)

    Koban, J.; Hofmann, W.

    1981-02-01

    For an exactly defined reactor arrangement (PCA = Pool Critical Assembly) neutron fluxes, neutron spectra and reaction rates for several neutron detectors were calculated by means of one and two dimensional transport codes. An international comparison proved the methods applied at KWU to be adequate. There were difficulties, however, in considering the three dimensions of the assembly which result mainly from its small dimension. This fact applies to all participants who didn't use three dimensional codes. (orig.) [de

  20. Concise four-vector scheme for neutron transport calculations

    International Nuclear Information System (INIS)

    Kulacsy, K.; Lukacs, B.; Racz, A.

    1995-01-01

    An explicit Riemannian geometrical form or the vectorial Neutron Streaming Term is presented. The method applies the full Riemannian technique of general covariance. There are cases when the symmetry of the neutron flux must be smaller than that of the arrangement. However, in coordinate space there are always solutions of the Neutron Transport Equation as symmetric as the arrangement, if the latter's symmetry is at least an affine collineation of the Euclidian 3-space. (author). 7 refs

  1. Fluid Physical and Transport Phenomena Studies aboard the International Space Station: Planned Experiments

    Science.gov (United States)

    Singh, Bhim S.

    1999-01-01

    This paper provides an overview of the microgravity fluid physics and transport phenomena experiments planned for the International Spare Station. NASA's Office of Life and Microgravity Science and Applications has established a world-class research program in fluid physics and transport phenomena. This program combines the vast expertise of the world research community with NASA's unique microgravity facilities with the objectives of gaining new insight into fluid phenomena by removing the confounding effect of gravity. Due to its criticality to many terrestrial and space-based processes and phenomena, fluid physics and transport phenomena play a central role in the NASA's Microgravity Program. Through widely publicized research announcement and well established peer-reviews, the program has been able to attract a number of world-class researchers and acquired a critical mass of investigations that is now adding rapidly to this field. Currently there arc a total of 106 ground-based and 20 candidate flight principal investigators conducting research in four major thrust areas in the program: complex flows, multiphase flow and phase change, interfacial phenomena, and dynamics and instabilities. The International Space Station (ISS) to be launched in 1998, provides the microgravity research community with a unprecedented opportunity to conduct long-duration microgravity experiments which can be controlled and operated from the Principal Investigators' own laboratory. Frequent planned shuttle flights to the Station will provide opportunities to conduct many more experiments than were previously possible. NASA Lewis Research Center is in the process of designing a Fluids and Combustion Facility (FCF) to be located in the Laboratory Module of the ISS that will not only accommodate multiple users but, allow a broad range of fluid physics and transport phenomena experiments to be conducted in a cost effective manner.

  2. Magnetic field devices for neutron spin transport and manipulation in precise neutron spin rotation measurements

    Energy Technology Data Exchange (ETDEWEB)

    Maldonado-Velázquez, M. [Posgrado en Ciencias Físicas, Universidad Nacional Autónoma de México, 04510 (Mexico); Barrón-Palos, L., E-mail: libertad@fisica.unam.mx [Instituto de Física, Universidad Nacional Autónoma de México, Apartado Postal 20-364, 01000 (Mexico); Crawford, C. [University of Kentucky, Lexington, KY 40506 (United States); Snow, W.M. [Indiana University, Bloomington, IN 47405 (United States)

    2017-05-11

    The neutron spin is a critical degree of freedom for many precision measurements using low-energy neutrons. Fundamental symmetries and interactions can be studied using polarized neutrons. Parity-violation (PV) in the hadronic weak interaction and the search for exotic forces that depend on the relative spin and velocity, are two questions of fundamental physics that can be studied via the neutron spin rotations that arise from the interaction of polarized cold neutrons and unpolarized matter. The Neutron Spin Rotation (NSR) collaboration developed a neutron polarimeter, capable of determining neutron spin rotations of the order of 10{sup −7} rad per meter of traversed material. This paper describes two key components of the NSR apparatus, responsible for the transport and manipulation of the spin of the neutrons before and after the target region, which is surrounded by magnetic shielding and where residual magnetic fields need to be below 100 μG. These magnetic field devices, called input and output coils, provide the magnetic field for adiabatic transport of the neutron spin in the regions outside the magnetic shielding while producing a sharp nonadiabatic transition of the neutron spin when entering/exiting the low-magnetic-field region. In addition, the coils are self contained, forcing the return magnetic flux into a compact region of space to minimize fringe fields outside. The design of the input and output coils is based on the magnetic scalar potential method.

  3. Neutron transport in Eulerian coordinates with bulk material motion

    Energy Technology Data Exchange (ETDEWEB)

    Baker, Randal S., E-mail: rsb@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Dahl, Jon A., E-mail: dahl@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Fichtl, Erin J., E-mail: efichtl@lanl.gov [Los Alamos National Laboratory, Computational Physics Group, Los Alamos, NM (United States); Morel, Jim E., E-mail: morel@tamu.edu [Department of Nuclear Engineering, Texas A& M University, College Station, TX (United States)

    2015-12-15

    A consistent, numerically stable algorithm for the solution of the neutron transport equation in the presence of a moving material background is presented for one-dimensional spherical geometry. Manufactured solutions are used to demonstrate the correctness and stability of our numerical algorithm. The importance of including moving material corrections is shown for the r-process in proto-neutron stars.

  4. Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

    International Nuclear Information System (INIS)

    Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

    1989-01-01

    Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

  5. A random walk approach to stochastic neutron transport

    International Nuclear Information System (INIS)

    Mulatier, Clelia de

    2015-01-01

    One of the key goals of nuclear reactor physics is to determine the distribution of the neutron population within a reactor core. This population indeed fluctuates due to the stochastic nature of the interactions of the neutrons with the nuclei of the surrounding medium: scattering, emission of neutrons from fission events and capture by nuclear absorption. Due to these physical mechanisms, the stochastic process performed by neutrons is a branching random walk. For most applications, the neutron population considered is very large, and all physical observables related to its behaviour, such as the heat production due to fissions, are well characterised by their average values. Generally, these mean quantities are governed by the classical neutron transport equation, called linear Boltzmann equation. During my PhD, using tools from branching random walks and anomalous diffusion, I have tackled two aspects of neutron transport that cannot be approached by the linear Boltzmann equation. First, thanks to the Feynman-Kac backward formalism, I have characterised the phenomenon of 'neutron clustering' that has been highlighted for low-density configuration of neutrons and results from strong fluctuations in space and time of the neutron population. Then, I focused on several properties of anomalous (non-exponential) transport, that can model neutron transport in strongly heterogeneous and disordered media, such as pebble-bed reactors. One of the novel aspects of this work is that problems are treated in the presence of boundaries. Indeed, even though real systems are finite (confined geometries), most of previously existing results were obtained for infinite systems. (author) [fr

  6. Tokamak fuelling with pellets: Effect of transport phenomena on the injection requirements

    International Nuclear Information System (INIS)

    Lengyel, L.L.

    1979-01-01

    Results of calculations on pellet-plasma interaction that take into account transport phenomena inherent in tokamak plasmas are analyzed. It is shown that the results obtained by different authors on the optimum pellet penetration depth and required pellet injection frequencies, which are partly contradictory, can be explained by means of the different transport processes taken into account or neglected in the calculations concerned. (orig.)

  7. Development of instrumentation in the transport phenomena research in thermal equipment

    International Nuclear Information System (INIS)

    Carvalho Tofani, P. de; Ladeira, L.C.D.

    1983-11-01

    The results obtained from the effort on the acquisition of know-how in experimental reactor thermal during the last years, through the approach of relevant aspects of basic research on transport phenomena applicable to nuclear reactor analysis and conventional thermal equipment based in the simultaneous development of instrumentation and experimental methods are presented. (E.G.) [pt

  8. Transport phenomena through porous screens and openings : from theory to greenhouse practice

    NARCIS (Netherlands)

    Miguel, A.A.F.

    1998-01-01

    The study of transport phenomena in multi-zone enclosures with permeable boundaries is fundamental for indoor climate control management. In this study, aspects concerning the air exchange through porous screens and openings, and heat transfer between the enclosure surface and inside air,

  9. Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem

    International Nuclear Information System (INIS)

    William Charlton

    2007-01-01

    Neutron radiography and computed tomography are commonly used techniques to non-destructively examine materials. Tomography refers to the cross-sectional imaging of an object from either transmission or reflection data collected by illuminating the object from many different directions

  10. A finite element method for neutron transport

    International Nuclear Information System (INIS)

    Ackroyd, R.T.

    1978-01-01

    A variational treatment of the finite element method for neutron transport is given based on a version of the even-parity Boltzmann equation which does not assume that the differential scattering cross-section has a spherical harmonic expansion. The theory of minimum and maximum principles is based on the Cauchy-Schwartz equality and the properties of a leakage operator G and a removal operator C. For systems with extraneous sources, two maximum and one minimum principles are given in boundary free form, to ease finite element computations. The global error of an approximate variational solution is given, the relationship of one the maximum principles to the method of least squares is shown, and the way in which approximate solutions converge locally to the exact solution is established. A method for constructing local error bounds is given, based on the connection between the variational method and the method of the hypercircle. The source iteration technique and a maximum principle for a system with extraneous sources suggests a functional for a variational principle for a self-sustaining system. The principle gives, as a consequence of the properties of G and C, an upper bound to the lowest eigenvalue. A related functional can be used to determine both upper and lower bounds for the lowest eigenvalue from an inspection of any approximate solution for the lowest eigenfunction. The basis for the finite element is presented in a general form so that two modes of exploitation can be undertaken readily. The model can be in phase space, with positional and directional co-ordinates defining points of the model, or it can be restricted to the positional co-ordinates and an expansion in orthogonal functions used for the directional co-ordinates. Suitable sets of functions are spherical harmonics and Walsh functions. The latter set is appropriate if a discrete direction representation of the angular flux is required. (author)

  11. Ordering phenomena in FeCo-films and Fe/Cr-multilayers: an X-ray and neutron scattering study

    Energy Technology Data Exchange (ETDEWEB)

    Nickel, B.

    2001-07-01

    The following topics are covered: critical phenomena in thin films, critical adsorption, finite size scaling, FeCo Ising model, kinematical scattering theory for thin films, FeCo thin films, growth and characterisation of single crystal FeCo thin films, X-ray study of ordering in FeCo films, antiferromagnetic coupling in Fe/Cr multilayers, neutron scattering on Fe/Cr multilayers (WL)

  12. Scanning transient current study of the I-V stabilization phenomena in silicon detectors irradiated by fast neutrons

    International Nuclear Information System (INIS)

    Eremin, V.; Verbitskaya, E.; Sidorov, A.; Fretwurst, E.; Lindstrom, G.

    1996-03-01

    Investigation of the I-V stabilization phenomena in neutron irradiated silicon detectors has been carried out using scanning transient current technique (STCT) on non-irradiated PP + -p-n + detectors. The PP + -p-n + detectors were used to simulate the PP + -n-n + detectors irradiated beyond the space charge sign inversion (SCSI). Two mechanisms partially responsible for the I- V stabilization have been identified

  13. Numerical simulation of the transport phenomena due to sudden heating in porous media

    Energy Technology Data Exchange (ETDEWEB)

    Lei, S.Y.; Zheng, G.Y.; Wang, B.X.; Yang, R.G.; Xia, C.M.

    1997-07-01

    Such process as wet porous media suddenly heated by hot fluids frequently occurs in nature and in industrial applications. The three-variable simulation model was developed to predict violent transport phenomena due to sudden heating in porous media. Two sets of independent variables were applied to different regions in porous media in the simulation. For the wet zone, temperature, wet saturation and air pressure were used as the independent variables. For the dry zone, the independent variables were temperature, vapor pressure and air pressure. The model simulated two complicated transport processes in wet unsaturated porous media which is suddenly heated by melting metal or boiling water. The effect of the gas pressure is also investigated on the overall transport phenomena.

  14. Lattice Boltzmann modeling of transport phenomena in fuel cells and flow batteries

    Science.gov (United States)

    Xu, Ao; Shyy, Wei; Zhao, Tianshou

    2017-06-01

    Fuel cells and flow batteries are promising technologies to address climate change and air pollution problems. An understanding of the complex multiscale and multiphysics transport phenomena occurring in these electrochemical systems requires powerful numerical tools. Over the past decades, the lattice Boltzmann (LB) method has attracted broad interest in the computational fluid dynamics and the numerical heat transfer communities, primarily due to its kinetic nature making it appropriate for modeling complex multiphase transport phenomena. More importantly, the LB method fits well with parallel computing due to its locality feature, which is required for large-scale engineering applications. In this article, we review the LB method for gas-liquid two-phase flows, coupled fluid flow and mass transport in porous media, and particulate flows. Examples of applications are provided in fuel cells and flow batteries. Further developments of the LB method are also outlined.

  15. Spin-related transport phenomena in HgTe-based quantum well structures

    International Nuclear Information System (INIS)

    Koenig, Markus

    2007-12-01

    Within the scope of this thesis, spin related transport phenomena have been investigated in HgTe/Hg 0.3 Cd 0.7 Te quantum well structures. In our experiments, the existence of the quantum spin Hall (QSH) state was successfully demonstrated for the first time and the presented results provide clear evidence for the charge transport properties of the QSH state. Our experiments provide the first direct observation of the Aharonov-Casher (AC) effect in semiconductor structures. In conclusion, HgTe quantum well structures have proven to be an excellent template for studying spin-related transport phenomena: The QSH relies on the peculiar band structure of the material and the existence of both the spin Hall effect and the AC effect is a consequence of the substantial spin-orbit interaction. (orig.)

  16. Spin-related transport phenomena in HgTe-based quantum well structures

    Energy Technology Data Exchange (ETDEWEB)

    Koenig, Markus

    2007-12-15

    Within the scope of this thesis, spin related transport phenomena have been investigated in HgTe/Hg{sub 0.3}Cd{sub 0.7}Te quantum well structures. In our experiments, the existence of the quantum spin Hall (QSH) state was successfully demonstrated for the first time and the presented results provide clear evidence for the charge transport properties of the QSH state. Our experiments provide the first direct observation of the Aharonov-Casher (AC) effect in semiconductor structures. In conclusion, HgTe quantum well structures have proven to be an excellent template for studying spin-related transport phenomena: The QSH relies on the peculiar band structure of the material and the existence of both the spin Hall effect and the AC effect is a consequence of the substantial spin-orbit interaction. (orig.)

  17. Transport coefficients in superfluid neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Tolos, Laura [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Frankfurt Institute for Advances Studies. Johann Wolfgang Goethe University, Ruth-Moufang-Str. 1, 60438 Frankfurt am Main (Germany); Manuel, Cristina [Instituto de Ciencias del Espacio (IEEC/CSIC) Campus Universitat Autònoma de Barcelona, Facultat de Ciències, Torre C5, E-08193 Bellaterra (Barcelona) (Spain); Sarkar, Sreemoyee [Tata Institute of Fundamental Research, Homi Bhaba Road, Mumbai-400005 (India); Tarrus, Jaume [Physik Department, Technische Universität München, D-85748 Garching (Germany)

    2016-01-22

    We study the shear and bulk viscosity coefficients as well as the thermal conductivity as arising from the collisions among phonons in superfluid neutron stars. We use effective field theory techniques to extract the allowed phonon collisional processes, written as a function of the equation of state and the gap of the system. The shear viscosity due to phonon scattering is compared to calculations of that coming from electron collisions. We also comment on the possible consequences for r-mode damping in superfluid neutron stars. Moreover, we find that phonon collisions give the leading contribution to the bulk viscosities in the core of the neutron stars. We finally obtain a temperature-independent thermal conductivity from phonon collisions and compare it with the electron-muon thermal conductivity in superfluid neutron stars.

  18. Neutron transport from targets to moderators

    International Nuclear Information System (INIS)

    Taylor, A.D.

    1981-06-01

    By appropriately choosing parameters such as temperature, decoupler, thickness and effective size it is possible to tailor the moderators of a pulsed spallation neutron source in such a way that the different characteristics regarding time structure and spectral distribution as requested for the different instruments can be met very closely. This enables a unique flexibility in the design of neutron spectrometers to be used at such a source. (author)

  19. Center for Electrocatalysis, Transport Phenomena, and Materials (CETM) for Innovative Energy Storage - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Soloveichik, Grigorii [GE Global Research, Niskayuna, New York (United States)

    2015-11-30

    EFRC vision. The direct use of organic hydrides in fuel cells as virtual hydrogen carriers that generate stable organic molecules, protons, and electrons upon electro-oxidation and can be electrochemically charged by re-hydrogenating the oxidized carrier was the major focus of the Center for Electrocatalysis, Transport Phenomena and Materials for Innovative Energy Storage (EFRC-ETM). Compared to a hydrogen-on-demand design that includes thermal decomposition of organic hydrides in a catalytic reactor, the proposed approach is much simpler and does not require additional dehydrogenation catalysts or heat exchangers. Further, this approach utilizes the advantages of a flow battery (i.e., separation of power and energy, ease of transport and storage of liquid fuels) with fuels that have system energy densities similar to current hydrogen PEM fuel cells. EFRC challenges. Two major EFRC challenges were electrocatalysis and transport phenomena. The electrocatalysis challenge addresses fundamental processes which occur at a single molecular catalyst (microscopic level) and involve electron and proton transfer between the hydrogen rich and hydrogen depleted forms of organic liquid fuel and the catalyst. To form stable, non-radical dehydrogenation products from the organic liquid fuel, it is necessary to ensure fast transport of at least two electrons and two protons (per double bond formation). The same is true for the reverse hydrogenation reaction. The transport phenomena challenge addresses transport of electrons to/from the electrocatalyst and the current collector as well as protons across the polymer membrane. Additionally it addresses prevention of organic liquid fuel, water and oxygen transport through the PEM. In this challenge, the transport of protons or molecules involves multiple sites or a continuum (macroscopic level) and water serves as a proton conducting medium for the majority of known sulfonic acid based PEMs. Proton transfer in the presence of

  20. Effect of granulation of geological samples in neutron transport measurements

    International Nuclear Information System (INIS)

    Woznicka, Urszula; Drozdowicz, Krzysztof; Gabanska, Barbara; Krynicka, Ewa; Igielski, Andrzej

    2001-01-01

    The thermal neutron absorption cross section is one of the parameters describing the transport of thermal neutrons in a medium. Theoretical descriptions and experiments which determine the absorption cross section have a wide literature for homogeneous media. The situation comes true e.g. for fluids or amorphous solids. There are many other media which should be treated as heterogeneous. Among others - geological materials. The material heterogeneity for the thermal neutron transport in a considered volume is understood here as an existence of many small regions which differ significantly in their macroscopic neutron diffusion parameters (defined by the absorption and transport cross sections). The final difference, which influences the neutron transport, comes from a combination of the absolute differences between the parameters and of sizes of regions (related to the neutron mean free paths). A rock can be naturally heterogeneous in the above meaning. Besides, it can happen that a preparation of the rock sample for a neutron measurement can increase its natural heterogeneity. (For example, when the rock material is crushed and the measured sample consists of the obtained grains). The question is which granulation is allowed to treat the sample material as still homogeneous, and from which size of the rock grains we have to consider a two-component medium. It has been experimentally proved that the effective absorption of thermal neutrons in a heterogeneous two-component material can significantly differ from the absorption in a homogeneous one which consists of the same elements. The final effect is dependent on a few factors: the macroscopic absorption cross sections of the components, their total mass contributions, and the size of the grains. The ratio of the effective absorption cross section of the heterogeneous material to the cross section of the equivalent homogeneous, is a measure of the heterogeneity effect on the thermal neutron absorption

  1. Final report, BWR drywell debris transport Phenomena Identification and Ranking Tables (PIRTs)

    International Nuclear Information System (INIS)

    Wilson, G.E.; Boyack, B.E.; Leonard, M.T.; Williams, K.A.; Wolf, L.T.

    1997-09-01

    The Nuclear Regulatory Commission has issued a Regulatory Bulletin and accompanying Regulatory Guide (1.82, Rev. 2) which requires licensees of boiling water reactors to develop a specific plan of action (including hardware backfits, if necessary) to preclude the possibility of early emergency core cooling system strainer blockage following a postulated loss-of-coolant-accident. The postulated mechanism for strainer blockage is destruction of piping insulation in the vicinity of the break and subsequent transport of fragmented insulation to the wetwell. In the absence of more definitive information, the Regulatory Guide recommends that licensees assume a drywell debris transport fraction of 1.0. Accordingly, the Nuclear Regulatory Commission initiated research focused toward developing a technical basis to provide insights useful to regulatory oversight of licensee submittals associated with resolution of the postulated strainer blockage issue. Part of this program was directed towards experimental and analytical research leading to a more realistic specification of the debris transport through the drywell to the wetwell. To help focus this development into a cost effective effort, a panel, with broad based knowledge and experience, was formed to address the relative importance of the various phenomena that can be expected in plant response to postulated accidents that may produce strainer blockage. The resulting phenomena identification and ranking tables reported herein were used to help guide research. The phenomena occurring in boiling water reactors drywells was the specific focus of the panel, although supporting experimental data and calculations of debris transport fractions were considered

  2. Neutron transport from targets to moderators

    International Nuclear Information System (INIS)

    Taylor, A.D.

    1980-01-01

    The title of this meeting is 'Targets for Neutron Beam Spallation Sources', but so far all the emphasis in the talks has been on how to produce the fast neutron flux. I would like to stress that that is just the beginning of the story. What we are required to produce are beams of thermal and epithermal neutrons with time and spectral characteristics tailored to the instrumental requirements. The real source of our neutrons is not uranium arrays or thorium cylinders but a small volume of hydrogenous material, some 10 x 10 x 5 cm 3 . This is really what the whole thing is about - the target produces a copious field of fast neutrons, but if we fail to moderate them with the right energy and time characteristics, we will not match to what is happening downstream. In this talk, I am going to deal specifically with what we have done for SNS to optimise the target-moderator-reflector and decoupler system in this respect. (orig.)

  3. Neutron and gamma-ray transport experiments in liquid air

    International Nuclear Information System (INIS)

    Farley, W.E.

    1976-01-01

    Accurate estimates of neutron and gamma radiations from a nuclear explosion and their subsequent transport through the atmosphere are vital to nuclear-weapon employment studies: i.e., for determining safety radii for aircraft crews, casualty and collateral-damage risk radii for tactical weapons, and the kill range from a high-yield defensive burst for a maneuvering reentry vehicle. Radiation transport codes, such as the Laboratory's TARTNP, are used to calculate neutron and gamma fluences. Experiments have been performed to check and update these codes. Recently, a 1.3-m-radius liquid-air (21 percent oxygen) sphere, with a pulsed source of 14-MeV neutrons at its center, was used to measure the fluence and spectra of emerging neutrons and secondary gamma rays. Comparison of measured radiation dose with TARTNP showed agreement within 10 percent

  4. The isotope density inverse problem in multigroup neutron transport

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1981-01-01

    The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms of the flux in a group - density of an element spatial correlation, leads to a set of integral functionals for the densities weighted by the group fluxes. Some methods of approximation to make the problem uniquently solvable are proposed. Particularly P 0 angular flux information and the spherically-symetrical geometry of an infinite medium are considered. The numerical calculation using this method related to sooner evaluated direct problem data gives promising agreement with primary densities. This approach would be the basis for further application in an elemental analysis of a medium, using an isotopic neutron source and a moving, energy-dependent neutron detector. (author)

  5. Transport of accelerator produced high energy neutrons though concrete

    International Nuclear Information System (INIS)

    Prabhakar Rao, G.; Sarkar, P.K.

    1996-01-01

    Development of a computational system for estimating the production and transport of high energy neutrons in particle accelerators is reported. The energy-angle distribution of neutrons from accelerated ions bombarding thick targets is calculated by a hybrid nuclear reaction model code, ALICE-91, modified to suit the purpose. Subsequent transmission of these neutrons through concrete slabs is treated using the anisotropic source-flux iteration technique (ASFIT) in the framework of a coupled neutron-gamma transport. Several parameters of both the codes have been optimized to obtain the transmitted dose through concrete. The calculations are found to be accurate and at the same time faster compared to the detailed Monte Carlo calculations. (author). 8 refs., 2 figs

  6. Neutron scattering investigation of Ce based heavy fermion systems. From magnetism to unconventional phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Geselbracht, Philipp

    2016-12-05

    In Ce based heavy fermion systems the hybridization of the 4f orbital of the Ce ion and the conduction band lead to unconventional phenomena such as quantum critical points or superconductivity. The aim of this thesis is to investigate and compare the magnetism on a microscopic scale within the heavy fermion families CeT{sub 2}X{sub 2} (X=Si,Ge) and CeTIn{sub 5}. To do so neutron scattering was used as the experimental method. For CeCu{sub 2}Ge{sub 2}, the antiferromagnetic order AF1 (vector τ=(0.285 0.285 0.544)) is well described as a spin density wave with reduced ordered moments in [1 anti 10] direction. The phase diagram with magnetic field applied along [1 anti 10] direction was investigated. Two new phases were observed: the elliptical helix phase AF2 with modified propagation vector vector τ=(0.34 0.27 0.55) and the C-phase with a yet unknown magnetic order. Above T{sub N}, in zero field, short range order was observed, hinting competition of AF1 and AF2. It is assumed that both structures are due to different nesting properties of the Fermi surface. The RKKY character of the electronic system leads to effective Lande factors in the AF1 (g{sup eff}=0.36) and AF2 (g{sup eff}=0.525) phases. From the zero field dispersion the strength of the next nearest neighbor RKKY interactions was extracted, yielding 2SJ{sub 1}=(-0.042±0.007) meV (basal plane) and 2SJ{sub 2}=(-0.18±0.01) meV (body diagonal). Comparing the RKKY interaction to CeCu{sub 2}Si{sub 2} and CeNi{sub 2}Ge{sub 2} reveals a strong enhancement of the interaction in the basal plane going from antiferromagnetism (CeCu{sub 2}Ge{sub 2}) to superconductivity (CeCu{sub 2}Si{sub 2}) and finally paramagnetism (CeNi{sub 2}Ge{sub 2}). This new finding appears to be an important puzzle piece for the understanding of the CeT{sub 2}X{sub 2} family as it suggests a dependence of the anisotropy of the RKKY interaction from the hybridization strength of the 4f orbital and the conduction band. The obtained phase

  7. Transport phenomena in alkaline direct ethanol fuel cells for sustainable energy production

    Science.gov (United States)

    An, L.; Zhao, T. S.

    2017-02-01

    Alkaline direct ethanol fuel cells (DEFC), which convert the chemical energy stored in ethanol directly into electricity, are one of the most promising energy-conversion devices for portable, mobile and stationary power applications, primarily because this type of fuel cell runs on a carbon-neutral, sustainable fuel and the electrocatalytic and membrane materials that constitute the cell are relatively inexpensive. As a result, the alkaline DEFC technology has undergone a rapid progress over the last decade. This article provides a comprehensive review of transport phenomena of various species in this fuel cell system. The past investigations into how the design and structural parameters of membrane electrode assemblies and the operating parameters affect the fuel cell performance are discussed. In addition, future perspectives and challenges with regard to transport phenomena in this fuel cell system are also highlighted.

  8. Transportable, Low-Dose Active Fast-Neutron Imaging

    Energy Technology Data Exchange (ETDEWEB)

    Mihalczo, John T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wright, Michael C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McConchie, Seth M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Archer, Daniel E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Palles, Blake A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    This document contains a description of the method of transportable, low-dose active fast-neutron imaging as developed by ORNL. The discussion begins with the technique and instrumentation and continues with the image reconstruction and analysis. The analysis discussion includes an example of how a gap smaller than the neutron production spot size and detector size can be detected and characterized depending upon the measurement time.

  9. On generating neutron transport tables with the NJOY system

    International Nuclear Information System (INIS)

    Caldeira, Alexandre D.; Claro, Luiz H.

    2013-01-01

    Incorrect values for the product of the average number of neutrons released per fission and the fission microscopic cross-section were detected in several energy groups of a neutron transport table generated with the most updated version of the NJOY system. It was verified that the problem persists when older versions of this system are utilized. Although this problem exists for, at least, ten years, it is still an open question. (author)

  10. Considerations in the design of an improved transportable neutron spectrometer

    CERN Document Server

    Williams, A M; Brushwood, J M; Beeley, P A

    2002-01-01

    The Transportable Neutron Spectrometer (TNS) has been used by the Ministry of Defence for over 15 years to characterise neutron fields in workplace environments and provide local correction factors for both area and personal dosimeters. In light of advances in neutron spectrometry, a programme to evaluate and improve TNS has been initiated. This paper describes TNS, presents its operation in known radioisotope fields and in a reactor environment. Deficiencies in the operation of the instrument are highlighted, together with proposals for updating the response functions and spectrum unfolding methodologies.

  11. Radiation transport phenomena and modeling. Part A: Codes; Part B: Applications with examples

    International Nuclear Information System (INIS)

    Lorence, L.J. Jr.; Beutler, D.E.

    1997-09-01

    This report contains the notes from the second session of the 1997 IEEE Nuclear and Space Radiation Effects Conference Short Course on Applying Computer Simulation Tools to Radiation Effects Problems. Part A discusses the physical phenomena modeled in radiation transport codes and various types of algorithmic implementations. Part B gives examples of how these codes can be used to design experiments whose results can be easily analyzed and describes how to calculate quantities of interest for electronic devices

  12. Measurements of anomalous neutron transport in bulk graphite

    International Nuclear Information System (INIS)

    Bowman, C.D.; Smith, G.A.; Vogelaar, B.; Howell, C.R.; Bilpuch, E.G.; Tornow, W.

    2003-01-01

    The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)

  13. Measurements of anomalous neutron transport in bulk graphite

    Energy Technology Data Exchange (ETDEWEB)

    Bowman, C.D.; Smith, G.A. [ADNA Corp., Los Alamos, NM (United States); Vogelaar, B. [Virginia Tech., Blacksburg, VA (United States); Howell, C.R.; Bilpuch, E.G.; Tornow, W. [Triangle Univ. Nuclear Lab., Duke Univ., Durham, NC (United States)

    2003-07-01

    The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)

  14. Review on modeling development for multiscale chemical reactions coupled transport phenomena in solid oxide fuel cells

    International Nuclear Information System (INIS)

    Andersson, Martin; Yuan, Jinliang; Sunden, Bengt

    2010-01-01

    A literature study is performed to compile the state-of-the-art, as well as future potential, in SOFC modeling. Principles behind various transport processes such as mass, heat, momentum and charge as well as for electrochemical and internal reforming reactions are described. A deeper investigation is made to find out potentials and challenges using a multiscale approach to model solid oxide fuel cells (SOFCs) and combine the accuracy at microscale with the calculation speed at macroscale to design SOFCs, based on a clear understanding of transport phenomena, chemical reactions and functional requirements. Suitable methods are studied to model SOFCs covering various length scales. Coupling methods between different approaches and length scales by multiscale models are outlined. Multiscale modeling increases the understanding for detailed transport phenomena, and can be used to make a correct decision on the specific design and control of operating conditions. It is expected that the development and production costs will be decreased and the energy efficiency be increased (reducing running cost) as the understanding of complex physical phenomena increases. It is concluded that the connection between numerical modeling and experiments is too rare and also that material parameters in most cases are valid only for standard materials and not for the actual SOFC component microstructures.

  15. The Lattice Boltzmann Method applied to neutron transport

    International Nuclear Information System (INIS)

    Erasmus, B.; Van Heerden, F. A.

    2013-01-01

    In this paper the applicability of the Lattice Boltzmann Method to neutron transport is investigated. One of the main features of the Lattice Boltzmann method is the simultaneous discretization of the phase space of the problem, whereby particles are restricted to move on a lattice. An iterative solution of the operator form of the neutron transport equation is presented here, with the first collision source as the starting point of the iteration scheme. A full description of the discretization scheme is given, along with the quadrature set used for the angular discretization. An angular refinement scheme is introduced to increase the angular coverage of the problem phase space and to mitigate lattice ray effects. The method is applied to a model problem to investigate its applicability to neutron transport and the results are compared to a reference solution calculated, using MCNP. (authors)

  16. A modular spherical harmonics approach to the neutron transport equation

    International Nuclear Information System (INIS)

    Inanc, F.; Rohach, A.F.

    1989-01-01

    A modular nodal method was developed for solving the neutron transport equation in 2-D xy coordinates. The spherical harmonic expansion was used for approximating the second-order even-parity form of the neutron transport equation. The boundary conditions of the spherical harmonics approximation were derived in a form to have forms analogous to the partial currents in the neutron diffusion equation. Relations were developed for generating both the second-order spherical harmonic equations and the boundary conditions in an automated computational algorithm. Nodes using different orders of the spherical harmonics approximation to the transport equation were interfaced through mixed-type boundary conditions. The determination of spherical harmonic orders implemented in the nodes were determined by the scheme in an automated manner. Results of the method compared favorably to benchmark problems. (author)

  17. Calculated characteristics of subcritical assembly with anisotropic transport of neutrons

    International Nuclear Information System (INIS)

    Gorin, N.V.; Lipilina, E.N.; Lyutov, V.D.; Saukov, A.I.

    2003-01-01

    There was considered possibility of creating enough sub-critical system that multiply neutron fluence from a primary source by many orders. For assemblies with high neutron tie between parts, it is impossible. That is why there was developed a construction consisting of many units (cascades) having weak feedback with preceding cascades. The feedback attenuation was obtained placing layers of slow neutron absorber and moderators between the cascades of fission material. Anisotropy of fast neutron transport through the layers was used. The system consisted of many identical cascades aligning one by another. Each cascade consists of layers of moderator, fissile material and absorber of slow neutrons. The calculations were carried out using the code MCNP.4a with nuclear data library ENDF/B5. In this construction neutrons spread predominantly in one direction multiplying in each next fissile layer, and they attenuate considerably in the opposite direction. In a calculated construction, multiplication factor of one cascade is about 1.5 and multiplication factor of whole construction composed of n cascades is 1.5 n . Calculated keff value is 0.9 for one cascade and does not exceed 0.98 for a system containing any number of cascades. Therefore the assembly is always sub-critical and therefore it is safe in respect of criticality. There was considered using such a sub-critical assembly to create a powerful neutron fluence for neutron boron-capturing therapy. The system merits and demerits were discussed. (authors)

  18. Thermo-fluidic devices and materials inspired from mass and energy transport phenomena in biological system

    Institute of Scientific and Technical Information of China (English)

    Jian XIAO; Jing LIU

    2009-01-01

    Mass and energy transport consists of one of the most significant physiological processes in nature, which guarantees many amazing biological phenomena and activ-ities. Borrowing such idea, many state-of-the-art thermo-fluidic devices and materials such as artificial kidneys, carrier erythrocyte, blood substitutes and so on have been successfully invented. Besides, new emerging technologies are still being developed. This paper is dedicated to present-ing a relatively complete review of the typical devices and materials in clinical use inspired by biological mass and energy transport mechanisms. Particularly, these artificial thermo-fluidic devices and materials will be categorized into organ transplantation, drug delivery, nutrient transport, micro operation, and power supply. Potential approaches for innovating conventional technologies were discussed, corresponding biological phenomena and physical mechan-isms were interpreted, future promising mass-and-energy-transport-based bionic devices were suggested, and prospects along this direction were pointed out. It is expected that many artificial devices based on biological mass and energy transport principle will appear to better improve vari-ous fields related to human life in the near future.

  19. Relaxation phenomena in and microscopic transport theories of deeply inelastic collisions between heavy ions

    International Nuclear Information System (INIS)

    Noerenberg, W.

    1976-01-01

    Relaxation phenomena in deeply inelastic collisions are qualitatively discussed and compared with precompound reactions. Different approaches for describing these processes are reviewed, in particular the microscopic transport theories, which can be understood from a generalized master equation for macroscopic variables. The Markoff approximation and the classical limit for the relative motion lead to two coupled equations, the classical equation of relative motion with friction and a Pauli master equation for the internal degrees of freedom. The master equation approximated by the corresponding Fokker-Planck equation for mass transfer and energy dissipation is discussed in detail. Simple analytic expressions are derived for the transport coefficients as functions of excitation energy, total mass, mass fragmentation and relative angular momentum. Calculated transport coefficients are compared with experimental values. Problems and future developments in microscopic transport theories are outlined. (orig.) [de

  20. A study of dissipative phenomena using Orion, a 4 π sectorized neutron detector

    International Nuclear Information System (INIS)

    Galin, J.; Guerreau, D.; Morjean, M.; Pouthas, J.; Saint-Laurent, F.; Sokolov, A.; Wang, X.M.; Piasecki, E.; Charvet, J.L.; CEA Centre d'Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette

    1990-01-01

    When studying the behavior of hot nuclei, the challenge is twofold: how are they formed in nucleus-nucleus collisions and how do they decay. For heavy and, thus neutron rich systems a large fraction of the thermalized energy is evacuated by neutron evaporation. Therefore the numbering, event-wise, of neutrons, over 4 π, gives a strong handle on energy dissipation for the different reaction channels. The first neutron measurements of this kind were performed using spherical detectors made of two hemispheres. Since then, a new and larger 4 π detector, ORION, has been designed in order to get information on the spatial distribution of the neutrons. The main characteristics of ORION are described and a few examples are given in order to illustrate the capabilities of such a detector in the study of dissipative collisions

  1. Hierarchical modeling of plasma and transport phenomena in a dielectric barrier discharge reactor

    Science.gov (United States)

    Bali, N.; Aggelopoulos, C. A.; Skouras, E. D.; Tsakiroglou, C. D.; Burganos, V. N.

    2017-12-01

    A novel dual-time hierarchical approach is developed to link the plasma process to macroscopic transport phenomena in the interior of a dielectric barrier discharge (DBD) reactor that has been used for soil remediation (Aggelopoulos et al 2016 Chem. Eng. J. 301 353-61). The generation of active species by plasma reactions is simulated at the microseconds (µs) timescale, whereas convection and thermal conduction are simulated at the macroscopic (minutes) timescale. This hierarchical model is implemented in order to investigate the influence of the plasma DBD process on the transport and reaction mechanisms during remediation of polluted soil. In the microscopic model, the variables of interest include the plasma-induced reactive concentrations, while in the macroscopic approach, the temperature distribution, and the velocity field both inside the discharge gap and within the polluted soil material as well. For the latter model, the Navier-Stokes and Darcy Brinkman equations for the transport phenomena in the porous domain are solved numerically using a FEM software. The effective medium theory is employed to provide estimates of the effective time-evolving and three-phase transport properties in the soil sample. Model predictions considering the temporal evolution of the plasma remediation process are presented and compared with corresponding experimental data.

  2. Finite-orbit-width effect and the radial electric field in neoclassical transport phenomena

    International Nuclear Information System (INIS)

    Satake, S.; Okamoto, M.; Nakajima, N.; Sugama, H.; Yokoyama, M.; Beidler, C.D.

    2005-01-01

    Modeling and detailed simulation of neoclassical transport phenomena both in 2D and 3D toroidal configurations are shown. The emphasis is put on the effect of finiteness of the drift-orbit width, which brings a non-local nature to neoclassical transport phenomena. Evolution of the self-consistent radial electric field in the framework of neoclassical transport is also investigated. The combination of Monte-Carlo calculation for ion transport and numerical solver of ripple-averaged kinetic equation for electrons makes it possible to calculate neoclassical fluxes and the time evolution of the radial electric field in the whole plasma region, including the finite-orbit-width (FOW) effects and global evolution of geodesic acoustic mode (GAM). The simulation results show that the heat conductivity around the magnetic axis is smaller than that obtained from standard neoclassical theory and that the evolution of GAM oscillation on each flux surface is coupled with other surfaces if the FOW effect is significant. A global simulation of radial electric field evolution in a non-axisymmetric plasma is also shown. (author)

  3. Analysis of transport phenomena and electrochemical reactions in a micro PEM fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Sadiq Al-Baghdadi, Maher A.R. [Fuel Cell Research Center, International Energy and Environment Foundation, Najaf, P.O.Box 39 (Iraq)

    2013-07-01

    Micro-fuel cells are considered as promising electrochemical power sources in portable electronic devices. The presence of microelectromechanical system (MEMS) technology makes it possible to manufacture the miniaturized fuel cell systems. The majority of research on micro-scale fuel cells is aimed at micro-power applications. Performance of micro-fuel cells are closely related to many factors, such as designs and operating conditions. CFD modeling and simulation for heat and mass transport in micro PEM fuel cells are being used extensively in researches and industrial applications to gain better understanding of the fundamental processes and to optimize the micro fuel cell designs before building a prototype for engineering application. In this research, full three-dimensional, non-isothermal computational fluid dynamics model of a micro proton exchange membrane (PEM) fuel cell has been developed. This comprehensive model accounts for the major transport phenomena such as convective and diffusive heat and mass transfer, electrode kinetics, transport and phase-change mechanism of water, and potential fields in a micro PEM fuel cell. The model explains many interacting, complex electrochemical, and transport phenomena that cannot be studied experimentally. Three-dimensional results of the species profiles, temperature distribution, potential distribution, and local current density distribution are presented and analysed, with the focus on the physical insight and fundamental understanding.

  4. Interfacial transport phenomena and stability in liquid-metal/water systems: scaling considerations

    International Nuclear Information System (INIS)

    Abdulla, S.; Liu, X.; Anderson, M.; Bonazza, R.; Corradini, M.; Cho, D.

    2001-01-01

    One concept being considered for steam generation in innovative nuclear reactor applications, involves water coming into direct contact with a circulating molten metal. The vigorous agitation of the two fluids, the direct liquid-liquid contact and the consequent large interfacial area give rise to very high heat transfer coefficients and rapid steam generation. For an optimum design of such direct contact heat exchange and vaporization systems, detailed knowledge is necessary of the various flow regimes, interfacial transport phenomena, heat transfer and operational stability. In this paper we describe current results from the first year of this research that studies the transport phenomena involved with the injection of water into molten metals (e.g., lead alloys). In particular, this work discusses scaling considerations related to direct contact heat exchange, our experimental plans for investigation and a test plan for the important experimental parameters; i.e., the water and liquid metal mass flow rates, the liquid metal pool temperature and the ambient pressure of the direct contact heat exchanger. Past experimental work and initial scaling results suggest that our experiments can directly represent the proper liquid metal pool temperature and the water subcooling. The experimental variation in water and liquid metal flow rates and system pressure (1-10 bar), although smaller than the current conceptual system designs, is sufficient to verify the expected scale effects to demonstrate the phenomena. (authors)

  5. In situ diagnostic of two-phase flow phenomena in polymer electrolyte fuel cells by neutron imaging

    International Nuclear Information System (INIS)

    Kramer, Denis; Zhang, Jianbo; Shimoi, Ryoichi; Lehmann, Eberhard; Wokaun, Alexander; Shinohara, Kazuhiko; Scherer, Guenther G.

    2005-01-01

    Neutron radiographical measurements have been performed on operating hydrogen-fueled polymer electrolyte fuel cells (PEFC). With the successful detection of liquid accumulation in flow field and gas diffusion layer (GDL) under various operating conditions a unique experimental approach for the investigation of two-phase flow phenomena in technical PEFC has been realized. The experimental setup will be described in detail. Algorithms for an enhanced quantitative evaluation of the obtained images are presented and successful application to the data demonstrated. Finally, results from PEFC investigations will be given. Different flow field geometries and their implications for liquid accumulation inside flow field and GDL are discussed

  6. Implementation of the quasi-static method for neutron transport

    International Nuclear Information System (INIS)

    Alcaro, Fabio; Dulla, Sandra; Ravetto, Piero; Le Tellier, Romain; Suteau, Christophe

    2011-01-01

    The study of the dynamic behavior of next generation nuclear reactors is a fundamental aspect for safety and reliability assessments. Despite the growing performances of modern computers, the full solution of the neutron Boltzmann equation in the time domain is still an impracticable task, thus several approximate dynamic models have been proposed for the simulation of nuclear reactor transients; the quasi-static method represents the standard tool currently adopted for the space-time solution of neutron transport problems. All the practical applications of this method that have been proposed contain a major limit, consisting in the use of isotropic quantities, such as scalar fluxes and isotropic external neutron sources, being the only data structures available in most deterministic transport codes. The loss of the angular information produces both inaccuracies in the solution of the kinetic model and the inconsistency of the quasi-static method itself. The present paper is devoted to the implementation of a consistent quasi-static method. The computational platform developed by CEA in Cadarache has been used for the creation of a kinetic package to be coupled with the existing SNATCH solver, a discrete-ordinate multi-dimensional neutron transport solver, employed for the solution of the steady-state Boltzmann equation. The work aims at highlighting the effects of the angular treatment of the neutron flux on the transient analysis, comparing the results with those produced by the previous implementations of the quasi-static method. (author)

  7. THE IMPORTANCE OF LIMIT SOLUTIONS & TEMPORAL AND SPATIAL SCALES IN THE TEACHING OF TRANSPORT PHENOMENA

    Directory of Open Access Journals (Sweden)

    SÁVIO LEANDRO BERTOLI

    2016-07-01

    Full Text Available In the engineering courses the field of Transport Phenomena is of significant importance and it is in several disciplines relating to Fluid Mechanics, Heat and Mass Transfer. In these disciplines, problems involving these phenomena are mathematically formulated and analytical solutions are obtained whenever possible. The aim of this paper is to emphasize the possibility of extending aspects of the teaching-learning in this area by a method based on time scales and limit solutions. Thus, aspects relative to the phenomenology naturally arise during the definition of the scales and / or by determining the limit solutions. Aspects concerning the phenomenology of the limit problems are easily incorporated into the proposed development, which contributes significantly to the understanding of physics inherent in the mathematical modeling of each limiting case studied. Finally the study aims to disseminate the use of the limit solutions and of the time scales in the general fields of engineering.

  8. A New Monte Carlo Neutron Transport Code at UNIST

    International Nuclear Information System (INIS)

    Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung

    2014-01-01

    Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results

  9. New methods in transport theory. Part of a coordinated programme on methods in neutron transport theory

    International Nuclear Information System (INIS)

    Stefanovic, D.

    1975-09-01

    The research work of this contract was oriented towards the study of different methods in neutron transport theory. Authors studied analytical solution of the neutron slowing down transport equation and extension of this solution to include the energy dependence of the anisotropy of neutron scattering. Numerical solution of the fast and resonance transport equation for the case of mixture of scatterers including inelastic effects were also reviewed. They improved the existing formalism for treating the scattering of neutrons on water molecules; Identifying modal analysis as the Galerkin method, general conditions for modal technique applications have been investigated. Inverse problems in transport theory were considered. They obtained the evaluation of an advanced level distribution function, made improvement of the standard formalism for treating the inelastic scattering and development of a cluster nuclear model for this evaluation. Authors studied the neutron transport treatment in space energy groups for criticality calculation of a reactor core, and development of the Monte Carlo sampling scheme from the neutron transport equation

  10. Finite element method for solving neutron transport problems

    International Nuclear Information System (INIS)

    Ferguson, J.M.; Greenbaum, A.

    1984-01-01

    A finite element method is introduced for solving the neutron transport equations. Our method falls into the category of Petrov-Galerkin solution, since the trial space differs from the test space. The close relationship between this method and the discrete ordinate method is discussed, and the methods are compared for simple test problems

  11. Neutron transport in two dissimilar media anisotropic scattering

    International Nuclear Information System (INIS)

    Burkart, A.R.; Ishiguro, Y.; Siewert, C.E.

    1976-01-01

    The elementary solution of the one-speed neutron-transport equation with linearly anisotropic scattering are used in conjunction with Chandrasekhar's invariance principles to solve in a concise manner the Milne problem for two adjoining half-spaces and the critical reactor problem for a reflected slab

  12. Neutron transport calculations of some fast critical assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Val Penalosa, J A

    1976-07-01

    To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs.

  13. Neutron transport calculations of some fast critical assemblies

    International Nuclear Information System (INIS)

    Martinez-Val Penalosa, J. A.

    1976-01-01

    To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs

  14. Characterization of transport phenomena in porous transport layers using X-ray microtomography

    Science.gov (United States)

    Hasanpour, S.; Hoorfar, M.; Phillion, A. B.

    2017-06-01

    Among different methods available for estimating the transport properties of porous transport layers (PTLs) of polymer electrolyte membrane fuel cells, X-ray micro computed tomography (X-μCT) imaging in combination with image-based numerical simulation has been recognized as a viable tool. In this study, four commercially-available single-layer and dual-layer PTLs are analyzed using this method in order to compare and contrast transport properties between different PTLs, as well as the variability within a single sheet. Complete transport property datasets are created for each PTL. The simulation predictions indicate that PTLs with high porosity show considerable variability in permeability and effective diffusivity, while PTLs with low porosity do not. Furthermore, it is seen that the Tomadakis-Sotirchos (TS) analytical expressions for porous media match the image-based simulations when porosity is relatively low but predict higher permeability and effective diffusivity for porosity values greater than 80%. Finally, the simulations show that cracks within MPL of dual-layer PTLs have a significant effect on the overall permeability and effective diffusivity of the PTLs. This must be considered when estimating the transport properties of dual-layer PTLs. These findings can be used to improve macro-scale models of product and reactant transport within fuel cells, and ultimately, fuel cell efficiency.

  15. Neutron Computed Tomography of Freeze/thaw Phenomena in Polymer Electrolyte Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Matthew M. Mech; Jack Brenizer; Kenan Unlu; A.K. Heller

    2008-12-12

    This report summarizes the final year's progress of the three-year NEER program. The overall objectives of this program were to 1) design and construct a sophisticated hight-resolution neutron computed tomography (NCT) facility, 2) develop novel and sophisticated liquid water and ice quantification analysis software for computed tomography, and 3) apply the advanced software and NCT capability to study liquid and ice distribution in polymer electrolyte fuel cells (PEFCs) under cold-start conditions. These objectives have been accomplished by the research team, enabling a new capability for advanced 3D image quantification with neutron imaging for fuel cell and other applications. The NCT water quantification methodology and software will greatly add to the capabilities of the neutron imaging community, and the quantified liquid water and ice distribution provided by its application to PEFCs will enhance understanding and guide design in the fuel cell community.

  16. Visualization and measurement of thermo-fluid phenomena by using neutron radiography

    Energy Technology Data Exchange (ETDEWEB)

    Mishima, Kaichiro [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1998-03-01

    Recently, neutron radiography is rapidly expanding its application to various fields incollaboration with the development of electronic imaging techniques. Particularly, in the field of thermal and fluid engineering, it has drawn much attention as an innovative and non-intrusive method for fluid-visualization and measurement using neutron beam as a probe. In this paper, some examples of applications are introduced on the high frame-rate imaging, quantification method (especially, determination of void fraction), and multidimensional measurement which are much interested in fluid measurement in relation to light water reactor safety and thermohydraulics. (author)

  17. Visualization and measurement of thermo-fluid phenomena by using neutron radiography

    International Nuclear Information System (INIS)

    Mishima, Kaichiro

    1998-01-01

    Recently, neutron radiography is rapidly expanding its application to various fields incollaboration with the development of electronic imaging techniques. Particularly, in the field of thermal and fluid engineering, it has drawn much attention as an innovative and non-intrusive method for fluid-visualization and measurement using neutron beam as a probe. In this paper, some examples of applications are introduced on the high frame-rate imaging, quantification method (especially, determination of void fraction), and multidimensional measurement which are much interested in fluid measurement in relation to light water reactor safety and thermohydraulics. (author)

  18. Optimization of a neutron detector design using adjoint transport simulation

    International Nuclear Information System (INIS)

    Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

    2012-01-01

    A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

  19. Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

    International Nuclear Information System (INIS)

    Thiagu Supramaniam

    2007-01-01

    The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

  20. Homogenization of the critically spectral equation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Allaire, G. [CEA Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie]|[Paris-6 Univ., 75 (France). Lab. d' Analyse Numerique; Bal, G. [Electricite de France (EDF), 92 - Clamart (France). Direction des Etudes et Recherches

    1998-07-01

    We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)

  1. Application of Walsh functions to neutron transport problems. I. Theory

    International Nuclear Information System (INIS)

    Seed, T.J.; Albrecht, R.W.

    1976-01-01

    An approximation to the neutron transport equation is made by representing the angular flux with an expansion of the angular dependence in the orthogonal, complete, and binary valued sets of Walsh function. The Walsh approximation is applied to the one-speed, isotropic-scattering, rectangular-geometry form of the neutron transport equation. Sets of partial differential equations for the expansion coefficients are derived along with appropriate boundary conditions for their solution. The sets of the Walsh expansion to one- and two-dimensional forms of the transport equation are also obtained. The two-dimensional expansion coefficient equations are shown to be not only hyperbolic but also transformable to a set of S/sub N/-like equations that are coupled only through the scattering term. Such transformal sets of equations are termed Walsh-derived quadrature sets

  2. Homogenization of the critically spectral equation in neutron transport

    International Nuclear Information System (INIS)

    Allaire, G.; Paris-6 Univ., 75; Bal, G.

    1998-01-01

    We address the homogenization of an eigenvalue problem for the neutron transport equation in a periodic heterogeneous domain, modeling the criticality study of nuclear reactor cores. We prove that the neutron flux, corresponding to the first and unique positive eigenvector, can be factorized in the product of two terms, up to a remainder which goes strongly to zero with the period. On terms is the first eigenvector of the transport equation in the periodicity cell. The other term is the first eigenvector of a diffusion equation in the homogenized domain. Furthermore, the corresponding eigenvalue gives a second order corrector for the eigenvalue of the heterogeneous transport problem. This result justifies and improves the engineering procedure used in practice for nuclear reactor cores computations. (author)

  3. Rapid Measurement of Neutron Dose Rate for Transport Index

    International Nuclear Information System (INIS)

    Morris, R.L.

    2000-01-01

    A newly available neutron dose equivalent remmeter with improved sensitivity and energy response has been put into service at Rocky Flats Environmental Technology Site (RFETS). This instrument is being used to expedite measurement of the Transport Index and as an ALARA tool to identify locations where slightly elevated neutron dose equivalent rates exist. The meter is capable of measuring dose rates as low as 0.2 μSv per hour (20 μrem per hour). Tests of the angular response and energy response of the instrument are reported. Calculations of the theoretical instrument response made using MCNPtrademark are reported for materials typical of those being shipped

  4. Discontinuous nodal schemes applied to the bidimensional neutron transport equation

    International Nuclear Information System (INIS)

    Delfin L, A.; Valle G, E. Del; Hennart B, J.P.

    1996-01-01

    In this paper several strong discontinuous nodal schemes are described, starting from the one that has only two interpolation parameters per cell to the one having ten. Their application to the spatial discretization of the neutron transport equation in X-Y geometry is also described, giving, for each one of the nodal schemes, the approximation for the angular neutron flux that includes the set of interpolation parameters and the corresponding polynomial space. Numerical results were obtained for several test problems presenting here the problem with the highest degree of difficulty and their comparison with published results 1,2 . (Author)

  5. A method for solving neutron transport equation

    International Nuclear Information System (INIS)

    Dimitrijevic, Z.

    1993-01-01

    The procedure for solving the transport equation by directly integrating for case one-dimensional uniform multigroup medium is shown. The solution is expressed in terms of linear combination of function H n (x,μ), and the coefficient is determined from given conditions. The solution is applied for homogeneous slab of critical thickness. (author)

  6. Proceedings of the Fifth Microgravity Fluid Physics and Transport Phenomena Conference

    Science.gov (United States)

    Singh, Bhim S. (Editor)

    2000-01-01

    The Fifth Microgravity Fluid Physics and Transport Phenomena Conference provided the scientific community the opportunity to view the current scope of the Microgravity Fluid Physics and Transport Phenomena Program and research opportunities and plans for the near future. Consistent with the conference theme "Microgravity Research an Agency-Wide Asset" the conference focused not only on fundamental research but also on applications of this knowledge towards enabling future space exploration missions. The conference included 14 invited plenary talks, 61 technical paper presentations, 61 poster presentations, exhibits and a forum on emerging research themes focusing on nanotechnology and biofluid mechanics. This web-based proceeding includes the presentation and poster charts provided by the presenters of technical papers and posters that were scanned at the conference site. Abstracts of all the papers and posters are included and linked to the presentations charts. The invited and plenary speakers were not required to provide their charts and are generally not available for scanning and hence not posted. The conference program is also included.

  7. Frontiers in transport phenomena research and education: Energy systems, biological systems, security, information technology and nanotechnology

    Energy Technology Data Exchange (ETDEWEB)

    Bergman, T.L.; Faghri, A. [Department of Mechanical Engineering, The University of Connecticut, Storrs, CT 06269-3139 (United States); Viskanta, R. [School of Mechanical Engineering, Purdue University, West Lafayette, IN 47907-2088 (United States)

    2008-09-15

    A US National Science Foundation-sponsored workshop entitled ''Frontiers in Transport Phenomena Research and Education: Energy Systems, Biological Systems, Security, Information Technology, and Nanotechnology'' was held in May of 2007 at the University of Connecticut. The workshop provided a venue for researchers, educators and policy-makers to identify frontier challenges and associated opportunities in heat and mass transfer. Approximately 300 invited participants from academia, business and government from the US and abroad attended. Based upon the final recommendations on the topical matter of the workshop, several trends become apparent. A strong interest in sustainable energy is evident. A continued need to understand the coupling between broad length (and time) scales persists, but the emerging need to better understand transport phenomena at the macro/mega scale has evolved. The need to develop new metrology techniques to collect and archive reliable property data persists. Societal sustainability received major attention in two of the reports. Matters involving innovation, entrepreneurship, and globalization of the engineering profession have emerged, and the responsibility to improve the technical literacy of the public-at-large is discussed. Integration of research thrusts and education activities is highlighted throughout. Specific recommendations, made by the panelists with input from the international heat transfer community and directed to the National Science Foundation, are included in several reports. (author)

  8. KAMCCO, a reactor physics Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.

    1976-06-01

    KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de

  9. Investigations of grain size dependent sediment transport phenomena on multiple scales

    Science.gov (United States)

    Thaxton, Christopher S.

    Sediment transport processes in coastal and fluvial environments resulting from disturbances such as urbanization, mining, agriculture, military operations, and climatic change have significant impact on local, regional, and global environments. Primarily, these impacts include the erosion and deposition of sediment, channel network modification, reduction in downstream water quality, and the delivery of chemical contaminants. The scale and spatial distribution of these effects are largely attributable to the size distribution of the sediment grains that become eligible for transport. An improved understanding of advective and diffusive grain-size dependent sediment transport phenomena will lead to the development of more accurate predictive models and more effective control measures. To this end, three studies were performed that investigated grain-size dependent sediment transport on three different scales. Discrete particle computer simulations of sheet flow bedload transport on the scale of 0.1--100 millimeters were performed on a heterogeneous population of grains of various grain sizes. The relative transport rates and diffusivities of grains under both oscillatory and uniform, steady flow conditions were quantified. These findings suggest that boundary layer formalisms should describe surface roughness through a representative grain size that is functionally dependent on the applied flow parameters. On the scale of 1--10m, experiments were performed to quantify the hydrodynamics and sediment capture efficiency of various baffles installed in a sediment retention pond, a commonly used sedimentation control measure in watershed applications. Analysis indicates that an optimum sediment capture effectiveness may be achieved based on baffle permeability, pond geometry and flow rate. Finally, on the scale of 10--1,000m, a distributed, bivariate watershed terain evolution module was developed within GRASS GIS. Simulation results for variable grain sizes and for

  10. Spatial domain decomposition for neutron transport problems

    International Nuclear Information System (INIS)

    Yavuz, M.; Larsen, E.W.

    1989-01-01

    A spatial Domain Decomposition method is proposed for modifying the Source Iteration (SI) and Diffusion Synthetic Acceleration (DSA) algorithms for solving discrete ordinates problems. The method, which consists of subdividing the spatial domain of the problem and performing the transport sweeps independently on each subdomain, has the advantage of being parallelizable because the calculations in each subdomain can be performed on separate processors. In this paper we describe the details of this spatial decomposition and study, by numerical experimentation, the effect of this decomposition on the SI and DSA algorithms. Our results show that the spatial decomposition has little effect on the convergence rates until the subdomains become optically thin (less than about a mean free path in thickness)

  11. Symmetrized neutron transport equation and the fast Fourier transform method

    International Nuclear Information System (INIS)

    Sinh, N.Q.; Kisynski, J.; Mika, J.

    1978-01-01

    The differential equation obtained from the neutron transport equation by the application of the source iteration method in two-dimensional rectangular geometry is transformed into a symmetrized form with respect to one of the angular variables. The discretization of the symmetrized equation leads to finite difference equations based on the five-point scheme and solved by use of the fast Fourier transform method. Possible advantages of the approach are shown on test calculations

  12. Deterministic methods to solve the integral transport equation in neutronic

    International Nuclear Information System (INIS)

    Warin, X.

    1993-11-01

    We present a synthesis of the methods used to solve the integral transport equation in neutronic. This formulation is above all used to compute solutions in 2D in heterogeneous assemblies. Three kinds of methods are described: - the collision probability method; - the interface current method; - the current coupling collision probability method. These methods don't seem to be the most effective in 3D. (author). 9 figs

  13. Mathematical models for volume rendering and neutron transport

    International Nuclear Information System (INIS)

    Max, N.

    1994-09-01

    This paper reviews several different models for light interaction with volume densities of absorbing, glowing, reflecting, or scattering material. They include absorption only, glow only, glow and absorption combined, single scattering of external illumination, and multiple scattering. The models are derived from differential equations, and illustrated on a data set representing a cloud. They are related to corresponding models in neutron transport. The multiple scattering model uses an efficient method to propagate the radiation which does not suffer from the ray effect

  14. Complex eigenvalues for neutron transport equation with quadratically anisotropic scattering

    International Nuclear Information System (INIS)

    Sjoestrand, N.G.

    1981-01-01

    Complex eigenvalues for the monoenergetic neutron transport equation in the buckling approximation have been calculated for various combinations of linearly and quadratically anisotropic scattering. The results are discussed in terms of the time-dependent case. Tables are given of complex bucklings for real decay constants and of complex decay constants for real bucklings. The results fit nicely into the pattern of real and purely imaginary eigenvalues obtained earlier. (author)

  15. Sensitivity of neutron air transport to nitrogen cross section uncertainties

    International Nuclear Information System (INIS)

    Niiler, A.; Beverly, W.B.; Banks, N.E.

    1975-01-01

    The sensitivity of the transport of 14-MeV neutrons in sea level air to uncertainties in the ENDF/B-III values of the various Nitrogen cross sections has been calculated using the correlated sampling Monte Carlo neutron transport code SAMCEP. The source consisted of a 14.0- to 14.9-MeV band of isotropic neutrons and the fluences (0.5 to 15.0 MeV) were calculated at radii from 50 to 1500 metres. The maximum perturbations, assigned to the ENDF/B-III or base cross section set in the 6.0- to 14.5-MeV energy range were; (1) 2 percent to the total, (2) 10 percent to the total elastic, (3) 40 percent to the inelastic and absorption and (4) 20 percent to the first Legendre coefficient and 10 percent to the second Legendre coefficient of the elastic angular distribtuions. Transport calculations were carried out using various physically realistic sets of perturbed cross sections, bounded by evaluator-assigned uncertainties, as well as the base set. Results show that in some energy intervals at 1500 metres, the differential fluence level with a perturbed set differed by almost a factor of two from the differential fluence level with the base set. 5 figures

  16. Transport of D-D fusion neutrons in thick concrete

    International Nuclear Information System (INIS)

    Ku, L.P.; Kolibal, J.G.

    1982-07-01

    By altering the collision mechanism in the numerical transport calculations, and by constructing an analytical model based on age-diffusion theory, the outstanding feature in the life history of D-D fusion neutrons penetrating deeply into ordinary concrete is shown to be the transport in the 2.3 MeV oxygen anti-resonance. This result is used to assess the impact of the cross-section uncertainties and the uncertainties due to variations in the D-D fusion spectrum and temperature

  17. Toward whole-core neutron transport without spatial homogenization

    International Nuclear Information System (INIS)

    Lewis, E. E.

    2009-01-01

    Full text of publication follows: A long-term goal of computational reactor physics is the deterministic analysis of power reactor core neutronics without incurring significant discretization errors in the energy, spatial or angular variables. In principle, given large enough parallel configurations with unlimited CPU time and memory, this goal could be achieved using existing three-dimensional neutron transport codes. In practice, however, solving the Boltzmann equation for neutrons over the six-dimensional phase space is made intractable by the nature of neutron cross-sections and the complexity and size of power reactor cores. Tens of thousands of energy groups would be required for faithful cross section representation. Likewise, the numerous material interfaces present in power reactor lattices require exceedingly fine spatial mesh structures; these ubiquitous interfaces preclude effective implementation of adaptive grid, mesh-less methods and related techniques that have been applied so successfully in other areas of engineering science. These challenges notwithstanding, substantial progress continues in the pursuit for more robust deterministic methods for whole-core neutronics analysis. This paper examines the progress over roughly the last decade, emphasizing the space-angle variables and the quest to eliminate errors attributable to spatial homogenization. As prolog we briefly assess 1990's methods used in light water reactor analysis and review the lessons learned from the C5G7 benchmark exercises which were originated in 1999 to appraise the ability of transport codes to perform core calculations without homogenization. We proceed by examining progress over the last decade much of which falls into three areas. These may be broadly characterized as reduced homogenization, dynamic homogenization and planar-axial synthesis. In the first, homogenization in three-dimensional calculations is reduced from the fuel assembly to the pin-cell level. In the second

  18. Assessment of corrosion phenomena in liquid lithium at T < 873 K. A Li(d,n) neutron source as case study

    Energy Technology Data Exchange (ETDEWEB)

    Knaster, J., E-mail: juan.knaster@ifmif.org [IFMIF/EVEDA Project Team (F4E), Rokkasho (Japan); Favuzza, P. [ENEA, Firenze (Italy)

    2017-05-15

    The corrosion induced by alkali metals in steels has been the subject of long decades of intense studies under both nuclear fission and fusion research programs. Li or its eutectic Pb-17Li is the liquid metal coolant choice for fusion blankets due to the tritium breeder capability of Li. Non-metal impurities enhance corrosion, but only N becomes potentially a problem given its high solubility in liquid Li and the depletion of Cr through ternary nitrides Li-Cr-N. The low solubility of C and O allow its cold trapping to values <10 wppm, however N can only be hot trapped demanding temperatures typically of 873 K. The inherent difficulties of experimentation on physicochemical kinetics related with alkali metals lead to a confusing divergence of results available in the literature; however, the understanding of the corrosion phenomena of RAFM steels exposed to flowing Li up to 873 K is mature. Next decade, 14 MeV neutrons will be available for fusion materials testing through Li(d,n) nuclear reactions. In such a facility, a concave RAFM steel backplate will be channelling 523 K flowing Li in the region where the 40 MeV deuteron beam will be impacting. If RAFM steels are considered, two main concurrent mechanisms will take place: a) mass transport of alloying elementsalong the loop and b) depletion of Cr through formation of Li{sub 9}CrN{sub 5}. Fortunately, the mass transport phenomena of Cr within the ΔT = 350 K in the loop is limited due to the poor solubility of Cr in liquid Li (0.21 wppm at 873 K). In turn, at 523 K Li the activity of N to form the ternary compound is negligible. However, the high solubility of Ni in Li (2144 wppm at 873 K), suggests the presence of mass transport phenomena of Ni from the stainless steel piping; unfortunately, the physicochemical kinetics are not fully understood. Lifus 6, in operation in Brasimone (ENEA) since the end 2015, will close in a definitive manner remaining open questions.

  19. Microscopic Linear Response Theory of Spin Relaxation and Relativistic Transport Phenomena in Graphene

    Directory of Open Access Journals (Sweden)

    Manuel Offidani

    2018-05-01

    Full Text Available We present a unified theoretical framework for the study of spin dynamics and relativistic transport phenomena in disordered two-dimensional Dirac systems with pseudospin-spin coupling. The formalism is applied to the paradigmatic case of graphene with uniform Bychkov-Rashba interaction and shown to capture spin relaxation processes and associated charge-to-spin interconversion phenomena in response to generic external perturbations, including spin density fluctuations and electric fields. A controlled diagrammatic evaluation of the generalized spin susceptibility in the diffusive regime of weak spin-orbit interaction allows us to show that the spin and momentum lifetimes satisfy the standard Dyakonov-Perel relation for both weak (Gaussian and resonant (unitary nonmagnetic disorder. Finally, we demonstrate that the spin relaxation rate can be derived in the zero-frequency limit by exploiting the SU(2 covariant conservation laws for the spin observables. Our results set the stage for a fully quantum-mechanical description of spin relaxation in both pristine graphene samples with weak spin-orbit fields and in graphene heterostructures with enhanced spin-orbital effects currently attracting much attention.

  20. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    Energy Technology Data Exchange (ETDEWEB)

    M. Anderson; M. Corradini; K.Y. Bank; R. Bonazza; D. Cho

    2005-04-26

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications.

  1. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    International Nuclear Information System (INIS)

    Anderson, M.; Corradini, M.; Bank, K.Y.; Bonazza, R.; Cho, D.

    2005-01-01

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications

  2. Modelling of melting and solidification transport phenomena during hypothetical NPP severe accidents

    International Nuclear Information System (INIS)

    Sarler, B.

    1992-01-01

    A physical and mathematical framework to deal with the transport phenomena occuring during melting and solidification of the hypothetical NPP severe accidents is presented. It concentrates on the transient temperature, velocity, and species concentration distributions during such events. The framework is based on the Mixture Continuum Formulation of the components and phases, cast in the boundary-domain integral shape structured by the fundamental solution of the Laplace equation. The formulation could cope with various solid-liquid sub-systems through the inclusion of the specific closure relations. The deduced system of boundary-domain integral equations for conservation of mass, energy, momentum, and species could be solved by the boundary element discrete approximative method. (author) [sl

  3. Transport phenomena and kinetic theory applications to gases, semiconductors, photons, and biological systems

    CERN Document Server

    Gabetta, Ester

    2007-01-01

    The study of kinetic equations related to gases, semiconductors, photons, traffic flow, and other systems has developed rapidly in recent years because of its role as a mathematical tool in many applications in areas such as engineering, meteorology, biology, chemistry, materials science, nanotechnology, and pharmacy. Written by leading specialists in their respective fields, this book presents an overview of recent developments in the field of mathematical kinetic theory with a focus on modeling complex systems, emphasizing both mathematical properties and their physical meaning. The overall presentation covers not only modeling aspects and qualitative analysis of mathematical problems, but also inverse problems, which lead to a detailed assessment of models in connection with their applications, and to computational problems, which lead to an effective link of models to the analysis of real-world systems. "Transport Phenomena and Kinetic Theory" is an excellent self-study reference for graduate students, re...

  4. Transport phenomena in high Tc superconductors. Resume of Ph.D thesis

    International Nuclear Information System (INIS)

    Crisan, I.A.

    1994-01-01

    This is an extended abstract of the Ph. D. thesis devoted to the transport phenomena in high-Tc superconductors. There are three chapters. The first chapter presents an overview of the essential theoretical aspects concerning the vortex dynamics particularly in ceramic superconductors. The chapter two gives a description of the preparation methods of superconductor samples used by the author as well as the measurement devices for volt-ampere characteristics and the associated electronic circuitry. In the third chapter there are presented the experimental data obtained from different samples prepared in different temperature and magnetic field conditions. The obtained results are finally interpreted in the frame of existent or original models. (M.I.C.). 31 Refs

  5. Numerical simulation of mass and energy transport phenomena in solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Arpino, F. [Dipartimento di Meccanica, Strutture, Ambiente e Territorio (DiMSAT), University of Cassino, via Di Biasio 43, Cassino (Italy); Massarotti, N. [Dipertimento per le Tecnologie (DiT), University of Naples ' ' Parthenope' ' , Centro Direzionale, isola C4, 80143 Napoli (Italy)

    2009-12-15

    Solid Oxide Fuel Cells (SOFCs) represent a very promising technology for near future energy conversion thanks to a number of advantages, including the possibility of using different fuels. In this paper, a detailed numerical model, based on a general mathematical description and on a finite element Characteristic based Split (CBS) algorithm code is employed to simulate mass and energy transport phenomena in SOFCs. The model predicts the thermodynamic quantity of interest in the fuel cell. Full details of the numerical solution obtained are presented both in terms of heat and mass transfer in the cell and in terms of electro-chemical reactions that occur in the system considered. The results obtained with the present algorithm is compared with the experimental data available in the literature for validation, showing an excellent agreement. (author)

  6. Transport Phenomena in Porous Media Aspects of MicroMacro Behaviour

    CERN Document Server

    Ichikawa, Yasuaki

    2012-01-01

    This monograph presents an integrated perspective of the wide range of phenomena and processes applicable to the study of transport of species in porous materials. In order to formulate the entire range of porous media and their uses, this book gives the basics of continuum mechanics, thermodynamics, seepage and consolidation and diffusion, including multiscale homogenization methods. The particular structure of the book has been chosen because it is essential to be aware of the true properties of porous materials particularly in terms of nano, micro and macro mechanisms.  This book is of pedagogical and practical importance to the fields covered by civil, environmental, nuclear and petroleum engineering and also in chemical physics and geophysics as it relates to radioactive waste disposal, geotechnical engineering, mining and petroleum engineering and chemical engineering.

  7. Ab-initio calculations of the hydrogen-uranium system: Surface phenomena, absorption, transport and trapping

    International Nuclear Information System (INIS)

    Taylor, Christopher D.; Scott Lillard, R.

    2009-01-01

    Density functional theory was applied to the initial steps of uranium hydriding: surface phenomena, absorption, bulk transport and trapping. H adsorbs exothermically to the (0 0 1) surface, yet H absorption into the bulk is endothermic, with off-center octahedral absorption having the lowest absorption energy of 0.39 eV, relative to molecular H 2 . H absorption in interstitial sites causes a local softening of the bulk modulus. Diffusion of H in unstrained α-U has a barrier of 0.6 eV. The energy of H absorption adjacent to the chemical impurities C, S, Si was lowered by an amount proportional to the size of the impurity atom, and the resulting lattice strain Si > S > C. Thus, impurities may promote hydriding by providing surfaces or prestrained zones for H uptake.

  8. Studies of the effect of radioactive waste on the transport phenomena in soil

    International Nuclear Information System (INIS)

    El-Reefy, A.I.A.F

    1992-01-01

    This thesis introduces a new concept in the field of soil mechanics. It is an integrated work between soil and radiation in the form of gamma-rays. Chapter II was introduced to cover the basics in geotechnical engineering so as to draw a clearer picture to radiologists. Similarly, Chapter III was introduced to enable geotechnical engineers to comprehend radioactive behaviour in general. Although these two chapters are for further reading they contain various points that will be referred to regularly in the latter pages. The aim of this work is to investigate: - The effect of γ -radiations on the transport phenomena in soil. This was carried out by studying the effect of the following factors on the transmission of γ -rays with different energies: 1) Soil sample thickness 2) Grain size 3)Water content 4) Degree of compaction. - The effect of γ -radiations on moisture movement through soil. -Using the -ray transmission method to determine the soil physical properties. - Improvement of soil to increase its ability to attenuate -radiations. Experimental work took place under strict conditions at the Hot Lab. Center located at Inchas. Soil sample was sought from a nearby site which eventually will be the actual radioactive disposal site. The physical properties of the soil sample were determined as well as its grain size distribution. Accurate and detailed data on the gamma rays transport phenomena in soils was obtained using an up to date γ -radiation measurement technique. Finally, the extensive data obtained throughout this research was recorded and analyzed to ultimately approach our aim

  9. Heat science and transport phenomena in fuel cells; Thermique et phenomenes de transport dans les piles a combustible

    Energy Technology Data Exchange (ETDEWEB)

    Liberatore, P.M.; Boillot, M. [Laboratoire des Sciences du Genie Chimique de Nancy, 54 - Vandoeuvre-les-Nancy (France); Bonnet, C.; Didieerjean, S.; Lapicque, F.; Deseure, J.; Lottin, O.; Maillet, D.; Oseen-Senda, J. [Laboratoire d' Energetique et de Mecanique Theorique et Appliquee, 54 - Vandoeuvre Les Nancy (France); Alexandre, A. [Laboratoire d' Etudes Thermiques, ENSMA, 86 Poitiers (France); Topin, F.; Occelli, R.; Daurelle, J.V. [IUSTI / Polytech' Marseille, Institut universitaire des Systemes Thermiques Industriels Ecole, 13 - Marseille (France); Pauchet, J.; Feidt, M. [CEA Grenoble, Groupement pour la recherche sur les echangeurs thermiques (Greth), 38 (France); Voarino, C. [CEA Centre d' Etudes du Ripault, 37 - Tours (France); Morel, B.; Laurentin, J.; Bultel, Y.; Lefebvre-Joud, F. [CEA Grenoble, LEPMI, 38 (France); Auvity, B.; Lasbet, Y.; Castelain, C.; Peerohossaini, H. [Ecole Centrale de Nantes, Laboratoire de Thermocinetique de Nantes (LTN), 44 - Nantes (France)

    2005-07-01

    In this work are gathered the transparencies of the lectures presented at the conference 'heat science and transport phenomena in fuel cells'. The different lectures have dealt with 1)the gas distribution in the bipolar plates of a fuel cell: experimental studies and computerized simulations 2)two-phase heat distributors in the PEMFC 3)a numerical study of the flow properties of the backing layers on the transfers in a PEMFC 4)modelling of the heat and mass transfers in a PEMFC 5)two-phase cooling of the PEMFC with pentane 6)stationary thermodynamic model of the SOFC in the GECOPAC system 7)modelling of the internal reforming at the anode of the SOFC 8)towards a new thermal design of the PEMFC bipolar plates. (O.M.)

  10. Continuous energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the neutron energy spectrum by target nuclei tagging

    Energy Technology Data Exchange (ETDEWEB)

    Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Grupo de Estudos Nucleares; Leite, Sergio Q. Bogado, E-mail: sbogado@ibest.com.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In this work a Monte Carlo simulator with continuous energy is used. This simulator distinguishes itself by using the sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum (for high energy neutrons) and the Maxwell-Boltzmann distribution (for thermal neutrons). The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. It is common practice in neutron transport calculations, e.g. multi-group transport, to consider that the neutrons only lose energy with each scattering reaction and then to use a thermal group with a Maxwellian distribution. Such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies, i.e. in the thermal energy region, in which it can be regarded as a Maxwell-Boltzmann distribution for thermal equilibrium. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution. It is then shown how this procedure can emulate the up-scattering effect by the increase in the neutron population kinetic energy. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process. This work contains some preliminary results obtained from a Monte Carlo simulator for neutron transport that is being developed at Federal University of Rio Grande do Sul. (author)

  11. Kinetic phenomena in charged particle transport in gases, swarm parameters and cross section data

    International Nuclear Information System (INIS)

    Petrovic, Z Lj; Suvakov, M; Nikitovic, Z; Dujko, S; Sasic, O; Jovanovic, J; Malovic, G; Stojanovic, V

    2007-01-01

    In this review we discuss the current status of the physics of charged particle swarms, mainly electrons. The whole field is analysed mainly through its relationship to plasma modelling and illustrated by some recent examples developed mainly by our group. The measurements of the swarm coefficients and the availability of the data are briefly discussed. More time is devoted to the development of complete electron-molecule cross section sets along with recent examples such as NO, CF 4 and HBr. We extend the discussion to the availability of ion and fast neutral data and how swarm experiments may serve to provide new data. As a point where new insight into the kinetics of charge particle transport is provided, the role of kinetic phenomena is discussed and recent examples are listed. We focus here on giving two examples on how non-conservative processes make dramatic effects in transport, the negative absolute mobility and the negative differential conductivity for positrons in argon. Finally we discuss the applicability of swarm data in plasma modelling and the relationship to other fields where swarm experiments and analysis make significant contributions. (topical review)

  12. Generalized diffusion theory for calculating the neutron transport scalar flux

    International Nuclear Information System (INIS)

    Alcouffe, R.E.

    1975-01-01

    A generalization of the neutron diffusion equation is introduced, the solution of which is an accurate approximation to the transport scalar flux. In this generalization the auxiliary transport calculations of the system of interest are utilized to compute an accurate, pointwise diffusion coefficient. A procedure is specified to generate and improve this auxiliary information in a systematic way, leading to improvement in the calculated diffusion scalar flux. This improvement is shown to be contingent upon satisfying the condition of positive calculated-diffusion coefficients, and an algorithm that ensures this positivity is presented. The generalized diffusion theory is also shown to be compatible with conventional diffusion theory in the sense that the same methods and codes can be used to calculate a solution for both. The accuracy of the method compared to reference S/sub N/ transport calculations is demonstrated for a wide variety of examples. (U.S.)

  13. Beam-transport optimization for cold-neutron spectrometer

    Directory of Open Access Journals (Sweden)

    Nakajima Kenji

    2015-01-01

    Full Text Available We report the design of the beam-transport system (especially the vertical geometry for a cold-neutron disk-chopper spectrometer AMATERAS at J-PARC. Based on the elliptical shape, which is one of the most effective geometries for a ballistic mirror, the design was optimized to obtain, at the sample position, a neutron beam with high flux without serious degrading in divergence and spacial homogeneity within the boundary conditions required from actual spectrometer construction. The optimum focal point was examined. An ideal elliptical shape was modified to reduce its height without serious loss of transmission. The final result was adapted to the construction requirements of AMATERAS. Although the ideas studied in this paper are considered for the AMATERAS case, they can be useful also to other spectrometers in similar situations.

  14. Colloids: a review of current knowledge with a view to application to phenomena of transportation within PWR

    International Nuclear Information System (INIS)

    Guinard, L.

    1996-01-01

    In an attempt to minimise dosimetry within the primary circuit of PWR units, research is being carried out into understanding the phenomena of transportation and deposition of corrosion products. It is therefore desirable to known the form of these corrosion products and the laws governing this form. It is generally considered that they are in soluble or particulate form. A third starts with a general presentation of colloids and goes on to define points which are useful, both on a theoretical and experimental level, in terms of application to phenomena of transportation within PWRs. (author). 69 refs., 30 figs., 6 tabs., 3 appends

  15. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  16. A three-dimensional neutron transport benchmark solution

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Kornreich, D.E.

    1993-01-01

    For one-group neutron transport theory in one dimension, several powerful analytical techniques have been developed to solve the neutron transport equation, including Caseology, Wiener-Hopf factorization, and Fourier and Laplace transform methods. In addition, after a Fourier transform in the transverse plane and formulation of a pseudo problem, two-dimensional (2-D) and three-dimensional (3-D) problems can be solved using the techniques specifically developed for the one-dimensional (1-D) case. Numerical evaluation of the resulting expressions requiring an inversion in the transverse plane have been successful for 2-D problems but becomes exceedingly difficult in the 3-D case. In this paper, we show that by using the symmetry along the beam direction, a 2-D problem can be transformed into a 3-D problem in an infinite medium. The numerical solution to the 3-D problem is then demonstrated. Thus, a true 3-D transport benchmark solution can be obtained from a well-established numerical solution to a 2-D problem

  17. Error reduction techniques for Monte Carlo neutron transport calculations

    International Nuclear Information System (INIS)

    Ju, J.H.W.

    1981-01-01

    Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas

  18. Novel Parallel Numerical Methods for Radiation and Neutron Transport

    International Nuclear Information System (INIS)

    Brown, P N

    2001-01-01

    In many of the multiphysics simulations performed at LLNL, transport calculations can take up 30 to 50% of the total run time. If Monte Carlo methods are used, the percentage can be as high as 80%. Thus, a significant core competence in the formulation, software implementation, and solution of the numerical problems arising in transport modeling is essential to Laboratory and DOE research. In this project, we worked on developing scalable solution methods for the equations that model the transport of photons and neutrons through materials. Our goal was to reduce the transport solve time in these simulations by means of more advanced numerical methods and their parallel implementations. These methods must be scalable, that is, the time to solution must remain constant as the problem size grows and additional computer resources are used. For iterative methods, scalability requires that (1) the number of iterations to reach convergence is independent of problem size, and (2) that the computational cost grows linearly with problem size. We focused on deterministic approaches to transport, building on our earlier work in which we performed a new, detailed analysis of some existing transport methods and developed new approaches. The Boltzmann equation (the underlying equation to be solved) and various solution methods have been developed over many years. Consequently, many laboratory codes are based on these methods, which are in some cases decades old. For the transport of x-rays through partially ionized plasmas in local thermodynamic equilibrium, the transport equation is coupled to nonlinear diffusion equations for the electron and ion temperatures via the highly nonlinear Planck function. We investigated the suitability of traditional-solution approaches to transport on terascale architectures and also designed new scalable algorithms; in some cases, we investigated hybrid approaches that combined both

  19. An application of reactor noise techniques to neutron transport problems in a random medium

    International Nuclear Information System (INIS)

    Sahni, D.C.

    1989-01-01

    Neutron transport problems in a random medium are considered by defining a joint Markov process describing the fluctuations of one neutron population and the random changes in the medium. Backward Chapman-Kolmogorov equations are derived which yield an adjoint transport equation for the average neutron density. It is shown that this average density also satisfied the direct transport equation as given by the phenomenological model. (author)

  20. MINARET: Towards a time-dependent neutron transport parallel solver

    International Nuclear Information System (INIS)

    Baudron, A.M.; Lautard, J.J.; Maday, Y.; Mula, O.

    2013-01-01

    We present the newly developed time-dependent 3D multigroup discrete ordinates neutron transport solver that has recently been implemented in the MINARET code. The solver is the support for a study about computing acceleration techniques that involve parallel architectures. In this work, we will focus on the parallelization of two of the variables involved in our equation: the angular directions and the time. This last variable has been parallelized by a (time) domain decomposition method called the para-real in time algorithm. (authors)

  1. Criticality problems in energy dependent neutron transport theory

    International Nuclear Information System (INIS)

    Victory, H.D. Jr.

    1979-01-01

    The criticality problem is considered for energy dependent neutron transport in an isotropically scattering, homogeneous slab. Under a positivity assumption on the scattering kernel, an expression can be found relating the thickness of the slab to a parameter characterizing production by fission. This is accomplished by exploiting the Perron-Frobenius-Jentsch characterization of positive operators (i.e. those leaving invariant a normal, reproducing cone in a Banach space). It is pointed out that those techniques work for classes of multigroup problems were the Case singular eigenfunction approach is not as feasible as in the one-group theory, which is also analyzed

  2. Reliability analysis of neutron transport simulation using Monte Carlo method

    International Nuclear Information System (INIS)

    Souza, Bismarck A. de; Borges, Jose C.

    1995-01-01

    This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs

  3. Structures of the fractional spaces generated by the difference neutron transport operator

    International Nuclear Information System (INIS)

    Ashyralyev, Allaberen; Taskin, Abdulgafur

    2015-01-01

    The initial boundary value problem for the neutron transport equation is considered. The first, second and third order of accuracy difference schemes for the approximate solution of this problem are presented. Highly accurate difference schemes for neutron transport equation based on Padé approximation are constructed. In applications, stability estimates for solutions of difference schemes for the approximate solution of the neutron transport equation are obtained.The positivity of the neutron transport operator in Slobodeckij spaces is proved. Numerical techniques are developed and algorithms are tested on an example in MATLAB

  4. Modeling diffusion-governed solidification of ternary alloys - Part 2: Macroscopic transport phenomena and macrosegregation.

    Science.gov (United States)

    Wu, M; Li, J; Ludwig, A; Kharicha, A

    2014-09-01

    Part 1 of this two-part investigation presented a multiphase solidification model incorporating the finite diffusion kinetics and ternary phase diagram with the macroscopic transport phenomena (Wu et al., 2013). In Part 2, the importance of proper treatment of the finite diffusion kinetics in the calculation of macrosegregation is addressed. Calculations for a two-dimensional (2D) square casting (50 × 50 mm 2 ) of Fe-0.45 wt.%C-1.06 wt.%Mn considering thermo-solutal convection and crystal sedimentation are performed. The modeling result indicates that the infinite liquid mixing kinetics as assumed by classical models (e.g., the Gulliver-Scheil or lever rule), which cannot properly consider the solute enrichment of the interdendritic or inter-granular melt at the early stage of solidification, might lead to an erroneous estimation of the macrosegregation. To confirm this statement, further theoretical and experimental evaluations are desired. The pattern and intensity of the flow and crystal sedimentation are dependent on the crystal morphologies (columnar or equiaxed); hence, the potential error of the calculated macrosegregation caused by the assumed growth kinetics depends on the crystal morphology. Finally, an illustrative simulation of an engineering 2.45-ton steel ingot is performed, and the results are compared with experimental results. This example demonstrates the model applicability for engineering castings regarding both the calculation efficiency and functionality.

  5. Study of phenomena of tracer transport and dispersion in fractured media

    International Nuclear Information System (INIS)

    Ippolito, Irene

    1993-01-01

    The objective of this research thesis is to present some transport phenomena according to two different approaches: firstly, the study of flows and tracing in a natural crack within a granitic site, and secondly, the study of flows of different geometries in model cracks, mainly by using techniques of tracer dispersion. The author first presents some properties of fractured media and elements of the theory of the phenomenon of dispersion. She notably discusses the reversibility of the Taylor dispersion which is the prevailing mechanism for simply connected geometries such as in the case of a flow between two continuous solid surfaces limiting a fracture. In the next chapters, the author reports the analysis of characteristics of local structures (mouths, roughnesses) of a single fracture by using echo dispersion. She reports experiments as well as Monte Carlo simulations performed on well defined geometries. In a parallel way, some characteristics measurements (rate-pressure, distribution of flows and tracing in transmission) and mechanical measurements of fracture deformation have been performed on a natural fracture in a granitic site [fr

  6. The application of the Chebyshev-spectral method in transport phenomena

    CERN Document Server

    Guo, Weidong; Narayanan, Ranga

    2012-01-01

    Transport phenomena problems that occur in engineering and physics are often multi-dimensional and multi-phase in character.  When taking recourse to numerical methods the spectral method is particularly useful and efficient. The book is meant principally to train students and non-specialists  to use the spectral method for solving problems that model fluid flow in closed geometries with heat or mass transfer.  To this aim the reader should bring a working knowledge of fluid mechanics and heat transfer and should be readily conversant with simple concepts of linear algebra including spectral decomposition of matrices as well as solvability conditions for inhomogeneous problems.  The book is neither meant to supply a ready-to-use program that is all-purpose nor to go through all manners of mathematical proofs.  The focus in this tutorial is on the use of the spectral methods for space discretization, because this is where most of the difficulty lies. While time dependent problems are also of great interes...

  7. Magneto-transport phenomena in metal/SiO2/n(p)-Si hybrid structures

    Science.gov (United States)

    Volkov, N. V.; Tarasov, A. S.; Rautskii, M. V.; Lukyanenko, A. V.; Bondarev, I. A.; Varnakov, S. N.; Ovchinnikov, S. G.

    2018-04-01

    Present review touches upon a subject of magnetotransport phenomena in hybrid structures which consist of ferromagnetic or nonmagnetic metal layer, layer of silicon oxide and silicon substrate with n- or p-type conductivity. Main attention will be paid to a number gigantic magnetotransport effects discovered in the devices fabricated on the base of the M/SiO2/n(p)-Si (M is ferromagnetic or paramagnetic metal) hybrid structures. These effects include bias induced dc magnetoresistance, gigantic magnetoimpedance, dc magnetoresistance induced by an optical irradiation and lateral magneto-photo-voltaic effect. The magnetoresistance ratio in ac and dc modes for some of our devices can exceed 106% in a magnetic field below 1 T. For lateral magneto-photo-voltaic effect, the relative change of photo-voltage in magnetic field can reach 103% at low temperature. Two types of mechanisms are responsible for sensitivity of the transport properties of the silicon based hybrid structures to magnetic field. One is related to transformation of the energy structure of the (donor) acceptor states including states near SiO2/n(p)-Si interface in magnetic field. Other mechanism is caused by the Lorentz force action. The features in behaviour of magnetotransport effects in concrete device depend on composition of the used structure, device topology and experimental conditions (bias voltage, optical radiation and others). Obtained results can be base for design of some electronic devices driven by a magnetic field. They can also provide an enhancement of the functionality for existing sensors.

  8. Modelling of melting and solidification transport phenomena during hypothetical NPP severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sarler, B [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    A physical and mathematical framework to deal with the transport phenomena occuring during melting and solidification of the hypothetical NPP severe accidents is presented. It concentrates on the transient temperature, velocity, and species concentration distributions during such events. The framework is based on the Mixture Continuum Formulation of the components and phases, cast in the boundary-domain integral shape structured by the fundamental solution of the Laplace equation. The formulation could cope with various solid-liquid sub-systems through the inclusion of the specific closure relations. The deduced system of boundary-domain integral equations for conservation of mass, energy, momentum, and species could be solved by the boundary element discrete approximative method. (author) [Slovenian] Predstavljeno je fizikalno in matematicno ogrodje za obravnavo prenosnih pojavov taljenja in strjevanja med hipoteticnimi tezkimi nezgodami v jedrskih elektrarnah. Osredotoceno je na popis neustaljene porazdelitve temperatur, hitrosti in koncentracij sestavin med taksnimi dogodki. Ogrodje temelji na formulaciji kontinuuma mesanice komponent in faz, v obliki robno obmocnih integralskih enacb, ki so sestavljena na podlagi fundamentalne resitve Laplace-ove enacbe. Formulacija lahko popisuje stevilne trdno-tekoce pod-sisteme na podlagi specificnih sklopitvenih relacij. Izpeljan sistem robno-obmocnih integralskih enacb za popis ohranitve mase, energije, gibalne kolicine in sestavin lahko resimo na podlagi diskretne aproksimativne metode robnih elementov. (author)

  9. Predictive modelling of edge transport phenomena in ELMy H-mode tokamak fusion plasmas

    International Nuclear Information System (INIS)

    Loennroth, J.-S.

    2009-01-01

    This thesis discusses a range of work dealing with edge plasma transport in magnetically confined fusion plasmas by means of predictive transport modelling, a technique in which qualitative predictions and explanations are sought by running transport codes equipped with models for plasma transport and other relevant phenomena. The focus is on high confinement mode (H-mode) tokamak plasmas, which feature improved performance thanks to the formation of an edge transport barrier. H-mode plasmas are generally characterized by the occurrence of edge localized modes (ELMs), periodic eruptions of particles and energy, which limit confinement and may turn out to be seriously damaging in future tokamaks. The thesis introduces schemes and models for qualitative study of the ELM phenomenon in predictive transport modelling. It aims to shed new light on the dynamics of ELMs using these models. It tries to explain various experimental observations related to the performance and ELM-behaviour of H-mode plasmas. Finally, it also tries to establish more generally the potential effects of ripple-induced thermal ion losses on H-mode plasma performance and ELMs. It is demonstrated that the proposed ELM modelling schemes can qualitatively reproduce the experimental dynamics of a number of ELM regimes. Using a theory-motivated ELM model based on a linear instability model, the dynamics of combined ballooning-peeling mode ELMs is studied. It is shown that the ELMs are most often triggered by a ballooning mode instability, which renders the plasma peeling mode unstable, causing the ELM to continue in a peeling mode phase. Understanding the dynamics of ELMs will be a key issue when it comes to controlling and mitigating the ELMs in future large tokamaks. By means of integrated modelling, it is shown that an experimentally observed increase in the ELM frequency and deterioration of plasma confinement triggered by external neutral gas puffing might be due to a transition from the second to

  10. Two-dimensional time dependent Riemann solvers for neutron transport

    International Nuclear Information System (INIS)

    Brunner, Thomas A.; Holloway, James Paul

    2005-01-01

    A two-dimensional Riemann solver is developed for the spherical harmonics approximation to the time dependent neutron transport equation. The eigenstructure of the resulting equations is explored, giving insight into both the spherical harmonics approximation and the Riemann solver. The classic Roe-type Riemann solver used here was developed for one-dimensional problems, but can be used in multidimensional problems by treating each face of a two-dimensional computation cell in a locally one-dimensional way. Several test problems are used to explore the capabilities of both the Riemann solver and the spherical harmonics approximation. The numerical solution for a simple line source problem is compared to the analytic solution to both the P 1 equation and the full transport solution. A lattice problem is used to test the method on a more challenging problem

  11. Criticality of neutron transport in a slab with finite reflectors

    International Nuclear Information System (INIS)

    Pao, C.V.

    1978-01-01

    The purpose of this paper is to investigate the subcriticality and the supercriticality for the neutron transport in a slab which is surrounded by two finite reflectors. The mathematical problem is to determine when the coupled boundary-value problem has or has no positive solution. It is shown under some explicit conditions on the material properties of the transport mediums and the size of the slab length that the coupled problem has a unique solution which insures the subcriticality of the system. It is also shown under some different conditions on the same physical quantities that the system cannot have a nonnegative solution when there is an external source, and it only has the trivial solution when there is no source in the system. This conclusion leads to the supercriticality of the system. Both upper and lower bounds for the critical length of the slab are explicitly given

  12. SUSD, Sensitivity and Uncertainty in Neutron Transport and Detector Response

    International Nuclear Information System (INIS)

    Furuta, Lazuo; Kondo, Shunsuke; Oka, Yoshika

    1991-01-01

    1 - Description of program or function: SUSD calculates sensitivity coefficients for one and two-dimensional transport problems. Variance and standard deviation of detector responses or design parameters can be obtained using cross-section covariance matrices. In neutron transport problems, this code is able to perform sensitivity-uncertainty analysis for secondary angular distribution (SAD) or secondary energy distribution (SED). 2 - Method of solution: The first-order perturbation theory is used to obtain sensitivity coefficients. The method described in the distributed report is employed to consider SAD/SED effect. 3 - Restrictions on the complexity of the problem: Variable dimension is used so that there is no limitation in each array size but the total core size

  13. Transport Phenomena Projects: Natural Convection between Porous, Concentric Cylinders--A Method to Learn and to Innovate

    Science.gov (United States)

    Saatadjian, Esteban; Lesage, Francois; Mota, Jose Paulo B.

    2013-01-01

    A project that involves the numerical simulation of transport phenomena is an excellent method to teach this subject to senior/graduate chemical engineering students. The subject presented here has been used in our senior/graduate course, it concerns the study of natural convection heat transfer between two concentric, horizontal, saturated porous…

  14. Research and Training of Using Neutron Scattering to Probe the Collective Phenomena in Doped Transition-Metal Oxides. Final report

    International Nuclear Information System (INIS)

    Zhang, Jiandi

    2008-01-01

    The objective of this funded research program include: explore and understand the microscopic origins of collective phenomena in doped transition-metal oxides (TMOs) using neutrons as one of the primary tools, and train new generation of neutron scatters and collaborate with Oak Ridge National Lab in both materials synthesis and characterization. The major physics issues focused on in this project consist of the microscopic correlations between lattice structure and magnetic ordering, the nature of elementary lattice and spin excitations, the origin of nanometer-scale phase separations, and the effects of dimensional confinement and broken symmetry. The main materials are doped TMOs grown as single crystals by a floating-zone technique at ORNL as well as multiplayer films grown with a laser-MBE facility at Florida International University (FIU). Our educational objective is the training of our graduate and undergraduate students, especially Hispanic and other minority students, to use neutrons as a probe for materials research by taking advantage of national neutron facilities and to grow novel materials by using the floating-zone and laser-MBE technique. The main achievements of the project include the systematic study of the spin dynamics, especially the spin wave excitations in ferromagnetic manganites; the discovery of the critical doping concentration for the magnetic phase separation of the charge-ordered state in Pr 1-x Ca x MnO 3 - ; the study of Σ 4 phonon softening associated with the lattice instability near the quantum critical point as well as the discovery of an anomalous mode in single-layered ruthenates. These results gain some important insights into the collective excitations in both spin and lattice degrees of freedom as well as their close coupling in these correlated TMO systems. Furthermore, this project also accomplished the synthesis and some characterization of the single crystals of a new material Ba 2-x Sr x CoO 4 , a compound in which

  15. Study of influence of transport performance of the neutron guide

    International Nuclear Information System (INIS)

    Li Xinxi; Wang Yan; Huang Chaoqiang; Chen Bo; Chen Liang

    2009-01-01

    For the sake of improving the performance of the neutron scattering instrument, usually we need use the neutron guide, it's very important to select the right type and optimizing of neutron guide. The papers calculate the focus neutron guide and the single channel neutron guide by numeric method. The results shows that the choice of neutron guide should consult the resolution requirement of neutron scattering instrument, and the length of the neutron guide should be optimized. The calculation results can be the theoretical reference for the design of neutron scattering instrument. (authors)

  16. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    International Nuclear Information System (INIS)

    Ruggles, A.E.; Cheng, L.Y.; Dimenna, R.A.; Griffith, P.; Wilson, G.E.

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis

  17. In situ diagnostic of two-phase flow phenomena in polymer electrolyte fuel cells by neutron imaging

    International Nuclear Information System (INIS)

    Zhang Jianbo; Kramer, Denis; Shimoi, Ryoichi; Ono, Yoshitaka; Lehmann, Eberhard; Wokaun, Alexander; Shinohara, Kazuhiko; Scherer, Guenther G.

    2006-01-01

    The formation of liquid water in operating polymer electrolyte fuel cells (PEFC) of industrial and laboratory size has been investigated by in situ neutron imaging. The influence of the materials chosen for the structural components of the cell on droplet formation and transport in flow fields and on liquid formation in gas diffusion layers has been studied. The changing of the cathodic gas diffusion layer material allowed the relationship between materials, liquid accumulation, and electrochemical performance to be examined. It has been shown that material choice has considerable bearing on the presence of liquid inside the porous structures and the electrochemical characteristics. A simplified quasi one-dimensional cell with an active area of 25 cm 2 was used for materials comparison, and the results were related to technically relevant operating conditions - where inhomogeneities have to be considered - by subsequent examination of cells with an active area of 100 cm 2

  18. Peculiarities of the radioactive particles transport phenomena in the facilitated diffusion processes

    International Nuclear Information System (INIS)

    Gavryushenko, D.A.; Sisojev, V.M.; Cherevko, K.V.; Vlasenko, T.S.

    2017-01-01

    The work is devoted to the up to date problem that is the description of the radioactive particle diffusion processes. One of the aims of the present study is to estimate the effects caused by the irradiation of the liquid systems on the ongoing transport processes. That can allow predicting the behavior of the liquid systems in the presence of the radioactive sources. The main objective of the present work is studying the radioactive particles diffusion phenomena with the possible facilitated diffusion processes being considered. The phenomena are studied based on the fundamental relations of the nonequilibrium statistical thermodynamics. The diffusive flows are evaluated with the special attention given to the accounting of the entropy effects due to the appearance of the new radioactive particles in the system. The developed approach is used to estimate the diffusive flow of the radioactive particles for the case of the plane-parallel pore with the semi-transparent walls. The choice of a model can be justified as it might be used to describe the production of the radioactive contaminated water when the radionuclide face the diffusion process after being washed from the radioactive wastes and the rests of the nuclear fuel. Within the suggested model it is shown that the diffusion coefficient depends on the structural properties of the liquid systems that might be changed under the influence of the irradiation. The obtained equations for calculating the diffusive flows show the definite stabilizing effect in respect to the concentration difference in between the boundaries of the plane-parallel pore. It leads to the decreased changes of the diffusive flow when the concentrations of the radioactive particles at the boundaries are changed in comparison with those observed for the constant diffusion coefficient. The observed behavior for the ideal solution model is explained by the entropy effects. The qualitative analysis of the possible influence of the changes in

  19. Transport calculations for a 14.8 MeV neutron beam in a water phantom

    International Nuclear Information System (INIS)

    Goetsch, S.J.

    1981-01-01

    A coupled neutron/photon Monte Carlo radiation transport code (MORSE-CG) has been used to calculate neutron and photon doses in a water phantom irradiated by 14.8 MeV neutrons from the Gas Target Neutron Source. The source-collimator-phantom geometry was carefully simulated. Results of calculations utilizing two different statistical estimators (next-collision and track-length) are presented

  20. Simulations of neutron transport at low energy: a comparison between GEANT and MCNP.

    Science.gov (United States)

    Colonna, N; Altieri, S

    2002-06-01

    The use of the simulation tool GEANT for neutron transport at energies below 20 MeV is discussed, in particular with regard to shielding and dose calculations. The reliability of the GEANT/MICAP package for neutron transport in a wide energy range has been verified by comparing the results of simulations performed with this package in a wide energy range with the prediction of MCNP-4B, a code commonly used for neutron transport at low energy. A reasonable agreement between the results of the two codes is found for the neutron flux through a slab of material (iron and ordinary concrete), as well as for the dose released in soft tissue by neutrons. These results justify the use of the GEANT/MICAP code for neutron transport in a wide range of applications, including health physics problems.

  1. Effect of fractional parameter on neutron transport in finite disturbed reactors with quadratic scattering

    International Nuclear Information System (INIS)

    Sallah, M.; Margeanu, C. A.

    2016-01-01

    The space-fractional neutron transport equation is used to describe the neutrons transport in finite disturbed reactors. It is approximated using the Pomraning-Eddington technique to yield two space-fractional differential equations, in terms of neutron density and net neutron flux. These resultant equations are coupled into a fractional diffusion-like equation for the neutron density whose solution is obtained by using Laplace transformation method. The solution is represented in terms of the Mittag-Leffler function and its different orders. The scattering is considered as quadratic scattering to offer a more realistic, compact representation of the system, and to increase the accuracy of the estimated neutronic parameters. The results are presented graphically to illustrate the fractional parameter effect in addition to the effect of radiative-transfer properties on the physical parameters of interest (reflection coefficient, transmission coefficient, neutron energy, and net neutron flux). The neutron transport problem in finite disturbed reactor with quadratic scattering is considered in investigating the shielding effectiveness, by using MAVRIC shielding module from SCALE6 programs package. The fractional parameter can be used to adjust the analysed data on neutron energy and flux, both for the theoretical model and the neutron transport application. (authors)

  2. TTF/TCNQ-based thin films and microcrystals. Growth and charge transport phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Solovyeva, Vita

    2011-05-26

    The thesis adresses several problems related to growth and charge transport phenomena in thin films of TTF-TCNQ and (BEDT-TTF)TCNQ. The following main new problems are addressed: - The influence of thin-film specific factors, such as the substrate material and growth-induced defects, on the Peierls transition temperature in TTF-TCNQ thin films was studied; - finite-size effects in TTF-TCNQ were investigated by considering transport properties in TTF-TCNQ microcrystals. The influence of the size of the crystal on the Peierls transition temperature was studied. In this context a new method of microcontact fabrication was employed to favor the measurements; - an analysis of radiation-induced defects in TTF-TCNQ thin films and microcrystals was performed. It was demonstrated than an electron beam can induce appreciable damage to the sample such that its electronic properties are strongly modified; - a bilayer growth method was established to fabricate (BEDT-TTF)TCNQ from the gas phase. This newly developed bilayer growth method was showed to be suitable for testing (BEDT-TTF)TCNQ charge-transfer phase formation; - the structure of the formed (BEDT-TTF)TCNQ charge-transfer compounds was analyzed by using a wide range of experimental techniques. An overview and the description of the basic physical principles underlying charge-transfer compounds is given in chapter 2. Experimental techniques used for the growth and characterization of thin films and microcrystals are presented in chapter 3. Chapter 4 gives an overview of the physical properties of the studied organic materials. Chapter 5 discussed the experimental study of TTF-TCNQ thin films. he Peierls transition in TTF-TCNQ is a consequence of the quasi-one-dimensional structure of the material and depends on different factors, studied in chapters 5 and 6. In contradistinction to TTF-TTCNQ, the (BEDT-TTF)TCNQ charge-transfer compound crystallizes in several different modifications with different physical properties

  3. The transport of neutrons and gamma-rays in the air

    International Nuclear Information System (INIS)

    Adamski, J.

    1980-01-01

    The transport of neutrons and gamma rays in the infinite homogeneous air has been investigated. For the calculations has been used the Multigroup One Dimensional Discrete Ordinates Transport Code ANISN-W. The calculations have been performed for three types of neutron sources. The neutrons and gamma ray doses in the air have been analyzed, and comparison to the other authors' results has been given. (author)

  4. Parallel computing for homogeneous diffusion and transport equations in neutronics

    International Nuclear Information System (INIS)

    Pinchedez, K.

    1999-06-01

    Parallel computing meets the ever-increasing requirements for neutronic computer code speed and accuracy. In this work, two different approaches have been considered. We first parallelized the sequential algorithm used by the neutronics code CRONOS developed at the French Atomic Energy Commission. The algorithm computes the dominant eigenvalue associated with PN simplified transport equations by a mixed finite element method. Several parallel algorithms have been developed on distributed memory machines. The performances of the parallel algorithms have been studied experimentally by implementation on a T3D Cray and theoretically by complexity models. A comparison of various parallel algorithms has confirmed the chosen implementations. We next applied a domain sub-division technique to the two-group diffusion Eigen problem. In the modal synthesis-based method, the global spectrum is determined from the partial spectra associated with sub-domains. Then the Eigen problem is expanded on a family composed, on the one hand, from eigenfunctions associated with the sub-domains and, on the other hand, from functions corresponding to the contribution from the interface between the sub-domains. For a 2-D homogeneous core, this modal method has been validated and its accuracy has been measured. (author)

  5. Finite element based composite solution for neutron transport problems

    International Nuclear Information System (INIS)

    Mirza, A.N.; Mirza, N.M.

    1995-01-01

    A finite element treatment for solving neutron transport problems is presented. The employs region-wise discontinuous finite elements for the spatial representation of the neutron angular flux, while spherical harmonics are used for directional dependence. Composite solutions has been obtained by using different orders of angular approximations in different parts of a system. The method has been successfully implemented for one dimensional slab and two dimensional rectangular geometry problems. An overall reduction in the number of nodal coefficients (more than 60% in some cases as compared to conventional schemes) has been achieved without loss of accuracy with better utilization of computational resources. The method also provides an efficient way of handling physically difficult situations such as treatment of voids in duct problems and sharply changing angular flux. It is observed that a great wealth of information about the spatial and directional dependence of the angular flux is obtained much more quickly as compared to Monte Carlo method, where most of the information in restricted to the locality of immediate interest. (author)

  6. Nodal methods for problems in fluid mechanics and neutron transport

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1985-01-01

    A new high-accuracy, coarse-mesh, nodal integral approach is developed for the efficient numerical solution of linear partial differential equations. It is shown that various special cases of this general nodal integral approach correspond to several high efficiency nodal methods developed recently for the numerical solution of neutron diffusion and neutron transport problems. The new approach is extended to the nonlinear Navier-Stokes equations of fluid mechanics; its extension to these equations leads to a new computational method, the nodal integral method which is implemented for the numerical solution of these equations. Application to several test problems demonstrates the superior computational efficiency of this new method over previously developed methods. The solutions obtained for several driven cavity problems are compared with the available experimental data and are shown to be in very good agreement with experiment. Additional comparisons also show that the coarse-mesh, nodal integral method results agree very well with the results of definitive ultra-fine-mesh, finite-difference calculations for the driven cavity problem up to fairly high Reynolds numbers

  7. Transport Phenomena in Nanowires, Nanotubes, and Other Low-Dimensional Systems

    KAUST Repository

    Montes, Enrique

    2017-01-01

    Nanoscale materials are not new in either nature or physics. However, the recent technological improvements have given scientists new tools to understand and quantify phenomena that occur naturally due to quantum confinement effects. In general, these phenomena induce remarkable optical, magnetic, and electronic properties in nanoscale materials in contrast to their bulk counterpart. In addition, scientists have recently developed the necessary tools to control and exploit these properties in electronic devices, in particular field effect transistors, magnetic memories, and gas sensors. In the present thesis we implement theoretical and computational tools for analyzing the ground state and electronic transport properties of nanoscale materials and their performance in electronic devices. The ground state properties are studied within density functional theory using the SIESTA code, whereas the transport properties are investigated using the non-equilibrium Green\\'s functions formalism implemented in the SMEAGOL code. First we study Si-based systems, as Si nanowires are believed to be important building blocks of the next generation of electronic devices. We derive the electron transport properties of Si nanowires connected to Au electrodes and their dependence on the nanowire growth direction, diameter, and length. At equilibrium Au-nanowire distance we find strong electronic coupling between electrodes and nanowire, resulting in low contact resistance. For the tunneling regime, the decay of the conductance with the nanowire length is rationalized using the complex band structure. The nanowires grown along the (110) direction show the smallest decay and the largest conductance and current. Due to the high spin coherence in Si, Si nanowires represent an interesting platform for spin devices. Therefore, we built a magnetic tunneling junction by connecting a (110) Si nanowire to ferromagnetic Fe electrodes. We have find a substantial low bias magnetoresistance of

  8. Neutron transport and Montecarlo method: analysis and revision

    International Nuclear Information System (INIS)

    Perlado, J.M.

    1982-01-01

    The resolution of the neutron transport equation by the Montecarlo method is presented. Coming from an extensive discussion on the best formulation of that equation in order to be treated through the mentioned method, the theoretical bases of the estimator and random-walk generation is extensively explained. The most general expression for the estimators in different physical situations, each with a diverse random-walk, is included in this basical theoretical part. Furthemore, a large revision on the variance reduction methods is made. Its theoretical presentation is claimed to be in connection with the need for each one of them. The use of the adjoint equation, as a part of the importance sampling, Russian Roulette, splitting, exponential transform, conditional and correlated Montecarlo, and one-collision and next-event extimators, are discussed. Finally, come comments in the presentation of the last works on the theoretical prediction of errors in the generation of estimators-random walks are made. (author)

  9. Approximate solution to neutron transport equation with linear anisotropic scattering

    International Nuclear Information System (INIS)

    Coppa, G.; Ravetto, P.; Sumini, M.

    1983-01-01

    A method to obtain an approximate solution to the transport equation, when both sources and collisions show a linearly anisotropic behavior, is outlined and the possible implications for numerical calculations in applied neutronics as well as shielding evaluations are investigated. The form of the differential system of equations taken by the method is quite handy and looks simpler and more manageable than any other today available technique. To go deeper into the efficiency of the method, some typical calculations concerning critical dimension of multiplying systems are then performed and the results are compared with the ones coming from the classical Ssub(N) approximations. The outcome of such calculations leads us to think of interesting developments of the method which could be quite useful in alternative to other today widespread approximate procedures, for any geometry, but especially for curved ones. (author)

  10. Numerical method for solving integral equations of neutron transport. II

    International Nuclear Information System (INIS)

    Loyalka, S.K.; Tsai, R.W.

    1975-01-01

    In a recent paper it was pointed out that the weakly singular integral equations of neutron transport can be quite conveniently solved by a method based on subtraction of singularity. This previous paper was devoted entirely to the consideration of simple one-dimensional isotropic-scattering and one-group problems. The present paper constitutes interesting extensions of the previous work in that in addition to a typical two-group anisotropic-scattering albedo problem in the slab geometry, the method is also applied to an isotropic-scattering problem in the x-y geometry. These results are compared with discrete S/sub N/ (ANISN or TWOTRAN-II) results, and for the problems considered here, the proposed method is found to be quite effective. Thus, the method appears to hold considerable potential for future applications. (auth)

  11. Massively parallel performance of neutron transport response matrix algorithms

    International Nuclear Information System (INIS)

    Hanebutte, U.R.; Lewis, E.E.

    1993-01-01

    Massively parallel red/black response matrix algorithms for the solution of within-group neutron transport problems are implemented on the Connection Machines-2, 200 and 5. The response matrices are dericed from the diamond-differences and linear-linear nodal discrete ordinate and variational nodal P 3 approximations. The unaccelerated performance of the iterative procedure is examined relative to the maximum rated performances of the machines. The effects of processor partitions size, of virtual processor ratio and of problems size are examined in detail. For the red/black algorithm, the ratio of inter-node communication to computing times is found to be quite small, normally of the order of ten percent or less. Performance increases with problems size and with virtual processor ratio, within the memeory per physical processor limitation. Algorithm adaptation to courser grain machines is straight-forward, with total computing time being virtually inversely proportional to the number of physical processors. (orig.)

  12. On the Solution of the Neutron Transport Equation

    Energy Technology Data Exchange (ETDEWEB)

    Depken, S

    1962-12-15

    The neutron transport equation has occupied the attention of many authors since Placzek, Wick and others made their first attempts to solve it, Even in the simple case of energy independent cross-sections, and disregarding the motion of the scattering nucleons, it is difficult to find a solution in an analytical form which is easily surveyable and fitted for numerical calculations. In Part I of this paper some new viewpoints will be introduced which enable the solution to be presented in its simplest possible form. Part II is devoted to an investigation of some functions introduced in Part I. In Part III the results are applied to the case of large energy lethargy, and the validity of derived formulas is discussed.

  13. Parallel processing of neutron transport in fuel assembly calculation

    International Nuclear Information System (INIS)

    Song, Jae Seung

    1992-02-01

    Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

  14. Computational analysis of interfacial attachment kinetics and transport phenomena during liquid phase epitaxy of mercury cadmium telluride

    Energy Technology Data Exchange (ETDEWEB)

    Rasin, Igal; Brandon, Simon [Dept. of Chemical Engineering, Technion, Haifa 32000 (Israel); Ben Dov, Anne; Grimberg, Ilana; Klin, Olga; Weiss, Eliezer [SCD-Semi-Conductor Devices, P.O. Box 2250/99, Haifa 31021 (Israel)

    2010-07-01

    Deposition of mercury cadmium telluride (MCT) thin films, on lattice matched cadmium zinc telluride substrates, is often achieved via Liquid Phase Epitaxy (LPE). The yield and quality of these films, required for the production of infrared detector devices, is to a large extent limited by lack of knowledge regarding details of physical phenomena underlying the deposition process. Improving the understanding of these phenomena and their impact on the quality of the resultant films is therefore an important goal which can be achieved through relevant computational and/or experimental studies. We present a combined computational and experimental effort aimed at elucidating physical phenomena underlying the LPE of MCT via a slider growth process. The focus of the presentation will be results generated by a time-dependent three-dimensional model of mass transport, fluid flow, and interfacial attachment kinetics, which we have developed and applied in the analysis of this LPE process. These results, combined with experimental analyses, lead to an improved understanding of the role of different transport and kinetic phenomena underlying this growth process.

  15. A spin-transport system for a longitudinally polarized epithermal neutron beam

    International Nuclear Information System (INIS)

    Crawford, B.E.; Bowman, J.D.; Penttilae, S.I.; Roberson, N.R.

    2001-01-01

    The TRIPLE (Time Reversal and Parity at Low Energies) collaboration uses a polarized epithermal neutron beam and a capture γ-ray detector to study parity violation in neutron-nucleus reactions. In order to preserve the spin polarization of the neutrons as they travel the 60-m path to the target, the beam pipes are wrapped with wire to produce a solenoidal magnetic field of about 10 G along the beam direction. The flanges and bellows between sections of the beam pipe cause gaps in the windings which in turn produce radial fields that can depolarize the neutron spins. A computer code has been developed that numerically evaluates the effect of these gaps on the polarization. A measurement of the neutron depolarization for neutrons in the actual spin-transport system agrees with a calculation of the neutron depolarization for the TRIPLE system. Features that will aid in designing similar spin-transport systems are discussed

  16. Transport calculation of medium-energy protons and neutrons by Monte Carlo method

    International Nuclear Information System (INIS)

    Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.

    1978-09-01

    A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)

  17. Safety improvement of start-up neutron source handling work by preparing new transport containers

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Sawahata, Hiroaki; Yanagida, Yoshinori; Shinohara, Masanori; Kawamoto, Taiki; Takada, Shoji

    2016-01-01

    The conventional transport containers that have been used in HTTR start-up neutron source replacement work are not specialized type for HTTR start-up neutron source. As the risks associated with the safe handling of neutron source holders due to the above fact, the following three risks have been confirmed: (1) exposure risk due to leakage of neutron source or gamma rays, (2) risk of erroneous fall of neutron source holders, and (3) accident due to incorrect handling of transport containers. This paper reports the risks confirmed in the handling of neutron source holders associated with transport containers and the risk reduction measures, as well as the fabrication of new transport containers. By employing the size-reduction and simple structure, new transport containers have been completed at the same level of costs compared with the continuous use of the conventional transport containers, while satisfying the criteria of Type A transport materials and serving as risk preventive measures. Thus, new transport containers aimed at the risk prevention measures for the handling work of neutron source holders have been completed, and the safety of operation has been improved. (A.O.)

  18. Influence of fission product transport on delayed neutron precursors and decay heat sources in LMFBR accidents

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1981-01-01

    A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted

  19. Experimental validation of GADRAS's coupled neutron-photon inverse radiation transport solver

    International Nuclear Information System (INIS)

    Mattingly, John K.; Mitchell, Dean James; Harding, Lee T.

    2010-01-01

    Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of α-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.

  20. Development of a transportable neutron radiography system for non-destructive tests application

    International Nuclear Information System (INIS)

    Silva, Ademir X. da; Crispim, Verginia R.

    1999-01-01

    This paper presents a study of a transportable neutron radiography system utilizing californium-252. Studies about moderation, collimation and shielding are showed. A Monte Carlo Code, MCNP3b, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio 7,5, for neutron flux up to 6 X 10 -6 cm -2 .s -1 per neutron source. (author)

  1. Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

    1987-02-01

    Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

  2. Mathematical modelling of transport phenomena in radioactive waste-cement-bentonite matrix

    International Nuclear Information System (INIS)

    Plecas, Ilija; Dimovic, Slavko

    2010-01-01

    Document available in extended abstract form only. The leaching rate of 137 Cs from spent mix bead (anion and cation) exchange resins in a cement-bentonite matrix has been studied. Transport phenomena involved in the leaching of a radioactive material from a cement-bentonite matrix are investigated using three methods based on theoretical equations. These are: the diffusion equation for a plane source an equation for diffusion coupled to a first-order equation and an empirical method employing a polynomial equation. The results presented in this paper are from a 25-year mortar and concrete testing project that will influence the design choices for radioactive waste packaging for a future Serbian radioactive waste disposal center. Radioactive waste is waste material containing radioactive chemical elements which does not have a practical purpose. It is often the product of a nuclear process, such as nuclear fission. Waste can also be generated from the processing of fuel for nuclear reactors or nuclear weapons. The main objective in managing and disposing of radioactive (or other) waste is to protect people and the environment. This means isolating or diluting the waste so that the rate or concentration of any radionuclides returned to the biosphere is harmless. Storage as the placement of waste in a nuclear facility where isolation, environmental protection and human control are provided with the intent that the waste will be retrieved at a later time. Disposal as the emplacement of waste in an approved, specified facility (e.g. near surface or geological repository) without the intention of retrieval. The processing of radioactive wastes may be done for economic reasons (e.g. to reduce the volume for storage or disposal, or to recover a 'resource' from the waste), or safety reasons (e.g. converting the waste to a more 'stable' form, such as one that will contain the radionuclide inventory for a long time). Typically processing involves reducing

  3. A CFD analysis of transport phenomena and electrochemical reactions in a tubular-shaped PEM fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Sadiq Al-Baghdadi, Maher A.R. [Fuel Cell Research Center, International Energy and Environment Foundation, Al-Najaf, P.O.Box 39 (Iraq)

    2013-07-01

    A fuel cell is most interesting new power source because it solves not only the environment problem but also natural resource exhaustion problem. CFD modeling and simulation for heat and mass transport in PEM fuel cells are being used extensively in researches and industrial applications to gain better understanding of the fundamental processes and to optimize fuel cell designs before building a prototype for engineering application. In this research, full three-dimensional, non-isothermal computational fluid dynamics model of a tubular-shaped proton exchange membrane (PEM) fuel cell has been developed. This comprehensive model accounts for the major transport phenomena such as convective and diffusive heat and mass transfer, electrode kinetics, transport and phase-change mechanism of water, and potential fields in a tubular-shaped PEM fuel cell. The model explains many interacting, complex electrochemical, and transport phenomena that cannot be studied experimentally. Three-dimensional results of the species profiles, temperature distribution, potential distribution, and local current density distribution are presented and analysed, with the focus on the physical insight and fundamental understanding.

  4. AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

  5. Numerical study of the particle transport in fast neutron detectors with conversion layer

    International Nuclear Information System (INIS)

    Sedlackova, K.; Zatko, B.; Necas, V.

    2012-01-01

    This paper deals with fast neutron and recoil proton transport simulation using statistical analysis of Monte Carlo radiation transport code (MCNPX). Its possibilities in the detector design and optimization are presented. MCNPX proved as a very advantageous self-contained simulation program for fast neutron and secondary proton tracking. Simulations of respective particle transport through conversion layer of HDPE and further in the active volume of detector let us to follow important characteristics as neutron/proton flux density, reaction rate of elastic scattering on hydrogen nuclei and deposited energy as well as their dependencies on incident neutron energy and conversion layer/active region thickness. The efficiency of neutrons to protons conversion has been calculated and its maximum was reached for 500 μm thick conversion layer. The minimum active region thickness has been estimated to be about 300 μm.(authors)

  6. Monte Carlo study of the mechanisms of transport of fast neutrons in various media

    International Nuclear Information System (INIS)

    Ku, L.

    1976-01-01

    The technique of analyzing Monte Carlo histories was used to study the details of neutron transport and slowing down mechanisms. The statistical properties of life histories of ''exceptional'' neutrons, i.e., those staying closer to the source or penetrating to larger distances from the source, were compared to those of the general population. The macroscopic behavior of ''exceptional'' neutrons was also related to the interaction mechanics and to the microscopic properties of the medium

  7. Transport phenomena of macro and micro flows behind orifice and flow accelerated corrosion

    International Nuclear Information System (INIS)

    Fujisawa, Nobuyuki; Hayase, Toshiyuki; Ohara, Taku; Ikohagi, Toshiaki

    2008-01-01

    This paper describes experiment and numerical simulations for macro and micro flows behind an orifice model in a square pipe, which are carried from the viewpoint of flow accelerated corrosion (FAC). The measurements of velocity field behind the orifice model were carried out using particle image velocimetry, and the variations of velocity field with respect to the accuracy of the orifice position were studied. It is found that the reattachment behavior of the flow is highly influenced by the orifice position, which is a critical problem for predicting the pipe thinning phenomena by FAC. The DNS simulation was also conducted for calculating the macro flow behind the orifice. The result suggests that the DNS simulation is applicable to the prediction of pipe thinning macro flow for highly aged nuclear plant. The micro flow simulation can predict the pipe thinning phenomena near the wall. (author)

  8. Transport calculation of neutron flux distribution in reflector of PW reactor

    International Nuclear Information System (INIS)

    Remec, I.

    1982-01-01

    Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

  9. Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1, 1974--August 31, 1975

    International Nuclear Information System (INIS)

    Bareiss, E.H.

    1975-01-01

    The objectives of the research remain the same as outlined in the original proposal. They are in short as follows: Develop mathematically and computationally founded criteria for the design of highly efficient and reliable multi-dimensional neutron transport codes to solve a variety of neutron migration and radiation problems and analyze existing and new methods for performance. (U.S.)

  10. FMCEIR: a Monte Carlo program for solving the stationary neutron and gamma transport equation

    International Nuclear Information System (INIS)

    Taormina, A.

    1978-05-01

    FMCEIR is a three-dimensional Monte Carlo program for solving the stationary neutron and gamma transport equation. It is used to study the problem of neutron and gamma streaming in the GCFR and HHT reactor channels. (G.T.H.)

  11. On the reciprocity-like relations in linear neutron transport theory

    International Nuclear Information System (INIS)

    Modak, R.S.; Sahni, D.C.

    1997-01-01

    The existence of certain reciprocity-like relations in neutron transport theory was shown earlier under some quite restrictive conditions. Here, these relations are shown to be valid in more general situations by using a different approach based on individual neutron trajectories. (author)

  12. Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1976--November 30, 1977

    International Nuclear Information System (INIS)

    Bareiss, E.H.

    1977-08-01

    The objectives of this research are to develop mathematically and computationally founded criteria for the design of highly efficient and reliable multidimensional neutron transport codes to solve a variety of neutron migration and radiation problems, and to analyze existing and new methods for performance

  13. Cooperative learning of neutron diffusion and transport theories

    International Nuclear Information System (INIS)

    Robinson, Michael A.

    1999-01-01

    A cooperative group instructional strategy is being used to teach a unit on neutron transport and diffusion theory in a first-year-graduate level, Reactor Theory course that was formerly presented in the traditional lecture/discussion style. Students are divided into groups of two or three for the duration of the unit. Class meetings are divided into traditional lecture/discussion segments punctuated by cooperative group exercises. The group exercises were designed to require the students to elaborate, summarize, or practice the material presented in the lecture/discussion segments. Both positive interdependence and individual accountability are fostered by adjusting individual grades on the unit exam by a factor dependent upon group achievement. Group collaboration was also encouraged on homework assignments by assigning each group a single grade on each assignment. The results of the unit exam have been above average in the two classes in which the cooperative group method was employed. In particular, the problem solving ability of the students has shown particular improvement. Further,the students felt that the cooperative group format was both more educationally effective and more enjoyable than the lecture/discussion format

  14. Neutron and photon transport calculations in fusion system. 2

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)

  15. Methodology of Continuous-Energy Adjoint Monte Carlo for Neutron, Photon, and Coupled Neutron-Photon Transport

    International Nuclear Information System (INIS)

    Hoogenboom, J. Eduard

    2003-01-01

    Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems

  16. Application of Trotter approximation for solving time dependent neutron transport equation

    International Nuclear Information System (INIS)

    Stancic, V.

    1987-01-01

    A method is proposed to solve multigroup time dependent neutron transport equation with arbitrary scattering anisotropy. The recurrence relation thus obtained is simple, numerically stable and especially suitable for treatment of complicated geometries. (author)

  17. TEMPS, 1-Group Time-Dependent Pulsed Source Neutron Transport

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    1988-01-01

    1 - Description of program or function: TEMPS numerically determines the scalar flux as given by the one-group neutron transport equation with a pulsed source in an infinite medium. Standard plane, point, and line sources are considered as well as a volume source in the negative half-space in plane geometry. The angular distribution of emitted neutrons can either be isotropic or mono-directional (beam) in plane geometry and isotropic in spherical and cylindrical geometry. A general anisotropic scattering Kernel represented in terms of Legendre polynomials can be accommodated with a time- dependent number of secondaries given by c(t)=c 0 (t/t 0 ) β , where β is greater than -1 and less than infinity. TEMPS is designed to provide the flux to a high degree of accuracy (4-5 digits) for use as a benchmark to which results from other numerical solutions or approximations can be compared. 2 - Method of solution: A semi-analytic Method of solution is followed. The main feature of this approach is that no discretization of the transport or scattering operators is employed. The numerical solution involves the evaluation of an analytical representation of the solution by standard numerical techniques. The transport equation is first reformulated in terms of multiple collisions with the flux represented by an infinite series of collisional components. Each component is then represented by an orthogonal Legendre series expansion in the variable x/t where the distance x and time t are measured in terms of mean free path and mean free time, respectively. The moments in the Legendre reconstruction are found from an algebraic recursion relation obtained from Legendre expansion in the direction variable mu. The multiple collision series is evaluated first to a prescribed relative error determined by the number of digits desired in the scalar flux. If the Legendre series fails to converge in the plane or point source case, an accelerative transformation, based on removing the

  18. PHISICS multi-group transport neutronic capabilities for RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)

    2012-07-01

    PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)

  19. Simulation of neutron transport equation using parallel Monte Carlo for deep penetration problems

    International Nuclear Information System (INIS)

    Bekar, K. K.; Tombakoglu, M.; Soekmen, C. N.

    2001-01-01

    Neutron transport equation is simulated using parallel Monte Carlo method for deep penetration neutron transport problem. Monte Carlo simulation is parallelized by using three different techniques; direct parallelization, domain decomposition and domain decomposition with load balancing, which are used with PVM (Parallel Virtual Machine) software on LAN (Local Area Network). The results of parallel simulation are given for various model problems. The performances of the parallelization techniques are compared with each other. Moreover, the effects of variance reduction techniques on parallelization are discussed

  20. Calculation of neutron and gamma transport at the FOA:type of problems and calculation methods

    International Nuclear Information System (INIS)

    Lefvert, T.

    1975-11-01

    Protection against the effects of nuclear warfare involves the analysis of the forms of results of a nuclear charge explosion producing neutron and gamma radiation. It brings out problems leading to the calculation of criticality, leakage, and deep transmission. Methods have been developed for various kinds of particle transport problems. Applications to radiation therapy, storage of fissile materials, and fast reactors are discussed. A list (with brief description) of all neutron and gamma transport programmes of the FOA is given. (J.S.)

  1. Application of direct discrete method (DDM) to multigroup neutron transport problems

    International Nuclear Information System (INIS)

    Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid

    2003-01-01

    The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)

  2. Fast transient transport phenomena measured by soft X-ray emission in TCV tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Furno, I. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-08-01

    Energy and particle transport during sawtooth activity in TCV plasmas has been studied in this thesis with high temporal resolution many chord diagnostics. We indicated the influence of sawteeth on plasma profiles in ohmic conditions and in the presence of auxiliary electron cyclotron resonance heating and current drive. A 2-dimensional model for heat transport, including localised heat source and a magnetic island, has been used to interpret the experimental observations. These results provided a new interpretation of a coupled heat and transport phenomenon which is potentially important for plasma confinement. The observations validate the applicability and show the possibility of improvement of a 2-dimensional theoretic a1 model for the study of heat transport in the presence of localised heat source and a magnetic island. Furthermore, the TCV results showed a new possibility for the interpretation of a coupled heat and particle transport phenomenon previously understood only in stellarators. (author)

  3. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    International Nuclear Information System (INIS)

    Iga, Kiminori; Takada, Hiroshi; Nagao, Tadashi.

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B 4 C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  4. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  5. Engineering a spin-fet: spin-orbit phenomena and spin transport induced by a gate electric field

    OpenAIRE

    Cardoso, J. L.; Hernández-Saldaña, H.

    2012-01-01

    In this work, we show that a gate electric field, applied in the base of the field-effect devices, leads to inducing spin-orbit interactions (Rashba and linear Dresselhauss) and confines the transport electrons in a two-dimensional electron gas. On the basis of these phenomena we solve analytically the Pauli equation when the Rashba strength and the linear Dresselhaus one are equal, for a tuning value of the gate electric field $\\mathcal{E}_g^*$. Using the transfer matrix approach, we provide...

  6. Concatenating algorithms for parallel numerical simulations coupling radiation hydrodynamics with neutron transport

    International Nuclear Information System (INIS)

    Mo Zeyao

    2004-11-01

    Multiphysics parallel numerical simulations are usually essential to simplify researches on complex physical phenomena in which several physics are tightly coupled. It is very important on how to concatenate those coupled physics for fully scalable parallel simulation. Meanwhile, three objectives should be balanced, the first is efficient data transfer among simulations, the second and the third are efficient parallel executions and simultaneously developments of those simulation codes. Two concatenating algorithms for multiphysics parallel numerical simulations coupling radiation hydrodynamics with neutron transport on unstructured grid are presented. The first algorithm, Fully Loosely Concatenation (FLC), focuses on the independence of code development and the independence running with optimal performance of code. The second algorithm. Two Level Tightly Concatenation (TLTC), focuses on the optimal tradeoffs among above three objectives. Theoretical analyses for communicational complexity and parallel numerical experiments on hundreds of processors on two parallel machines have showed that these two algorithms are efficient and can be generalized to other multiphysics parallel numerical simulations. In especial, algorithm TLTC is linearly scalable and has achieved the optimal parallel performance. (authors)

  7. Continuous energy adjoint Monte Carlo for coupled neutron-photon transport

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.

    2001-07-01

    Although the theory for adjoint Monte Carlo calculations with continuous energy treatment for neutrons as well as for photons is known, coupled neutron-photon transport problems present fundamental difficulties because of the discrete energies of the photons produced by neutron reactions. This problem was solved by forcing the energy of the adjoint photon to the required discrete value by an adjoint Compton scattering reaction or an adjoint pair production reaction. A mathematical derivation shows the exact procedures to follow for the generation of an adjoint neutron and its statistical weight. A numerical example demonstrates that correct detector responses are obtained compared to a standard forward Monte Carlo calculation. (orig.)

  8. Angular neutron transport investigation in the HZETRN free-space ion and nucleon transport and shielding computer program

    International Nuclear Information System (INIS)

    Singleterry, R.C. Jr.; Wilson, J.W.

    1997-01-01

    Extension of the high charge and energy (HZE) transport computer program HZETRN for angular transport of neutrons is considered. For this paper, only light ion transport, He 4 and lighter, will be analyzed using a pure solar proton source. The angular transport calculator is the ANISN/PC program which is being controlled by the HZETRN program. The neutron flux values are compared for straight-ahead transport and angular transport in one dimension. The shield material is aluminum and the target material is water. The thickness of these materials is varied; however, only the largest model calculated is reported which is 50 gm/cm 2 of aluminum and 100 gm/cm 2 of water. The flux from the ANISN/PC calculation is about two orders of magnitude lower than the flux from HZETRN for very low energy neutrons. It is only a magnitude lower for the neutrons in the 10 to 20 MeV range in the aluminum and two orders lower in the water. The major reason for this difference is in the transport modes: straight-ahead versus angular. The angular treatment allows a longer path length than the straight-ahead approximation. Another reason is the different cross section sets used by the ANISN/PC-BUGLE-80 mode and the HZETRN mode. The next step is to investigate further the differences between the two codes and isolate the differences to just the angular versus straight-ahead transport mode. Then, create a better coupling between the angular neutron transport and the charged particle transport

  9. The Application of Neutron Transport Green's Functions to Threat Scenario Simulation

    Science.gov (United States)

    Thoreson, Gregory G.; Schneider, Erich A.; Armstrong, Hirotatsu; van der Hoeven, Christopher A.

    2015-02-01

    Radiation detectors provide deterrence and defense against nuclear smuggling attempts by scanning vehicles, ships, and pedestrians for radioactive material. Understanding detector performance is crucial to developing novel technologies, architectures, and alarm algorithms. Detection can be modeled through radiation transport simulations; however, modeling a spanning set of threat scenarios over the full transport phase-space is computationally challenging. Previous research has demonstrated Green's functions can simulate photon detector signals by decomposing the scenario space into independently simulated submodels. This paper presents decomposition methods for neutron and time-dependent transport. As a result, neutron detector signals produced from full forward transport simulations can be efficiently reconstructed by sequential application of submodel response functions.

  10. Neutron transport simulation in high speed moving media using Geant4

    Science.gov (United States)

    Li, G.; Ciungu, B.; Harrisson, G.; Rogge, R. B.; Tun, Z.; van der Ende, B. M.; Zwiers, I.

    2017-12-01

    A method using Geant4 to simulate neutron transport in moving media is described. The method is implanted in the source code of the software since Geant4 does not intrinsically support a moving object. The simulation utilizes the existing physical model and data library in Geant4, combined with frame transformations to account for the effect of relative velocity between neutrons and the moving media. An example is presented involving a high speed rotating cylinder to verify this method and show the effect of moving media on neutron transport.

  11. Modeling conversion and transport phenomena in solid-state fermentation: a review and perspectives

    NARCIS (Netherlands)

    Rahardjo, Y.S.P.; Tramper, J.; Rinzema, A.

    2006-01-01

    Solid-state fermentation (SSF) is accompanied inevitably by development of concentration and temperature gradients within the substrate particles and microbial biofilms. These gradients are needed for driving the transport of substrates and products. In addition, concentration gradients have been

  12. Experimental and analytical study of interfacial area transport phenomena in a vertical two-phase flow

    International Nuclear Information System (INIS)

    Huh, Byung-Gil; Euh, Dong-Jin; Yun, Byong-Jo; Youn, Young-Jung; Yoon, Han-Yeong; Song, Chul-Hwa

    2005-03-01

    The number density transport equations for various bubble groups are used to predict the void fraction and the interfacial area concentration. As the closure relations for number density transport equation, the coalescence due to random collisions and the breakup due to the impact of turbulent eddies is modified based on the previous studies and the bubble expansion term due to the pressure reduction is considered. Also, the coalescence due to a wake entrainment is modeled newly to apply to the number density transport equation. In order to predict the local experimental data, the code is developed that the two-fluid model is coupled systematically with the number density transport equation for each bubble group. As for the results of the numerical analysis, the void fraction and interfacial area concentration are predicted well by the developed code and models although some deviations exist in the values between the prediction and experiment, especially, for the high void fraction conditions

  13. Mathematical interpretation of Brownian motor model: Limit cycles and directed transport phenomena

    Science.gov (United States)

    Yang, Jianqiang; Ma, Hong; Zhong, Suchuang

    2018-03-01

    In this article, we first suggest that the attractor of Brownian motor model is one of the reasons for the directed transport phenomenon of Brownian particle. We take the classical Smoluchowski-Feynman (SF) ratchet model as an example to investigate the relationship between limit cycles and directed transport phenomenon of the Brownian particle. We study the existence and variation rule of limit cycles of SF ratchet model at changing parameters through mathematical methods. The influences of these parameters on the directed transport phenomenon of a Brownian particle are then analyzed through numerical simulations. Reasonable mathematical explanations for the directed transport phenomenon of Brownian particle in SF ratchet model are also formulated on the basis of the existence and variation rule of the limit cycles and numerical simulations. These mathematical explanations provide a theoretical basis for applying these theories in physics, biology, chemistry, and engineering.

  14. Experimental and analytical study of interfacial area transport phenomena in a vertical two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Byung-Gil; Euh, Dong-Jin; Yun, Byong-Jo; Youn, Young-Jung; Yoon, Han-Yeong; Song, Chul-Hwa

    2005-03-01

    The number density transport equations for various bubble groups are used to predict the void fraction and the interfacial area concentration. As the closure relations for number density transport equation, the coalescence due to random collisions and the breakup due to the impact of turbulent eddies is modified based on the previous studies and the bubble expansion term due to the pressure reduction is considered. Also, the coalescence due to a wake entrainment is modeled newly to apply to the number density transport equation. In order to predict the local experimental data, the code is developed that the two-fluid model is coupled systematically with the number density transport equation for each bubble group. As for the results of the numerical analysis, the void fraction and interfacial area concentration are predicted well by the developed code and models although some deviations exist in the values between the prediction and experiment, especially, for the high void fraction conditions.

  15. Modeling of the Transport Phenomena in Passive Direct Methanol Fuel Cells Using a Two-Phase Anisotropic Model

    Directory of Open Access Journals (Sweden)

    Zheng Miao

    2014-04-01

    Full Text Available The transport phenomena in a passive direct methanol fuel cell (DMFC were numerically simulated by the proposed two-dimensional two-phase nonisothermal mass transport model. The anisotropic transport characteristic and deformation of the gas diffusion layer (GDL were considered in this model. The natural convection boundary conditions were adopted for the transport of methanol, oxygen, and heat at the GDL outer surface. The effect of methanol concentration in the reservoir on cell performance was examined. The distribution of multiphysical fields in the membrane electrode assembly (MEA, especially in the catalyst layers (CLs, was obtained and analyzed. The results indicated that transport resistance for the methanol mainly existed in the MEA while that for oxygen and heat was primarily due to natural convection at the GDL outer surface. Because of the relatively high methanol concentration, the local reaction rate in CLs was mainly determined by the overpotential. Methanol concentration between 3 M and 4 M was recommended for passive liquid feed DMFC in order to achieve a balance between the cell performance and the methanol crossover.

  16. Parallel computing solution of Boltzmann neutron transport equation

    International Nuclear Information System (INIS)

    Ansah-Narh, T.

    2010-01-01

    The focus of the research was on developing parallel computing algorithm for solving Eigen-values of the Boltzmam Neutron Transport Equation (BNTE) in a slab geometry using multi-grid approach. In response to the problem of slow execution of serial computing when solving large problems, such as BNTE, the study was focused on the design of parallel computing systems which was an evolution of serial computing that used multiple processing elements simultaneously to solve complex physical and mathematical problems. Finite element method (FEM) was used for the spatial discretization scheme, while angular discretization was accomplished by expanding the angular dependence in terms of Legendre polynomials. The eigenvalues representing the multiplication factors in the BNTE were determined by the power method. MATLAB Compiler Version 4.1 (R2009a) was used to compile the MATLAB codes of BNTE. The implemented parallel algorithms were enabled with matlabpool, a Parallel Computing Toolbox function. The option UseParallel was set to 'always' and the default value of the option was 'never'. When those conditions held, the solvers computed estimated gradients in parallel. The parallel computing system was used to handle all the bottlenecks in the matrix generated from the finite element scheme and each domain of the power method generated. The parallel algorithm was implemented on a Symmetric Multi Processor (SMP) cluster machine, which had Intel 32 bit quad-core x 86 processors. Convergence rates and timings for the algorithm on the SMP cluster machine were obtained. Numerical experiments indicated the designed parallel algorithm could reach perfect speedup and had good stability and scalability. (au)

  17. Computational Modeling of a Time-Independent, Heterogeneous Reactor Core Using Simplified Discrete Ordinates Neutron Transport Techniques

    National Research Council Canada - National Science Library

    Labowski, Kristofer

    2001-01-01

    The Linear Characteristic (LC) method on rectangular boxoid meshes is a discrete ordinate neutron transport technique that uses both zeroth and first moments of the angular neutron flux to construct a relatively accurate...

  18. Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

    International Nuclear Information System (INIS)

    Brenner, D.J.; Prael, R.E.; Little, R.C.

    1987-01-01

    Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs

  19. Influence of anomalous transport phenomena on onset of Neoclassical Tearing Modes in tokamaks

    International Nuclear Information System (INIS)

    Konovalov, S.V.; Mikhailovskii, A.B.; Shirokov, M.S.; Ozeki, T.; Takizuka, T.; Hayashi, N.

    2005-01-01

    Influence of anomalous perpendicular heat transport and anomalous ion perpendicular viscosity on conditions of Neoclassical Tearing Mode (NTM) onset is studied theoretically. Series of various parallel transport mechanisms competitive to anomalous cross-island heat transport in formation of the perturbed electron and ion temperature profiles within the island are considered. Analytical solutions to respective heat balance equations were found and perturbed temperature profiles were calculated rigorously. The partial contributions from the plasma electron and ion temperature perturbations in the bootstrap drive of the mode and magnetic curvature effect were then accounted in construction of a generalized transport threshold model of NTMs. Taking into account the curvature effect weakening in the generalized transport threshold model predicts notable improvement of NTM stability. The anomalous perpendicular ion viscosity was shown to modify collisionality dependence of polarization current effect reducing it to the low collisionality limit. The bootstrap drive of NTM in the presence of anomalous perpendicular ion viscosity was found to be dependent on the island rotation frequency and direction. For island rotating in direction of the electron diamagnetic drift viscosity effect was shown to be stabilizing. The role of viscosity effect grows rapidly with rise of the plasma ion temperature. (author)

  20. Efficient modeling of reactive transport phenomena by a multispecies random walk coupled to chemical equilibrium

    International Nuclear Information System (INIS)

    Pfingsten, W.

    1996-01-01

    Safety assessments for radioactive waste repositories require a detailed knowledge of physical, chemical, hydrological, and geological processes for long time spans. In the past, individual models for hydraulics, transport, or geochemical processes were developed more or less separately to great sophistication for the individual processes. Such processes are especially important in the near field of a waste repository. Attempts have been made to couple at least two individual processes to get a more adequate description of geochemical systems. These models are called coupled codes; they couple predominantly a multicomponent transport model with a chemical reaction model. Here reactive transport is modeled by the sequentially coupled code MCOTAC that couples one-dimensional advective, dispersive, and diffusive transport with chemical equilibrium complexation and precipitation/dissolution reactions in a porous medium. Transport, described by a random walk of multispecies particles, and chemical equilibrium calculations are solved separately, coupled only by an exchange term. The modular-structured code was applied to incongruent dissolution of hydrated silicate gels, to movement of multiple solid front systems, and to an artificial, numerically difficult heterogeneous redox problem. These applications show promising features with respect to applicability to relevant problems and possibilities of extensions

  1. Thermodynamics, transport phenomena, and electrochemistry of external field-assisted nonthermal food technologies.

    Science.gov (United States)

    Misra, N N; Martynenko, Alex; Chemat, Farid; Paniwnyk, Larysa; Barba, Francisco J; Jambrak, Anet Režek

    2018-07-24

    Interest in the development and adoption of nonthermal technologies is burgeoning within the food and bioprocess industry, the associated research community, and among the consumers. This is evident from not only the success of some innovative nonthermal technologies at industrial scale, but also from the increasing number of publications dealing with these topics, a growing demand for foods processed by nonthermal technologies and use of natural ingredients. A notable feature of the nonthermal technologies such as cold plasma, electrohydrodynamic processing, pulsed electric fields, and ultrasound is the involvement of external fields, either electric or sound. Therefore, it merits to study the fundamentals of these technologies and the associated phenomenon with a unified approach. In this review, we revisit the fundamental physical and chemical phenomena governing the selected technologies, highlight similarities, and contrasts, describe few successful applications, and finally, identify the gaps in research.

  2. Fission product transport in the primary system, important phenomena, and code status

    International Nuclear Information System (INIS)

    Gieseke, J.A.; Jordan, H.; Kuhlman, M.R.

    1990-01-01

    The purpose of this paper is to identify important issues concerning the transport and deposition of radionuclides in the reactor coolant system (RCS) under accident conditions and to examine how such issues are being treated or should be treated by the various available computer codes. In general, the RCS is a very important section of the transport pathway along which radionuclides move and by which they are attenuated as they travel after being released from the fuel. The RCS can serve as a sink for radionuclides that may deposit from the gas and react with surfaces, or can serve as a repository for materials deposited from the gas which are then available for later release into the transporting gas stream. The RCS may also have thermal hydraulic conditions that foster aerosol growth by condensation or agglomeration, and may provide an environment in which gas phase or heterogeneous chemical reactions may occur

  3. Various Transport Phenomena and Modeling in a Methane Reformer Duct for PEMFCs

    International Nuclear Information System (INIS)

    Jinliang Yuan; Fuan Ren; Jinliang Yuan; Bengt Sunden

    2006-01-01

    There are various physical processes (such as mass, heat and momentum transport) integrated with catalytic chemical reactions in a methane steam reforming duct. It is often found that endothermic and exothermic reactions in the ducts are strongly coupled by heat transfer from adjacent catalytic combustion ducts. In this paper, a three-dimensional calculation method is developed to simulate and analyze steam reforming of methane, and the effects on various transport processes in a steam reforming duct. The reformer conditions such as mass balances associated with the reforming reactions and gas permeation to/from the porous catalyst layer are applied in the analysis. The predicted results are presented and discussed for a composite duct consisting of a porous catalyst reaction area, the gas flow duct and solid layers. Parametric studies are conducted and the results show that the variables, such as fuel reformer temperatures and catalyst loadings, have significant effects on the transport processes and reformer performance. (authors)

  4. Transport phenomena in the edge of Alcator C-Mod plasmas

    International Nuclear Information System (INIS)

    Terry, J.L.; Basse, N.P.; Cziegler, I.; Greenwald, M.; LaBombard, B.; Edlund, E.M.; Hughes, J.W.; Lin, L.; Lin, Y.; Porkolab, M.; Veto, B.; Wukitch, S.J.; Grulke, O.; Zweben, S.J.; Sampsell, M.

    2005-01-01

    Two aspects of edge turbulence and transport in Alcator C-Mod are explored. The quasi-coherent mode, an edge fluctuation present in Enhanced Da H-mode plasmas, is examined with regard to its role in the enhanced particle transport found in these plasmas, its in/out asymmetry, its poloidal wave number, and its radial width and location. It is shown to play a dominant role in the perpendicular particle transport. The QCM is not observed at the inboard midplane, indicating that its amplitude there is significantly smaller than on the outboard side. The peak amplitude of the QCM is found just inside the separatrix, with a radial width ≥5 mm, leading to a non-zero amplitude outside the separatrix and qualitatively consistent with its transport enhancement. Also examined are the characteristics of the intermittent convective transport, associated with 'blobs' and typically occurring in the scrape-off-layer. The blobs are qualitatively similar in L- and H-mode. When their sizes, occurrence frequencies, and magnitudes are compared, it is found that the blob size may be somewhat smaller in ELMfree H-Mode, and blob frequency is similar. A clear difference is seen in the blob magnitude in the far SOL, with ELMfree H-mode showing a smaller perturbation there than L-mode. As the Greenwald density limit is approached (n/n GW ≥0.7), blobs are seen inside the separatrix, consistent with the observation that the high cross-field transport region, normally found in the far scrape-off, penetrates the closed flux surfaces at high n/n GW . (author)

  5. Laser field effects on the transport phenomena: Energy loss and stopping power

    International Nuclear Information System (INIS)

    Torres Silva, H.; Sakanaka, P.H.

    1990-01-01

    The energy loss method has been applied to a large variety of transport problems in optics, solid-state and fusion research. In these papers, however, the transport equations were linearized, so there are no multiphoton interaction. On the other hand, Bivona et al. (1982) [2] have shown that, for a one-component plasma, the strong field effects would be only of academic interest. On the basis of the center of mass approach [3], a generalization of the energy loss rate which is in accordance with the recent results of Arista et al. (1989) [4] is obtained. (Author)

  6. Studies of transport phenomena in tokamaks with nonstationary intervention into the discharge

    International Nuclear Information System (INIS)

    Kalmykov, S.G.

    1993-01-01

    Together with detailed plasma parameter measurements, an experimental basis is provided to deduce radial profiles of local transport coefficients, to obtain their temporal evolution in the transient phase of the discharge. The equations of heat and particle balance were used as proper instrument to perform the coefficients calculation. The majority of the experiments deals with heat transport processes in the electron component of plasma. A problem in getting ohmic heat deposit radial distribution arise with use of the electron population heat balance equation. For its solution, numerical simulation of the plasma column loop voltage based on poloidal magnetic field classical diffusion supposition was used. (L.C.J.A.)

  7. Report by the AERES on the unit: Unit of researches on neutron transport and radioactivity confinement in nuclear installations under the supervision of the establishments and bodies: IRSN

    International Nuclear Information System (INIS)

    2010-07-01

    This report is a kind of audit report on a research laboratory whose activity is organized in three departments: neutron transport and criticality (themes: numerical methods, maths and statistics related to the simulation of neutral particle propagation, nuclear data, uncertainty propagation and bias estimation, code qualification and associated experimental programs, neutron transport in reactors and fuel cycle, criticality accidents), radionuclide transfer in radioactive waste disposals (site identification strategy, hydro-mechanical phenomena affecting storage performance, physical-chemical evolution factors, storage modelling), and metrology and confinement of radioactive gases and aerosols. The authors discuss an assessment of the unit activities in terms of strengths and opportunities, aspects to be improved and recommendations, productions and publications. A more detailed assessment is presented for each department in terms of scientific quality, influence and attractiveness (awards, recruitment capacity, capacity to obtain financing and to tender, participation to international programs), strategy and governance, and project

  8. Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

    International Nuclear Information System (INIS)

    Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

    2012-01-01

    10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

  9. Special Features of the Air to Space Neutron Transport Problem

    Science.gov (United States)

    2017-09-14

    an atmosphere model. Radioactive Decay Free neutrons are not stable elementary particles. They decay radioactively with a half- life of around ten...milliseconds to seconds, so that radioactive decay of neutrons is negligible. (The probability of decay in 100 milliseconds with a 10 minute half- life is...the bottom and top of a layer are 1bZ - and bZ respectively. The methods developed here apply to any planet with an atmosphere and an orbiting

  10. Resolution of the neutron transport equation by massively parallel computer in the Cronos code

    International Nuclear Information System (INIS)

    Zardini, D.M.

    1996-01-01

    The feasibility of neutron transport problems parallel resolution by CRONOS code's SN module is here studied. In this report we give the first data about the parallel resolution by angular variable decomposition of the transport equation. Problems about parallel resolution by spatial variable decomposition and memory stage limits are also explained here. (author)

  11. The infinite medium Green's function for neutron transport in plane geometry 40 years later

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    1993-01-01

    In 1953, the first of what was supposed to be two volumes on neutron transport theory was published. The monograph, entitled open-quotes Introduction to the Theory of Neutron Diffusionclose quotes by Case et al., appeared as a Los Alamos National Laboratory report and was to be followed by a second volume, which never appeared as intended because of the death of Placzek. Instead, Case and Zweifel collaborated on the now classic work entitled Linear Transport Theory 2 in which the underlying mathematical theory of linear transport was presented. The initial monograph, however, represented the coming of age of neutron transport theory, which had its roots in radiative transfer and kinetic theory. In addition, it provided the first benchmark results along with the mathematical development for several fundamental neutron transport problems. In particular, one-dimensional infinite medium Green's functions for the monoenergetic transport equation in plane and spherical geometries were considered complete with numerical results to be used as standards to guide code development for applications. Unfortunately, because of the limited computational resources of the day, some numerical results were incorrect. Also, only conventional mathematics and numerical methods were used because the transport theorists of the day were just becoming acquainted with more modern mathematical approaches. In this paper, Green's function solution is revisited in light of modern numerical benchmarking methods with an emphasis on evaluation rather than theoretical results. The primary motivation for considering the Green's function at this time is its emerging use in solving finite and heterogeneous media transport problems

  12. TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: TDTORT solves the time-dependent, three-dimensional neutron transport equation with explicit representation of delayed neutrons to estimate the fission yield from fissionable material transients. This release includes a modified version of TORT from the C00650MFMWS01 DOORS3.1 code package plus the time-dependent TDTORT code. GIP is also included for cross-section preparation. TORT calculates the flux or fluence of particles due to particles incident upon the system from extraneous sources or generated internally as a result of interaction with the system in two- or three-dimensional geometric systems. The principle application is to the deep-penetration transport of neutrons and photons. Reactor eigenvalue problems can also be solved. Numerous printed edits of the results are available, and results can be transferred to output files for subsequent analysis. TDTORT reads ANISN-format cross-section libraries, which are not included in the package. Users may choose from several available in RSICC's data library collection which can be identified by the keyword 'ANISN FORMAT'. 2 - Methods:The time-dependent spatial flux is expressed as a product of a space-, energy-, and angle-dependent shape function, which is usually slowly varying in time and a purely time-dependent amplitude function. The shape equation is solved for the shape using TORT; and the result is used to calculate the point kinetics parameters (e.g., reactivity) by using their inner product definitions, which are then used to solve the time-dependent amplitude and precursor equations. The amplitude function is calculated by solving the kinetics equations using the LSODE solver. When a new shape calculation is needed, the flux is calculated using the newly computed amplitude function. The Boltzmann transport equation is solved using the method of discrete ordinates to treat the directional variable and weighted finite-difference methods, in addition to Linear Nodal

  13. Application of neutron/gamma transport codes for the design of explosive detection systems

    International Nuclear Information System (INIS)

    Elias, E.; Shayer, Z.

    1994-01-01

    Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs

  14. Comparison of analytical transport and stochastic solutions for neutron slowing down in an infinite medium

    International Nuclear Information System (INIS)

    Jahshan, S.N.; Wemple, C.A.; Ganapol, B.D.

    1993-01-01

    A comparison of the numerical solutions of the transport equation describing the steady neutron slowing down in an infinite medium with constant cross sections is made with stochastic solutions obtained from tracking successive neutron histories in the same medium. The transport equation solution is obtained using a numerical Laplace transform inversion algorithm. The basis for the algorithm is an evaluation of the Bromwich integral without analytical continuation. Neither the transport nor the stochastic solution is limited in the number of scattering species allowed. The medium may contain an absorption component as well. (orig.)

  15. Computational complexity in multidimensional neutron transport theory calculations. Progress report, September 1, 1975--August 31, 1976

    International Nuclear Information System (INIS)

    Bareiss, E.H.

    1976-05-01

    The objectives of the work are to develop mathematically and computationally founded for the design of highly efficient and reliable multidimensional neutron transport codes to solve a variety of neutron migration and radiation problems, and to analyze existing and new methods for performance. As new analytical insights are gained, new numerical methods are developed and tested. Significant results obtained include implementation of the integer-preserving Gaussian elimination method (two-step method) in a CDC 6400 computer code, modes analysis for one-dimensional transport solutions, and a new method for solving the 1-T transport equation. Some of the work dealt with the interface and corner problem in diffusion theory

  16. Constitutive laws for the neutron density current

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Morales-Sandoval, Jaime B.; Vazquez-Rodriguez, Rodolfo; Espinosa-Martinez, Erick-G.

    2008-01-01

    In this technical note, a fractional wave equation for the average neutron motion in nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fickian and non-Fickian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional exponent that can be manipulated to obtain the best representation of the neutron transport phenomena. The detrended fluctuation analysis (DFA) method is presented in this paper to estimate the fractional exponent

  17. Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen

    2013-01-01

    Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)

  18. Design studies for a high-resolution, transportable neutron radiography/radioscopy system

    International Nuclear Information System (INIS)

    Gillespie, G.H.; Micklich, B.J.; McMichael, G.E.

    1996-01-01

    A preliminary design has been developed for a high-resolution, transportable neutron radiology system (TNRS) concept. The primary system requirement is taken to be a thermal neutron flux of 10[sup 6] n/(cm[sup 2]-sec) with a L/D ratio of 100. The approach is to use an accelerator-driven neutron source, with a radiofrequency quadrupole (RFQ) as the primary accelerator component. Initial concepts for all of the major components of the system have been developed,and selected key parts have been examined further. An overview of the system design is presented, together with brief summaries of the concepts for the ion source, low energy beam transport (LEBT), RFQ, high energy beam transport (HEBT), target, moderator, collimator, image collection, power, cooling, vacuum, structure, robotics, control system, data analysis, transport vehicle, and site support. The use of trade studies for optimizing the TNRS concept are also described

  19. Coupled transport phenomena in a clay from a Callovo-Oxfordian formation

    International Nuclear Information System (INIS)

    Paszkuta, M.

    2005-06-01

    Low permeability materials containing clay play an important role in practical life and natural environment. Indeed, the ability of clay soils to act as semi permeable membranes, that inhibit the passage of electrolytes, is of great interest. The major objective of this thesis is to evaluate the transport properties of natural clays and in particular coupled transports when a pressure gradient, an electrical field, a concentration gradient and a temperature gradient interact. The material is a compact argillite extracted in East France from a Callovo-Oxfordian formation which was supplied to us by ANDRA. NaCl was used as the main solute. Two series of experiments were performed to measure permeability, diffusion, conductivity, the electro-osmotic coefficient and the Soret coefficient. (author)

  20. New transport phenomena probed by dielectric spectroscopy of oxidized and non-oxidized porous silicon

    Energy Technology Data Exchange (ETDEWEB)

    Urbach, B.; Axelrod, E.; Sa' ar, A. [Racah Institute of Physics and the Center for Nanoscience and Nanotechnology, the Hebrew University of Jerusalem, Jerusalem 91904 (Israel)

    2007-05-15

    Dielectric spectroscopy accompanied by infrared (IR) and photoluminescence (PL) spectroscopy have been utilized to reveal the correlation between transport, optical and structural properties of oxidized porous silicon (PS). Three relaxation processes at low-, mid- and high-temperatures were observed, including dc-conductivity at high-temperatures. Both the low-T relaxation and the dc conductivity were found to be thermally activated processes that involve tunneling and hopping in between the nanocrystals in oxidized PS. We have found that the dc-conductivity is limited by geometrical constrictions along the transport channels, which are not effected by the oxidation process and are characterized by activation energies of about {proportional_to}0.85 eV. The low-T relaxation process involves thermal activation followed by tunneling in between neighbor nanocrystals, with somewhat lower activation energies. (copyright 2007 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  1. Analysis of transport phenomena and electrochemical reactions in a micro PEM fuel cell

    OpenAIRE

    Maher A.R. Sadiq Al-Baghdadi

    2014-01-01

    Micro-fuel cells are considered as promising electrochemical power sources in portable electronic devices. The presence of microelectromechanical system (MEMS) technology makes it possible to manufacture the miniaturized fuel cell systems. The majority of research on micro-scale fuel cells is aimed at micro-power applications. Performance of micro-fuel cells are closely related to many factors, such as designs and operating conditions. CFD modeling and simulation for heat and mass transport i...

  2. COMSOL-PHREEQC: a tool for high performance numerical simulation of reactive transport phenomena

    International Nuclear Information System (INIS)

    Nardi, Albert; Vries, Luis Manuel de; Trinchero, Paolo; Idiart, Andres; Molinero, Jorge

    2012-01-01

    Document available in extended abstract form only. Comsol Multiphysics (COMSOL, from now on) is a powerful Finite Element software environment for the modelling and simulation of a large number of physics-based systems. The user can apply variables, expressions or numbers directly to solid and fluid domains, boundaries, edges and points, independently of the computational mesh. COMSOL then internally compiles a set of equations representing the entire model. The availability of extremely powerful pre and post processors makes COMSOL a numerical platform well known and extensively used in many branches of sciences and engineering. On the other hand, PHREEQC is a freely available computer program for simulating chemical reactions and transport processes in aqueous systems. It is perhaps the most widely used geochemical code in the scientific community and is openly distributed. The program is based on equilibrium chemistry of aqueous solutions interacting with minerals, gases, solid solutions, exchangers, and sorption surfaces, but also includes the capability to model kinetic reactions with rate equations that are user-specified in a very flexible way by means of Basic statements directly written in the input file. Here we present COMSOL-PHREEQC, a software interface able to communicate and couple these two powerful simulators by means of a Java interface. The methodology is based on Sequential Non Iterative Approach (SNIA), where PHREEQC is compiled as a dynamic subroutine (iPhreeqc) that is called by the interface to solve the geochemical system at every element of the finite element mesh of COMSOL. The numerical tool has been extensively verified by comparison with computed results of 1D, 2D and 3D benchmark examples solved with other reactive transport simulators. COMSOL-PHREEQC is parallelized so that CPU time can be highly optimized in multi-core processors or clusters. Then, fully 3D detailed reactive transport problems can be readily simulated by means of

  3. Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

    Energy Technology Data Exchange (ETDEWEB)

    Modak, R S; Kumar, Vinod; Menon, S V.G. [Theoretical Physics Div., Bhabha Atomic Research Centre, Mumbai (India); Gupta, Anurag [Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai (India)

    2005-09-15

    The numerical solution of linear multi-energy-group neutron transport equation is required in several analyses in nuclear reactor physics and allied areas. Computer codes based on the discrete ordinates (Sn) method are commonly used for this purpose. These codes solve external source problem and K-eigenvalue problem. The overall solution technique involves solution of source problem in each energy group as intermediate procedures. Such a single-group source problem is solved by the so-called Source Iteration (SI) method. As is well-known, the SI-method converges very slowly for optically thick and highly scattering regions, leading to large CPU times. Over last three decades, many schemes have been tried to accelerate the SI; the most prominent being the Diffusion Synthetic Acceleration (DSA) scheme. The DSA scheme, however, often fails and is also rather difficult to implement. In view of this, in 1997, Ramone and others have developed a new acceleration scheme called Transport Synthetic Acceleration (TSA) which is much more robust and easy to implement. This scheme has been recently incorporated in 2-D and 3-D in-house codes at BARC. This report presents studies on the utility of TSA scheme for fairly general test problems involving many energy groups and anisotropic scattering. The scheme is found to be useful for problems in Cartesian as well as Cylindrical geometry. (author)

  4. Discrete-ordinates electron transport calculations using standard neutron transport codes

    International Nuclear Information System (INIS)

    Morel, J.E.

    1979-01-01

    The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

  5. Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

    International Nuclear Information System (INIS)

    Modak, R.S.; Vinod Kumar; Menon, S.V.G.; Gupta, Anurag

    2005-09-01

    The numerical solution of linear multi-energy-group neutron transport equation is required in several analyses in nuclear reactor physics and allied areas. Computer codes based on the discrete ordinates (Sn) method are commonly used for this purpose. These codes solve external source problem and K-eigenvalue problem. The overall solution technique involves solution of source problem in each energy group as intermediate procedures. Such a single-group source problem is solved by the so-called Source Iteration (SI) method. As is well-known, the SI-method converges very slowly for optically thick and highly scattering regions, leading to large CPU times. Over last three decades, many schemes have been tried to accelerate the SI; the most prominent being the Diffusion Synthetic Acceleration (DSA) scheme. The DSA scheme, however, often fails and is also rather difficult to implement. In view of this, in 1997, Ramone and others have developed a new acceleration scheme called Transport Synthetic Acceleration (TSA) which is much more robust and easy to implement. This scheme has been recently incorporated in 2-D and 3-D in-house codes at BARC. This report presents studies on the utility of TSA scheme for fairly general test problems involving many energy groups and anisotropic scattering. The scheme is found to be useful for problems in Cartesian as well as Cylindrical geometry. (author)

  6. Transient transport phenomena induced by cold pulses in W7-AS

    International Nuclear Information System (INIS)

    Walter, H.; Stroth, U.; Bleuel, J.; Burhenn, R.; Geist, T.; Giannone, L.; Hartfuss, H.; Koponen, J.P.T.; Ledl, L.; Pereverzev, G.

    1998-01-01

    Cold-pulse experiments were carried out in the W7-AS stellarator for the first time. Carbon was injected by a laser blow-off system into the plasma edge. The electron density increase due to the injected carbon is found to be responsible for the edge electron temperature drop. In all cases, the propagation of the edge temperature perturbation to the plasma centre could be modelled with a local parameter-dependent electron heat diffusivity. Unlike in tokamaks, non-local transport effects were not observed in these experiments. (author)

  7. Transient transport phenomena induced by cold pulses in W7-AS

    Energy Technology Data Exchange (ETDEWEB)

    Walter, H; Stroth, U; Bleuel, J; Burhenn, R; Geist, T; Giannone, L.; Hartfuss, H; Koponen, J P.T.; Ledl, L; Pereverzev, G [Max-Planck-Institut fuer Plasmaphysik, EURATOM-IPP Association, 85748 Garching (Germany)

    1998-09-01

    Cold-pulse experiments were carried out in the W7-AS stellarator for the first time. Carbon was injected by a laser blow-off system into the plasma edge. The electron density increase due to the injected carbon is found to be responsible for the edge electron temperature drop. In all cases, the propagation of the edge temperature perturbation to the plasma centre could be modelled with a local parameter-dependent electron heat diffusivity. Unlike in tokamaks, non-local transport effects were not observed in these experiments. (author)

  8. About relaxation phenomena and transport processing in a fully ionized non-ideal plasma

    International Nuclear Information System (INIS)

    Baimbetov, F.B.; Iztleuov, N.T.

    1999-01-01

    In this report correlation effects for non-ideal plasma are accounted in the so called pseudopotentials. The accounting of high order correlation influences in the pseudopotential lead to the strongly screened potential. Kinetic equation with pseudopotential is cited. The equations which describe the relaxation of the difference of directed velocities of plasma particles, and frequency of particle collision which determines relaxation of temperature for non-ideal plasma are obtained. On basis of mentioned kinetic equation the transport equation is obtained as well

  9. Novel phenomena in one-dimensional non-linear transport in long quantum wires

    International Nuclear Information System (INIS)

    Morimoto, T; Hemmi, M; Naito, R; Tsubaki, K; Park, J-S; Aoki, N; Bird, J P; Ochiai, Y

    2006-01-01

    We have investigated the non-linear transport properties of split-gate quantum wires of various channel lengths. In this report, we present results on a resonant enhancement of the non-linear conductance that is observed near pinch-off under a finite source-drain bias voltage. The resonant phenomenon exhibits a strong dependence on temperature and in-plane magnetic field. We discuss the possible relationship of this phenomenon to the spin-polarized manybody state that has recently been suggested to occur in quasi-one dimensional systems

  10. Theoretical Interpretation of Pulsed Neutron Phenomena; Interpretation Theorique des Phenomenes dus aux Neutrons Pulses; Teoreticheskaya interpretatsiya yavlenij, svyazannykh s impul'snymi nejtronami; Interpretacion Teorica de los Fenomenos Debidos a Neutrones Pulsados

    Energy Technology Data Exchange (ETDEWEB)

    Corngold, N. [Brookhaven National Laboratory, Upton, NY (United States)

    1965-08-15

    The author discusses the present understanding of transient effects in the variation of neutron distributions in space, time and energy. In the analysis, particular attention is given to the pulsed neutron experiment in non-multiplying media, and to its interpretation in terms of space-energy modes and (time) decay constants. Theoretical research in this field, much of which has come from the Brookhaven National Laboratory, has been built upon the exact solution of several special problems, and approximate, numerical solutions in more realistic situations. Recently, a unified point of view, based on the Van Hove-Glauber theory of the scattering of slow neutrons, has been proposed. It has been found that the long-time behaviour of the pair correlation function, which enables one to distinguish between different types of moderating materials, also controls the type of decay-constant spectrum that one finds. In his discussion the author emphasizes the case of infinite medium solutions that are proportional to exp(iB * r), in media having 1/v absorption. The analytical work is concerned with modal solutions, proportional to exp({lambda}t) as well, while the numerical work is directed at the full, ''life-history'' solution, as well as the modal solutions. The author also remarks on phenomena encountered when the absorption is not 1/v and when the moderator sample is so small that the long-time solutions are no longer separable in space and energy. Finally, the paper discusses the bearing of these recent developments upon the interpretation of pulsed experiments. Particular reference is made to the measurement of diffusion and diffusion cooling coefficients, and to the nature of the approach to the asymptotic distribution. (author) [French] L'auteur examine la conception actuelle des effets transitoires dans la variation des distributions des neutrons dans l'espace, dans le temps et selon l'energie. Dans son etude, il accorde une attention particuliere aux experiences

  11. Interfacing MCNPX and McStas for simulation of neutron transport

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

    2013-01-01

    Stas[4, 5, 6, 7]. The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as Mc...... geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides....

  12. How to polarise all neutrons in one beam: a high performance polariser and neutron transport system

    Science.gov (United States)

    Rodriguez, D. Martin; Bentley, P. M.; Pappas, C.

    2016-09-01

    Polarised neutron beams are used in disciplines as diverse as magnetism,soft matter or biology. However, most of these applications often suffer from low flux also because the existing neutron polarising methods imply the filtering of one of the spin states, with a transmission of 50% at maximum. With the purpose of using all neutrons that are usually discarded, we propose a system that splits them according to their polarisation, flips them to match the spin direction, and then focuses them at the sample. Monte Carlo (MC) simulations show that this is achievable over a wide wavelength range and with an outstanding performance at the price of a more divergent neutron beam at the sample position.

  13. Fast-ion transport during repetitive burst phenomena of toroidal Alfven eigenmodes in the Large Helical Device

    International Nuclear Information System (INIS)

    Nishiura, M.; Isobe, M.; Yamamoto, S.

    2008-10-01

    Alfven instabilities induced fast-ion losses have been directly observed for the first time by a newly developed scintillator lost ion probe (SLIP) in the Large Helical Device (LHD). The SLIP can measure the pitch angle and gyro radius of escaped fast ions toward loss region. Neutral beam driven Alfven Eigenmodes (AEs) are excited under the reactor relevant conditions: the ratio of fast ion (beam) speed υ b and Alfven speed υ A is more than 0.3 - 4.0. The beta value for fast ions is considered roughly to be ∼10%. Non-linear phenomena related to Alfven instabilities are observed under such conditions. During repetitive Toroidal Alfven Eigenmode (TAE) bursts, synchronized fast ion losses are observed by SLIP. From the orbit calculation the measured fast ion with pitch angle of 130 degrees and beam energy of 150 keV surely pass through the locations of TAE gaps. The orbit analysis found that the observed fast ions interact strongly with the excited TAEs. This result becomes the first experimental evidence of radial transport of fast ions predicted theoretically during TAE activities. In addition, from the correlation between stored energy degradation and fast-ion loss rate, it is found that fast-ion losses induced by TAE activities with low toroidal mode numbers categorize two phenomena without and with fast- ion loss enhancements, which indicate the fast-ion redistribution and loss. (author)

  14. An Equation-Type Approach for the Numerical Solution of the Partial Differential Equations Governing Transport Phenomena in Porous Media

    KAUST Repository

    Sun, Shuyu; Salama, Amgad; El-Amin, Mohamed

    2012-01-01

    A new technique for the numerical solution of the partial differential equations governing transport phenomena in porous media is introduced. In this technique, the governing equations as depicted from the physics of the problem are used without extra manipulations. In other words, there is no need to reduce the number of governing equations by some sort of mathematical manipulations. This technique enables the separation of the physics part of the problem and the solver part, which makes coding more robust and could be used in several other applications with little or no modifications (e.g., multi-phase flow in porous media). In this method, one abandons the need to construct the coefficient matrix for the pressure equation. Alternatively, the coefficients are automatically generated within the solver routine. We show examples of using this technique to solving several flow problems in porous media.

  15. An Equation-Type Approach for the Numerical Solution of the Partial Differential Equations Governing Transport Phenomena in Porous Media

    KAUST Repository

    Sun, Shuyu

    2012-06-02

    A new technique for the numerical solution of the partial differential equations governing transport phenomena in porous media is introduced. In this technique, the governing equations as depicted from the physics of the problem are used without extra manipulations. In other words, there is no need to reduce the number of governing equations by some sort of mathematical manipulations. This technique enables the separation of the physics part of the problem and the solver part, which makes coding more robust and could be used in several other applications with little or no modifications (e.g., multi-phase flow in porous media). In this method, one abandons the need to construct the coefficient matrix for the pressure equation. Alternatively, the coefficients are automatically generated within the solver routine. We show examples of using this technique to solving several flow problems in porous media.

  16. Modelling the transport phenomena and texture changes of chicken breast meat during the roasting in a convective oven

    DEFF Research Database (Denmark)

    Rabeler, Felix; Feyissa, Aberham Hailu

    2018-01-01

    A numerical 3D model of coupled transport phenomena and texture changes during the roasting of chicken breast meat in a convection oven was developed. The model is based on heat and mass transfer coupled with the kinetics of temperature induced texture changes of chicken breast meat. The partial...... experimentally values. The developed model enables the prediction of the texture development inside the chicken meat as function of the process parameters. The model predictions and measured values show the clear effect of changing process settings on the texture profiles during the roasting process. Overall......, the developed model provides deep insights into the local and spatial texture changes of chicken breast meat during the roasting process that cannot be gained by experimentation alone....

  17. Neutron transport calculation for Activation Evaluation for Decommissioning of PET cyclotron Facility

    Science.gov (United States)

    Nobuhara, Fumiyoshi; Kuroyanagi, Makoto; Masumoto, Kazuyoshi; Nakamura, Hajime; Toyoda, Akihiro; Takahashi, Katsuhiko

    2017-09-01

    In order to evaluate the state of activation in a cyclotron facility used for the radioisotope production of PET diagnostics, we measured the neutron flux by using gold foils and TLDs. Then, the spatial distribution of neutrons and induced activity inside the cyclotron vault were simulated with the Monte Calro calculation code for neutron transport and DCHAIN-SP for activation calculation. The calculated results are in good agreement with measured values within factor 3. Therefore, the adaption of the advanced evaluation procedure for activation level is proved to be important for the planning of decommissioning of these facilities.

  18. A domian Decomposition Method for Transient Neutron Transport with Pomrning-Eddington Approximation

    International Nuclear Information System (INIS)

    Hendi, A.A.; Abulwafa, E.E.

    2008-01-01

    The time-dependent neutron transport problem is approximated using the Pomraning-Eddington approximation. This approximation is two-flux approximation that expands the angular intensity in terms of the energy density and the net flux. This approximation converts the integro-differential Boltzmann equation into two first order differential equations. The A domian decomposition method that used to solve the linear or nonlinear differential equations is used to solve the resultant two differential equations to find the neutron energy density and net flux, which can be used to calculate the neutron angular intensity through the Pomraning-Eddington approximation

  19. Transient analysis of carbon monoxide transport phenomena and adsorption kinetics in HT-PEMFC during dynamic current extraction

    International Nuclear Information System (INIS)

    Raj, Kamal Abdul Rasheedj; Chan, Siew Hwa

    2015-01-01

    Highlights: • Increasing the fuel cell temperature reduces outlet CO concentration. • Increasing the CO inlet (initial) concentration increases outlet CO concentration. • Increasing current density step and dwell time increases outlet CO concentration. • Increasing in the CL and GDL porosities reduces outlet CO concentration. - Abstract: This paper investigates the transport phenomena of carbon monoxide (CO) and adsorption kinetics, in a high-temperature proton exchange membrane fuel cell (HT-PEMFC) during step-wise current extraction (step-change in current extraction). Step-wise current extraction is a common process done to accommodate a sudden power surge during an operation. Since HT-PEMFCs are capable of handling high impurity of CO, hydrogen fuel that is contaminated with trace amount of CO is usually considered for commercial benefits. Thus, a transient three-dimensional isothermal anodic electro-kinetic numerical model is developed to determine the effect of operating parameters such as fuel cell temperature, CO inlet (initial) concentration, step-change of current density and dwell time on the transport phenomena of CO and adsorption kinetics. In addition, geometrical factors such as the catalyst layer (CL) and gas diffusion layer (GDL) porosity are also varied as well. The results show that the above-mentioned operating parameters can affect the maximum CO concentration at the CL, especially at the outlet of the channel. Specifically, a reduction of fuel cell temperature can significantly increase the CO concentration near the outlet, while increasing CO inlet (initial) concentration, step-change amplitude of current density and current density dwell time can cause an increase in CO concentration at the outlet, albeit to different extent. In addition, the increase in the porosity of CL and GDL, results in the reduction of the maximum CO concentration at the outlet, albeit to different extent. In addition, the CO and hydrogen surface coverage

  20. Investigation of high-p{sub T} phenomena within a partonic transport model

    Energy Technology Data Exchange (ETDEWEB)

    Fochler, Oliver

    2011-10-26

    In the work presented herein the microscopic transport model BAMPS (Boltzmann Approach to Multi-Parton Scatterings) is applied to simulate the time evolution of the hot partonic medium that is created in Au+Au collisions at the Relativistic Heavy Ion Collider (RHIC) and in Pb+Pb collisions at the recently started Large Hadron Collider (LHC). The study is especially focused on the investigation of the nuclear modification factor R{sub AA}, that quantifies the suppression of particle yields at large transverse momentum with respect to a scaled proton+proton reference, and the simultaneous description of the collective properties of the medium in terms of the elliptic flow v{sub 2} within a common framework. (orig.)

  1. Investigation of high-pT phenomena within a partonic transport model

    International Nuclear Information System (INIS)

    Fochler, Oliver

    2011-01-01

    In the work presented herein the microscopic transport model BAMPS (Boltzmann Approach to Multi-Parton Scatterings) is applied to simulate the time evolution of the hot partonic medium that is created in Au+Au collisions at the Relativistic Heavy Ion Collider (RHIC) and in Pb+Pb collisions at the recently started Large Hadron Collider (LHC). The study is especially focused on the investigation of the nuclear modification factor R AA , that quantifies the suppression of particle yields at large transverse momentum with respect to a scaled proton+proton reference, and the simultaneous description of the collective properties of the medium in terms of the elliptic flow v 2 within a common framework. (orig.)

  2. Study of a transportable neutron radiography system; Estudo de um sistema neutrongrafico transportavel

    Energy Technology Data Exchange (ETDEWEB)

    Souza, S N.A. de

    1991-05-01

    This work presents a study a transportable neutron radiography system for a 185 GBq {sup 241} Am-Be ({alpha}, {eta}) source with a neutron yield roughly 1,25 x 10{sup 7} n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio of 14, for neutron fluxes up to 4,09 x 10{sup 2} n.cm{sup -2}.s{sup -1}. Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10{sup 2} n.cm{sup -2}.s{sup -1} and 2,65 x 10{sup 2} n.cm{sup -2}.s{sup -1}, respectively. (author). 30 refs, 39 figs, 9 tabs.

  3. Imaging transport phenomena during lysozyme protein crystal growth by the hanging drop technique

    Science.gov (United States)

    Sethia Gupta, Anamika; Gupta, Rajive; Panigrahi, P. K.; Muralidhar, K.

    2013-06-01

    The present study reports the transport process that occurs during the growth of lysozyme protein crystals by the hanging drop technique. A rainbow schlieren technique has been employed for imaging changes in salt concentration. A one dimensional color filter is used to record the deflection of the light beam. An optical microscope and an X-ray crystallography unit are used to characterize the size, tetragonal shape and Bravais lattice constants of the grown crystals. A parametric study on the effect of drop composition, drop size, reservoir height and number of drops on the crystal size and quality is reported. Changes in refractive index are not large enough to create a meaningful schlieren image in the air gap between the drop and the reservoir. However, condensation of fresh water over the reservoir solution creates large changes in the concentration of NaCl, giving rise to clear color patterns in the schlieren images. These have been analyzed to obtain salt concentration profiles near the free surface of the reservoir solution as a function of time. The diffusion of fresh water into the reservoir solution at the early stages of crystal growth followed by the mass flux of salt from the bulk solution towards the free surface has been recorded. The overall crystal growth process can be classified into two regimes, as demarcated by the changes in slope of salt concentration within the reservoir. The salt concentration in the reservoir equilibrates at long times when the crystallization process is complete. Thus, transport processes in the reservoir emerge as the route to monitor protein crystal growth in the hanging drop configuration. Results show that crystal growth rate is faster for a higher lysozyme concentration, smaller drops, and larger reservoir heights.

  4. An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

    International Nuclear Information System (INIS)

    Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

    2014-01-01

    Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

  5. Solution and study of nodal neutron transport equation applying the LTSN-DiagExp method

    International Nuclear Information System (INIS)

    Hauser, Eliete Biasotto; Pazos, Ruben Panta; Vilhena, Marco Tullio de; Barros, Ricardo Carvalho de

    2003-01-01

    In this paper we report advances about the three-dimensional nodal discrete-ordinates approximations of neutron transport equation for Cartesian geometry. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtained the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. (author)

  6. Least-squares finite element discretizations of neutron transport equations in 3 dimensions

    Energy Technology Data Exchange (ETDEWEB)

    Manteuffel, T.A [Univ. of Colorado, Boulder, CO (United States); Ressel, K.J. [Interdisciplinary Project Center for Supercomputing, Zurich (Switzerland); Starkes, G. [Universtaet Karlsruhe (Germany)

    1996-12-31

    The least-squares finite element framework to the neutron transport equation introduced in is based on the minimization of a least-squares functional applied to the properly scaled neutron transport equation. Here we report on some practical aspects of this approach for neutron transport calculations in three space dimensions. The systems of partial differential equations resulting from a P{sub 1} and P{sub 2} approximation of the angular dependence are derived. In the diffusive limit, the system is essentially a Poisson equation for zeroth moment and has a divergence structure for the set of moments of order 1. One of the key features of the least-squares approach is that it produces a posteriori error bounds. We report on the numerical results obtained for the minimum of the least-squares functional augmented by an additional boundary term using trilinear finite elements on a uniform tesselation into cubes.

  7. Spallation neutron production and the current intra-nuclear cascade and transport codes

    International Nuclear Information System (INIS)

    Filges, D.; Goldenbaum, F.

    2001-01-01

    A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models. (orig.)

  8. Spallation neutron production and the current intra-nuclear cascade and transport codes

    Science.gov (United States)

    Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.

    A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.

  9. Analytical benchmarks for nuclear engineering applications. Case studies in neutron transport theory

    International Nuclear Information System (INIS)

    2008-01-01

    The developers of computer codes involving neutron transport theory for nuclear engineering applications seldom apply analytical benchmarking strategies to ensure the quality of their programs. A major reason for this is the lack of analytical benchmarks and their documentation in the literature. The few such benchmarks that do exist are difficult to locate, as they are scattered throughout the neutron transport and radiative transfer literature. The motivation for this benchmark compendium, therefore, is to gather several analytical benchmarks appropriate for nuclear engineering applications under one cover. We consider the following three subject areas: neutron slowing down and thermalization without spatial dependence, one-dimensional neutron transport in infinite and finite media, and multidimensional neutron transport in a half-space and an infinite medium. Each benchmark is briefly described, followed by a detailed derivation of the analytical solution representation. Finally, a demonstration of the evaluation of the solution representation includes qualified numerical benchmark results. All accompanying computer codes are suitable for the PC computational environment and can serve as educational tools for courses in nuclear engineering. While this benchmark compilation does not contain all possible benchmarks, by any means, it does include some of the most prominent ones and should serve as a valuable reference. (author)

  10. Silicon based nanogap device for studying electrical transport phenomena in molecule-nanoparticle hybrids

    International Nuclear Information System (INIS)

    Strobel, Sebastian; Hernandez, Rocio Murcia; Hansen, Allan G; Tornow, Marc

    2008-01-01

    We report the fabrication and characterization of vertical nanogap electrode devices using silicon-on-insulator substrates. Using only standard silicon microelectronic process technology, nanogaps down to 26 nm electrode separation were prepared. Transmission electron microscopy cross-sectional analysis revealed the well defined material architecture of the nanogap, comprising two electrodes of dissimilar geometrical shape. This asymmetry is directly reflected in transport measurements on molecule-nanoparticle hybrid systems formed by self-assembling a monolayer of mercaptohexanol on the electrode surface and the subsequent dielectrophoretic trapping of 30 nm diameter Au nanoparticles. The observed Coulomb staircase I-V characteristic measured at T = 4.2 K is in excellent agreement with theoretical modelling, whereby junction capacitances of the order of a few 10 -18 farad and asymmetric resistances of 30 and 300 MΩ, respectively, are also supported well by our independent estimates for the formed double barrier tunnelling system. We propose our nanoelectrode system for integrating novel functional electronic devices such as molecular junctions or nanoparticle hybrids into existing silicon microelectronic process technology

  11. Investigations into co-transport phenomena in PUREX relevant solutions by use of an analytical ultracentrifuge

    International Nuclear Information System (INIS)

    Gauglitz, R.; Marx, G.

    1991-01-01

    The diffusion of the elements uranium, neodymium, chromium, iron, ruthenium was studied in aqueous and aqueous nitric acid solutions. The diffusion of these elements was measured not only under a concentration gradient but also with respect to simultanious nitric and diffusion. Measurements with schlierenoptics and with uv/vis.-optic were carried out at the same time. Besides the diffusion of metal nitrates and potassium dichromate also nitric acid diffusion was investigated. Also in this case apparent diffusion coefficients were determined under concentration gradients and under the influence of superposing element gradients. The diffusion of the elements uranium, neodymium, ruthenium, neptunium and plutonium was also determined in organic systems. At first the transport of the elements was measured under an element gradient, in presence of various amounts of water and nitric acid. These experiments were followed by those on simultanious diffusion of water of nitric acid and elements in TBP/dodecane mixture. Furthermore TBP/dodecane solutions were oxidized with potassium dichromate. This oxidation was measured using a gaschromatograph. From the obtained results the formation rate for the oxidation products could be calculated which proved that higher nitric acid concentrations increased the rate. (orig.) With 16 refs., 100 tabs., 120 figs [de

  12. Transport phenomena in the cathode of a molten carbonate fuel cell

    International Nuclear Information System (INIS)

    Berg, P.; Findlay, J.

    2009-01-01

    'Full text': A Molten Carbonate Fuel Cell (MCFC) is an electro-chemical energy conversion technology that runs on natural gas and employs a molten salt electrolyte. In order to keep the electrolyte in this state, the cell must be kept at a temperature above 500 C, eliminating the need for noble catalysts. There has been only a limited amount of research on modelling the transport processes inside this device, mainly due to its limited ability for mobile applications. A model for the reaction-diffusion processes within the cathode of a MCFC is developed using Fick's Law for diffusion and incorporating Darcy's Law for convection. A model for Binary Diffusion is also discussed and compared to those for Fickian diffusion. It can be shown that there exists a limiting case for diffusion across the cathode that depends on the conductivity for the liquid potential, for which there exists an analytical solution. Results are also discussed for varying diffusivities and permeabilities. Ultimately, this research focuses on the optimization of the electrode porosity to increase the power output of the fuel cell. The porosity is considered as a function of position, and is optimized using the software package MATLAB. (author)

  13. Phenomena of the ionic transport in the stress corrosion of metals

    International Nuclear Information System (INIS)

    Gravano, S.M.

    1986-07-01

    For the study of electrochemical conditions of propagation, a model which calculates the concentrations and potential profiles inside cracks or localized corrosion cavities, was developed. Considering transport by difussion and migration it was applied to pure metals (Zn, Fe) in solutions where pitting occurs (NaCl or Na2SO4, with borate buffer), and also extended to systems where stress corrosion cracking is present, such as Cu and yellow brass in NaNO2. Physical bases of the 'constant intermediate elongation rate technique' to predict stress corrosion cracking susceptibility was analized, studying by mathematical models: 1) dissolution current, that should be the result of superposition of repassivation transients on the fresh metal, exposed to corrosive medium by strain, with the same rate of that of a static specimen; 2) ohmic drop, that in some systems could be quite important and it must be considered in the overpotential evaluation; and 3) metallic ion concentration that, instead of what happens in a crack, never attains saturation in the analized cases. For repassivation transient according to the crak propagation models proposed by Scully and Ford it was found that, at the tip of the crack, it is unlikely that the same repassivation transients occur as in the constant intermediate elongation rate experiments. (M.E.L.)

  14. Forty years of 9Sr in situ migration: importance of soil characterization in modeling transport phenomena

    International Nuclear Information System (INIS)

    Fernandez, J.M.; Piault, E.; Macouillard, D.; Juncos, C.

    2006-01-01

    In 1960 experiments were carried out on the transfer of 9 Sr between soil, grapes and wine. The experiments were conducted in situ on a piece of land limited by two control strips. The 9 Sr migration over the last 40 years was studied by performing radiological and physico-chemical characterizations of the soil on eight 70 cm deep cores. The vertical migration modeling of 9 Sr required the definition of a triple layer conceptual model integrating the rainwater infiltration at constant flux as the only external factor of influence. Afterwards the importance of a detailed soil characterization for modeling was discussed and satisfactory simulation of the 9 Sr vertical transport was obtained and showed a calculated migration rate of about 1.0 cm year -1 in full agreement with the in situ measured values. The discussion was regarding some of the key parameters such as granulometry, organic matter content (in the Van Genuchten parameter determination), Kd and the efficient rainwater infiltration. Besides the experimental data, simplifying assumptions in modeling such as water-soil redistribution calculation and factual discontinuities in conceptual model were examined

  15. Charge Transport Phenomena in Detectors of the Cryogenic Dark Matter Search

    Science.gov (United States)

    Sundqvist, Kyle

    2008-03-01

    The Cryogenic Dark Matter Search (CDMS) seeks to detect putative weakly-interacting massive particles (WIMPS), which could explain the dark matter problem in cosmology and particle physics. By simultaneously measuring the number of charge carriers and the energy in athermal phonons created by particle interactions in intrinsic Ge and Si crystals at a temperature of 40 mK, a signature response for each event is produced. This response, combined with phonon pulse-shape information, allows CDMS to actively discriminate candidate WIMP interactions with nuclei apart from electromagnetic radioactive background which interacts with electrons. The challenges associated with these techniques are unique. Carrier drift-fields are maintained at only a few V/cm, else drift-emitted Luke-Neganov phonons would dominate the phonons of the original interaction. Under such conditions, carrier scattering is dominated by zero-point fluctuations of the lattice ions. It has been an open question how well the 8 Kelvin data prominent in the literature depicts this case. We compare the simulated transport properties of electrons and holes in Ge at 40 mK and at 8 K, and apply this understanding to our detectors.

  16. Silicon based nanogap device for studying electrical transport phenomena in molecule-nanoparticle hybrids

    Energy Technology Data Exchange (ETDEWEB)

    Strobel, Sebastian; Hernandez, Rocio Murcia [Walter Schottky Institut, Technische Universitaet Muenchen, Am Coulombwall 3, 85748 Garching (Germany); Hansen, Allan G; Tornow, Marc [Institut fuer Halbleitertechnik, Technische Universitaet Braunschweig, Hans-Sommer-Strasse 66, 38106 Braunschweig (Germany)], E-mail: m.tornow@tu-bs.de

    2008-09-17

    We report the fabrication and characterization of vertical nanogap electrode devices using silicon-on-insulator substrates. Using only standard silicon microelectronic process technology, nanogaps down to 26 nm electrode separation were prepared. Transmission electron microscopy cross-sectional analysis revealed the well defined material architecture of the nanogap, comprising two electrodes of dissimilar geometrical shape. This asymmetry is directly reflected in transport measurements on molecule-nanoparticle hybrid systems formed by self-assembling a monolayer of mercaptohexanol on the electrode surface and the subsequent dielectrophoretic trapping of 30 nm diameter Au nanoparticles. The observed Coulomb staircase I-V characteristic measured at T = 4.2 K is in excellent agreement with theoretical modelling, whereby junction capacitances of the order of a few 10{sup -18} farad and asymmetric resistances of 30 and 300 M{omega}, respectively, are also supported well by our independent estimates for the formed double barrier tunnelling system. We propose our nanoelectrode system for integrating novel functional electronic devices such as molecular junctions or nanoparticle hybrids into existing silicon microelectronic process technology.

  17. Silicon based nanogap device for studying electrical transport phenomena in molecule-nanoparticle hybrids.

    Science.gov (United States)

    Strobel, Sebastian; Hernández, Rocío Murcia; Hansen, Allan G; Tornow, Marc

    2008-09-17

    We report the fabrication and characterization of vertical nanogap electrode devices using silicon-on-insulator substrates. Using only standard silicon microelectronic process technology, nanogaps down to 26 nm electrode separation were prepared. Transmission electron microscopy cross-sectional analysis revealed the well defined material architecture of the nanogap, comprising two electrodes of dissimilar geometrical shape. This asymmetry is directly reflected in transport measurements on molecule-nanoparticle hybrid systems formed by self-assembling a monolayer of mercaptohexanol on the electrode surface and the subsequent dielectrophoretic trapping of 30 nm diameter Au nanoparticles. The observed Coulomb staircase I-V characteristic measured at T = 4.2 K is in excellent agreement with theoretical modelling, whereby junction capacitances of the order of a few 10(-18) farad and asymmetric resistances of 30 and 300 MΩ, respectively, are also supported well by our independent estimates for the formed double barrier tunnelling system. We propose our nanoelectrode system for integrating novel functional electronic devices such as molecular junctions or nanoparticle hybrids into existing silicon microelectronic process technology.

  18. The spectral element approach for the solution of neutron transport problems

    International Nuclear Information System (INIS)

    Barbarino, A.; Dulla, S.; Ravetto, P.; Mund, E.H.

    2011-01-01

    In this paper a possible application of the Spectral Element Method to neutron transport problems is presented. The basic features of the numerical scheme on the one-dimensional diffusion equation are illustrated. Then, the AN model for neutron transport is introduced, and the basic steps for the construction of a bi-dimensional solver are described. The AN equations are chosen for their structure, involving a system of coupled elliptic-type equations. Some calculations are carried out on typical benchmark problems and results are compared with the Finite Element Method, in order to evaluate their performances. (author)

  19. Multigroup neutron transport equation in the diffusion and P{sub 1} approximation

    Energy Technology Data Exchange (ETDEWEB)

    Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1970-07-01

    Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)

  20. Spectroscopic imaging studies of nanoscale polarity and mass transport phenomena in self-assembled organic nanotubes.

    Science.gov (United States)

    Xu, Hao; Nagasaka, Shinobu; Kameta, Naohiro; Masuda, Mitsutoshi; Ito, Takashi; Higgins, Daniel A

    2017-08-02

    Synthetic organic nanotubes self-assembled from bolaamphiphile surfactants are now being explored for use as drug delivery vehicles. In this work, several factors important to their implementation in drug delivery are explored. All experiments are performed with the nanotubes immersed in ethanol. First, Nile Red (NR) and a hydroxylated Nile Red derivative (NR-OH) are loaded into the nanotubes and spectroscopic fluorescence imaging methods are used to determine the apparent dielectric constant of their local environment. Both are found in relatively nonpolar environments, with the NR-OH molecules preferring regions of relatively higher dielectric constant compared to NR. Unique two-color imaging fluorescence correlation spectroscopy (imaging FCS) measurements are then used along with the spectroscopic imaging results to deduce the dielectric properties of the environments sensed by mobile and immobile populations of probe molecules. The results reveal that mobile NR molecules pass through less polar regions, likely within the nanotube walls, while immobile NR molecules are found in more polar regions, possibly near the nanotube surfaces. In contrast, mobile and immobile NR-OH molecules are found to locate in environments of similar polarity. The imaging FCS results also provide quantitative data on the apparent diffusion coefficient for each dye. The mean diffusion coefficient for the NR dye was approximately two-fold larger than that of NR-OH. Slower diffusion by the latter could result from its additional hydrogen bonding interactions with polar triglycine, amine, and glucose moieties near the nanotube surfaces. The knowledge gained in these studies will allow for the development of nanotubes that are better engineered for applications in the controlled transport and release of uncharged, dipolar drug molecules.

  1. Use of the light-water neutron scattering kernel in the study of neutron transport processes in mixtures of light and heavy water

    International Nuclear Information System (INIS)

    Tewari, S.P.

    1975-01-01

    A method of studying neutron transport properties in H 2 O-D 2 O mixtures, both liquid and solid, which extrapolates the neutron thermalization parameters of H 2 O is described. The decay of pulsed neutrons in the media has been investigated as an example of the application of the method. The results of the calcutions agree with the experiment for concentrations up to 50 percent D 2 O. (1 figure) (U.S.)

  2. New methods in linear transport theory. Part of a coordinated programme on methods in neutron transport theory

    International Nuclear Information System (INIS)

    Mika, J.

    1975-09-01

    Originally the work was oriented towards two main topics: a) difference and integral methods in neutron transport theory. Two computers were used for numerical calculations GIER and CYBER-72. During the first year the main effort was shifted towards basic theoretical investigations. At the first step the ANIS code was adopted and later modified to check various finite difference approaches against each other. Then the general finite element method and the singular perturbation method were developed. The analysis of singularities of the one-dimensional neutron transport equation in spherical geometry has been done and presented. Later the same analysis for the case of cylindrical symmetry has been carried out. The second and the third year programme included the following topics: 1) finite difference methods in stationary neutron transport theory; 2)mathematical fundamentals of approximate methods for solving the transport equation; 3) singular perturbation method for the time-dependent transport equation; 4) investigation of various iterative procedures in reactor calculations. This investigation will help to better understanding of the mathematical basis for existing and developed numerical methods resulting in more effective algorithms for reactor computer codes

  3. ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration

    International Nuclear Information System (INIS)

    Devillers, C.

    1973-01-01

    1 - Nature of physical problem solved: The ANISN system treats neutron and gamma transport in one-dimensional plane, spherical and cylinder geometry. The multigroup cross sections prepared by the programs LIANE and SUPERTOG are processed by the program RETTOG, which produces a binary library with Legendre expansions. The binary library can be updated and edited with the program LGR/B. The photon multigroup cross sections are created with the program GAMLEG/A. If the bulk of the data is too large, the program TAPEMA produces a special group-by-group library. The volume sources are calculated from a reduced set of input data and punched in a format suitable for input to ANISN, using the program PRESOU. The program ANISN calculates fluxes by groups, space intervals, angle and any number of reaction rates. The energy and space dependent fluxes are stored on tape and can be reprocessed, edited and plotted with the program ANISEX, which also permits to calculate supplementary reaction rates. The program ANISN can condense cross sections into a reduced number of groups. The ANISN system is used as a reference system for the evaluation of approximation methods (space-diffusion or point- kernel) or for the preparation of multigroup libraries for 2- dimensional transport codes (DOT). In particular it is used for shielding problems with high attenuation in water reactors and fast reactors. 2 - Method of solution: Method of discrete ordinates. The program has been designed to treat deep penetration with detailed calculation of spectrum as function of angle. Tests for pointwise convergence have also been introduced. 3 - Restrictions on the complexity of the problem: The complexity of the problem is limited by the storage size

  4. Effects of fuel particle size distributions on neutron transport in stochastic media

    International Nuclear Information System (INIS)

    Liang, Chao; Pavlou, Andrew T.; Ji, Wei

    2014-01-01

    Highlights: • Effects of fuel particle size distributions on neutron transport are evaluated. • Neutron channeling is identified as the fundamental reason for the effects. • The effects are noticeable at low packing and low optical thickness systems. • Unit cells of realistic reactor designs are studied for different size particles. • Fuel particle size distribution effects are not negligible in realistic designs. - Abstract: This paper presents a study of the fuel particle size distribution effects on neutron transport in three-dimensional stochastic media. Particle fuel is used in gas-cooled nuclear reactor designs and innovative light water reactor designs loaded with accident tolerant fuel. Due to the design requirements and fuel fabrication limits, the size of fuel particles may not be perfectly constant but instead follows a certain distribution. This brings a fundamental question to the radiation transport computation community: how does the fuel particle size distribution affect the neutron transport in particle fuel systems? To answer this question, size distribution effects and their physical interpretations are investigated by performing a series of neutron transport simulations at different fuel particle size distributions. An eigenvalue problem is simulated in a cylindrical container consisting of fissile fuel particles with five different size distributions: constant, uniform, power, exponential and Gaussian. A total of 15 parametric cases are constructed by altering the fissile particle volume packing fraction and its optical thickness, but keeping the mean chord length of the spherical fuel particle the same at different size distributions. The tallied effective multiplication factor (k eff ) and the spatial distribution of fission power density along axial and radial directions are compared between different size distributions. At low packing fraction and low optical thickness, the size distribution shows a noticeable effect on neutron

  5. Mathematical Modeling of Transport Phenomena in Polymer Electrolyte and Direct Methanol Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Birgersson, Erik

    2004-02-01

    This thesis deals with modeling of two types of fuel cells: the polymer electrolyte fuel cell (PEFC) and the direct methanol fuel cell (DMFC), for which we address four major issues: a) mass transport limitations; b) water management (PEFC); c) gas management (DMFC); d) thermal management. Four models have been derived and studied for the PEFC, focusing on the cathode. The first exploits the slenderness of the cathode for a two-dimensional geometry, leading to a reduced model, where several non dimensional parameters capture the behavior of the cathode. The model was extended to three dimensions, where four different flow distributors were studied for the cathode. A quantitative comparison shows that the interdigitated channels can sustain the highest current densities. These two models, comprising isothermal gas phase flow, limit the studies to (a). Returning to a two-dimensional geometry of the PEFC, the liquid phase was introduced via a separate flow model approach for the cathode. In addition to conservation of mass, momentum and species, the model was extended to consider simultaneous charge and heat transfer for the whole cell. Different thermal, flow fields, and hydrodynamic conditions were studied, addressing (a), (b) and (d). A scale analysis allowed for predictions of the cell performance prior to any computations. Good agreement between experiments with a segmented cell and the model was obtained. A liquid-phase model, comprising conservation of mass, momentum and species, was derived and analyzed for the anode of the DMFC. The impact of hydrodynamic, electrochemical and geometrical features on the fuel cell performance were studied, mainly focusing on (a). The slenderness of the anode allows the use of a narrow-gap approximation, leading to a reduced model, with benefits such as reduced computational cost and understanding of the physical trends prior to any numerical computations. Adding the gas-phase via a multiphase mixture approach, the gas

  6. Neutron Transport Methods for Accelerator-Driven Systems

    International Nuclear Information System (INIS)

    Nicholas Tsoulfanidis; Elmer Lewis

    2005-01-01

    The objective of this project has been to develop computational methods that will enable more effective analysis of Accelerator Driven Systems (ADS). The work is centered at the University of Missouri at Rolla, with a subcontract at Northwestern University, and close cooperation with the Nuclear Engineering Division at Argonne National Laboratory. The work has fallen into three categories. First, the treatment of the source for neutrons originating from the spallation target which drives the neutronics calculations of the ADS. Second, the generalization of the nodal variational method to treat the R-Z geometry configurations frequently needed for scoping calculations in Accelerator Driven Systems. Third, the treatment of void regions within variational nodal methods as needed to treat the accelerator beam tube

  7. A Monte Carlo Green's function method for three-dimensional neutron transport

    International Nuclear Information System (INIS)

    Gamino, R.G.; Brown, F.B.; Mendelson, M.R.

    1992-01-01

    This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution

  8. The study of neutron transport by oscillation method

    International Nuclear Information System (INIS)

    Raievski, V.

    1959-01-01

    The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [fr

  9. PELAN - a transportable, neutron-based UXO identification technique

    International Nuclear Information System (INIS)

    Vourvopoulos, G.

    1998-01-01

    An elemental characterization method is used to differentiate between inert projectiles and UXO's. This method identifies in a non-intrusive, nondestructive manner, the elemental composition of the projectile contents. Most major and minor chemical elements within the interrogated object (hydrogen, carbon, nitrogen, oxygen, fluorine, phosphorus, chlorine, arsenic, etc.) are identified and quantified. The method is based on PELAN - Pulsed Elemental Analysis with Neutrons. PELAN uses pulsed neutrons produced from a compact, sealed tube neutron generator. Using an automatic analysis computer program, the quantities of each major and minor chemical element are determined. A decision-making tree identifies the object by comparing its elemental composition with stored elemental composition libraries of substances that could be contained within the projectile. In a series of blind tests, PELAN was able to identify without failure, the contents of each shell placed in front of it. The PELAN probe does not need to be in contact with the interrogated projectile. If the object is buried, the interrogation can take place in situ provided the probe can be inserted a few centimeters from the object's surface. (author)

  10. Direct integration multiple collision integral transport analysis method for high energy fusion neutronics

    International Nuclear Information System (INIS)

    Koch, K.R.

    1985-01-01

    A new analysis method specially suited for the inherent difficulties of fusion neutronics was developed to provide detailed studies of the fusion neutron transport physics. These studies should provide a better understanding of the limitations and accuracies of typical fusion neutronics calculations. The new analysis method is based on the direct integration of the integral form of the neutron transport equation and employs a continuous energy formulation with the exact treatment of the energy angle kinematics of the scattering process. In addition, the overall solution is analyzed in terms of uncollided, once-collided, and multi-collided solution components based on a multiple collision treatment. Furthermore, the numerical evaluations of integrals use quadrature schemes that are based on the actual dependencies exhibited in the integrands. The new DITRAN computer code was developed on the Cyber 205 vector supercomputer to implement this direct integration multiple-collision fusion neutronics analysis. Three representative fusion reactor models were devised and the solutions to these problems were studied to provide suitable choices for the numerical quadrature orders as well as the discretized solution grid and to understand the limitations of the new analysis method. As further verification and as a first step in assessing the accuracy of existing fusion-neutronics calculations, solutions obtained using the new analysis method were compared to typical multigroup discrete ordinates calculations

  11. Arbitrary quadrature: its application in the solution of one-dimensional, planar neutron transport problems

    International Nuclear Information System (INIS)

    Sanchez, J.

    2010-10-01

    A standard numerical procedure for the solution of singular integral equations is applied to the one-dimensional transport equation for monoenergetic neutrons. As is usual in quadrature methods, the procedure yields an Eigen system whose solution provide, for the critical slab, both the eigenvalue which is proportional to the number of secondary neutrons per collision, and the density as a function of position. The results obtained with two versions of the procedure, differing only in the extent of the basic region to which they are applied, are compared with analytically derived results available for benchmarking. The procedures considered yield consistent results for the calculated neutron densities and eigenvalues. Since the one-dimensional transport kernel and its spatial moments are integrable and their integrals can be put in terms of exponential integral functions, the resulting approximations to the neutron density yield somewhat lengthy but closed, forms. These approximate expressions of the neutron density can be used to render, after they are operated on, closed-form formulas for build-up factors, extrapolation distances or angular densities or employed for other purposes that require an analytical expression of the neutron density. As an example of this latter capability, the results of the calculation of the angular density at the surface of the slab are provided. (Author)

  12. Description of a neutron field perturbed by a probe using coupled Monte Carlo and discrete ordinates radiation transport calculations

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1984-01-01

    This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs

  13. Quantifying moisture transport in cementitious materials using neutron radiography

    Science.gov (United States)

    Lucero, Catherine L.

    A portion of the concrete pavements in the US have recently been observed to have premature joint deterioration. This damage is caused in part by the ingress of fluids, like water, salt water, or deicing salts. The ingress of these fluids can damage concrete when they freeze and expand or can react with the cementitious matrix causing damage. To determine the quality of concrete for assessing potential service life it is often necessary to measure the rate of fluid ingress, or sorptivity. Neutron imaging is a powerful method for quantifying fluid penetration since it can describe where water has penetrated, how quickly it has penetrated and the volume of water in the concrete or mortar. Neutrons are sensitive to light atoms such as hydrogen and thus clearly detect water at high spatial and temporal resolution. It can be used to detect small changes in moisture content and is ideal for monitoring wetting and drying in mortar exposed to various fluids. This study aimed at developing a method to accurately estimate moisture content in mortar. The common practice is to image the material dry as a reference before exposing to fluid and normalizing subsequent images to the reference. The volume of water can then be computed using the Beer-Lambert law. This method can be limiting because it requires exact image alignment between the reference image and all subsequent images. A model of neutron attenuation in a multi-phase cementitious composite was developed to be used in cases where a reference image is not available. The attenuation coefficients for water, un-hydrated cement, and sand were directly calculated from the neutron images. The attenuation coefficient for the hydration products was then back-calculated. The model can estimate the degree of saturation in a mortar with known mixture proportions without using a reference image for calculation. Absorption in mortars exposed to various fluids (i.e., deionized water and calcium chloride solutions) were investigated

  14. Radiation transport calculations for the ANS [Advanced Neutron Source] beam tubes

    International Nuclear Information System (INIS)

    Engle, W.W. Jr.; Lillie, R.A.; Slater, C.O.

    1988-01-01

    The Advanced Neutron Source facility (ANS) will incorporate a large number of both radial and no-line-of-sight (NLS) beam tubes to provide very large thermal neutron fluxes to experimental facilities. The purpose of this work was to obtain comparisons for the ANS single- and split-core designs of the thermal and damage neutron and gamma-ray scalar fluxes in these beams tubes. For experimental locations far from the reactor cores, angular flux data are required; however, for close-in experimental locations, the scalar fluxes within each beam tube provide a credible estimate of the various signal to noise ratios. In this paper, the coupled two- and three-dimensional radiation transport calculations employed to estimate the scalar neutron and gamma-ray fluxes will be described and the results from these calculations will be discussed. 6 refs., 2 figs

  15. TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222

    International Nuclear Information System (INIS)

    Shen, H.; Li, Z.; Wang, K.; Yu, G.

    2010-01-01

    A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)

  16. ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

    International Nuclear Information System (INIS)

    1985-01-01

    1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

  17. Quadrature with arbitrary weight for the numerical solution of the critical slab Neutron Transport Equation

    International Nuclear Information System (INIS)

    Sanchez G, J.

    2007-01-01

    A standard procedure for the solution of singular integral equations is applied to the one-dimensional transport equation for monoenergetic neutrons. The results obtained with two versions of the procedure, differing only in the extent of the basic region to which they are applied, are compared with analytically derived results available for benchmarking. The procedures considered yield consistent results for the calculated neutron densities and eigenvalues. Several approximate expressions of the neutron density are used to render closed-form formulas for the densities which can then be analytically operated on to obtain expressions for extrapolation distances or angular densities or serve other purposes that require an analytical expression of the neutron density. (Author)

  18. Monoenergetic time-dependent neutron transport in an infinite medium with time-varying cross sections

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    1987-01-01

    For almost 20 yr, the main thrust of the author's research has been the generation of as many benchmark solutions to the time-dependent monoenergetic neutron transport equation as possible. The major motivation behind this effort has been to provide code developers with highly accurate numerical solutions to serve as standards in the assessment of numerical transport algorithms. In addition, these solutions provide excellent educational tools since the important physical features of neutron transport are still present even though the problems solved are idealized. A secondary motivation, though of equal importance, is the intellectual stimulation and understanding provided by the combination of the analytical, numerical, and computational techniques required to obtain these solutions. Therefore, to further the benchmark development, the added complication of time-dependent cross sections in the one-group transport equation is considered here

  19. The study of neutron transport by oscillation method; Etude du transport des neutrons par la methode de modulation

    Energy Technology Data Exchange (ETDEWEB)

    Raievski, V

    1959-07-01

    The oscillation method is of very general use for studying the behavior of thermal neutrons in media. The main experiments are described and a general theory of them is given. This theory, which is presented in the first part, is established using the two-group approximation which has proved its efficiency in the case of thermal neutron piles. The validity of the two-group approximation is recalled. This allows definition of the meaning of the parameters used in the theory and which are measured in these experiments. The experiments carried out by this method are described, especially those performed at the Centre d'Etudes Nucleaires de Saclay where the method has been extensively used. These experiments are interpreted by means of the general theory given previously. In this way, the identity of parameters measured by this method and those given by the theory is proved. This is particularly conclusive is the case of the mean life of neutrons in a pile. (author) [French] La methode de modulation est un procede tres general d'etude des proprietes neutroniques des milieux contenant des neutrons thermiques. Le present rapport a pour but de decrire les principales de ces experiences et d'en donner une theorie generale. Cette theorie, exposee dans la premiere partie, est etablie dons le cadre de l'approximation a deux groupes de vitesse qui a prouve son efficacite dons le cas des piles a neutrons thermiques. Le domaine de validite de l'approximation a deux groupes est rappele au debut, ce qui permet de definir avec precision la signification des parametres qui entrent dons la theorie et qui font l'objet de ces mesures. La deuxieme partie decrit les experiences realisees, en particulier celles effectuees au Centre d'Etudes Nucleaires de Saclay ou la methode a ete considerablement developpee. Ces experiences sont interpretees dans le cadre de la theorie generale exposee precedemment. On prouve ainsi l'identite des parametres mesures par cette methode et de ceux figurant

  20. Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons

    International Nuclear Information System (INIS)

    Sedlačková, Katarína; Zaťko, Bohumír; Šagátová, Andrea; Nečas, Vladimír

    2013-01-01

    Several Monte Carlo all-particle transport codes are under active development around the world. In this paper we focused on the capabilities of the MCNPX code (Monte Carlo N-Particle eXtended) to follow the particle transport in semiconductor detector of fast neutrons. Semiconductor detector based on semi-insulating GaAs was the object of our investigation. As converter material capable to produce charged particles from the (n, p) interaction, a high-density polyethylene (HDPE) was employed. As the source of fast neutrons, the 239 Pu–Be neutron source was used in the model. The simulations were performed using the MCNPX code which makes possible to track not only neutrons but also recoiled protons at all interesting energies. Hence, the MCNPX code enables seamless particle transport and no other computer program is needed to process the particle transport. The determination of the optimal thickness of the conversion layer and the minimum thickness of the active region of semiconductor detector as well as the energy spectra simulation were the principal goals of the computer modeling. Theoretical detector responses showed that the best detection efficiency can be achieved for 500 μm thick HDPE converter layer. The minimum detector active region thickness has been estimated to be about 400 μm. -- Highlights: ► Application of the MCNPX code for fast neutron detector design is demonstrated. ► Simulations of the particle transport through conversion film of HDPE are presented. ► Simulations of the particle transport through detector active region are presented. ► The optimal thickness of the HDPE conversion film has been calculated. ► Detection efficiency of 0.135% was reached for 500 μm thick HDPE conversion film

  1. Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

    International Nuclear Information System (INIS)

    Shafii, Mohammad Ali; Meidianti, Rahma; Wildian,; Fitriyani, Dian; Tongkukut, Seni H. J.; Arkundato, Artoto

    2014-01-01

    Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation

  2. Numerical solution of the time dependent neutron transport equation by the method of the characteristics

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2013-01-01

    This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps

  3. Numerical solution of the time dependent neutron transport equation by the method of the characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto, E-mail: alby@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States)

    2013-05-01

    This study presents three numerical algorithms to solve the time dependent neutron transport equation by the method of the characteristics. The algorithms have been developed taking into account delayed neutrons and they have been implemented into the novel MCART code, which solves the neutron transport equation for two-dimensional geometry and an arbitrary number of energy groups. The MCART code uses regular mesh for the representation of the spatial domain, it models up-scattering, and takes advantage of OPENMP and OPENGL algorithms for parallel computing and plotting, respectively. The code has been benchmarked with the multiplication factor results of a Boiling Water Reactor, with the analytical results for a prompt jump transient in an infinite medium, and with PARTISN and TDTORT results for cross section and source transients. The numerical simulations have shown that only two numerical algorithms are stable for small time steps.

  4. Theoretical analysis of integral neutron transport equation using collision probability method with quadratic flux approach

    Energy Technology Data Exchange (ETDEWEB)

    Shafii, Mohammad Ali, E-mail: mashafii@fmipa.unand.ac.id; Meidianti, Rahma, E-mail: mashafii@fmipa.unand.ac.id; Wildian,, E-mail: mashafii@fmipa.unand.ac.id; Fitriyani, Dian, E-mail: mashafii@fmipa.unand.ac.id [Department of Physics, Andalas University Padang West Sumatera Indonesia (Indonesia); Tongkukut, Seni H. J. [Department of Physics, Sam Ratulangi University Manado North Sulawesi Indonesia (Indonesia); Arkundato, Artoto [Department of Physics, Jember University Jember East Java Indonesia (Indonesia)

    2014-09-30

    Theoretical analysis of integral neutron transport equation using collision probability (CP) method with quadratic flux approach has been carried out. In general, the solution of the neutron transport using the CP method is performed with the flat flux approach. In this research, the CP method is implemented in the cylindrical nuclear fuel cell with the spatial of mesh being conducted into non flat flux approach. It means that the neutron flux at any point in the nuclear fuel cell are considered different each other followed the distribution pattern of quadratic flux. The result is presented here in the form of quadratic flux that is better understanding of the real condition in the cell calculation and as a starting point to be applied in computational calculation.

  5. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  6. The importance of anisotropic scattering in high energy neutron transport problems

    International Nuclear Information System (INIS)

    Prillinger, G.; Mattes, M.

    1984-01-01

    To describe the highly anisotropic scattering of very fast neutrons adequately the transport code ANISN has been improved. Fokker-Planck terms have been introduced into the transport equation which accurately describe the small changes in energy and angle. The new code has been tested for a d(50)-Be neutron source in a deep penetration iron problem. The influence of the forward peaked elastic scattering on the fast neutron spectrum is shown to be significant and can be handled efficiently in the new ANISN version. Since common cross-section libraries are limited by Legendre expansion, or by their upper energy boundary, or exclude elastic scattering above 20 MeV a special library has been created. (Auth.)

  7. Analytical synthetic methods of solution of neutron transport equation with diffusion theory approaches energy multigroup

    International Nuclear Information System (INIS)

    Moraes, Pedro Gabriel B.; Leite, Michel C.A.; Barros, Ricardo C.

    2013-01-01

    In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation

  8. Radiation Transport Simulation for Boron Neutron Capture Therapy (BNCT)

    Energy Technology Data Exchange (ETDEWEB)

    Ziegner, M.; Blaickner, M. [AIT Austrian Institute of Technology GmbH, Health and Environment Department, Molecular Medicine, Muthgasse 11, 1190 Wien (Austria); Ziegner, M.; Khan, R.; Boeck, H. [Vienna University of Technology, Institute of Atomic and Subatomic Physics, Stadionallee 2, 1020 Wien (Austria); Bortolussi, S.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, National Institute of Nuclear Physics (INFN) Pavia Section, Pavia (Italy); Schmitz, T.; Hampel, G. [Nuclear Chemistry, University of Mainz, Fritz Strassmann Weg 2, 55099 Mainz (Germany)

    2011-07-01

    This work is part of a larger project initiated by the University of Mainz and aiming to use the university's TRIGA reactor to develop a treatment for liver metastases based on Boron Neutron Capture Therapy (BNCT). Diffuse distribution of cancerous cells within the organ makes complete resection difficult and the vicinity to radiosensitive organs impedes external irradiation. Therefore the method of 'autotransplantation', first established at the University of Pavia, is used. The liver is taken out of the body, irradiated in the thermal column of the reactor, therewith purged of metastases and then reimplanted. A highly precise dosimetry system is to be developed by means of measurements at the University of Mainz and computational calculations at the AIT. The stochastic MCNP-5 Monte Carlo-Code, developed by Los Alamos Laboratories, is applied. To verify the calculations of the flux and the absorbed dose in matter a number of measurements are performed irradiating different phantoms and liver sections in a 20cm x 20cm beam tube, which was created by removing graphite blocks from the thermal column of the reactor. The detector material consists of L- {alpha} -alanine pellets which are thought to be the most suitable because of their good tissue equivalence, small size and their wide response range. Another experiment focuses on the determination of the relative biological effectiveness (RBE-factor) of the neutron and photon dose for liver cells. Therefore cell culture plates with the cell medium enriched with {sup 157}Gd and {sup 10}B at different concentrations are irradiated. With regard to the alanine pellets MCNP-5 calculations give stable results. Nevertheless the absorbed dose is underestimated compared to the measurements, a phenomenon already observed in previous works. The cell culture calculations showed the enormous impact of the added isotopes with high thermal neutron cross sections, especially {sup 157}Gd, on the absorbed dose

  9. MC++: A parallel, portable, Monte Carlo neutron transport code in C++

    International Nuclear Information System (INIS)

    Lee, S.R.; Cummings, J.C.; Nolen, S.D.

    1997-01-01

    MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms

  10. Two-group neutron transport theory in adjacent space with lineary anisotropic scattering

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    1978-01-01

    A solution method for two-group neutron transport theory with anisotropic scattering is introduced by the combination of case method (expansion method of self singular function) and the invariant imbedding (invariance principle). The numerical results for the Milne problem in light water and borated water is presented to demonstrate the avalibility of the method [pt

  11. The neutron transport code DTF-Traca users manual and input data

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C

    1979-07-01

    This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.

  12. General-purpose Monte Carlo codes for neutron and photon transport calculations. MVP version 3

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu

    2017-01-01

    JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described. (author)

  13. In situ neutron depth profiling: A powerful method to probe lithium transport in micro-batteries

    NARCIS (Netherlands)

    Oudenhoven, J.F.M.; Labohm, F.; Mulder, M.; Niessen, R.A.H.; Mulder, F.M.; Notten, P.H.L.

    2011-01-01

    In situ neutron depth profiling (NDP) offers the possibility to observe lithium transport inside micro-batteries during battery operation. It is demonstrated that NDP results are consistent with the results of electrochemical measurements, and that the use of an enriched6LiCoO2 cathode offers more

  14. The neutron transport code DTF-Traca users manual and input data

    International Nuclear Information System (INIS)

    Ahnert, C.

    1979-01-01

    This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs

  15. The neutron transport code DTF-TRACA. User's manual and input data

    International Nuclear Information System (INIS)

    Anhert, C.

    1979-01-01

    A user's manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data description is given. The new options developped at JEN are included too. (author)

  16. Developing the theory of nonstationary neutron transport in a homogeneous infinite medium. γk coefficients

    International Nuclear Information System (INIS)

    Trukhanov, G.Ya.

    2005-01-01

    Time-dependent neutron transport theory of G.Ya. Trukhanov and S.A. Podosenov is developed. Errors of calculating of power series expansion coefficients, γ k , in this theory were estimated. It has been found that power series convergence radius R=|χ 1,2 |= 0.9595. Power series convergence speed were estimated [ru

  17. Procedure for obtaining neutron diffusion coefficients from neutron transport Monte Carlo calculations (AWBA Development Program)

    International Nuclear Information System (INIS)

    Gast, R.C.

    1981-08-01

    A procedure for defining diffusion coefficients from Monte Carlo calculations that results in suitable ones for use in neutron diffusion theory calculations is not readily obtained. This study provides a survey of the methods used to define diffusion coefficients from deterministic calculations and provides a discussion as to why such traditional methods cannot be used in Monte Carlo. This study further provides the empirical procedure used for defining diffusion coefficients from the RCP01 Monte Carlo program

  18. Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

    Energy Technology Data Exchange (ETDEWEB)

    Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France); Méchin, Laurence [CNRS, UCBN, Groupe de Recherche en Informatique, Image, Automatique et Instrumentation de Caen, 14050 Caen (France); Hamel, Matthieu [CEA, LIST, Laboratoire Capteurs Architectures Electroniques, 91191 Gif-sur-Yvette (France)

    2016-08-21

    Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

  19. Sensitive and transportable gadolinium-core plastic scintillator sphere for neutron detection and counting

    International Nuclear Information System (INIS)

    Dumazert, Jonathan; Coulon, Romain; Carrel, Frédérick; Corre, Gwenolé; Normand, Stéphane; Méchin, Laurence; Hamel, Matthieu

    2016-01-01

    Neutron detection forms a critical branch of nuclear-related issues, currently driven by the search for competitive alternative technologies to neutron counters based on the helium-3 isotope. The deployment of plastic scintillators shows a high potential for efficient detectors, safer and more reliable than liquids, more easily scalable and cost-effective than inorganic. In the meantime, natural gadolinium, through its 155 and mostly 157 isotopes, presents an exceptionally high interaction probability with thermal neutrons. This paper introduces a dual system including a metal gadolinium core inserted at the center of a high-scale plastic scintillator sphere. Incident fast neutrons are thermalized by the scintillator shell and then may be captured with a significant probability by gadolinium 155 and 157 nuclei in the core. The deposition of a sufficient fraction of the capture high-energy prompt gamma signature inside the scintillator shell will then allow discrimination from background radiations by energy threshold, and therefore neutron detection. The scaling of the system with the Monte Carlo MCNPX2.7 code was carried out according to a tradeoff between the moderation of incident fast neutrons and the probability of slow neutron capture by a moderate-cost metal gadolinium core. Based on the parameters extracted from simulation, a first laboratory prototype for the assessment of the detection method principle has been synthetized. The robustness and sensitivity of the neutron detection principle are then assessed by counting measurement experiments. Experimental results confirm the potential for a stable, highly sensitive, transportable and cost-efficient neutron detector and orientate future investigation toward promising axes.

  20. Minaret, a deterministic neutron transport solver for nuclear core calculations

    International Nuclear Information System (INIS)

    Moller, J-Y.; Lautard, J-J.

    2011-01-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  1. Minaret, a deterministic neutron transport solver for nuclear core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

    2011-07-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  2. Exact analytical solution of time-independent neutron transport equation, and its applications to systems with a point source

    International Nuclear Information System (INIS)

    Mikata, Y.

    2014-01-01

    Highlights: • An exact solution for the one-speed neutron transport equation is obtained. • This solution as well as its derivation are believed to be new. • Neutron flux for a purely absorbing material with a point neutron source off the origin is obtained. • Spherically as well as cylindrically piecewise constant cross sections are studied. • Neutron flux expressions for a point neutron source off the origin are believed to be new. - Abstract: An exact analytical solution of the time-independent monoenergetic neutron transport equation is obtained in this paper. The solution is applied to systems with a point source. Systematic analysis of the solution of the time-independent neutron transport equation, and its applications represent the primary goal of this paper. To the best of the author’s knowledge, certain key results on the scalar neutron flux as well as their derivations are new. As an application of these results, a scalar neutron flux for a purely absorbing medium with a spherically piecewise constant cross section and an isotropic point neutron source off the origin as well as that for a cylindrically piecewise constant cross section with a point neutron source off the origin are obtained. Both of these results are believed to be new

  3. A Comparison of Numerical Strategies for Modeling the Transport Phenomena in High-Energy Laser Surface Alloying Process

    Directory of Open Access Journals (Sweden)

    Dipankar Chatterjee

    2017-06-01

    Full Text Available A comparative assessment is done on the effectiveness of some developed and reported macroscopic and mesoscopic models deployed for addressing the three-dimensional thermo-fluidic transport during high-power laser surface alloying process. The macroscopic models include the most celebrated k–ε turbulence model and the large eddy simulation (LES model, whereas a kinetic theory-based lattice Boltzmann (LB approach is invoked under the mesoscopic paradigm. The time-dependent Navier–Stokes equations are transformed into the k–ε turbulence model by performing the Reynolds averaging technique, whereas a spatial filtering operation is used to produce the LES model. The models are suitably modified to address the turbulent melt-pool convection by using a modified eddy viscosity expression including a damping factor in the form of square root of the liquid fraction. The LB scheme utilizes three separate distribution functions to monitor the underlying hydrodynamic, thermal and compositional fields. Accordingly, the kinematic viscosity, thermal and mass diffusivities are adjusted independently. A single domain fixed-grid enthalpy-porosity approach is utilized to model the phase change phenomena in conjunction with an appropriate enthalpy updating closure scheme. The performance of these models is recorded by capturing the characteristic nature of the thermo-fluidic transport during the laser material processing. The maximum values of the pertinent parameters in the computational domain obtained from several modeling efforts are compared to assess their capabilities. The comparison shows that the prediction from the k–ε turbulence model is higher than the LES and LB models. In addition, the results from all three models are compared with the available experimental results in the form of dimensionless composition of the alloyed layer along the dimensionless depth of the pool. The comparison reveals that the LB and the LES approaches are better

  4. Thermochemical conversion of biomass in smouldering combustion across scales: The roles of heterogeneous kinetics, oxygen and transport phenomena.

    Science.gov (United States)

    Huang, Xinyan; Rein, Guillermo

    2016-05-01

    The thermochemical conversion of biomass in smouldering combustion is investigated here by combining experiments and modeling at two scales: matter (1mg) and bench (100g) scales. Emphasis is put on the effect of oxygen (0-33vol.%) and oxidation reactions because these are poorly studied in the literature in comparison to pyrolysis. The results are obtained for peat as a representative biomass for which there is high-quality experimental data published previously. Three kinetic schemes are explored, including various steps of drying, pyrolysis and oxidation. The kinetic parameters are found using the Kissinger-Genetic Algorithm method, and then implemented in a one-dimensional model of heat and mass transfer. The predictions are validated with thermogravimetric and bench-scale experiments and then analyzed to unravel the role of heterogeneous reaction. This is the first time that the influence of oxygen on biomass smouldering is explained in terms of both chemistry and transport phenomena across scales. Copyright © 2016 The Authors. Published by Elsevier Ltd.. All rights reserved.

  5. Investigation of the Mesoporous Metal-Organic Framework as a New Platform To Study the Transport Phenomena of Biomolecules.

    Science.gov (United States)

    Chen, Yao; Hong, Seongmin; Fu, Chung-Wei; Hoang, Tran; Li, Xiao; Valencia, Veronica; Zhang, Zhenjie; Perman, Jason A; Ma, Shengqian

    2017-03-29

    Mesoporous materials, Tb-mesoMOF and MCM-41, were used to study the transport phenomena of biomolecules entering the interior pores from solution. Vitamins B 12 and B 2 were successfully encapsulated into these mesoporous materials, whereas Tb-mesoMOF (0.33 g of B 12 /g, 0.01 g of B 2 /g) adsorbed a higher amount of vitamin per mass than MCM-41 (0.21 g of B 12 /g, 0.002 g of B 2 /g). The diffusion mechanism of the biomolecules entering Tb-mesoMOF was evaluated using a mathematical model. The Raman spectroscopy studies showed vitamin B 12 has been encapsulated within Tb-mesoMOF's pores, and evaluation of the peak shifts indicated strong interactions linking vitamin B 12 's pyrroline moiety with Tb-mesoMOF's triazine and benzoate rings. Because of these stronger interactions between the vitamins and Tb-mesoMOF, longer egress times were observed than with MCM-41.

  6. Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Kornreich, D.E.

    1997-01-01

    Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade

  7. Numerical solution of neutron transport equations in discrete ordinates and slab geometry

    International Nuclear Information System (INIS)

    Serrano Pedraza, F.

    1985-01-01

    An unified formalism to solve numerically, between other equation, the neutron transport in discrete ordinates, slab geometry, several energy groups and independents of time, has been developed recently. Such a formalism cover some of the conventional schemes as diamond difference, (WDD) characteristic step (SC) lineal characteristic (LC), quadratic characteristic (QC) and lineal discontinuous. Unified formation gives before hand the convergence order of the previously selected scheme. In fact it allows besides to generate a big amount of numerical schemes, with which is also possible to solve numerical equations as soon as neutron transport. The essential purpose of this work was to solve the neutron transport equations in slab geometry and discrete ordinates considering several energy groups without to take under advisement time dependence based in the above mentioned unified formalism. To reach this purpose it was necesary to design a computer code with the name TNOD1 (Neutron transport in discrete ordinates and 1 dimension) which includes each one of the schemes already pointed out. there exist two numerical schemes, also recently developed, quadratic continuous (QC) and cubic continuous (CN), although covered by unified formalism, it has been possible to include them inside this computer code without make substantial changes in its structure. In chapter I, derivative of neutron transport equation independent of time is taken, for angular flux, including boundary conditions and discontinuity. In chapter II the neutron transport equations are obtained in multigroups, independents of time, for approximation of discrete ordinates. Description of theory related with unified formalism and its relationship with mentioned discretization schemes is presented in chapter III. Chapter IV describes the computer code developed and finally, in chapter V different numerical results obtained with TNOD1 program are shown. In Appendix A theorems and mathematical arguments used

  8. Spatially adaptive hp refinement approach for PN neutron transport equation using spectral element method

    International Nuclear Information System (INIS)

    Nahavandi, N.; Minuchehr, A.; Zolfaghari, A.; Abbasi, M.

    2015-01-01

    Highlights: • Powerful hp-SEM refinement approach for P N neutron transport equation has been presented. • The method provides great geometrical flexibility and lower computational cost. • There is a capability of using arbitrary high order and non uniform meshes. • Both posteriori and priori local error estimation approaches have been employed. • High accurate results are compared against other common adaptive and uniform grids. - Abstract: In this work we presented the adaptive hp-SEM approach which is obtained from the incorporation of Spectral Element Method (SEM) and adaptive hp refinement. The SEM nodal discretization and hp adaptive grid-refinement for even-parity Boltzmann neutron transport equation creates powerful grid refinement approach with high accuracy solutions. In this regard a computer code has been developed to solve multi-group neutron transport equation in one-dimensional geometry using even-parity transport theory. The spatial dependence of flux has been developed via SEM method with Lobatto orthogonal polynomial. Two commonly error estimation approaches, the posteriori and the priori has been implemented. The incorporation of SEM nodal discretization method and adaptive hp grid refinement leads to high accurate solutions. Coarser meshes efficiency and significant reduction of computer program runtime in comparison with other common refining methods and uniform meshing approaches is tested along several well-known transport benchmarks

  9. Removal, transportation and disposal of the Millstone 2 neutron thermal shield

    International Nuclear Information System (INIS)

    Snedeker, D.F.; Thomas, L.S.; Schmoker, D.S.; Cade, M.S.

    1985-01-01

    Some PWR reactors equipped with neutron thermal shields (NTS) have experienced severe neutron shield degradation to the extent that removal and disposal of these shields has become necessary. Due to the relative size and activation levels of the thermal shield, disposal techniques, remote material handling and transportation equipment must be carefully evaluated to minimize plant down time and maintain disposal costs at a minimum. This paper describes the techniques, equipment and methodology employed in the removal, transportation and disposal of the NTS at the Millstone 2 Nuclear Generating Station, a PWR facility owned and operated by Northeast Utilities of Hartford, CT. Specific areas addressed include: (1) remote underwater equipment and tooling for use in segmenting and loading the thermal shield in a disposal liner; (2) adaptation of the General Electric IF-300 Irradiated Fuel Cask for transportation of the NTS for disposal; (3) equipment and techniques used for cask handling and liner burial at the Low Level Radioactive Waste (LLRW) disposal facility

  10. Hybrid variational principles and synthesis method for finite element neutron transport calculations

    International Nuclear Information System (INIS)

    Ackroyd, R.T.; Nanneh, M.M.

    1990-01-01

    A family of hybrid variational principles is derived using a generalised least squares method. Neutron conservation is automatically satisfied for the hybrid principles employing two trial functions. No interfaces or reflection conditions need to be imposed on the independent even-parity trial function. For some hybrid principles a single trial function can be employed by relating one parity trial function to the other, using one of the parity transport equation in relaxed form. For other hybrid principles the trial functions can be employed sequentially. Synthesis of transport solutions, starting with the diffusion theory approximation, has been used as a way of reducing the scale of the computation that arises with established finite element methods for neutron transport. (author)

  11. Development of a CAD-based neutron transport code with the method of characteristics

    International Nuclear Information System (INIS)

    Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing

    2012-01-01

    The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)

  12. Numerical simulation for fractional order stationary neutron transport equation using Haar wavelet collocation method

    Energy Technology Data Exchange (ETDEWEB)

    Saha Ray, S., E-mail: santanusaharay@yahoo.com; Patra, A.

    2014-10-15

    Highlights: • A stationary transport equation has been solved using the technique of Haar wavelet collocation method. • This paper intends to provide the great utility of Haar wavelets to nuclear science problem. • In the present paper, two-dimensional Haar wavelets are applied. • The proposed method is mathematically very simple, easy and fast. - Abstract: In this paper the numerical solution for the fractional order stationary neutron transport equation is presented using Haar wavelet Collocation Method (HWCM). Haar wavelet collocation method is efficient and powerful in solving wide class of linear and nonlinear differential equations. This paper intends to provide an application of Haar wavelets to nuclear science problems. This paper describes the application of Haar wavelets for the numerical solution of fractional order stationary neutron transport equation in homogeneous medium with isotropic scattering. The proposed method is mathematically very simple, easy and fast. To demonstrate about the efficiency and applicability of the method, two test problems are discussed.

  13. Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

    International Nuclear Information System (INIS)

    Chen, G.S.

    1997-01-01

    We apply and compare the preconditioned generalized conjugate gradient methods to solve the linear system equation that arises in the two-dimensional neutron and photon transport equation in this paper. Several subroutines are developed on the basis of preconditioned generalized conjugate gradient methods for time-independent, two-dimensional neutron and photon transport equation in the transport theory. These generalized conjugate gradient methods are used. TFQMR (transpose free quasi-minimal residual algorithm), CGS (conjuage gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These sub-routines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. (author)

  14. An assessment of the evaporation and condensation phenomena of lithium during the operation of a Li(d,xn fusion relevant neutron source

    Directory of Open Access Journals (Sweden)

    J. Knaster

    2016-12-01

    Full Text Available The flowing lithium target of a Li(d,xn fusion relevant neutron source must evacuate the deuteron beam power and generate in a stable manner a flux of neutrons with a broad peak at 14 MeV capable to cause similar phenomena as would undergo the structural materials of plasma facing components of a DEMO like reactors. Whereas the physics of the beam-target interaction are understood and the stability of the lithium screen flowing at the nominal conditions of IFMIF (25 mm thick screen with +/–1 mm surface amplitudes flowing at 15 m/s and 523 K has been demonstrated, a conclusive assessment of the evaporation and condensation of lithium during operation was missing. First attempts to determine evaporation rates started by Hertz in 1882 and have since been subject of continuous efforts driven by its practical importance; however intense surface evaporation is essentially a non-equilibrium process with its inherent theoretical difficulties. Hertz-Knudsen-Langmuir (HKL equation with Schrage’s ‘accommodation factor’ η = 1.66 provide excellent agreement with experiments for weak evaporation under certain conditions, which are present during a Li(d,xn facility operation. An assessment of the impact under the known operational conditions for IFMIF (574 K and 10−3Pa on the free surface, with the sticking probability of 1 inherent to a hot lithium gas contained in room temperature steel walls, is carried out. An explanation of the main physical concepts to adequately place needed assumptions is included.

  15. OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

    Energy Technology Data Exchange (ETDEWEB)

    Hou, Jason, E-mail: jason.hou@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Ivanov, Kostadin N. [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Boyarinov, Victor F.; Fomichenko, Peter A. [National Research Centre “Kurchatov Institute”, Kurchatov Sq. 1, Moscow (Russian Federation)

    2017-06-15

    Highlights: • A time-dependent homogenization-free neutron transport benchmark was created. • The first phase, known as the kinetics phase, was described in this work. • Preliminary results for selected 2-D transient exercises were presented. - Abstract: A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for the time-dependent neutron transport calculations without spatial homogenization has been established in order to facilitate the development and assessment of numerical methods for solving the space-time neutron kinetics equations. The benchmark has been named the OECD/NEA C5G7-TD benchmark, and later extended with three consecutive phases each corresponding to one modelling stage of the multi-physics transient analysis of the nuclear reactor core. This paper provides a detailed introduction of the benchmark specification of Phase I, known as the “kinetics phase”, including the geometry description, supporting neutron transport data, transient scenarios in both two-dimensional (2-D) and three-dimensional (3-D) configurations, as well as the expected output parameters from the participants. Also presented are the preliminary results for the initial state 2-D core and selected transient exercises that have been obtained using the Monte Carlo method and the Surface Harmonic Method (SHM), respectively.

  16. The spectral element method for static neutron transport in AN approximation. Part I

    International Nuclear Information System (INIS)

    Barbarino, A.; Dulla, S.; Mund, E.H.; Ravetto, P.

    2013-01-01

    Highlights: ► Spectral elements methods (SEMs) are extended for the neutronics of nuclear reactor cores. ► The second-order, A N formulation of neutron trasport is adopted. ► Results for classical benchmark cases in 2D are presented and compared to finite elements. ► The advantages of SEM in terms of precision and convergence rate are illustrated. ► SEM consitutes a promising approach for the solution of neutron transport problems. - Abstract: Spectral elements methods provide very accurate solutions of elliptic problems. In this paper we apply the method to the A N (i.e. SP 2N−1 ) approximation of neutron transport. Numerical results for classical benchmark cases highlight its performance in comparison with finite element computations, in terms of accuracy per degree of freedom and convergence rate. All calculations presented in this paper refer to two-dimensional problems. The method can easily be extended to three-dimensional cases. The results illustrate promising features of the method for more complex transport problems

  17. Asymptotic equivalence of neutron diffusion and transport in time-independent reactor systems

    International Nuclear Information System (INIS)

    Borysiewicz, M.; Mika, J.; Spiga, G.

    1982-01-01

    Presented in this paper is the asymptotic analysis of the time-independent neutron transport equation in the second-order variational formulation. The small parameter introduced into the equation is an estimate of the ratio of absorption and leakage to scattering in the system considered. When the ratio tends to zero, the weak solution to the transport problem tends to the weak solution of the diffusion problem, including properly defined boundary conditions. A formula for the diffusion coefficient different from that based on averaging the transport mean-free-path is derived

  18. Practical adjoint Monte Carlo technique for fixed-source and eigenfunction neutron transport problems

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1981-01-01

    An adjoint Monte Carlo technique is described for the solution of neutron transport problems. The optimum biasing function for a zero-variance collision estimator is derived. The optimum treatment of an analog of a non-velocity thermal group has also been derived. The method is extended to multiplying systems, especially for eigenfunction problems to enable the estimate of averages over the unknown fundamental neutron flux distribution. A versatile computer code, FOCUS, has been written, based on the described theory. Numerical examples are given for a shielding problem and a critical assembly, illustrating the performance of the FOCUS code. 19 refs

  19. UN Method For The Critical Slab Problem In One-Speed Neutron Transport Theory

    International Nuclear Information System (INIS)

    Oeztuerk, Hakan; Guengoer, Sueleyman

    2008-01-01

    The Chebyshev polynomial approximation (U N method) is used to solve the critical slab problem in one-speed neutron transport theory using Marshak boundary condition. The isotropic scattering kernel with the combination of forward and backward scattering is chosen for the neutrons in a uniform finite slab. Numerical results obtained by the U N method are presented in the tables together with the results obtained by the well-known P N method for comparison. It is shown that the method converges rapidly with its easily executable equations.

  20. The neutron transport with general boundary conditions (II)

    International Nuclear Information System (INIS)

    Boulanouar, Mohamed

    2012-01-01

    This Note deals with the one-dimensional transport operator, on an unbounded domain, endowed with general boundary conditions. We show the generation of a strongly continuous semigroup and we study its spectral properties. In particular, we prove the existence of a leading eigenvalue. (author)

  1. Non-Fick ian law for the neutron density current

    International Nuclear Information System (INIS)

    Espinosa P, G.; Vazquez R, R.; Morales S, J.

    2008-01-01

    In this paper, a fractional wave equation for the average neutron motion in a nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fick ian and non-Fick ian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion. The relaxation time associated with a rapid variation in the neutron flux contains an adjustable parameter, which can be manipulated to obtain the best representation of the neutron transport phenomena. (Author)

  2. An integral equation arising in two group neutron transport theory

    International Nuclear Information System (INIS)

    Cassell, J S; Williams, M M R

    2003-01-01

    An integral equation describing the fuel distribution necessary to maintain a flat flux in a nuclear reactor in two group transport theory is reduced to the solution of a singular integral equation. The formalism developed enables the physical aspects of the problem to be better understood and its relationship with the corresponding diffusion theory model is highlighted. The integral equation is solved by reducing it to a non-singular Fredholm equation which is then evaluated numerically

  3. Comparison of neutronic transport equation resolution nodal methods

    International Nuclear Information System (INIS)

    Zamonsky, O.M.; Gho, C.J.

    1990-01-01

    In this work, some transport equation resolution nodal methods are comparatively studied: the constant-constant (CC), linear-nodal (LN) and the constant-quadratic (CQ). A nodal scheme equivalent to finite differences has been used for its programming, permitting its inclusion in existing codes. Some bidimensional problems have been solved, showing that linear-nodal (LN) are, in general, obtained with accuracy in CPU shorter times. (Author) [es

  4. Methodology for coupling computational fluid dynamics and integral transport neutronics

    International Nuclear Information System (INIS)

    Thomas, J. W.; Zhong, Z.; Sofu, T.; Downar, T. J.

    2004-01-01

    The CFD code STAR-CD was coupled to the integral transport code DeCART in order to provide high-fidelity, full physics reactor simulations. An interface program was developed to perform the tasks of mapping the STAR-CD mesh to the DeCART mesh, managing all communication between STAR-CD and DeCART, and monitoring the convergence of the coupled calculations. The interface software was validated by comparing coupled calculation results with those obtained using an independently developed interface program. An investigation into the convergence characteristics of coupled calculations was performed using several test models on a multiprocessor LINUX cluster. The results indicate that the optimal convergence of the coupled field calculation depends on several factors, to include the tolerance of the STAR-CD solution and the number of DeCART transport sweeps performed before exchanging data between codes. Results for a 3D, multi-assembly PWR problem on 12 PEs of the LINUX cluster indicate the best performance is achieved when the STAR-CD tolerance and number of DeCART transport sweeps are chosen such that the two fields converge at approximately the same rate. (authors)

  5. A Numerical Simulation of Transport Phenomena During the Horizontal Single Belt Casting Process Using an Inclined Feeding System

    Science.gov (United States)

    Xu, Mianguang; Isac, Mihaiela; Guthrie, Roderick I. L.

    2018-06-01

    Horizontal single belt casting (HSBC) has great potential to replace current conventional continuous casting (CCC) processes for sheet metal production, by directly casting 3 to 1 mm sheet for the automobile industry. In the present paper, two-dimensional mathematical models were developed to study transport phenomena, for the case of an inclined wall feeding system for a liquid aluminum wrought alloy (AA6111). Based on the commercial software ANSYS FLUENT 14.5 and user-defined functions, a two-layer turbulence model was used to examine the fluid flow emanating from a slot nozzle set above a water-cooled, high-speed, steel belt. The Volume of Fluid (VOF) method was used to predict the shape of the melt-air interface. A transformed coordinate system ( x', y') was established in order to analyze the fluid flow on the inclined wall of the feeding system. The total pressure gauge gradient ( ∂p total/ ∂x) was used to describe the behavior of the melt film inside the slot nozzle of the head box. The modeling results show that during the melt film falling process, the total gauge pressure varies within the slot nozzle, which can decrease the stability of the falling film. The first impingement between the falling film and the inclined refractory wall of the feeding system gives rise to a local oscillation, and this influences the stability of the melt film moving downwards. At the rear meniscus position between the inclined wall and the moving belt, there is a clear vibration of the air-melt interface, together with a recirculation zone. The weak vibration of the air-melt interface could be induced by the periodic variation of the melt-air interface. Moreover, the formation of tiny air pockets is predicted. Finally, on the inclined wall of the feeding system, a suitable length of the transition area is needed to avoid over-acceleration of the melt film due to the force of gravity.

  6. A Numerical Simulation of Transport Phenomena During the Horizontal Single Belt Casting Process Using an Inclined Feeding System

    Science.gov (United States)

    Xu, Mianguang; Isac, Mihaiela; Guthrie, Roderick I. L.

    2018-03-01

    Horizontal single belt casting (HSBC) has great potential to replace current conventional continuous casting (CCC) processes for sheet metal production, by directly casting 3 to 1 mm sheet for the automobile industry. In the present paper, two-dimensional mathematical models were developed to study transport phenomena, for the case of an inclined wall feeding system for a liquid aluminum wrought alloy (AA6111). Based on the commercial software ANSYS FLUENT 14.5 and user-defined functions, a two-layer turbulence model was used to examine the fluid flow emanating from a slot nozzle set above a water-cooled, high-speed, steel belt. The Volume of Fluid (VOF) method was used to predict the shape of the melt-air interface. A transformed coordinate system (x', y') was established in order to analyze the fluid flow on the inclined wall of the feeding system. The total pressure gauge gradient (∂p total/∂x) was used to describe the behavior of the melt film inside the slot nozzle of the head box. The modeling results show that during the melt film falling process, the total gauge pressure varies within the slot nozzle, which can decrease the stability of the falling film. The first impingement between the falling film and the inclined refractory wall of the feeding system gives rise to a local oscillation, and this influences the stability of the melt film moving downwards. At the rear meniscus position between the inclined wall and the moving belt, there is a clear vibration of the air-melt interface, together with a recirculation zone. The weak vibration of the air-melt interface could be induced by the periodic variation of the melt-air interface. Moreover, the formation of tiny air pockets is predicted. Finally, on the inclined wall of the feeding system, a suitable length of the transition area is needed to avoid over-acceleration of the melt film due to the force of gravity.

  7. Coarse-mesh rebalance methods compatible with the spherical harmonic fictitious source in neutron transport calculations

    International Nuclear Information System (INIS)

    Miller, W.F. Jr.

    1975-10-01

    The coarse-mesh rebalance method, based on neutron conservation, is used in discrete ordinates neutron transport codes to accelerate convergence of the within-group scattering source. Though very powerful for this application, the method is ineffective in accelerating the iteration on the discrete-ordinates-to-spherical-harmonics fictitious sources used for ray-effect elimination. This is largely because this source makes a minimum contribution to the neutron balance equation. The traditional rebalance approach is derived in a variational framework and compared with new rebalance approaches tailored to be compatible with the fictitious source. The new approaches are compared numerically to determine their relative advantages. It is concluded that there is little incentive to use the new methods. (3 tables, 5 figures)

  8. A time-dependent neutron transport model and its coupling to thermal-hydraulics

    International Nuclear Information System (INIS)

    Pautz, A.

    2001-01-01

    A new neutron transport code for time-dependent analyses of nuclear systems has been developed. The code system is based on the well-known Discrete Ordinates code DORT, which solves the steady-state neutron/photon transport equation in two dimensions for an arbitrary number of energy groups and the most common regular geometries. For the implementation of time-dependence a fully implicit first-order scheme was employed to minimize errors due to temporal discretization. This requires various modifications to the transport equation as well as the extensive use of elaborated acceleration mechanisms. The convergence criteria for fluxes, fission rates etc. had to be strongly tightened to ensure the reliability of results. To perform coupled analyses, an interface to the GRS system code ATHLET has been developed. The nodal power densities from the neutron transport code are passed to ATHLET to calculate thermal-hydraulic system parameters, e.g. fuel and coolant temperatures. These are in turn used to generate appropriate nuclear cross sections by interpolation of pre-calculated data sets for each time step. Finally, to demonstrate the transient capabilities of the coupled code system, the research reactor FRM-II has been analysed. Several design basis accidents were modelled, like the loss of off site power, loss of secondary heat sink and unintended control rod withdrawal. (author)

  9. Paranormal phenomena

    Science.gov (United States)

    Gaina, Alex

    1996-08-01

    Critical analysis is given of some paranormal phenomena events (UFO, healers, psychokinesis (telekinesis))reported in Moldova. It is argued that correct analysis of paranormal phenomena should be made in the framework of electromagnetism.

  10. Neutron scattering and muon spin rotation as probes of light interstitial transport

    International Nuclear Information System (INIS)

    Brown, D.W.

    1985-01-01

    The transport of light interstitials, specifically of hydrogen isotopes and the positive muon, is studied with the help of microscopic transport models. The principal observables are the differential neutron scattering cross section of the hydrogen isotopes and the muon spin rotation signal of the positive muon. The transport feature of primary interest is coherence arising as a result of persistence of quantum mechanical phase memory. Evaluation of observables is based on the generalized master equation, or alternatively, the stochastic Liouville equation. The latter is applied to obtain the neutron scattering lineshapes for local tunneling systems as well as for extended Bravais and non-Bravais lattices. It is found that the usual form of the stochastic Liouville equation does not address adequately transport among non-degenerate site-states. An appropriate modification is suggested and employed to obtain scattering lineshapes applicable to recent experiments on impurity-trapped hydrogen. The muon spin rotation signal is formulated under the assumption that spin interactions constitute a negligible source of scattering for muon transport. The depolarization function is evaluated for the cases of local tunneling systems and simple models of spatially extended transport. The former addresses consequences of coherence and both address the consequences of the spatial extent of the muon wavefunction. It is found that the depolarization function is sensitive to the wave function extent, and the detail attributable to it is characterized

  11. Multi-group transport methods for high-resolution neutron activation analysis

    International Nuclear Information System (INIS)

    Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

    2009-01-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

  12. Anisotropic kernel p(μ → μ') for transport calculations of elastically scattered neutrons

    International Nuclear Information System (INIS)

    Stevenson, B.

    1985-01-01

    Literature in the area of anisotropic neutron scattering is by no means lacking. Attention, however, is usually devoted to solution of some particular neutron transport problem and the model employed is at best approximate. The present approach to the problem in general is classically exact and may be of some particular value to individuals seeking exact numerical results in transport calculations. For attempts neutrons originally directed toward the unit vector Omega, it attempts the evaluation of p(theta'), defined such that p(theta') d theta' is that fraction of scattered neutrons that emerges in the vicinity of a cone i.e., having been scattered to between angles theta' and theta' + d theta' with the axis of preferred orientation i; Omega makes an angle theta with i. The relative simplicity of the final form of the solution for hydrogen, in spite of the complicated nature of the limits involved, is a trade-off that truly is not necessary. The exact general solution presented here in integral form, has exceedingly simple limits, i.e., 0 ≤ theta' ≤ π regardless of the material involved; but the form of the final solution is extraordinarily complicated

  13. Importance estimation in Monte Carlo modelling of neutron and photon transport

    International Nuclear Information System (INIS)

    Mickael, M.W.

    1992-01-01

    The estimation of neutron and photon importance in a three-dimensional geometry is achieved using a coupled Monte Carlo and diffusion theory calculation. The parameters required for the solution of the multigroup adjoint diffusion equation are estimated from an analog Monte Carlo simulation of the system under investigation. The solution of the adjoint diffusion equation is then used as an estimate of the particle importance in the actual simulation. This approach provides an automated and efficient variance reduction method for Monte Carlo simulations. The technique has been successfully applied to Monte Carlo simulation of neutron and coupled neutron-photon transport in the nuclear well-logging field. The results show that the importance maps obtained in a few minutes of computer time using this technique are in good agreement with Monte Carlo generated importance maps that require prohibitive computing times. The application of this method to Monte Carlo modelling of the response of neutron porosity and pulsed neutron instruments has resulted in major reductions in computation time. (Author)

  14. Neutron transport by collision probability method in complicated geometries

    International Nuclear Information System (INIS)

    Constantin, Marin

    2000-01-01

    For the first flight collision probability (FFCP) method a rapidly increasing of the memory requirements and execution time with the number of discrete regions occurs. Generally, the use of the method is restricted at cell/supercell level. However, the amazing developments both in computer hardware and computer architecture allow a real extending of the problems' domain and a more detailed treatment of the geometry. Two ways are discussed into the paper: the direct design of new codes and the improving of the mainframe old versions. The author's experience is focused on the performances' improving of the 3D integral transport code PIJXYZ (from an old version to a modern one) and on the design and developing of the 2D transport code CP 2 D in the last years. In the first case an optimization process have been performed before the parallelization. In the second a modular design and the newest techniques (factorization of the geometry, the macrobands method, the mobile set of chords, the automatic calculation of the integration error, optimal algorithms for the innermost programming level, the mixed method for tracking process and CPs calculation, etc.) were adopted. In both cases the parallelization uses a PCs network system. Some short examples for CP 2 D and PIJXYZ calculation are presented: reactivity void effect in typical CANDU cells using a multistratified coolant model, a problem of some adjacent fuel assemblies, CANDU reactivity devices 3D simulation. (author)

  15. Three-dimensional multi-phase flow computational fluid dynamics model for analysis of transport phenomena and thermal stresses in PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Maher, A.R.; Al-Baghdadi, S. [International Technological Univ., London (United Kingdom). Dept. of Mechanical Engineering; Haroun, A.K.; Al-Janabi, S. [Babylon Univ., Babylon (Iraq). Dept. of Mechanical Engineering

    2007-07-01

    Fuel cell technology is expected to play an important role in meeting the growing demand for distributed generation because it can convert the chemical energy of a clean fuel directly into electrical energy. An operating fuel cell has varying local conditions of temperature, humidity, and power generation across the active area of the fuel cell in 3D. This paper presented a model that was developed to improve the basic understanding of the transport phenomena and thermal stresses in PEM fuel cells, and to investigate the behaviour of polymer membrane under hygro and thermal stresses during the cell operation. This comprehensive 3D, multiphase, non-isothermal model accounts for the major transport phenomena in a PEM fuel cell, notably convective and diffusive heat and mass transfer; electrode kinetics; transport and phase change mechanism of water; and potential fields. The model accounts for the liquid water flux inside the gas diffusion layers by viscous and capillary forces and can therefore predict the amount of liquid water inside the gas diffusion layers. This study also investigated the key parameters affecting fuel cell performance including geometry, materials and operating conditions. The model considers the many interacting, complex electrochemical, transport phenomena, thermal stresses and deformation that cannot be studied experimentally. It was concluded that the model can provide a computer-aided tool for the design and optimization of future fuel cells with much higher power density and lower cost. 21 refs., 2 tabs., 14 figs.

  16. Resolution of the neutron transport equation by a three-dimensional least square method

    International Nuclear Information System (INIS)

    Varin, Elisabeth

    2001-01-01

    The knowledge of space and time distribution of neutrons with a certain energy or speed allows the exploitation and control of a nuclear reactor and the assessment of the irradiation dose about an irradiated nuclear fuel storage site. The neutron density is described by a transport equation. The objective of this research thesis is to develop a software for the resolution of this stationary equation in a three-dimensional Cartesian domain by means of a deterministic method. After a presentation of the transport equation, the author gives an overview of the different deterministic resolution approaches, identifies their benefits and drawbacks, and discusses the choice of the Ressel method. The least square method is precisely described and then applied. Numerical benchmarks are reported for validation purposes

  17. Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)

    International Nuclear Information System (INIS)

    Pellegrino, Esteban

    2011-01-01

    Monte Carlo simulation is well suited for solving the Boltzmann neutron transport equation in an inhomogeneous media for complicated geometries. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop PC. The interest in adopting Graphics Processing Units (GPUs) for Monte Carlo acceleration is rapidly growing. This is due to the massive parallelism provided by the latest GPU technologies which is the most promising solution to the challenge of performing full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem were developed for GPU environments in order to evaluate issues associated with computational speedup using GPUs. Results obtained in this work suggest that a speedup of several orders of magnitude is possible using the state-of-the-art GPU technologies. (author) [es

  18. A coordinate transform method for one-speed neutron transport in composite slabs

    International Nuclear Information System (INIS)

    Haidar, N.H.S.

    1988-01-01

    The optical path transformation is applied to reduce the one-speed neutron transport equation for a class of composite subcritical slabs to single-region problems. The class idealises, within the uncertainty of the one-speed model, a variety of practical situations such as U-D 2 O-C-Zr-Pb or Pu-U-Na-Fe symmetric reactor assemblies; which may possibly contain a symmetrically anisotropic neutron source. A closed form double series solution, which turns out to be quite convenient for design and optimisation purposes, has been obtained, in terms of discontinuous functions for the multi-regional angular flux by application of a double finite Legendre transform. Disadvantage factor evaluations for a U-C lattice cell resulting from a low-order P 0 P 1 approximation of this method are found to be in full agreement with hybrid diffusion-transport estimates. (author)

  19. Effect of Fast Neutron Irradiation on Current Transport Properties of HTS Materials

    CERN Document Server

    Ballarino, A; Kruglov, V S; Latushkin, S T; Lubimov, A N; Ryazanov, A I; Shavkin, S V; Taylor, T M; Volkov, P V

    2004-01-01

    The effect of fast neutron irradiation with energy up to 35 MeV and integrated fluence of up to 5 x 10**15 cm-2 on the current transport properties of HTS materials Bi-2212 and Bi-2223 has been studied, both at liquid nitrogen and at room temperatures. The samples irradiated were selected after verification of the stability of their superconducting properties after temperature cycling in the range of 77 K - 293 K. It has been found that the irradiation by fast neutrons up to the above dose does not produce a significant degradation of critical current. The effect of room temperature annealing on the recovery of transport properties of the irradiated samples is also reported, as is a preliminary microstructure investigation of the effect of irradiation on the soldered contacts.

  20. Improved method for solving the neutron transport problem by discretization of space and energy variables

    International Nuclear Information System (INIS)

    Bosevski, T.

    1971-01-01

    The polynomial interpolation of neutron flux between the chosen space and energy variables enabled transformation of the integral transport equation into a system of linear equations with constant coefficients. Solutions of this system are the needed values of flux for chosen values of space and energy variables. The proposed improved method for solving the neutron transport problem including the mathematical formalism is simple and efficient since the number of needed input data is decreased both in treating the spatial and energy variables. Mathematical method based on this approach gives more stable solutions with significantly decreased probability of numerical errors. Computer code based on the proposed method was used for calculations of one heavy water and one light water reactor cell, and the results were compared to results of other very precise calculations. The proposed method was better concerning convergence rate, decreased computing time and needed computer memory. Discretization of variables enabled direct comparison of theoretical and experimental results

  1. Presentation of some methods for the solution of the monoenergetic neutrons transport equation

    International Nuclear Information System (INIS)

    Valle G, E. del.

    1978-01-01

    The neutrons transport theory problems whose solution has been reached were collected in order to show that the transport equation is so complicated that different techniques were developed so as to give approximative numerical solutions to problems concerning the practical application. Such a technique, which had not been investigated in the literature dealing with these problems, is described here. The results which were obtained through this technique in undimensional problems of criticity are satisfactory and speaking in a conceptual way this method is extremely simple because it times. There is no limitation to deal with problems related neutrons sources with an arbitrary distribution and in principle the application of this technique can be extended to unhomogeneous environments. (author)

  2. On discontinuous Galerkin and discrete ordinates approximations for neutron transport equation and the critical eigenvalue

    International Nuclear Information System (INIS)

    Asadzadeh, M.; Thevenot, L.

    2010-01-01

    The objective of this paper is to give a mathematical framework for a fully discrete numerical approach for the study of the neutron transport equation in a cylindrical domain (container model,). More specifically, we consider the discontinuous Galerkin (D G) finite element method for spatial approximation of the mono-energetic, critical neutron transport equation in an infinite cylindrical domain ??in R3 with a polygonal convex cross-section ? The velocity discretization relies on a special quadrature rule developed to give optimal estimates in discrete ordinate parameters compatible with the quasi-uniform spatial mesh. We use interpolation spaces and derive optimal error estimates, up to maximal available regularity, for the fully discrete scalar flux. Finally we employ a duality argument and prove superconvergence estimates for the critical eigenvalue.

  3. Neutron transport solver parallelization using a Domain Decomposition method

    International Nuclear Information System (INIS)

    Van Criekingen, S.; Nataf, F.; Have, P.

    2008-01-01

    A domain decomposition (DD) method is investigated for the parallel solution of the second-order even-parity form of the time-independent Boltzmann transport equation. The spatial discretization is performed using finite elements, and the angular discretization using spherical harmonic expansions (P N method). The main idea developed here is due to P.L. Lions. It consists in having sub-domains exchanging not only interface point flux values, but also interface flux 'derivative' values. (The word 'derivative' is here used with quotes, because in the case considered here, it in fact consists in the Ω.∇ operator, with Ω the angular variable vector and ∇ the spatial gradient operator.) A parameter α is introduced, as proportionality coefficient between point flux and 'derivative' values. This parameter can be tuned - so far heuristically - to optimize the method. (authors)

  4. Colloids: a review of current knowledge with a view to application to phenomena of transportation within PWR; Colloides: point de vue sur les connaissances actuelles en vue d`une application aux phenomenes de transport dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Guinard, L.

    1996-12-31

    In an attempt to minimise dosimetry within the primary circuit of PWR units, research is being carried out into understanding the phenomena of transportation and deposition of corrosion products. It is therefore desirable to known the form of these corrosion products and the laws governing this form. It is generally considered that they are in soluble or particulate form. A third starts with a general presentation of colloids and goes on to define points which are useful, both on a theoretical and experimental level, in terms of application to phenomena of transportation within PWRs. (author). 69 refs., 30 figs., 6 tabs., 3 appends.

  5. Normal and adjoint integral and integrodifferential neutron transport equations. Pt. 2

    International Nuclear Information System (INIS)

    Velarde, G.

    1976-01-01

    Using the simplifying hypotheses of the integrodifferential Boltzmann equations of neutron transport, given in JEN 334 report, several integral equations, and theirs adjoint ones, are obtained. Relations between the different normal and adjoint eigenfunctions are established and, in particular, proceeding from the integrodifferential Boltzmann equation it's found out the relation between the solutions of the adjoint equation of its integral one, and the solutions of the integral equation of its adjoint one (author)

  6. Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

    International Nuclear Information System (INIS)

    Kirk, B.L.; West, J.T.

    1984-06-01

    The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

  7. Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program

    International Nuclear Information System (INIS)

    Moskowitz, B.S.

    2000-01-01

    This paper describes the modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program. This effort represents a complete 'white sheet of paper' rewrite of the code. In this paper, the motivation driving this project, the design objectives for the new version of the program, and the design choices and their consequences will be discussed. The design itself will also be described, including the important subsystems as well as the key classes within those subsystems

  8. The properties and transport phenomena in oxide films on iron, nickel, chromium and their alloys in aqueous environments

    Energy Technology Data Exchange (ETDEWEB)

    Laitinen, T.; Bojinov, M.; Betova, I.; Maekelae, K.; Saario, T. [VTT Manufacturing Technology, Espoo (Finland)

    1999-01-01

    . In addition, the films are to a large extent influenced by the kinetic factors determining their growth rate and steady state thickness. Thus a thermodynamic consideration of the film is not sufficient to model and predict its growth and dissolution. Instead, kinetic models based on in situ experimental data are required. The kinetic models presented in the literature for both ambient and high-temperature aqueous oxidation of metals lack a correlation between the structure of the oxide films and their electronic and ionic properties. Also, a quantitative treatment and thus the capability to predict material behaviour in varying conditions is lacking. A comprehensive understanding of the correlation between applied water chemistry, the behaviour of oxide films and optimum performance of the plant is thus also lacking. The situation calls for more experimental work combined with comprehensive modelling of the behaviour of both the compact and the porous part of the oxide film formed on a metal surface. This will make it possible to recognise the rate-limiting steps of the processes in the film, and thus to influence the rate of activity incorporation and different corrosion phenomena related to transport of species in the film. (author) 210 refs.

  9. The properties and transport phenomena in oxide films on iron, nickel, chromium and their alloys in aqueous environments

    International Nuclear Information System (INIS)

    Laitinen, T.; Bojinov, M.; Betova, I.; Maekelae, K.; Saario, T.

    1999-01-01

    . In addition, the films are to a large extent influenced by the kinetic factors determining their growth rate and steady state thickness. Thus a thermodynamic consideration of the film is not sufficient to model and predict its growth and dissolution. Instead, kinetic models based on in situ experimental data are required. The kinetic models presented in the literature for both ambient and high-temperature aqueous oxidation of metals lack a correlation between the structure of the oxide films and their electronic and ionic properties. Also, a quantitative treatment and thus the capability to predict material behaviour in varying conditions is lacking. A comprehensive understanding of the correlation between applied water chemistry, the behaviour of oxide films and optimum performance of the plant is thus also lacking. The situation calls for more experimental work combined with comprehensive modelling of the behaviour of both the compact and the porous part of the oxide film formed on a metal surface. This will make it possible to recognise the rate-limiting steps of the processes in the film, and thus to influence the rate of activity incorporation and different corrosion phenomena related to transport of species in the film. (author)

  10. The Properties Of And Transport Phenomena In Oxide Films On Iron, Nickel, Chromium And Their Alloys In Aqueous Environments

    International Nuclear Information System (INIS)

    Saario, T.; Laitinen, T.; Maekelae, K.; Bojinov, M.; Betova, I.

    1998-07-01

    . In addition, the films are to a large extent influenced by the kinetic factors determining their growth rate and steady state thickness. Thus a thermodynamic consideration of the film is not sufficient to model and predict its growth and dissolution. Instead, kinetic models based on in situ experimental data are required. The kinetic models presented in the literature for both ambient and high-temperature aqueous oxidation of metals lack a correlation between the structure of the oxide films and their electronic and ionic properties. Also, a quantitative treatment and thus the capability to predict material behaviour in varying conditions is lacking. A comprehensive understanding of the correlation between applied water chemistry, the behaviour of oxide films and optimum performance of the plant is thus also lacking. The situation calls for more experimental work combined with comprehensive modelling of the behaviour of both the dense and the porous part of the oxide film formed on a metal surface. This will make it possible to recognise the rate-limiting steps of the processes in the film, and thus to influence the rate of activity incorporation and different corrosion phenomena related to transport of species in the film. (author)

  11. Developing and investigating a pure Monte-Carlo module for transient neutron transport analysis

    International Nuclear Information System (INIS)

    Mylonakis, Antonios G.; Varvayanni, M.; Grigoriadis, D.G.E.; Catsaros, N.

    2017-01-01

    Highlights: • Development and investigation of a Monte-Carlo module for transient neutronic analysis. • A transient module developed on the open-source Monte-Carlo static code OpenMC. • Treatment of delayed neutrons is inserted. • Simulation of precursors’ decay process is performed. • Transient analysis of simplified test-cases. - Abstract: In the field of computational reactor physics, Monte-Carlo methodology is extensively used in the analysis of static problems while the transient behavior of the reactor core is mostly analyzed using deterministic algorithms. However, deterministic algorithms make use of various approximations mainly in the geometric and energetic domain that may induce inaccuracy. Therefore, Monte-Carlo methodology which generally does not require significant approximations seems to be an attractive candidate tool for the analysis of transient phenomena. One of the most important constraints towards this direction is the significant computational cost; however since nowadays the available computational resources are continuously increasing, the potential use of the Monte-Carlo methodology in the field of reactor core transient analysis seems feasible. So far, very few attempts to employ Monte-Carlo methodology to transient analysis have been reported. Even more, most of those few attempts make use of several approximations, showing the existence of an “open” research field of great interest. It is obvious that comparing to static Monte-Carlo, a straight-forward physical treatment of a transient problem requires the temporal evolution of the simulated neutrons; but this is not adequate. In order to be able to properly analyze transient reactor core phenomena, the proper simulation of delayed neutrons together with other essential extensions and modifications is necessary. This work is actually the first step towards the development of a tool that could serve as a platform for research and development on this interesting but also

  12. Post-merger evolution of a neutron star-black hole binary with neutrino transport

    Science.gov (United States)

    Foucart, Francois; O'Connor, Evan; Roberts, Luke; Duez, Matthew D.; Haas, Roland; Kidder, Lawrence E.; Ott, Christian D.; Pfeiffer, Harald P.; Scheel, Mark A.; Szilagyi, Bela

    2015-06-01

    We present a first simulation of the post-merger evolution of a black hole-neutron star binary in full general relativity using an energy-integrated general-relativistic truncated moment formalism for neutrino transport. We describe our implementation of the moment formalism and important tests of our code, before studying the formation phase of an accretion disk after a black hole-neutron star merger. We use as initial data an existing general-relativistic simulation of the merger of a neutron star of mass 1.4 M⊙ with a black hole of mass 7 M⊙ and dimensionless spin χBH=0.8 . Comparing with a simpler leakage scheme for the treatment of the neutrinos, we find noticeable differences in the neutron-to-proton ratio in and around the disk, and in the neutrino luminosity. We find that the electron neutrino luminosity is much lower in the transport simulations, and that both the disk and the disk outflows are less neutron rich. The spatial distribution of the neutrinos is significantly affected by relativistic effects, due to large velocities and curvature in the regions of strongest emission. Over the short time scale evolved, we do not observe purely neutrino-driven outflows. However, a small amount of material (3 ×10-4M⊙ ) is ejected in the polar region during the circularization of the disk. Most of that material is ejected early in the formation of the disk, and is fairly neutron rich (electron fraction Ye˜0.15 - 0.25 ). Through r-process nucleosynthesis, that material should produce high-opacity lanthanides in the polar region, and could thus affect the light curve of radioactively powered electromagnetic transients. We also show that by the end of the simulation, while the bulk of the disk remains neutron rich (Ye˜0.15 - 0.2 and decreasing), its outer layers have a higher electron fraction: 10% of the remaining mass has Ye>0.3 . As that material would be the first to be unbound by disk outflows on longer time scales, and as composition evolution is

  13. Solution to the monoenergetic time-dependent neutron transport equation with a time-varying source

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    1986-01-01

    Even though fundamental time-dependent neutron transport problems have existed since the inception of neutron transport theory, it has only been recently that a reliable numerical solution to one of the basic problems has been obtained. Experience in generating numerical solutions to time-dependent transport equations has indicated that the multiple collision formulation is the most versatile numerical technique for model problems. The formulation coupled with a moment reconstruction of each collided flux component has led to benchmark-quality (four- to five-digit accuracy) numerical evaluation of the neutron flux in plane infinite geometry for any degree of scattering anisotropy and for both pulsed isotropic and beam sources. As will be shown in this presentation, this solution can serve as a Green's function, thus extending the previous results to more complicated source situations. Here we will be concerned with a time-varying source at the center of an infinite medium. If accurate, such solutions have both pedagogical and practical uses as benchmarks against which other more approximate solutions designed for a wider class of problems can be compared

  14. Advances in the solution of three-dimensional nodal neutron transport equation

    International Nuclear Information System (INIS)

    Pazos, Ruben Panta; Hauser, Eliete Biasotto; Vilhena, Marco Tullio de

    2003-01-01

    In this paper we study the three-dimensional nodal discrete-ordinates approximations of neutron transport equation in a convex domain with piecewise smooth boundaries. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtaining the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. We give numerical results obtained with an algebraic computer system (for N up to 8) and with a code for higher values of N. We compare our results for the geometry of a box with a source in a vertex and a leakage zone in the opposite with others techniques used in this problem. (author)

  15. Synergism of the method of characteristics and CAD technology for neutron transport calculation

    International Nuclear Information System (INIS)

    Chen, Z.; Wang, D.; He, T.; Wang, G.; Zheng, H.

    2013-01-01

    The method of characteristics (MOC) is a very popular methodology in neutron transport calculation and numerical simulation in recent decades for its unique advantages. One of the key problems determining whether the MOC can be applied in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. Most of the existing MOC codes describe the geometry by lines and arcs with extensive input data, such as circles, ellipses, regular polygons and combination of them. Thus they have difficulty in geometry modeling, background meshing and ray tracing for complicated geometry domains. In this study, a new idea making use of a CAD solid modeler MCAM which is a CAD/Image-based Automatic Modeling Program for Neutronics and Radiation Transport developed by FDS Team in China was introduced for geometry modeling and ray tracing of particle transport to remove these geometrical limitations mentioned above. The diamond-difference scheme was applied to MOC to reduce the spatial discretization error of the flat flux approximation in theory. Based on MCAM and MOC, a new MOC code was developed and integrated into SuperMC system, which is a Super Multi-function Computational system for neutronics and radiation simulation. The numerical testing results demonstrated the feasibility and effectiveness of the new idea for geometry treatment in SuperMC. (authors)

  16. SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo

    International Nuclear Information System (INIS)

    2003-01-01

    1 - Nature of physical problem solved: The SAM-CE system comprises two Monte Carlo codes, SAM-F and SAM-A. SAM-F supersedes the forward Monte Carlo code, SAM-C. SAM-A is an adjoint Monte Carlo code designed to calculate the response due to fields of primary and secondary gamma radiation. The SAM-CE system is a FORTRAN Monte Carlo computer code designed to solve the time-dependent neutron and gamma-ray transport equations in complex three-dimensional geometries. SAM-CE is applicable for forward neutron calculations and for forward as well as adjoint primary gamma-ray calculations. In addition, SAM-CE is applicable for the gamma-ray stage of the coupled neutron-secondary gamma ray problem, which may be solved in either the forward or the adjoint mode. Time-dependent fluxes, and flux functionals such as dose, heating, count rates, etc., are calculated as functions of energy, time and position. Multiple scoring regions are permitted and these may be either finite volume regions or point detectors or both. Other scores of interest, e.g., collision and absorption densities, etc., are also made. 2 - Method of solution: A special feature of SAM-CE is its use of the 'combinatorial geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All nuclear interaction cross section data (derived from the ENDF for neutrons and from the UNC-format library for gamma-rays) are tabulated in point energy meshes. The energy meshes for neutrons are internally derived, based on built-in convergence criteria and user- supplied tolerances. Tabulated neutron data for each distinct nuclide are in unique and appropriate energy meshes. Both resolved and unresolved resonance parameters from ENDF data files are treated automatically, and extremely precise and detailed descriptions of cross section behaviour is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux

  17. Criticality problems for slabs and spheres in energy dependent neutron transport theory

    International Nuclear Information System (INIS)

    Victory, H.D. Jr.

    1980-01-01

    The steady-state equation for energy-dependent neutron transport in isotropically scattering slabs and spheres is formulated as an integral equation. The Perron-Frobenius-Jentzsch theory of positive operators is used to analyze criticality problems for transport in slab and spherical media consisting of core and reflector. In addition, with an adroit selection of diffusion-like solutions, this theory is used to obtain an expression relating the critical radius of a homogeneous sphere to a parameter characterizing fission production. 21 refs

  18. Neutron transport study based on assembly modular ray tracing MOC method

    International Nuclear Information System (INIS)

    Tian Chao; Zheng Youqi; Li Yunzhao; Li Shuo; Chai Xiaoming

    2015-01-01

    It is difficulty for the MOC method based on Cell Modular Ray Tracing to deal with the irregular geometry such as the water gap between the PWR lattices. Hence, the neutron transport code NECP-Medlar based on Assembly Modular Ray Tracing is developed. CMFD method is used to accelerate the transport calculation. The numerical results of the 2D C5G7 benchmark and typical PWR lattice prove that NECP-Medlar has an excellent performance in terms of accuracy and efficiency. Besides, NECP-Medlar can describe clearly the flux distribution of the lattice with water gap. (authors)

  19. Neutron and gamma ray transport calculations in shielding system

    Energy Technology Data Exchange (ETDEWEB)

    Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)

  20. A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

    International Nuclear Information System (INIS)

    Lee, S.R.; Cummings, J.C.; Nolen, S.D.

    1997-01-01

    We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed

  1. Modelling of near-field radionuclide transport phenomena in a KBS-3V type of repository for nuclear waste with Goldsim Code - and verification against previous methods

    International Nuclear Information System (INIS)

    Pulkkanen, V.-M.; Nordman, H.

    2010-03-01

    Traditional radionuclide transport models overestimate significantly some phenomena, or completely ignore them. This motivates the development of new more precise models. As a result, this work is a description of commissioning of a new KBS-3V near-field radionuclide transport model, which has been done with a commercial software called GoldSim. According to earlier models, GoldSim model uses rz coordinates, but the solubilities of radionuclides have been treated more precisely. To begin with, the physical phenomena concerning near-field transport have been introduced according to GoldSim way of thinking. Also, the computational methods of GoldSim have been introduced and compared to methods used earlier. The actual verification of GoldSim model has been carried out by comparing the GoldSim results from simple cases to the corresponding results obtained with REPCOM, a software developed by VTT and used in several safety assessments. The results agree well. Finally, a few complicated cases were studied. In these cases, the REPCOM's limitations in handling of some phenomena become evident. The differences in the results are caused especially by the extension of the solubility limit to the whole computational domain, and the element-wise treatment of the solubilities which was used instead of nuclide-wise treatment. This work has been carried out as a special assignment to the former laboratory of Advanced Energy Systems in Helsinki University of Technology. The work was done at VTT. (orig.)

  2. 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)

  3. Some results on the neutron transport and the coupling of equations

    International Nuclear Information System (INIS)

    Bal, G.

    1997-01-01

    Neutron transport in nuclear reactors is well modeled by the linear Boltzmann transport equation. Its resolution is relatively easy but very expensive. To achieve whole core calculations, one has to consider simpler models, such as diffusion or homogeneous transport equations. However, the solutions may become inaccurate in particular situations (as accidents for instance). That is the reason why we wish to solve the equations on small area accurately and more coarsely on the remaining part of the core. It is than necessary to introduce some links between different discretizations or modelizations. In this note, we give some results on the coupling of different discretizations of all degrees of freedom of the integral-differential neutron transport equation (two degrees for the angular variable, on for the energy component, and two or three degrees for spatial position respectively in 2D (cylindrical symmetry) and 3D). Two chapters are devoted to the coupling of discrete ordinates methods (for angular discretization). The first one is theoretical and shows the well posing of the coupled problem, whereas the second one deals with numerical applications of practical interest (the results have been obtained from the neutron transport code developed at the R and D, which has been modified for introducing the coupling). Next, we present the nodal scheme RTN0, used for the spatial discretization. We show well posing results for the non-coupled and the coupled problems. At the end, we deal with the coupling of energy discretizations for the multigroup equations obtained by homogenization. Some theoretical results of the discretization of the velocity variable (well-posing of problems), which do not deal directly with the purposes of coupling, are presented in the annexes. (author)

  4. The potential of detecting intermediate-scale biomass and canopy interception in a coniferous forest using cosmic-ray neutron intensity measurements and neutron transport modeling

    Science.gov (United States)

    Andreasen, M.; Looms, M. C.; Bogena, H. R.; Desilets, D.; Zreda, M. G.; Sonnenborg, T. O.; Jensen, K. H.

    2014-12-01

    The water stored in the various compartments of the terrestrial ecosystem (in snow, canopy interception, soil and litter) controls the exchange of the water and energy between the land surface and the atmosphere. Therefore, measurements of the water stored within these pools are critical for the prediction of e.g. evapotranspiration and groundwater recharge. The detection of cosmic-ray neutron intensity is a novel non-invasive method for the quantification of continuous intermediate-scale soil moisture. The footprint of the cosmic-ray neutron probe is a hemisphere of a few hectometers and subsurface depths of 10-70 cm depending on wetness. The cosmic-ray neutron method offers measurements at a scale between the point-scale measurements and large-scale satellite retrievals. The cosmic-ray neutron intensity is inversely correlated to the hydrogen stored within the footprint. Overall soil moisture represents the largest pool of hydrogen and changes in the soil moisture clearly affect the cosmic-ray neutron signal. However, the neutron intensity is also sensitive to variations of hydrogen in snow, canopy interception and biomass offering the potential to determine water content in such pools from the signal. In this study we tested the potential of determining canopy interception and biomass using cosmic-ray neutron intensity measurements within the framework of the Danish Hydrologic Observatory (HOBE) and the Terrestrial Environmental Observatories (TERENO). Continuous measurements at the ground and the canopy level, along with profile measurements were conducted at towers at forest field sites. Field experiments, including shielding the cosmic-ray neutron probes with cadmium foil (to remove lower-energy neutrons) and measuring reference intensity rates at complete water saturated conditions (on the sea close to the HOBE site), were further conducted to obtain an increased understanding of the physics controlling the cosmic-ray neutron transport and the equipment used

  5. Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

    CERN Document Server

    Žukauskaite, A; Plukiene, R; Plukis, A

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

  6. 48{sup th} Annual meeting on nuclear technology (AMNT 2017). Key topic / Outstanding know-how and sustainable innovations. Technical session: Reactor physics, thermo and fluid dynamics. Neutron flux oscillations phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Herb, Joachim [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany). Abt. Kuehlkreislauf

    2018-01-15

    The Technical Session about Neutron Flux Oscillation Phenomena was chaired by Joachim Herb (Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) GmbH) and well attended by approx. 50 listeners. It comprised of three keynotes and two technical presentations. The main topics were the significant changes of the neutron flux noise levels in different German and foreign pressurized water reactors (PWRs). For about ten years an increase in neutron noise levels has been observed in German PWRs. During the following five years the noise levels have been decreasing again. In principle, a correlation of the neutron noise levels to the use of certain fuel element types was observed and the phenomenon of neutron flux oscillations had been known since decades. Nevertheless, no self-consistent physical theory exists so far, which can explain the observed changes and the absolute levels of the observed neutron flux noise levels. Therefore, safety authorities, technical support organizations (TSO), utilities as well as research organizations showed increased interest in this topic during the last years. The results of the corresponding work as well as an outlook into soon-starting research projects were given in this session.

  7. Fast rigorous numerical method for the solution of the anisotropic neutron transport problem and the NITRAN system for fusion neutronics application. Pt. 1

    International Nuclear Information System (INIS)

    Takahashi, A.; Rusch, D.

    1979-07-01

    Some recent neutronics experiments for fusion reactor blankets show that the precise treatment of anisotropic secondary emissions for all types of neutron scattering is needed for neutron transport calculations. In the present work new rigorous methods, i.e. based on non-approximative microscopic neutron balance equations, are applied to treat the anisotropic collision source term in transport equations. The collision source calculation is free from approximations except for the discretization of energy, angle and space variables and includes the rigorous treatment of nonelastic collisions, as far as nuclear data are given. Two methods are presented: first the Ii-method, which relies on existing nuclear data files and then, as an ultimate goal, the I*-method, which aims at the use of future double-differential cross section data, but which is also applicable to the present single-differential data basis to allow a smooth transition to the new data type. An application of the Ii-method is given in the code system NITRAN which employs the Ssub(N)-method to solve the transport equations. Both rigorous methods, the Ii- and the I*-method, are applicable to all radiation transport problems and they can be used also in the Monte-Carlo-method to solve the transport problem. (orig./RW) [de

  8. Calculation of neutron spectra for a 252Cf transport cask using ANISN running on a PC

    International Nuclear Information System (INIS)

    West, L.; Akin, B.P.; Lemley, E.C.

    1995-01-01

    Neutron spectra have been calculated using the ANISN one-dimensional discrete ordinates code for the case of a 152 Cf source in a transport cask of a particular design. All computations were done on personal computers (PCs) (mostly 486 models) with the ANISN-ORNL (486 version) computer code. With a source of 252 Cf fission neutrons, the neutron flux spectrum in the cask cannot be characterized as open-quotes moderated.close quotes Concern about an appropriate choice for the cross-section data set has led to a comparison, for this application, of three different cross-section libraries: DABL, HILO, and BUGLE-80. Although the cross-section sets were not originally designed for PC use, the libraries have been successfully employed for PC computations. Work with yet another data library, BUGLE-93, is incomplete at this stage. From neutron flux spectra on the surface of the cask, personnel dosimetric quantities (such as dose equivalent) have been determined for the DABL, HILO, and BUGLE-80 ANISN calculations

  9. Positive solution of a time and energy dependent neutron transport problem

    International Nuclear Information System (INIS)

    Pao, C.V.

    1975-01-01

    A constructive method is given for the determination of a solution and an existence--uniqueness theorem for some nonlinear time and energy dependent neutron transport problems, including the linear transport system. The geometry of the medium under consideration is allowed to be either bounded or unbounded which includes the geometry of a finite or infinite cylinder, a half-space and the whole space R/subm/ (m=1,2,center-dotcenter-dotcenter-dot). Our approach to the problem is by successive approximation which leads to various recursion formulas for the approximations in terms of explicit integrations. It is shown under some Lipschitz conditions on the nonlinear functions, which describe the process of neutrons absorption, fission, and scattering, that the sequence of approximations converges to a unique positive solution. Since these conditions are satisfied by the linear transport equation, all the results for the nonlinear system are valid for the linear transport problem. In the general nonlinear problem, the existence of both local and global solutions are discussed, and an iterative process for the construction of the solution is given

  10. 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

  11. Method for calculating anisotropic neutron transport using scattering kernel without polynomial expansion

    International Nuclear Information System (INIS)

    Takahashi, Akito; Yamamoto, Junji; Ebisuya, Mituo; Sumita, Kenji

    1979-01-01

    A new method for calculating the anisotropic neutron transport is proposed for the angular spectral analysis of D-T fusion reactor neutronics. The method is based on the transport equation with new type of anisotropic scattering kernels formulated by a single function I sub(i) (μ', μ) instead of polynomial expansion, for instance, Legendre polynomials. In the calculation of angular flux spectra by using scattering kernels with the Legendre polynomial expansion, we often observe the oscillation with negative flux. But in principle this oscillation disappears by this new method. In this work, we discussed anisotropic scattering kernels of the elastic scattering and the inelastic scatterings which excite discrete energy levels. The other scatterings were included in isotropic scattering kernels. An approximation method, with use of the first collision source written by the I sub(i) (μ', μ) function, was introduced to attenuate the ''oscillations'' when we are obliged to use the scattering kernels with the Legendre polynomial expansion. Calculated results with this approximation showed remarkable improvement for the analysis of the angular flux spectra in a slab system of lithium metal with the D-T neutron source. (author)

  12. FPGA hardware acceleration for high performance neutron transport computation based on agent methodology - 318

    International Nuclear Information System (INIS)

    Shanjie, Xiao; Tatjana, Jevremovic

    2010-01-01

    The accurate, detailed and 3D neutron transport analysis for Gen-IV reactors is still time-consuming regardless of advanced computational hardware available in developed countries. This paper introduces a new concept in addressing the computational time while persevering the detailed and accurate modeling; a specifically designed FPGA co-processor accelerates robust AGENT methodology for complex reactor geometries. For the first time this approach is applied to accelerate the neutronics analysis. The AGENT methodology solves neutron transport equation using the method of characteristics. The AGENT methodology performance was carefully analyzed before the hardware design based on the FPGA co-processor was adopted. The most time-consuming kernel part is then transplanted into the FPGA co-processor. The FPGA co-processor is designed with data flow-driven non von-Neumann architecture and has much higher efficiency than the conventional computer architecture. Details of the FPGA co-processor design are introduced and the design is benchmarked using two different examples. The advanced chip architecture helps the FPGA co-processor obtaining more than 20 times speed up with its working frequency much lower than the CPU frequency. (authors)

  13. Analysis of the neutrons dispersion in a semi-infinite medium based in transport theory and the Monte Carlo method

    International Nuclear Information System (INIS)

    Arreola V, G.; Vazquez R, R.; Guzman A, J. R.

    2012-10-01

    In this work a comparative analysis of the results for the neutrons dispersion in a not multiplicative semi-infinite medium is presented. One of the frontiers of this medium is located in the origin of coordinates, where a neutrons source in beam form, i.e., μο=1 is also. The neutrons dispersion is studied on the statistical method of Monte Carlo and through the unidimensional transport theory and for an energy group. The application of transport theory gives a semi-analytic solution for this problem while the statistical solution for the flow was obtained applying the MCNPX code. The dispersion in light water and heavy water was studied. A first remarkable result is that both methods locate the maximum of the neutrons distribution to less than two mean free trajectories of transport for heavy water, while for the light water is less than ten mean free trajectories of transport; the differences between both methods is major for the light water case. A second remarkable result is that the tendency of both distributions is similar in small mean free trajectories, while in big mean free trajectories the transport theory spreads to an asymptote value and the solution in base statistical method spreads to zero. The existence of a neutron current of low energy and toward the source is demonstrated, in contrary sense to the neutron current of high energy coming from the own source. (Author)

  14. Coupling of neutron transport equations. First results; Couplage d`equations en transport neutronique. premiere approche 1D monocinetique

    Energy Technology Data Exchange (ETDEWEB)

    Bal, G.

    1995-07-01

    To achieve whole core calculations of the neutron transport equation, we have to follow this 2 step method: space and energy homogenization of the assemblies; resolution of the homogenized equation on the whole core. However, this is no more valid when accidents occur (for instance depressurization causing locally strong heterogeneous media). One solution consists then in coupling two kinds of resolutions: a fine computation on the damaged cell (fine mesh, high number of energy groups) coupled with a coarse one everywhere else. We only deal here with steady state solutions (which already live in 6D spaces). We present here two such methods: The coupling by transmission of homogenized sections and the coupling by transmission of boundary conditions. To understand what this coupling is, we first restrict ourselves to 1D with respect to space in one energy group. The first two chapters deal with a recall of basic properties of the neutron transport equation. We give at chapter 3 some indications of the behaviour of the flux with respect to the cross sections. We present at chapter 4 some couplings and give some properties. Chapter 5 is devoted to a presentation of some numerical applications. (author). 9 refs., 7 figs.

  15. SITHA program for simulating hadron transport in the 100-1010eV energy range. Simulation of neutron transport with E6 eV

    International Nuclear Information System (INIS)

    Daniehl', A.V.; Dushin, V.N.

    1987-01-01

    The methods for simulation of neutron transport with Z<20 MeV used in the SITHA (simulation transport hadron) program, the original library of group microconstants (175 groups) with subgroup description of resonance range and a set of programs for its creation are described. The results of a number of integral experiments are discussed

  16. Neutron and Gamma Shielding Evaluation for KN-12 Spent Nuclear Fuel Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cho, I. J.; Min, D. K.; Lee, J. C.; You, G. S.; Yoon, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, G. H.; Jeong, Y. C.; Ko, Y. W. [Korea Hydro and Nuclear Power Co., LTD., Kori (Korea, Republic of)

    2007-07-01

    The CASTOR KN-12 is designed to transport 12 intact PWR spent fuel assemblies for dry and wet transportation conditions. The overall cask length is 480.1 cm with a wall thickness 37.5 cm. Shield for the KN-12 is maintained by the thick walled cask body and the lid. For neutron shielding, polyethylene rods (PE) are arranged in longitudinal boreholes in the vessel wall and PE-plates are inserted between the cask lid and lid side shock absorber and between the cask bottom and bottom steel plate. The shielding evaluation of the cask has been performed with MCNP to confirm the shielding integrity of cask for pre-service inspection of transport cask.

  17. Response matrix method for neutron transport in reactor lattices using group symmetry properties

    International Nuclear Information System (INIS)

    Mund, E.H.

    1991-01-01

    This paper describes a response matrix method for the approximate solution of one-velocity, multi-dimensional transport problems in reactor lattices, with isotropic neutron scattering. The transport equation is solved on a homogeneous cell by using a Petrov-Galerkin technique based on a set of trial and test functions (including polynomials and exponential functions) closely related to transport problems in infinite media. The number of non-zero elements of the response matrices reduces to a minimum when the symmetry properties of the cell are included ab initio in the span of the basis functions. To include these properties, use is made of projection operations which are performed very efficiently on symbolic manipulation programs. Numerical results of model problems in square geometry show a good agreement with reference solutions

  18. The use of symbolic computation in radiative, energy, and neutron transport calculations

    Science.gov (United States)

    Frankel, J. I.

    This investigation uses symbolic computation in developing analytical methods and general computational strategies for solving both linear and nonlinear, regular and singular, integral and integro-differential equations which appear in radiative and combined mode energy transport. This technical report summarizes the research conducted during the first nine months of the present investigation. The use of Chebyshev polynomials augmented with symbolic computation has clearly been demonstrated in problems involving radiative (or neutron) transport, and mixed-mode energy transport. Theoretical issues related to convergence, errors, and accuracy have also been pursued. Three manuscripts have resulted from the funded research. These manuscripts have been submitted to archival journals. At the present time, an investigation involving a conductive and radiative medium is underway. The mathematical formulation leads to a system of nonlinear, weakly-singular integral equations involving the unknown temperature and various Legendre moments of the radiative intensity in a participating medium. Some preliminary results are presented illustrating the direction of the proposed research.

  19. Solution of neutron transport equation using Daubechies' wavelet expansion in the angular discretization

    International Nuclear Information System (INIS)

    Cao Liangzhi; Wu Hongchun; Zheng Youqi

    2008-01-01

    Daubechies' wavelet expansion is introduced to discretize the angular variables of the neutron transport equation when the neutron angular flux varies very acutely with the angular directions. An improvement is made by coupling one-dimensional wavelet expansion and discrete ordinate method to make two-dimensional angular discretization efficient and stable. The angular domain is divided into several subdomains for treating the vacuum boundary condition exactly in the unstructured geometry. A set of wavelet equations coupled with each other is obtained in each subdomain. An iterative method is utilized to decouple the wavelet moments. The numerical results of several benchmark problems demonstrate that the wavelet expansion method can provide more accurate results by lower-order expansion than other angular discretization methods

  20. Spherical harmonics solutions of multi-dimensional neutron transport equation by finite Fourier transformation

    International Nuclear Information System (INIS)

    Kobayashi, Keisuke

    1977-01-01

    A method of solution of a monoenergetic neutron transport equation in P sub(L) approximation is presented for x-y and x-y-z geometries using the finite Fourier transformation. A reactor system is assumed to consist of multiregions in each of which the nuclear cross sections are spatially constant. Since the unknown functions of this method are the spherical harmonics components of the neutron angular flux at the material boundaries alone, the three- and two-dimensional equations are reduced to two- and one-dimensional equations, respectively. The present approach therefore gives fewer unknowns than in the usual series expansion method or in the finite difference method. Some numerical examples are shown for the criticality problem. (auth.)

  1. One-speed neutron transport in spheres with totally absorbing cores

    International Nuclear Information System (INIS)

    Sjoestrand, N.G.

    1988-01-01

    Stationary and time-dependent transport of neutrons of one speed has been studied in spheres with totally absorbing cores. For stationary, critical reactors the number of secondaries per collision has been calculated numerically for various inner and outer radii. In the time-dependent case, the decay constant has been calculated for spherical shells of different inner radii and thicknesses. For a fixed ratio between shell thickness and inner radius, the curve of the decay constant versus shell thickness crosses the Corngold limit in the same way as the curve for a homogeneous sphere. When the ratio goes to zero the curve approaches that for an infinite slab. The behaviour is discussed in view of a new result from collision theory, viz. that the following condition must be fulfilled for a body at the point where the decay constant curve crosses the Corngold limit: the average exit distance of the neutrons is equal to the mean free path for scattering

  2. Theory of threshold phenomena

    International Nuclear Information System (INIS)

    Hategan, Cornel

    2002-01-01

    Theory of Threshold Phenomena in Quantum Scattering is developed in terms of Reduced Scattering Matrix. Relationships of different types of threshold anomalies both to nuclear reaction mechanisms and to nuclear reaction models are established. Magnitude of threshold effect is related to spectroscopic factor of zero-energy neutron state. The Theory of Threshold Phenomena, based on Reduced Scattering Matrix, does establish relationships between different types of threshold effects and nuclear reaction mechanisms: the cusp and non-resonant potential scattering, s-wave threshold anomaly and compound nucleus resonant scattering, p-wave anomaly and quasi-resonant scattering. A threshold anomaly related to resonant or quasi resonant scattering is enhanced provided the neutron threshold state has large spectroscopic amplitude. The Theory contains, as limit cases, Cusp Theories and also results of different nuclear reactions models as Charge Exchange, Weak Coupling, Bohr and Hauser-Feshbach models. (author)

  3. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    1979-11-01

    The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables

  4. Small scale laboratory studies of flow and transport phenomena in pores and fractures: Phase II. Progress report, 3rd year continuation proposal, and work plan

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, J.L.

    1994-05-01

    Small scale laboratory experiments, equipped with an ability to actually observe behavior on the pore level using microscopy, provide an economical and easily understood scientific tool to help us validate concepts and assumptions about the transport of contaminants, and offers the propensity to discover heretofore unrecognized phenomena or behavior. The main technique employs etched glass micromodels, composed of two etched glass plates, sintered together, to form a two dimensional network of three dimensional pores. Flow and transport behavior is observed on a pore or pore network level, and recorder on film and video tape. This technique is coupled with related column studies. These techniques have been used to study multiphase flow, colloid transport and most recently bacteria transport. The project has recently moved to the Bacteria Transport Subprogram, and efforts have been redirected to support that Subprogram and its collaborative field experiment. We proposed to study bacteria transport factors of relevance to the field experiment, using micromodels and other laboratory techniques. Factors that may be addressed include bacteria characteristics (eg, hydrophobicity), pore size and shape, permeability heterogeneity, surface chemistry (eg, iron oxide coatings), surface chemistry heterogeneity, active versus resting cell bacteria, and mixed bacteria populations. In other work we will continue to examine the effects of fluid-fluid interfaces on bacteria transport, and develop a new assay for bacteria hydrophobicity. Finally we will collaborate on characterization of the field site, and the design, operation, and interpretation of the field experiment.

  5. GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport

    International Nuclear Information System (INIS)

    Clarno, K.; De Almeida, V.; D'Azevedo, E.; De Oliveira, C.; Hamilton, S.

    2006-01-01

    A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)

  6. properties of the SN - equivalent integral transport operator in slab geometry and the iterative acceleration of neutron transport methods

    International Nuclear Information System (INIS)

    Massimiliano, Rosa; Azmy, Y.Y.; Morel, J.E.

    2005-01-01

    solution of the neutron transport equation. (authors)

  7. EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

    International Nuclear Information System (INIS)

    Jones, D.B.

    1986-01-01

    EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

  8. Application of the three-dimensional transport code to analysis of the neutron streaming experiment

    International Nuclear Information System (INIS)

    Chatani, K.; Slater, C.O.

    1990-01-01

    The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan

  9. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

    International Nuclear Information System (INIS)

    Collins, Benjamin; Stimpson, Shane; Kelley, Blake W.; Young, Mitchell T.H.; Kochunas, Brendan; Graham, Aaron; Larsen, Edward W.; Downar, Thomas; Godfrey, Andrew

    2016-01-01

    A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

  10. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Benjamin, E-mail: collinsbs@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Stimpson, Shane, E-mail: stimpsonsg@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Kelley, Blake W., E-mail: kelleybl@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Young, Mitchell T.H., E-mail: youngmit@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Kochunas, Brendan, E-mail: bkochuna@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Graham, Aaron, E-mail: aarograh@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Larsen, Edward W., E-mail: edlarsen@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Downar, Thomas, E-mail: downar@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Godfrey, Andrew, E-mail: godfreyat@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

    2016-12-01

    A consistent “2D/1D” neutron transport method is derived from the 3D Boltzmann transport equation, to calculate fuel-pin-resolved neutron fluxes for realistic full-core Pressurized Water Reactor (PWR) problems. The 2D/1D method employs the Method of Characteristics to discretize the radial variables and a lower order transport solution to discretize the axial variable. This paper describes the theory of the 2D/1D method and its implementation in the MPACT code, which has become the whole-core deterministic neutron transport solver for the Consortium for Advanced Simulations of Light Water Reactors (CASL) core simulator VERA-CS. Several applications have been performed on both leadership-class and industry-class computing clusters. Results are presented for whole-core solutions of the Watts Bar Nuclear Power Station Unit 1 and compared to both continuous-energy Monte Carlo results and plant data.

  11. Application of synthetic diffusion method in the numerical solution of the equations of neutron transport in slab geometry

    International Nuclear Information System (INIS)

    Valdes Parra, J.J.

    1986-01-01

    One of the main problems in reactor physics is to determine the neutron distribution in reactor core, since knowing that, it is possible to calculate the rapidity of occurrence of different nuclear reaction inside the reactor core. Within different theories existing in nuclear reactor physics, is neutron transport the one in which equation who govern the exact behavior of neutronic distribution are developed even inside the proper neutron transport theory, there exist different methods of solution which are approximations to exact solution; still more, with the purpose to reach a more precise solution, the majority of methods have been approached to the obtention of solutions in numerical form with the aim of take the advantages of modern computers, and for this reason a great deal of effort is dedicated to numerical solution of the equations of neutron transport. In agreement with the above mentioned, in this work has been developed a computer program which uses a relatively new techniques known as 'acceleration of synthetic diffusion' which has been applied to solve the neutron transport equation with 'classical schemes of spatial integration' obtaining results with a smaller quantity of interactions, if they compare to done without using such equation (Author)

  12. Simulation of neutron transport process, photons and charged particles within the Monte Carlo method

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Artamonov, S.N.; Bolonkina, G.V.; Lomtev, V.L.; Pupko, S.V.

    1991-01-01

    Description is given to the program system BRAND designed for the accurate solution of non-stationary transport equation of neutrons, photons and charged particles in the conditions of real three-dimensional geometry. An extensive set of local and non-local estimates provides an opportunity of calculating a great set of linear functionals normally being of interest in the calculation of reactors, radiation protection and experiment simulation. The process of particle interaction with substance is simulated on the basis of individual non-group data on each isotope of the composition. 24 refs

  13. Transport coefficients in neutron star cores in BHF approach. Comparison of different nucleon potentials

    Science.gov (United States)

    Shternin, P. S.; Baldo, M.; Schulze, H.-J.

    2017-12-01

    Thermal conductivity and shear viscosity of npeµ matter in non-superfluid neutron star cores are considered in the framework of Brueckner-Hartree-Fock many-body theory. We extend our previous work (Shternin et al 2013 PRC 88 065803) by analysing different nucleon-nucleon potentials and different three-body forces. We find that the use of different potentials leads up to one order of magnitude variations in the values of the nucleon contribution to transport coefficients. The nucleon contribution dominates the thermal conductivity, but for all considered models the shear viscosity is dominated by leptons.

  14. Comparison of two Ssub(infinity) methods for solving the neutron transport equation

    International Nuclear Information System (INIS)

    Mennig, J.; Brandt, D.; Haelg, W.

    1978-01-01

    A semianalytic method (S 0 sub(infinity)) is presented for solving the monoenergetic multi-region transport equation. This method is compared with results from S 1 sub(infinity)-theory given in the literature. Application of S 1 sub(infinity)-theory to reactor shields may lead to negative neutron fluxes and to flux oscillations. These unphysical effects are completely avoided by the new method. Numerical results demonstrate the limitations of S 1 sub(infinity) and confirm the numerical stability of (S 0 sub(infinity)). (Auth.)

  15. Solution of the neutron transport equation by means of Hermite-Ssub(infinity)-theory

    International Nuclear Information System (INIS)

    Brandt, D.; Haelg, W.; Mennig, J.

    1979-01-01

    A stable numerical approximation Hsub(α)-Ssub(infinity) is obtained through the use of Hermite's method of order α(Hsub(α)) in the spatial integration of the ID neutron transport equation. The theory for α = 1 is applied to a one-group shielding problem. Numerical calculations show the new method to converge much faster than earlier versions of Ssub(infinity)-theory. Comparison of H 1 - Ssub(infinity) with the well-known Ssub(N)-code ANISN indicates a large gain in computing time for the former. (Auth.)

  16. Asymptotic formulae for solutions of the two-group integral neutron-transport equation

    International Nuclear Information System (INIS)

    Duracz, T.

    1976-01-01

    The steady-state, two-group integral neutron-transport equation is considered for two cases. First, for plane geometry, formulae for the asymptotic flux are obtained, under assumptions of homogeneous medium with isotropic scattering, extended to infinity (whole space and half-space), with sources vanishing at infinity as 0(esup(-IXI)). Next, for spherical geometry, the Milne problem is considered and formulae for the asymptotic flux are obtained. These formulae have the form of asymptotic expansions for small and large radii of the black sphere. (orig.) [de

  17. Solution of the neutron transport problem with anisotropic scattering in cylindrical geometry by the decomposition method

    International Nuclear Information System (INIS)

    Goncalves, G.A.; Bogado Leite, S.Q.; Vilhena, M.T. de

    2009-01-01

    An analytical solution has been obtained for the one-speed stationary neutron transport problem, in an infinitely long cylinder with anisotropic scattering by the decomposition method. Series expansions of the angular flux distribution are proposed in terms of suitably constructed functions, recursively obtainable from the isotropic solution, to take into account anisotropy. As for the isotropic problem, an accurate closed-form solution was chosen for the problem with internal source and constant incident radiation, obtained from an integral transformation technique and the F N method

  18. Whole core neutronics modeling of a TRIGA reactor using integral transport theory

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.; Toffer, H.

    1990-01-01

    An innovative analysis approach for performing whole core reactor physics calculations for TRIGA reactors has been employed recently at the Westinghouse Hanford Company. A deterministic transport theory model with sufficient geometric complexity to evaluate asymmetric loading patterns was used. Calculations of this complexity have been performed in the past using Monte Carlo simulation, such as the MCNP code. However, the Monte Carlo calculations are more difficult to prepare and require more computer time. On the Hanford Site CRAY XMP-18 computer, the new methods required less than one-third of the central processing unit time per calculation as compared to an MCNP calculation using 100,000 neutron histories

  19. Adjacent-cell Preconditioners for solving optically thick neutron transport problems

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1994-01-01

    We develop, analyze, and test a new acceleration scheme for neutron transport methods, the Adjacent-cell Preconditioner (AP) that is particularly suited for solving optically thick problems. Our method goes beyond Diffusion Synthetic Acceleration (DSA) methods in that it's spectral radius vanishes with increasing cell thickness. In particular, for the ID case the AP method converges immediately, i.e. in one iteration, to 10 -4 pointwise relative criterion in problems with dominant cell size of 10 mfp or thicker. Also the AP has a simple formalism and is cell-centered hence, multidimensional and high order extensions are easier to develop, and more efficient to implement

  20. Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

    CERN Document Server

    Žukauskaitėa, A; Plukienė, R; Ridikas, D

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.