Computer codes for three dimensional mass transport with non-linear sorption

International Nuclear Information System (INIS)

Noy, D.J.

1985-03-01

The report describes the mathematical background and data input to finite element programs for three dimensional mass transport in a porous medium. The transport equations are developed and sorption processes are included in a general way so that non-linear equilibrium relations can be introduced. The programs are described and a guide given to the construction of the required input data sets. Concluding remarks indicate that the calculations require substantial computer resources and suggest that comprehensive preliminary analysis with lower dimensional codes would be important in the assessment of field data. (author)

A new two dimensional spectral/spatial multi-diagonal code for noncoherent optical code division multiple access (OCDMA) systems

Science.gov (United States)

Kadhim, Rasim Azeez; Fadhil, Hilal Adnan; Aljunid, S. A.; Razalli, Mohamad Shahrazel

2014-10-01

A new two dimensional codes family, namely two dimensional multi-diagonal (2D-MD) codes, is proposed for spectral/spatial non-coherent OCDMA systems based on the one dimensional MD code. Since the MD code has the property of zero cross correlation, the proposed 2D-MD code also has this property. So that, the multi-access interference (MAI) is fully eliminated and the phase induced intensity noise (PIIN) is suppressed with the proposed code. Code performance is analyzed in terms of bit error rate (BER) while considering the effect of shot noise, PIIN, and thermal noise. The performance of the proposed code is compared with the related MD, modified quadratic congruence (MQC), two dimensional perfect difference (2D-PD) and two dimensional diluted perfect difference (2D-DPD) codes. The analytical and the simulation results reveal that the proposed 2D-MD code outperforms the other codes. Moreover, a large number of simultaneous users can be accommodated at low BER and high data rate.

Library system for a one dimensional tokamak transport code: (LIBJT60), 1

International Nuclear Information System (INIS)

Hirayama, Toshio

1982-12-01

A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

BERMUDA-1DG: a one-dimensional photon transport code

International Nuclear Information System (INIS)

Suzuki, Tomoo; Hasegawa, Akira; Nakashima, Hiroshi; Kaneko, Kunio.

1984-10-01

A one-dimensional photon transport code BERMUDA-1DG has been developed for spherical and infinite slab geometries. The purpose of development is to equip the function of gamma rays calculation for the BERMUDA code system, which was developed by 1983 only for neutron transport calculation as a preliminary version. A group constants library has been prepared for 30 nuclides, and it now consists of the 36-group total cross sections and secondary gamma ray yields by the 120-group neutron flux. For the Compton scattering, group-angle transfer matrices are accurately obtained by integrating the Klein-Nishina formula taking into account the energy and scattering angle correlation. The pair production cross sections are now calculated in the code from atomic number and midenergy of each group. To obtain angular flux distribution, the transport equation is solved in the same way as in case of neutron, using the direct integration method in a multigroup model. Both of an independent gamma ray source problem and a neutron-gamma source problem are possible to be solved. This report is written as a user's manual with a brief description of the calculational method. (author)

Extensions of the 3-dimensional plasma transport code E3D

International Nuclear Information System (INIS)

Runov, A.; Schneider, R.; Kasilov, S.; Reiter, D.

2004-01-01

One important aspect of modern fusion research is plasma edge physics. Fluid transport codes extending beyond the standard 2-D code packages like B2-Eirene or UEDGE are under development. A 3-dimensional plasma fluid code, E3D, based upon the Multiple Coordinate System Approach and a Monte Carlo integration procedure has been developed for general magnetic configurations including ergodic regions. These local magnetic coordinates lead to a full metric tensor which accurately accounts for all transport terms in the equations. Here, we discuss new computational aspects of the realization of the algorithm. The main limitation to the Monte Carlo code efficiency comes from the restriction on the parallel jump of advancing test particles which must be small compared to the gradient length of the diffusion coefficient. In our problems, the parallel diffusion coefficient depends on both plasma and magnetic field parameters. Usually, the second dependence is much more critical. In order to allow long parallel jumps, this dependence can be eliminated in two steps: first, the longitudinal coordinate x 3 of local magnetic coordinates is modified in such a way that in the new coordinate system the metric determinant and contra-variant components of the magnetic field scale along the magnetic field with powers of the magnetic field module (like in Boozer flux coordinates). Second, specific weights of the test particles are introduced. As a result of increased parallel jump length, the efficiency of the code is about two orders of magnitude better. (copyright 2004 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

THREEDANT: A code to perform three-dimensional, neutral particle transport calculations

International Nuclear Information System (INIS)

Alcouffe, R.E.

1994-01-01

The THREEDANT code solves the three-dimensional neutral particle transport equation in its first order, multigroup, discrate ordinate form. The code allows an unlimited number of groups (depending upon the cross section set), angular quadrature up to S-100, and unlimited Pn order again depending upon the cross section set. The code has three options for spatial differencing, diamond with set-to-zero fixup, adaptive weighted diamond, and linear modal. The geometry options are XYZ and RZΘ with a special XYZ option based upon a volume fraction method. This allows objects or bodies of any shape to be modelled as input which gives the code as much geometric description flexibility as the Monte Carlo code MCNP. The transport equation is solved by source iteration accelerated by the DSA method. Both inner and outer iterations are so accelerated. Some results are presented which demonstrate the effectiveness of these techniques. The code is available on several types of computing platforms

Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

International Nuclear Information System (INIS)

Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

2006-01-01

The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such

The one-dimensional transport codes MAKOKOT. Presentation and directions for use

International Nuclear Information System (INIS)

Capes, H.; Mercier, C.; Morera, J.P.

1986-06-01

In this note are presented the different one-dimensional evolution codes available to date under the generic name MAKOKOT. They are six principal codes: - TRANS: for ion and electron transport; -NEUTRE: for neutrals; -IMPUR: for impurities; -ECRH: for electron cyclotron resonance; -DENT: for sawtooth modelling and analysis; -BILAN: for global verification of conservation. One supplementary code is added which is an impurity evolution code; it takes in account, in 1-D geometry, the buffer zone generated by the limiter between the hot plasma and the wall. An abundant bibliography is given. A comprehensive manner of using is given which underlines the use versatility of these codes [fr

Development of three-dimensional transport code by the double finite element method

International Nuclear Information System (INIS)

Fujimura, Toichiro

1985-01-01

Development of a three-dimensional neutron transport code by the double finite element method is described. Both of the Galerkin and variational methods are adopted to solve the problem, and then the characteristics of them are compared. Computational results of the collocation method, developed as a technique for the vaviational one, are illustrated in comparison with those of an Ssub(n) code. (author)

K-FIX: a computer program for transient, two-dimensional, two-fluid flow. THREED: an extension of the K-FIX code for three-dimensional calculations

International Nuclear Information System (INIS)

Rivard, W.C.; Torrey, M.D.

1978-10-01

The transient, two-dimensional, two-fluid code K-FIX has been extended to perform three-dimensional calculations. This capability is achieved by adding five modification sets of FORTRAN statements to the basic two-dimensional code. The modifications are listed and described, and a complete listing of the three-dimensional code is provided. Results of an example problem are provided for verification

Development of new two-dimensional spectral/spatial code based on dynamic cyclic shift code for OCDMA system

Science.gov (United States)

Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria

2017-07-01

In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.

Approximate solutions of the two-dimensional integral transport equation by collision probability methods

International Nuclear Information System (INIS)

Sanchez, Richard

1977-01-01

A set of approximate solutions for the isotropic two-dimensional neutron transport problem has been developed using the Interface Current formalism. The method has been applied to regular lattices of rectangular cells containing a fuel pin, cladding and water, or homogenized structural material. The cells are divided into zones which are homogeneous. A zone-wise flux expansion is used to formulate a direct collision probability problem within a cell. The coupling of the cells is made by making extra assumptions on the currents entering and leaving the interfaces. Two codes have been written: the first uses a cylindrical cell model and one or three terms for the flux expansion; the second uses a two-dimensional flux representation and does a truly two-dimensional calculation inside each cell. In both codes one or three terms can be used to make a space-independent expansion of the angular fluxes entering and leaving each side of the cell. The accuracies and computing times achieved with the different approximations are illustrated by numerical studies on two benchmark pr

ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes

International Nuclear Information System (INIS)

Halbleib, J.A.; Mehlhorn, T.A.

1985-01-01

The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence

Verification of three dimensional triangular prismatic discrete ordinates transport code ENSEMBLE-TRIZ by comparison with Monte Carlo code GMVP

International Nuclear Information System (INIS)

Homma, Y.; Moriwaki, H.; Ikeda, K.; Ohdi, S.

2013-01-01

This paper deals with the verification of the 3 dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with the multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at the beginning of cycle of an initial core and at the beginning and the end of cycle of an equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multiplication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity. (authors)

Multitasking the three-dimensional transport code TORT on CRAY platforms

International Nuclear Information System (INIS)

Azmy, Y.Y.

1996-01-01

The multitasking options in the three-dimensional neutral particle transport code TORT originally implemented for Cray's CTSS operating system are revived and extended to run on Cray Y/MP and C90 computers using the UNICOS operating system. These include two coarse-grained domain decompositions; across octants, and across directions within an octant, termed Octant Parallel (OP), and Direction Parallel (DP), respectively. Parallel performance of the DP is significantly enhanced by increasing the task grain size and reducing load imbalance via dynamic scheduling of the discrete angles among the participating tasks. Substantial Wall Clock speedup factors, approaching 4.5 using 8 tasks, have been measured in a time-sharing environment, and generally depend on the test problem specifications, number of tasks, and machine loading during execution

Bounds on the Capacity of Weakly constrained two-dimensional Codes

DEFF Research Database (Denmark)

Forchhammer, Søren

2002-01-01

Upper and lower bounds are presented for the capacity of weakly constrained two-dimensional codes. The maximum entropy is calculated for two simple models of 2-D codes constraining the probability of neighboring 1s as an example. For given models of the coded data, upper and lower bounds...... on the capacity for 2-D channel models based on occurrences of neighboring 1s are considered....

Improvement of the efficiency of two-dimensional multigroup transport calculations assuming isotropic reflection with multilevel spatial discretisation

International Nuclear Information System (INIS)

Stankovski, Z.; Zmijarevic, I.

1987-06-01

This paper presents two approximations used in multigroup two-dimensional transport calculations in large, very homogeneous media: isotropic reflection together with recently proposed group-dependent spatial representations. These approximations are implemented as standard options in APOLLO 2 assembly transport code. Presented example calculations show that significant savings in computational costs are obtained while preserving the overall accuracy

Two-dimensional full-core transport theory Benchmarks for the WWER reactors

International Nuclear Information System (INIS)

Petkov, P.T.

2002-01-01

Several two-dimensional full-core real geometry many-group steady-state problems for the WWER-440 and WWER-1000 reactors have been solved by the MARIKO code, based on the method of characteristics. The reference transport theory solutions include assembly-wise and pin-wise power distributions. Homogenized two-group diffusion parameters and discontinuity factors have been calculated by MARIKO for each assembly type both for the whole assembly and for each cell in the smallest sector of symmetry, using the B1 method for calculation of the critical spectrum. Accurate albedo-type boundary conditions have been calculated by MARIKO for the core-reflector and core-absorber boundaries, both for each outer assembly face and for each outer cell face. Comparison with the reference solutions of the two-group nodal diffusion code SPPS-1.6 and the few-group fine-mesh diffusion codes HEX2DA and HEX2DB are presented (Authors)

Development of three-dimensional neoclassical transport simulation code with high performance Fortran on a vector-parallel computer

International Nuclear Information System (INIS)

Satake, Shinsuke; Okamoto, Masao; Nakajima, Noriyoshi; Takamaru, Hisanori

2005-11-01

A neoclassical transport simulation code (FORTEC-3D) applicable to three-dimensional configurations has been developed using High Performance Fortran (HPF). Adoption of computing techniques for parallelization and a hybrid simulation model to the δf Monte-Carlo method transport simulation, including non-local transport effects in three-dimensional configurations, makes it possible to simulate the dynamism of global, non-local transport phenomena with a self-consistent radial electric field within a reasonable computation time. In this paper, development of the transport code using HPF is reported. Optimization techniques in order to achieve both high vectorization and parallelization efficiency, adoption of a parallel random number generator, and also benchmark results, are shown. (author)

Solving the two-dimensional stationary transport equation with the aid of the nodal method

International Nuclear Information System (INIS)

Mesina, M.

1976-07-01

In this document the two-dimensional stationary transport equation for the geometry of a fuel assembly or for a system of square boxes has been formulated as an algebraic eigenvalue problem, and the solution was achieved with the computer code NODE 2 which was developed for this purpose. (orig.) [de

A one-dimensional transport code for the simulation of D-T burning tokamak plasma

International Nuclear Information System (INIS)

Tone, Tatsuzo; Maki, Koichi; Kasai, Masao; Nishida, Hidetsugu

1980-11-01

A one-dimensional transport code for D-T burning tokamak plasma has been developed, which simulates the spatial behavior of fuel ions(D, T), alpha particles, impurities, temperatures of ions and electrons, plasma current, neutrals, heating of alpha and injected beam particles. The basic transport equations are represented by one generalized equation so that the improvement of models and the addition of new equations may be easily made. A model of burn control using a variable toroidal field ripple is employed. This report describes in detail the simulation model, numerical method and the usage of the code. Some typical examples to which the code has been applied are presented. (author)

Development of a neutron transport code many-group two-dimensional heterogeneous calculations by the method of characteristics

International Nuclear Information System (INIS)

Petkov, P.T.

2000-01-01

The method of characteristics (MOC) is gaining increased popularity in the reactor physics community all over the world because it gives a new degree of freedom in nuclear reactor analysis. The MARIKO code solves the neutron transport equation by the MOC in two-dimensional real geometry. The domain of solution can be a rectangle or right hexagon with periodic boundary conditions on the outer boundary. Any reasonable symmetry inside the domain can be fully accounted for. The geometry is described in three levels-macro-cells, cells, and regions. The macro-cells and cells can be any polygon. The outer boundary of a region can be any combination of straight line and circular arc segments. Any level of embedded regions is allowed. Procedures for automatic geometry description of hexagonal fuel assemblies and reflector macro-cells have been developed. The initial ray tracing procedure is performed for the full rectangular or hexagonal domain, but only azimuthal angles in the smallest symmetry interval are tracked. (Authors)

Coupled geochemical and solute transport code development

International Nuclear Information System (INIS)

Morrey, J.R.; Hostetler, C.J.

1985-01-01

A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code

SCORCH - a zero dimensional plasma evolution and transport code for use in small and large tokamak systems

International Nuclear Information System (INIS)

Clancy, B.E.; Cook, J.L.

1984-12-01

The zero-dimensional code SCORCH determines number density and temperature evolution in plasmas using concepts derived from the Hinton and Hazeltine transport theory. The code uses the previously reported ADL-1 data library

DANTSYS: A diffusion accelerated neutral particle transport code system

Energy Technology Data Exchange (ETDEWEB)

Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.

1995-06-01

The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.

DANTSYS: A diffusion accelerated neutral particle transport code system

International Nuclear Information System (INIS)

Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O'Dell, R.D.; Walters, W.F.

1995-06-01

The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZΘ symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing

A two-dimensional, two-phase mass transport model for liquid-feed DMFCs

International Nuclear Information System (INIS)

Yang, W.W.; Zhao, T.S.

2007-01-01

A two-dimensional, isothermal two-phase mass transport model for a liquid-feed direct methanol fuel cell (DMFC) is presented in this paper. The two-phase mass transport in the anode and cathode porous regions is formulated based on the classical multiphase flow in porous media without invoking the assumption of constant gas pressure in the unsaturated porous medium flow theory. The two-phase flow behavior in the anode flow channel is modeled by utilizing the drift-flux model, while in the cathode flow channel the homogeneous mist-flow model is used. In addition, a micro-agglomerate model is developed for the cathode catalyst layer. The model also accounts for the effects of both methanol and water crossover through the membrane. The comprehensive model formed by integrating those in the different regions is solved numerically using a home-written computer code and validated against the experimental data in the literature. The model is then used to investigate the effects of various operating and structural parameters, such as methanol concentration, anode flow rate, porosities of both anode and cathode electrodes, the rate of methanol crossover, and the agglomerate size, on cell performance

Neutronics code VALE for two-dimensional triagonal (hexagonal) and three-dimensional geometries

International Nuclear Information System (INIS)

Vondy, D.R.; Fowler, T.B.

1981-08-01

This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180 0 rotational symmetry, and the rotational symmetry conditions for the 30 0 , 60 0 , and 120 0 triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source

Data exchange between zero dimensional code and physics platform in the CFETR integrated system code

Energy Technology Data Exchange (ETDEWEB)

Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)

2016-11-01

Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.

Some efficient Lagrangian mesh finite elements encoded in ZEPHYR for two dimensional transport calculations

International Nuclear Information System (INIS)

Mordant, Maurice.

1981-04-01

To solve a multigroup stationary neutron transport equation in two-dimensional geometries (X-Y), (R-O) or (R-Z) generally on uses discrete ordinates and rectangular meshes. The way to do it is then well known, well documented and somewhat obvious. If one needs to treat awkward geometries or distorted meshes, things are not so easy and the way to do it is no longer straightforward. We have studied this problem at Limeil Nuclear Center and as an alternative to Monte Carlo methods and code we have implemented in ZEPHYR code at least two efficient finite element solutions for Lagrangian meshes involving any kind of triangles and quadrilaterals

Plasmator. A numerical code for simulation of plasma transport in Tokamaks

International Nuclear Information System (INIS)

Guasp, J.

1979-01-01

Plasmator is a flexible monodimensional numerical code for plasma transport in Tokamaks of circular cross-section, it allows neutral particle transport and impurity effects. The code leaves a total freedom in the analytical form of transport coefficients. It has been writen in Fortran-V for the UNIVAC-1100/80 from JEN and allows for the possibility of graphics for radial profiles and temporal evolution of the main plasma magnitudes, as well in three-dimensional as in two-dimensional representation either on a Calcomp plotter or in the printer. (author)

The three-dimensional, discrete ordinates neutral particle transport code TORT: An overview

International Nuclear Information System (INIS)

Azmy, Y.Y.

1996-01-01

The centerpiece of the Discrete Ordinates Oak Ridge System (DOORS), the three-dimensional neutral particle transport code TORT is reviewed. Its most prominent features pertaining to large applications, such as adjustable problem parameters, memory management, and coarse mesh methods, are described. Advanced, state-of-the-art capabilities including acceleration and multiprocessing are summarized here. Future enhancement of existing graphics and visualization tools is briefly presented

In-facility transport code review

International Nuclear Information System (INIS)

Spore, J.W.; Boyack, B.E.; Bohl, W.R.

1996-07-01

The following computer codes were reviewed by the In-Facility Transport Working Group for application to the in-facility transport of radioactive aerosols, flammable gases, and/or toxic gases: (1) CONTAIN, (2) FIRAC, (3) GASFLOW, (4) KBERT, and (5) MELCOR. Based on the review criteria as described in this report and the versions of each code available at the time of the review, MELCOR is the best code for the analysis of in-facility transport when multidimensional effects are not significant. When multi-dimensional effects are significant, GASFLOW should be used

New approach in two-dimensional fluid modeling of edge plasma transport with high intermittency due to blobs and edge localized modes

International Nuclear Information System (INIS)

Pigarov, A. Yu.; Krasheninnikov, S. I.; Rognlien, T. D.

2011-01-01

A new approach is proposed to simulate intermittent, non-diffusive plasma transport (via blobs and filaments of edge localized modes (ELMs)) observed in the tokamak edge region within the framework of two-dimensional transport codes. This approach combines the inherently three-dimensional filamentary structures associated with an ensemble of blobs into a macro-blob in the two-dimensional poloidal cross-section and advects the macro-blob ballistically across the magnetic field, B. Intermittent transport is represented as a sequence of macro-blobs appropriately seeded in the edge plasma according to experimental statistics. In this case, the code is capable of reproducing both the long-scale temporal evolution of the background plasma and the fast spatiotemporal dynamics of blobs. We report the results from a two-dimensional edge plasma code modeling of a single macro-blob dynamics, and its interaction with initially stationary background plasma as well as with material surfaces. The mechanisms of edge plasma particle and energy losses from macro-blobs are analyzed. The effects of macro-blob sizes and advection velocity on edge plasma profiles are studied. The macro-blob impact on power loading and sputtering rates on the chamber wall and on inner and outer divertor plates is discussed. Temporal evolution of particle inventory of the edge plasma perturbed by macro-blobs is analyzed. Application of macro-blobs to ELM modeling is highlighted.

TUTANK a two-dimensional neutron kinetics code

International Nuclear Information System (INIS)

Watts, M.G.; Halsall, M.J.; Fayers, F.J.

1975-04-01

TUTANK is a two-dimensional neutron kinetics code which treats two neutron energy groups and up to six groups of delayed neutron precursors. A 'theta differencing' method is used to integrate the time dependence of the equations. A position dependent exponential transformation on the time variable is available as an option, which in many circumstances can remove much of the time dependence, and thereby allow longer time steps to be taken. A further manipulation is made to separate the solutions of the neutron fluxes and the precursor concentrations. The spatial equations are based on standard diffusion theory, and their solution is obtained from alternating direction sweeps with a transverse buckling - the so-called ADI-B 2 method. Other features of the code include an elementary temperature feedback and heat removal treatment, automatic time step adjustment, a flexible method of specifying cross-section and heat transfer coefficient variations during a transient, and a restart facility which requires a minimal data specification. Full details of the code input are given. An example of the solution of a NEACRP benchmark for an LWR control rod withdrawal is given. (author)

Application of space-angle synthesis to two-dimensional neutral-particle transport problems of weapon physics

International Nuclear Information System (INIS)

Roberds, R.M.

1975-01-01

A space-angle synthesis (SAS) method has been developed for treating the steady-state, two-dimensional transport of neutrons and gamma rays from a point source of simulated nuclear weapon radiation in air. The method was validated by applying it to the problem of neutron transport from a point source in air over a ground interface, and then comparing the results to those obtained by DOT, a state-of-the-art, discrete-ordinates code. In the SAS method, the energy dependence of the Boltzmann transport equation was treated in the standard multigroup manner. The angular dependence was treated by expanding the flux in specially tailored trial functions and applying the method of weighted residuals which analytically integrated the transport equation over all angles. The weighted-residual approach was analogous to the conventional spherical-harmonics (P/sub N/) method with the exception that the tailored expansion allowed for more rapid convergence than a spherical-harmonics P 1 expansion and resulted in a greater degree of accuracy. The trial functions used in the expansion were odd and even combinations of selected trial solutions, the trial solutions being shaped ellipsoids which approximated the angular distribution of the neutron flux in one-dimensional space. The parameters which described the shape of the ellipsoid varied with energy group and the spatial medium, only, and were obtained from a one-dimensional discrete-ordinates calculation. Thus, approximate transport solutions were made available for all two-dimensional problems of a certain class by using tabulated parameters obtained from a single, one-dimensional calculation

Bounds on the capacity of constrained two-dimensional codes

DEFF Research Database (Denmark)

Forchhammer, Søren; Justesen, Jørn

2000-01-01

Bounds on the capacity of constrained two-dimensional (2-D) codes are presented. The bounds of Calkin and Wilf apply to first-order symmetric constraints. The bounds are generalized in a weaker form to higher order and nonsymmetric constraints. Results are given for constraints specified by run-l...

Two-dimensional disruption thermal analysis code DREAM

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.

1988-08-01

When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)

A DETERMINISTIC METHOD FOR TRANSIENT, THREE-DIMENSIONAL NUETRON TRANSPORT

International Nuclear Information System (INIS)

S. GOLUOGLU, C. BENTLEY, R. DEMEGLIO, M. DUNN, K. NORTON, R. PEVEY I.SUSLOV AND H.L. DODDS

1998-01-01

A deterministic method for solving the time-dependent, three-dimensional Boltzmam transport equation with explicit representation of delayed neutrons has been developed and evaluated. The methodology used in this study for the time variable of the neutron flux is known as the improved quasi-static (IQS) method. The position, energy, and angle-dependent neutron flux is computed deterministically by using the three-dimensional discrete ordinates code TORT. This paper briefly describes the methodology and selected results. The code developed at the University of Tennessee based on this methodology is called TDTORT. TDTORT can be used to model transients involving voided and/or strongly absorbing regions that require transport theory for accuracy. This code can also be used to model either small high-leakage systems, such as space reactors, or asymmetric control rod movements. TDTORT can model step, ramp, step followed by another step, and step followed by ramp type perturbations. It can also model columnwise rod movement can also be modeled. A special case of columnwise rod movement in a three-dimensional model of a boiling water reactor (BWR) with simple adiabatic feedback is also included. TDTORT is verified through several transient one-dimensional, two-dimensional, and three-dimensional benchmark problems. The results show that the transport methodology and corresponding code developed in this work have sufficient accuracy and speed for computing the dynamic behavior of complex multidimensional neutronic systems

One-dimensional transport code for one-group problems in plane geometry

International Nuclear Information System (INIS)

Bareiss, E.H.; Chamot, C.

1970-09-01

Equations and results are given for various methods of solution of the one-dimensional transport equation for one energy group in plane geometry with inelastic scattering and an isotropic source. After considerable investigation, a matrix method of solution was found to be faster and more stable than iteration procedures. A description of the code is included which allows for up to 24 regions, 250 points, and 16 angles such that the product of the number of angles and the number of points is less than 600

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

Evaluation of four-dimensional nonbinary LDPC-coded modulation for next-generation long-haul optical transport networks.

Science.gov (United States)

Zhang, Yequn; Arabaci, Murat; Djordjevic, Ivan B

2012-04-09

Leveraging the advanced coherent optical communication technologies, this paper explores the feasibility of using four-dimensional (4D) nonbinary LDPC-coded modulation (4D-NB-LDPC-CM) schemes for long-haul transmission in future optical transport networks. In contrast to our previous works on 4D-NB-LDPC-CM which considered amplified spontaneous emission (ASE) noise as the dominant impairment, this paper undertakes transmission in a more realistic optical fiber transmission environment, taking into account impairments due to dispersion effects, nonlinear phase noise, Kerr nonlinearities, and stimulated Raman scattering in addition to ASE noise. We first reveal the advantages of using 4D modulation formats in LDPC-coded modulation instead of conventional two-dimensional (2D) modulation formats used with polarization-division multiplexing (PDM). Then we demonstrate that 4D LDPC-coded modulation schemes with nonbinary LDPC component codes significantly outperform not only their conventional PDM-2D counterparts but also the corresponding 4D bit-interleaved LDPC-coded modulation (4D-BI-LDPC-CM) schemes, which employ binary LDPC codes as component codes. We also show that the transmission reach improvement offered by the 4D-NB-LDPC-CM over 4D-BI-LDPC-CM increases as the underlying constellation size and hence the spectral efficiency of transmission increases. Our results suggest that 4D-NB-LDPC-CM can be an excellent candidate for long-haul transmission in next-generation optical networks.

Recent developments in the Los Alamos radiation transport code system

International Nuclear Information System (INIS)

Forster, R.A.; Parsons, K.

1997-01-01

A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results

Method for coupling two-dimensional to three-dimensional discrete ordinates calculations

International Nuclear Information System (INIS)

Thompson, J.L.; Emmett, M.B.; Rhoades, W.A.; Dodds, H.L. Jr.

1985-01-01

A three-dimensional (3-D) discrete ordinates transport code, TORT, has been developed at the Oak Ridge National Laboratory for radiation penetration studies. It is not feasible to solve some 3-D penetration problems with TORT, such as a building located a large distance from a point source, because (a) the discretized 3-D problem is simply too big to fit on the computer or (b) the computing time (and corresponding cost) is prohibitive. Fortunately, such problems can be solved with a hybrid approach by coupling a two-dimensional (2-D) description of the point source, which is assumed to be azimuthally symmetric, to a 3-D description of the building, the region of interest. The purpose of this paper is to describe this hybrid methodology along with its implementation and evaluation in the DOTTOR (Discrete Ordinates to Three-dimensional Oak Ridge Transport) code

Contaminant transport in fracture networks with heterogeneous rock matrices. The Picnic code

International Nuclear Information System (INIS)

Barten, Werner; Robinson, Peter C.

2001-02-01

In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flow paths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling. To make use of these investigations the 'PICNIC project' was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCH MD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales. The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flow path and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely different

Contaminant transport in fracture networks with heterogeneous rock matrices. The Picnic code

Energy Technology Data Exchange (ETDEWEB)

Barten, Werner [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Robinson, Peter C. [QuantiSci Limited, Henley-on-Thames (United Kingdom)

2001-02-01

In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flow paths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling. To make use of these investigations the 'PICNIC project' was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCH MD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales. The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flow path and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely

Three-dimensional tokamak equilibria and stellarators with two-dimensional magnetic symmetry

International Nuclear Information System (INIS)

Garabedian, P.R.

1997-01-01

Three-dimensional computer codes have been developed to simulate equilibrium, stability and transport in tokamaks and stellarators. Bifurcated solutions of the tokamak problem suggest that three-dimensional effects may be more important than has generally been thought. Extensive calculations have led to the discovery of a stellarator configuration with just two field periods and with aspect ratio 3.2 that has a magnetic field spectrum B mn with toroidal symmetry. Numerical studies of equilibrium, stability and transport for this new device, called the Modular Helias-like Heliac 2 (MHH2), will be presented. (author)

Surface harmonics method for two-dimensional time-dependent neutron transport problems of square-lattice nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Boyarinov, V. F.; Kondrushin, A. E.; Fomichenko, P. A. [National Research Centre Kurchatov Institute, Kurchatov Sq. 1, Moscow (Russian Federation)

2013-07-01

Time-dependent equations of the Surface Harmonics Method (SHM) have been derived from the time-dependent neutron transport equation with explicit representation of delayed neutrons for solving the two-dimensional time-dependent problems. These equations have been realized in the SUHAM-TD code. The TWIGL benchmark problem has been used for verification of the SUHAM-TD code. The results of the study showed that computational costs required to achieve necessary accuracy of the solution can be an order of magnitude less than with the use of the conventional finite difference method (FDM). (authors)

Sensitivity analysis explains quasi-one-dimensional current transport in two-dimensional materials

DEFF Research Database (Denmark)

Boll, Mads; Lotz, Mikkel Rønne; Hansen, Ole

2014-01-01

We demonstrate that the quasi-one-dimensional (1D) current transport, experimentally observed in graphene as measured by a collinear four-point probe in two electrode configurations A and B, can be interpreted using the sensitivity functions of the two electrode configurations (configurations...... A and B represents different pairs of electrodes chosen for current sources and potential measurements). The measured sheet resistance in a four-point probe measurement is averaged over an area determined by the sensitivity function. For a two-dimensional conductor, the sensitivity functions for electrode...... configurations A and B are different. But when the current is forced to flow through a percolation network, e.g., graphene with high density of extended defects, the two sensitivity functions become identical. This is equivalent to a four-point measurement on a line resistor, hence quasi-1D transport...

TWO-DIMENSIONAL CORE-COLLAPSE SUPERNOVA MODELS WITH MULTI-DIMENSIONAL TRANSPORT

International Nuclear Information System (INIS)

Dolence, Joshua C.; Burrows, Adam; Zhang, Weiqun

2015-01-01

We present new two-dimensional (2D) axisymmetric neutrino radiation/hydrodynamic models of core-collapse supernova (CCSN) cores. We use the CASTRO code, which incorporates truly multi-dimensional, multi-group, flux-limited diffusion (MGFLD) neutrino transport, including all relevant O(v/c) terms. Our main motivation for carrying out this study is to compare with recent 2D models produced by other groups who have obtained explosions for some progenitor stars and with recent 2D VULCAN results that did not incorporate O(v/c) terms. We follow the evolution of 12, 15, 20, and 25 solar-mass progenitors to approximately 600 ms after bounce and do not obtain an explosion in any of these models. Though the reason for the qualitative disagreement among the groups engaged in CCSN modeling remains unclear, we speculate that the simplifying ''ray-by-ray'' approach employed by all other groups may be compromising their results. We show that ''ray-by-ray'' calculations greatly exaggerate the angular and temporal variations of the neutrino fluxes, which we argue are better captured by our multi-dimensional MGFLD approach. On the other hand, our 2D models also make approximations, making it difficult to draw definitive conclusions concerning the root of the differences between groups. We discuss some of the diagnostics often employed in the analyses of CCSN simulations and highlight the intimate relationship between the various explosion conditions that have been proposed. Finally, we explore the ingredients that may be missing in current calculations that may be important in reproducing the properties of the average CCSNe, should the delayed neutrino-heating mechanism be the correct mechanism of explosion

Two-dimensional core calculation research for fuel management optimization based on CPACT code

International Nuclear Information System (INIS)

Chen Xiaosong; Peng Lianghui; Gang Zhi

2013-01-01

Fuel management optimization process requires rapid assessment for the core layout program, and the commonly used methods include two-dimensional diffusion nodal method, perturbation method, neural network method and etc. A two-dimensional loading patterns evaluation code was developed based on the three-dimensional LWR diffusion calculation program CPACT. Axial buckling introduced to simulate the axial leakage was searched in sub-burnup sections to correct the two-dimensional core diffusion calculation results. Meanwhile, in order to get better accuracy, the weight equivalent volume method of the control rod assembly cross-section was improved. (authors)

Two-dimensional heat conducting simulation of plasma armatures

International Nuclear Information System (INIS)

Huerta, M.A.; Boynton, G.

1991-01-01

This paper reports on our development of a two-dimensional MHD code to simulate internal motions in a railgun plasma armature. The authors use the equations of resistive MHD, with Ohmic heating, and radiation heat transport. The authors use a Flux Corrected Transport code to advance all quantities in time. Our runs show the development of complex flows, subsequent shedding of secondary arcs, and a drop in the acceleration of the armature

Comparison of two-dimensional and three-dimensional MHD equilibrium and stability codes

International Nuclear Information System (INIS)

Herrnegger, F.; Merkel, P.; Johnson, J.L.

1986-02-01

Stability results obtained with the fully three-dimensional magnetohydrodynamic code BETA, the helically invariant code HERA, and the asymptotic stellarator expansion code STEP agree well for a straight l = 2, M = 5 stellarator model. This good agreement between the BETA and STEP codes persists as toroidal curvature is introduced. This validation provides justification for confidence in work with these models. 20 refs., 11 figs

Analysis of the OPERA-15 two-dimensional voiding experiment using the SAS4A code

International Nuclear Information System (INIS)

Briggs, L.L.

1984-01-01

Overall, SAS4A appears to do a good job for simulating the OPERA-15 experiment. For most of the experiment parameters, the code calculations compare quite well with the experimental data. The lack of a multi-dimensional voiding model has the effect of extending the flow coastdown time until voiding starts; otherwise, the code simulates the accident progression satisfactorily. These results indicate a need for further work in this area in the form of a tandem analysis by a two-dimensional flow code and a one-dimensional version of that code to confirm the observations derived from the SAS4A analysis

ADREA-I: A transient three dimensional transport code for atmospheric and other applications - some preliminary results

International Nuclear Information System (INIS)

Bartzis, G.

1985-02-01

In this work a general description of the ADREA-I code is presented and some preliminary results are discussed. The ADREA-I is a transient three dimensional computer code aimed at transport analysis with particular emphasis on atmospheric dispersion under any realistic terrain conditions (complex or not) applicable to the planetary boundary layer in a distance extending up to a hundred kilometers or more. The complex geometry applications and the reasonable results obtained constitute a solid indication of the broad capability of the code. (author)

User's guide for TWOHEX: a code package for two-dimensional, neutral-particle transport in equilateral triangular meshes

International Nuclear Information System (INIS)

Walters, W.F.; Brinkley, F.W.; Marr, D.R.

1984-10-01

TWOHEX solves the two-dimensional multigroup transport equation on an equilateral triangular mesh in the x,y plane. Both regular and adjoint, inhomogeneous (fixed source) and homogeneous problems are solved. Three problem domains are treated by TWOHEX. The whole core domain is a 60 0 parallelogram with vacuum boundary conditions on each face. The third core domain is a 120 0 parallelogram with two vacuum and two rotational boundary conditions. The sixth core domain is a 60 0 parallelogram with two vacuum and two rotational boundary conditions. General anisotropic scattering is allowed and an anisotropic inhomogeneous source may be input as cell averages

GRAVE: An Interactive Geometry Construction and Visualization Software System for the TORT Nuclear Radiation Transport Code

International Nuclear Information System (INIS)

Blakeman, E.D.

2000-01-01

A software system, GRAVE (Geometry Rendering and Visual Editor), has been developed at the Oak Ridge National Laboratory (ORNL) to perform interactive visualization and development of models used as input to the TORT three-dimensional discrete ordinates radiation transport code. Three-dimensional and two-dimensional visualization displays are included. Display capabilities include image rotation, zoom, translation, wire-frame and translucent display, geometry cuts and slices, and display of individual component bodies and material zones. The geometry can be interactively edited and saved in TORT input file format. This system is an advancement over the current, non-interactive, two-dimensional display software. GRAVE is programmed in the Java programming language and can be implemented on a variety of computer platforms. Three- dimensional visualization is enabled through the Visualization Toolkit (VTK), a free-ware C++ software library developed for geometric and data visual display. Future plans include an extension of the system to read inputs using binary zone maps and combinatorial geometry models containing curved surfaces, such as those used for Monte Carlo code inputs. Also GRAVE will be extended to geometry visualization/editing for the DORT two-dimensional transport code and will be integrated into a single GUI-based system for all of the ORNL discrete ordinates transport codes

POST: a postprocessor computer code for producing three-dimensional movies of two-phase flow in a reactor vessel

International Nuclear Information System (INIS)

Taggart, K.A.; Liles, D.R.

1977-08-01

The development of the TRAC computer code for analysis of LOCAs in light-water reactors involves the use of a three-dimensional (r-theta-z), two-fluid hydrodynamics model to describe the two-phase flow of steam and water through the reactor vessel. One of the major problems involved in interpreting results from this code is the presentation of three-dimensional flow patterns. The purpose of the report is to present a partial solution to this data display problem. A first version of a code which produces three-dimensional movies of flow in the reactor vessel has been written and debugged. This code (POST) is used as a postprocessor in conjunction with a stand alone three-dimensional two-phase hydrodynamics code (CYLTF) which is a test bed for the three-dimensional algorithms to be used in TRAC

Analysis of time-of-flight experiment on lithium-oxide assemblies by a two-dimensional transport code DOT3.5

International Nuclear Information System (INIS)

Oyama, Yukio; Yamaguchi, Seiya; Maekawa, Hiroshi

1985-03-01

Calculational analyses were made on the time-of-flight experiment of neutron leakage spectra from lithium-oxide slabs. The uncertainties in the calculation due to modelling were examined and it was estimated to be 1-2 %. The calculational results were compared with the experimental ones. The calculations were carried out by a two-dimensional transport code DOT3.5 using ENDF/B-4 nuclear data file. The comparison of energy-integrated fluxes in C/E from made it clear that the tendency of discrepancy between both results depended on the thickness of assembly and leaking angle. The discrepancy of C/E was about 40 % at the maximum. The effect due to the cross section change to a new data of 7 Li(n,n't) 4 He was also examined. This type of comparison is useful for the systematic assesments. From the comparison, it was suggested that the angular distribution of secondary neutron should be improved in the calculation, and the correct differential data of cross section are required. (author)

Two-dimensional Simulations of Correlation Reflectometry in Fusion Plasmas

International Nuclear Information System (INIS)

Valeo, E.J.; Kramer, G.J.; Nazikian, R.

2001-01-01

A two-dimensional wave propagation code, developed specifically to simulate correlation reflectometry in large-scale fusion plasmas is described. The code makes use of separate computational methods in the vacuum, underdense and reflection regions of the plasma in order to obtain the high computational efficiency necessary for correlation analysis. Simulations of Tokamak Fusion Test Reactor (TFTR) plasma with internal transport barriers are presented and compared with one-dimensional full-wave simulations. It is shown that the two-dimensional simulations are remarkably similar to the results of the one-dimensional full-wave analysis for a wide range of turbulent correlation lengths. Implications for the interpretation of correlation reflectometer measurements in fusion plasma are discussed

Electronic Transport in Two-Dimensional Materials

Science.gov (United States)

Sangwan, Vinod K.; Hersam, Mark C.

2018-04-01

Two-dimensional (2D) materials have captured the attention of the scientific community due to the wide range of unique properties at nanometer-scale thicknesses. While significant exploratory research in 2D materials has been achieved, the understanding of 2D electronic transport and carrier dynamics remains in a nascent stage. Furthermore, because prior review articles have provided general overviews of 2D materials or specifically focused on charge transport in graphene, here we instead highlight charge transport mechanisms in post-graphene 2D materials, with particular emphasis on transition metal dichalcogenides and black phosphorus. For these systems, we delineate the intricacies of electronic transport, including band structure control with thickness and external fields, valley polarization, scattering mechanisms, electrical contacts, and doping. In addition, electronic interactions between 2D materials are considered in the form of van der Waals heterojunctions and composite films. This review concludes with a perspective on the most promising future directions in this fast-evolving field.

Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

Science.gov (United States)

Smith, L. M.; Hochstedler, R. D.

1997-02-01

Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).

Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

International Nuclear Information System (INIS)

Smith, L.M.; Hochstedler, R.D.

1997-01-01

Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)

Two-Dimensional Charge Transport in Disordered Organic Semiconductors

NARCIS (Netherlands)

Brondijk, J. J.; Roelofs, W. S. C.; Mathijssen, S. G. J.; Shehu, A.; Cramer, T.; Biscarini, F.; Blom, P. W. M.; de Leeuw, D. M.

2012-01-01

We analyze the effect of carrier confinement on the charge-transport properties of organic field-effect transistors. Confinement is achieved experimentally by the use of semiconductors of which the active layer is only one molecule thick. The two-dimensional confinement of charge carriers provides

Development of one-dimensional computational fluid dynamics code 'GFLOW' for groundwater flow and contaminant transport analysis

International Nuclear Information System (INIS)

Rahatgaonkar, P. S.; Datta, D.; Malhotra, P. K.; Ghadge, S. G.

2012-01-01

Prediction of groundwater movement and contaminant transport in soil is an important problem in many branches of science and engineering. This includes groundwater hydrology, environmental engineering, soil science, agricultural engineering and also nuclear engineering. Specifically, in nuclear engineering it is applicable in the design of spent fuel storage pools and waste management sites in the nuclear power plants. Ground water modeling involves the simulation of flow and contaminant transport by groundwater flow. In the context of contaminated soil and groundwater system, numerical simulations are typically used to demonstrate compliance with regulatory standard. A one-dimensional Computational Fluid Dynamics code GFLOW had been developed based on the Finite Difference Method for simulating groundwater flow and contaminant transport through saturated and unsaturated soil. The code is validated with the analytical model and the benchmarking cases available in the literature. (authors)

Incorporation of coupled nonequilibrium chemistry into a two-dimensional nozzle code (SEAGULL)

Science.gov (United States)

Ratliff, A. W.

1979-01-01

A two-dimensional multiple shock nozzle code (SEAGULL) was extended to include the effects of finite rate chemistry. The basic code that treats multiple shocks and contact surfaces was fully coupled with a generalized finite rate chemistry and vibrational energy exchange package. The modified code retains all of the original SEAGULL features plus the capability to treat chemical and vibrational nonequilibrium reactions. Any chemical and/or vibrational energy exchange mechanism can be handled as long as thermodynamic data and rate constants are available for all participating species.

Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

International Nuclear Information System (INIS)

Arien, B.; Daniels, J.

1986-12-01

CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

LABAN-PEL: a two-dimensional, multigroup diffusion, high-order response matrix code

International Nuclear Information System (INIS)

Mueller, E.Z.

1991-06-01

The capabilities of LABAN-PEL is described. LABAN-PEL is a modified version of the two-dimensional, high-order response matrix code, LABAN, written by Lindahl. The new version extends the capabilities of the original code with regard to the treatment of neutron migration by including an option to utilize full group-to-group diffusion coefficient matrices. In addition, the code has been converted from single to double precision and the necessary routines added to activate its multigroup capability. The coding has also been converted to standard FORTRAN-77 to enhance the portability of the code. Details regarding the input data requirements and calculational options of LABAN-PEL are provided. 13 refs

Study of bifurcation behavior of two-dimensional turbo product code decoders

International Nuclear Information System (INIS)

He Yejun; Lau, Francis C.M.; Tse, Chi K.

2008-01-01

Turbo codes, low-density parity-check (LDPC) codes and turbo product codes (TPCs) are high performance error-correction codes which employ iterative algorithms for decoding. Under different conditions, the behaviors of the decoders are different. While the nonlinear dynamical behaviors of turbo code decoders and LDPC decoders have been reported in the literature, the dynamical behavior of TPC decoders is relatively unexplored. In this paper, we investigate the behavior of the iterative algorithm of a two-dimensional TPC decoder when the input signal-to-noise ratio (SNR) varies. The quantity to be measured is the mean square value of the posterior probabilities of the information bits. Unlike turbo decoders or LDPC decoders, TPC decoders do not produce a clear 'waterfall region'. This is mainly because the TPC decoding algorithm does not converge to 'indecisive' fixed points even at very low SNR values

Two-dimensional transport of tokamak plasmas

International Nuclear Information System (INIS)

Hirshman, S.P.; Jardin, S.C.

1979-01-01

A reduced set of two-fluid transport equations is obtained from the conservation equations describing the time evolution of the differential particle number, entropy, and magnetic fluxes in an axisymmetric toroidal plasma with nested magnetic surfaces. Expanding in the small ratio of perpendicular to parallel mobilities and thermal conductivities yields as solubility constraints one-dimensional equations for the surface-averaged thermodynamic variables and magnetic fluxes. Since Ohm's law E +u x B =R', where R' accounts for any nonideal effects, only determines the particle flow relative to the diffusing magnetic surfaces, it is necessary to solve a single two-dimensional generalized differential equation, (partial/partialt) delpsi. (delp - J x B) =0, to find the absolute velocity of a magnetic surface enclosing a fixed toroidal flux. This equation is linear but nonstandard in that it involves flux surface averages of the unknown velocity. Specification of R' and the cross-field ion and electron heat fluxes provides a closed system of equations. A time-dependent coordinate transformation is used to describe the diffusion of plasma quantities through magnetic surfaces of changing shape

Mesoscopic current transport in two-dimensional materials with grain boundaries: Four-point probe resistance and Hall effect

DEFF Research Database (Denmark)

Lotz, Mikkel Rønne; Boll, Mads; Østerberg, Frederik Westergaard

2016-01-01

-configurations depends on the dimensionality of the current transport (i.e., one- or two-dimensional). At low grain density or low grain boundary resistivity, two-dimensional transport is observed. In contrast, at moderate grain density and high grain resistivity, one-dimensional transport is seen. Ultimately...

Application of the three-dimensional Oak Ridge transport code

International Nuclear Information System (INIS)

Rhoades, W.A.; Childs, R.L.; Emmett, M.B.; Cramer, S.N.

1984-01-01

TORT, a 3-d extension of the DOT discrete ordinates transport code is now in production use for studies of radiation penetration into large concrete and masonry structures. This paper discusses certain features of the new code and shows representative results, including comparisons with Monte Carlo calculations

CTCN: Colloid transport code -- nuclear

International Nuclear Information System (INIS)

Jain, R.

1993-01-01

This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential-algebraic equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential-algebraic systems

Implementation of the kinetics in the transport code AZTRAN

International Nuclear Information System (INIS)

Duran G, J. A.; Del Valle G, E.; Gomez T, A. M.

2017-09-01

This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S n for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

Study of bifurcation behavior of two-dimensional turbo product code decoders

Energy Technology Data Exchange (ETDEWEB)

He Yejun [Department of Electronic and Information Engineering, Hong Kong Polytechnic University, Hunghom, Hong Kong (China); Lau, Francis C.M. [Department of Electronic and Information Engineering, Hong Kong Polytechnic University, Hunghom, Hong Kong (China)], E-mail: encmlau@polyu.edu.hk; Tse, Chi K. [Department of Electronic and Information Engineering, Hong Kong Polytechnic University, Hunghom, Hong Kong (China)

2008-04-15

Turbo codes, low-density parity-check (LDPC) codes and turbo product codes (TPCs) are high performance error-correction codes which employ iterative algorithms for decoding. Under different conditions, the behaviors of the decoders are different. While the nonlinear dynamical behaviors of turbo code decoders and LDPC decoders have been reported in the literature, the dynamical behavior of TPC decoders is relatively unexplored. In this paper, we investigate the behavior of the iterative algorithm of a two-dimensional TPC decoder when the input signal-to-noise ratio (SNR) varies. The quantity to be measured is the mean square value of the posterior probabilities of the information bits. Unlike turbo decoders or LDPC decoders, TPC decoders do not produce a clear 'waterfall region'. This is mainly because the TPC decoding algorithm does not converge to 'indecisive' fixed points even at very low SNR values.

Parallel processing of two-dimensional Sn transport calculations

International Nuclear Information System (INIS)

Uematsu, M.

1997-01-01

A parallel processing method for the two-dimensional S n transport code DOT3.5 has been developed to achieve a drastic reduction in computation time. In the proposed method, parallelization is achieved with angular domain decomposition and/or space domain decomposition. The calculational speed of parallel processing by angular domain decomposition is largely influenced by frequent communications between processing elements. To assess parallelization efficiency, sample problems with up to 32 x 32 spatial meshes were solved with a Sun workstation using the PVM message-passing library. As a result, parallel calculation using 16 processing elements, for example, was found to be nine times as fast as that with one processing element. As for parallel processing by geometry segmentation, the influence of processing element communications on computation time is small; however, discontinuity at the segment boundary degrades convergence speed. To accelerate the convergence, an alternate sweep of angular flux in conjunction with space domain decomposition and a two-step rescaling method consisting of segmentwise rescaling and ordinary pointwise rescaling have been developed. By applying the developed method, the number of iterations needed to obtain a converged flux solution was reduced by a factor of 2. As a result, parallel calculation using 16 processing elements was found to be 5.98 times as fast as the original DOT3.5 calculation

Benchmark test of drift-kinetic and gyrokinetic codes through neoclassical transport simulations

International Nuclear Information System (INIS)

Satake, S.; Sugama, H.; Watanabe, T.-H.; Idomura, Yasuhiro

2009-09-01

Two simulation codes that solve the drift-kinetic or gyrokinetic equation in toroidal plasmas are benchmarked by comparing the simulation results of neoclassical transport. The two codes are the drift-kinetic δf Monte Carlo code (FORTEC-3D) and the gyrokinetic full- f Vlasov code (GT5D), both of which solve radially-global, five-dimensional kinetic equation with including the linear Fokker-Planck collision operator. In a tokamak configuration, neoclassical radial heat flux and the force balance relation, which relates the parallel mean flow with radial electric field and temperature gradient, are compared between these two codes, and their results are also compared with the local neoclassical transport theory. It is found that the simulation results of the two codes coincide very well in a wide rage of plasma collisionality parameter ν * = 0.01 - 10 and also agree with the theoretical estimations. The time evolution of radial electric field and particle flux, and the radial profile of the geodesic acoustic mode frequency also coincide very well. These facts guarantee the capability of GT5D to simulate plasma turbulence transport with including proper neoclassical effects of collisional diffusion and equilibrium radial electric field. (author)

Preparation of functions of computer code GENGTC and improvement for two-dimensional heat transfer calculations for irradiation capsules

International Nuclear Information System (INIS)

Nomura, Yasushi; Someya, Hiroyuki; Ito, Haruhiko.

1992-11-01

Capsules for irradiation tests in the JMTR (Japan Materials Testing Reactor), consist of irradiation specimens surrounded by a cladding tube, holders, an inner tube and a container tube (from 30mm to 65mm in diameter). And the annular gaps between these structural materials in the capsule are filled with liquids or gases. Cooling of the capsule is done by reactor primary coolant flowing down outside the capsule. Most of the heat generated by fission in fuel specimens and gamma absorption in structural materials is directed radially to the capsule container outer surface. In thermal performance calculations for capsule design, an one(r)-dimensional heat transfer computer code entitled (Generalyzed Gap Temperature Calculation), GENGTC, originally developed in Oak Ridge National Laboratory, U.S.A., has been frequently used. In designing a capsule, are needed many cases of parametric calculations with respect to changes materials and gap sizes. And in some cases, two(r,z)-dimensional heat transfer calculations are needed for irradiation test capsules with short length fuel rods. Recently the authors improved the original one-dimensional code GENGTC, (1) to simplify preparation of input data, (2) to perform automatic calculations for parametric survey based on design temperatures, ect. Moreover, the computer code has been improved to perform r-z two-dimensional heat transfer calculation. This report describes contents of the preparation of the one-dimensional code GENGTC and the improvement for the two-dimensional code GENGTC-2, together with their code manuals. (author)

A numerical method for two-dimensional anisotropic transport problem in cylindrical geometry

International Nuclear Information System (INIS)

Du Mingsheng; Feng Tiekai; Fu Lianxiang; Cao Changshu; Liu Yulan

1988-01-01

The authors deal with the triangular mesh-discontinuous finite element method for solving the time-dependent anisotropic neutron transport problem in two-dimensional cylindrical geometry. A prior estimate of the numerical solution is given. Stability is proved. The authors have computed a two dimensional anisotropic neutron transport problem and a Tungsten-Carbide critical assembly problem by using the numerical method. In comparision with DSN method and the experimental results obtained by others both at home and abroad, the method is satisfactory

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)

Linking the plasma code EDGE2D to the neutral code NIMBUS for a self consistent transport model of the boundary

International Nuclear Information System (INIS)

De Matteis, A.

1987-01-01

This report describes the fully automatic linkage between the finite difference, two-dimensional code EDGE2D, based on the classical Braginskii partial differential equations of ion transport, and the Monte Carlo code NIMBUS, which solves the integral form of the stationary, linear Boltzmann equation for neutral transport in a plasma. The coupling has been performed for the real poloidal geometry of JET with two belt-limiters and real magnetic configurations with or without a single-null point. The new integrated system starts from the magnetic geometry computed by predictive or interpretative equilibrium codes and yields the plasma and neutrals characteristics in the edge

Two-dimensional impurity transport calculations for a high recycling divertor

International Nuclear Information System (INIS)

Brooks, J.N.

1986-04-01

Two dimensional analysis of impurity transport in a high recycling divertor shows asymmetric particle fluxes to the divertor plate, low helium pumping efficiency, and high scrapeoff zone shielding for sputtered impurities

The TORT three-dimensional discrete ordinates neutron/photon transport code (TORT version 3)

Energy Technology Data Exchange (ETDEWEB)

Rhoades, W.A.; Simpson, D.B.

1997-10-01

TORT calculates the flux or fluence of neutrons and/or photons throughout three-dimensional systems due to particles incident upon the system`s external boundaries, due to fixed internal sources, or due to sources generated by interaction with the system materials. The transport process is represented by the Boltzman transport equation. The method of discrete ordinates is used to treat the directional variable, and a multigroup formulation treats the energy dependence. Anisotropic scattering is treated using a Legendre expansion. Various methods are used to treat spatial dependence, including nodal and characteristic procedures that have been especially adapted to resist numerical distortion. A method of body overlay assists in material zone specification, or the specification can be generated by an external code supplied by the user. Several special features are designed to concentrate machine resources where they are most needed. The directional quadrature and Legendre expansion can vary with energy group. A discontinuous mesh capability has been shown to reduce the size of large problems by a factor of roughly three in some cases. The emphasis in this code is a robust, adaptable application of time-tested methods, together with a few well-tested extensions.

Two-dimensional cross-section sensitivity and uncertainty analysis of the LBM [Lithium Blanket Module] experiments at LOTUS

International Nuclear Information System (INIS)

Davidson, J.W.; Dudziak, D.J.; Pelloni, S.; Stepanek, J.

1988-01-01

In a recent common Los Alamos/PSI effort, a sensitivity and nuclear data uncertainty path for the modular code system AARE (Advanced Analysis for Reactor Engineering) was developed. This path includes the cross-section code TRAMIX, the one-dimensional finite difference S/sub N/-transport code ONEDANT, the two-dimensional finite element S/sub N/-transport code TRISM, and the one- and two-dimensional sensitivity and nuclear data uncertainty code SENSIBL. Within the framework of the present work a complete set of forward and adjoint two-dimensional TRISM calculations were performed both for the bare, as well as for the Pb- and Be-preceeded, LBM using MATXS8 libraries. Then a two-dimensional sensitivity and uncertainty analysis for all cases was performed. The goal of this analysis was the determination of the uncertainties of a calculated tritium production per source neutron from lithium along the central Li 2 O rod in the LBM. Considered were the contributions from 1 H, 6 Li, 7 Li, 9 Be, /sup nat/C, 14 N, 16 O, 23 Na, 27 Al, /sup nat/Si, /sup nat/Cr, /sup nat/Fe, /sup nat/Ni, and /sup nat/Pb. 22 refs., 1 fig., 3 tabs

Transport behavior of water molecules through two-dimensional nanopores

International Nuclear Information System (INIS)

Zhu, Chongqin; Li, Hui; Meng, Sheng

2014-01-01

Water transport through a two-dimensional nanoporous membrane has attracted increasing attention in recent years thanks to great demands in water purification and desalination applications. However, few studies have been reported on the microscopic mechanisms of water transport through structured nanopores, especially at the atomistic scale. Here we investigate the microstructure of water flow through two-dimensional model graphene membrane containing a variety of nanopores of different size by using molecular dynamics simulations. Our results clearly indicate that the continuum flow transits to discrete molecular flow patterns with decreasing pore sizes. While for pores with a diameter ≥15 Å water flux exhibits a linear dependence on the pore area, a nonlinear relationship between water flux and pore area has been identified for smaller pores. We attribute this deviation from linear behavior to the presence of discrete water flow, which is strongly influenced by the water-membrane interaction and hydrogen bonding between water molecules

Two-dimensional shielding benchmarks for iron at YAYOI, (1)

International Nuclear Information System (INIS)

Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

TRIDENT: a two-dimensional, multigroup, triangular mesh discrete ordinates, explicit neutron transport code

International Nuclear Information System (INIS)

Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.

1977-03-01

TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided

Two-dimensional full-wave code for reflectometry simulations in TJ-II

International Nuclear Information System (INIS)

Blanco, E.; Heuraux, S.; Estrada, T.; Sanchez, J.; Cupido, L.

2004-01-01

A two-dimensional full-wave code in the extraordinary mode has been developed to simulate reflectometry in TJ-II. The code allows us to study the measurement capabilities of the future correlation reflectometer that is being installed in TJ-II. The code uses the finite-difference-time-domain technique to solve Maxwell's equations in the presence of density fluctuations. Boundary conditions are implemented by a perfectly matched layer to simulate free propagation. To assure the stability of the code, the current equations are solved by a fourth-order Runge-Kutta method. Density fluctuation parameters such as fluctuation level, wave numbers, and correlation lengths are extrapolated from those measured at the plasma edge using Langmuir probes. In addition, realistic plasma shape, density profile, magnetic configuration, and experimental setup of TJ-II are included to determine the plasma regimes in which accurate information may be obtained

TRIPOLI-4: Monte Carlo transport code functionalities and applications; TRIPOLI-4: code de transport Monte Carlo fonctionnalites et applications

Energy Technology Data Exchange (ETDEWEB)

Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), Service d' Etudes de Reacteurs et de Modelisation Avancee, 91 - Gif sur Yvette (France)

2003-07-01

Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)

MOCUM: A two-dimensional method of characteristics code based on constructive solid geometry and unstructured meshing for general geometries

International Nuclear Information System (INIS)

Yang Xue; Satvat, Nader

2012-01-01

Highlight: ► A two-dimensional numerical code based on the method of characteristics is developed. ► The complex arbitrary geometries are represented by constructive solid geometry and decomposed by unstructured meshing. ► Excellent agreement between Monte Carlo and the developed code is observed. ► High efficiency is achieved by parallel computing. - Abstract: A transport theory code MOCUM based on the method of characteristics as the flux solver with an advanced general geometry processor has been developed for two-dimensional rectangular and hexagonal lattice and full core neutronics modeling. In the code, the core structure is represented by the constructive solid geometry that uses regularized Boolean operations to build complex geometries from simple polygons. Arbitrary-precision arithmetic is also used in the process of building geometry objects to eliminate the round-off error from the commonly used double precision numbers. Then, the constructed core frame will be decomposed and refined into a Conforming Delaunay Triangulation to ensure the quality of the meshes. The code is fully parallelized using OpenMP and is verified and validated by various benchmarks representing rectangular, hexagonal, plate type and CANDU reactor geometries. Compared with Monte Carlo and deterministic reference solution, MOCUM results are highly accurate. The mentioned characteristics of the MOCUM make it a perfect tool for high fidelity full core calculation for current and GenIV reactor core designs. The detailed representation of reactor physics parameters can enhance the safety margins with acceptable confidence levels, which lead to more economically optimized designs.

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

Design of sampling tools for Monte Carlo particle transport code JMCT

International Nuclear Information System (INIS)

Shangguan Danhua; Li Gang; Zhang Baoyin; Deng Li

2012-01-01

A class of sampling tools for general Monte Carlo particle transport code JMCT is designed. Two ways are provided to sample from distributions. One is the utilization of special sampling methods for special distribution; the other is the utilization of general sampling methods for arbitrary discrete distribution and one-dimensional continuous distribution on a finite interval. Some open source codes are included in the general sampling method for the maximum convenience of users. The sampling results show sampling correctly from distribution which are popular in particle transport can be achieved with these tools, and the user's convenience can be assured. (authors)

ANTHEM: a two-dimensional multicomponent self-consistent hydro-electron transport code for laser-matter interaction studies

International Nuclear Information System (INIS)

Mason, R.J.

1982-01-01

The ANTHEM code for the study of CO 2 -laser-generated transport is outlined. ANTHEM treats the background plasma as coupled Eulerian thermal and ion fluids, and the suprathermal electrons as either a third fluid or a body of evolving collisional PIC particles. The electrons scatter off the ions; the suprathermals drag against the thermal background. Self-consistent E- and B-fields are computed by the Implicit Moment Method. The current status of the code is described. Typical output from ANTHEM is discussed with special application to Augmented-Return-Current CO 2 -laser-driven targets

Two-dimensional cross-section sensitivity and uncertainty analysis of the LBM experience at LOTUS

International Nuclear Information System (INIS)

Davidson, J.W.; Dudziak, D.J.; Pelloni, S.; Stepanek, J.

1989-01-01

In recent years, the LOTUS fusion blanket facility at IGA-EPF in Lausanne provided a series of irradiation experiments with the Lithium Blanket Module (LBM). The LBM has both realistic fusion blanket and materials and configuration. It is approximately an 80-cm cube, and the breeding material is Li 2 . Using as the D-T neutron source the Haefely Neutron Generator (HNG) with an intensity of about 5·10 12 n/s, a series of experiments with the bare LBM as well as with the LBM preceded by Pb, Be and ThO 2 multipliers were carried out. In a recent common Los Alamos/PSI effort, a sensitivity and nuclear data uncertainty path for the modular code system AARE (Advanced Analysis for Reactor Engineering) was developed. This path includes the cross-section code TRAMIX, the one-dimensional finite difference S n -transport code ONEDANT, the two-dimensional finite element S n -transport code TRISM, and the one- and two-dimensional sensitivity and nuclear data uncertainty code SENSIBL. For the nucleonic transport calculations, three 187-neutron-group libraries are presently available: MATXS8A and MATXS8F based on ENDF/B-V evaluations and MAT187 based on JEF/EFF evaluations. COVFILS-2, a 74-group library of neutron cross-sections, scattering matrices and covariances, is the data source for SENSIBL; the 74-group structure of COVFILS-2 is a subset of the Los Alamos 187-group structure. Within the framework of the present work a complete set of forward and adjoint two-dimensional TRISM calculations were performed both for the bare, as well as for the Pb- and Be-preceded, LBM using MATXS8 libraries. Then a two-dimensional sensitivity and uncertainty analysis for all cases was performed

Colloid transport code-nuclear user's manual

International Nuclear Information System (INIS)

Jain, R.

1992-01-01

This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems

MULTI2D - a computer code for two-dimensional radiation hydrodynamics

Science.gov (United States)

Ramis, R.; Meyer-ter-Vehn, J.; Ramírez, J.

2009-06-01

required. Nature of problem: In inertial confinement fusion and related experiments with lasers and particle beams, energy transport by thermal radiation becomes important. Under these conditions, the radiation field strongly interacts with the hydrodynamic motion through emission and absorption processes. Solution method: The equations of radiation transfer coupled with Lagrangian hydrodynamics, heat diffusion and beam tracing (laser or ions) are solved, in two-dimensional axial-symmetric geometry ( R-Z coordinates) using a fractional step scheme. Radiation transfer is solved with angular resolution. Matter properties are either interpolated from tables (equations-of-state and opacities) or computed by user routines (conductivities and beam attenuation). Restrictions: The code has been designed for typical conditions prevailing in inertial confinement fusion (ns time scale, matter states close to local thermodynamical equilibrium, negligible radiation pressure, …). Although a wider range of situations can be treated, extrapolations to regions beyond this design range need special care. Unusual features: A special computer language, called r94, is used at top levels of the code. These parts have to be converted to standard C by a translation program (supplied as part of the package). Due to the complexity of code (hydro-code, grid generation, user interface, graphic post-processor, translator program, installation scripts) extensive manuals are supplied as part of the package. Running time: 567 seconds for the example supplied.

Proton transport in a membrane protein channel: two-dimensional infrared spectrum modeling.

NARCIS (Netherlands)

Liang, C.; Knoester, J.; Jansen, T.L.Th.A.

2012-01-01

We model the two-dimensional infrared (2DIR) spectrum of a proton channel to investigate its applicability as a spectroscopy tool to study the proton transport process in biological systems. Proton transport processes in proton channels are involved in numerous fundamental biochemical reactions.

Non-standard model for electron heat transport for multidimensional hydrodynamic codes

Energy Technology Data Exchange (ETDEWEB)

Nicolai, Ph.; Busquet, M.; Schurtz, G. [CEA/DAM-Ile de France, 91 - Bruyeres Le Chatel (France)

2000-07-01

In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

Non-standard model for electron heat transport for multidimensional hydrodynamic codes

International Nuclear Information System (INIS)

Nicolai, Ph.; Busquet, M.; Schurtz, G.

2000-01-01

In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

Thermoelectric transport in two-dimensional giant Rashba systems

Science.gov (United States)

Xiao, Cong; Li, Dingping; Ma, Zhongshui; Niu, Qian

Thermoelectric transport in strongly spin-orbit coupled two-dimensional Rashba systems is studied using the analytical solution of the linearized Boltzmann equation. To highlight the effects of inter-band scattering, we assume point-like potential impurities, and obtain the band-and energy-dependent transport relaxation times. Unconventional transport behaviors arise when the Fermi level lies near or below the band crossing point (BCP), such as the non-Drude electrical conducivity below the BCP, the failure of the standard Mott relation linking the Peltier coefficient to the electrical conductivity near the BCP, the enhancement of diffusion thermopower and figure of merit below the BCP, the zero-field Hall coefficient which is not inversely proportional to and not a monotonic function of the carrier density, the enhanced Nernst coefficient below the BCP, and the enhanced current-induced spin-polarization efficiency.

Development of a three-dimensional neutron transport code DFEM based on the double finite element method

International Nuclear Information System (INIS)

Fujimura, Toichiro

1996-01-01

A three-dimensional neutron transport code DFEM has been developed by the double finite element method to analyze reactor cores with complex geometry as large fast reactors. Solution algorithm is based on the double finite element method in which the space and angle finite elements are employed. A reactor core system can be divided into some triangular and/or quadrangular prism elements, and the spatial distribution of neutron flux in each element is approximated with linear basis functions. As for the angular variables, various basis functions are applied, and their characteristics were clarified by comparison. In order to enhance the accuracy, a general method is derived to remedy the truncation errors at reflective boundaries, which are inherent in the conventional FEM. An adaptive acceleration method and the source extrapolation method were applied to accelerate the convergence of the iterations. The code structure is outlined and explanations are given on how to prepare input data. A sample input list is shown for reference. The eigenvalue and flux distribution for real scale fast reactors and the NEA benchmark problems were presented and discussed in comparison with the results of other transport codes. (author)

BALDUR: a one-dimensional plasma transport code

International Nuclear Information System (INIS)

Singer, C.E.; Post, D.E.; Mikkelsen, D.R.

1986-07-01

The purpose of BALDUR is to calculate the evolution of plasma parameters in an MHD equilibrium which can be approximated by concentric circular flux surfaces. Transport of up to six species of ionized particles, of electron and ion energy, and of poloidal magnetic flux is computed. A wide variety of source terms are calculated including those due to neutral gas, fusion, and auxiliary heating. The code is primarily designed for modeling tokamak plasmas but could be adapted to other toroidal confinement systems

Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries

CERN Document Server

Ilic, R D; Stankovic, S J

2002-01-01

This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...

Use of the APOLLO2 transport code for PWR assembly studies

International Nuclear Information System (INIS)

Belhaffaf, D.; Coste, M.; Lenain, R.; Mathonniere, G.; Sanchez, R.; Stankovski, Z.; Zmijarevic, I.

1992-01-01

This paper presents some validation and application aspects of the APOLLO2, user oriented, modular code for multigroup transport assembly calculation which is developed at the French Commissariat a l'Energies Atomique. The main points approached in this paper are: the two dimensional collision probability convergence, critical leakage calculation schemes, self-shielding spatial discretization, and the equivalence procedure

Experimental transport analysis code system in JT-60

International Nuclear Information System (INIS)

Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

1988-03-01

Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

Colloid transport code-nuclear user`s manual

Energy Technology Data Exchange (ETDEWEB)

Jain, R. [New Mexico Univ., Albuquerque, NM (United States)

1992-04-03

This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems.

Fast algorithm for two-dimensional data table use in hydrodynamic and radiative-transfer codes

International Nuclear Information System (INIS)

Slattery, W.L.; Spangenberg, W.H.

1982-01-01

A fast algorithm for finding interpolated atomic data in irregular two-dimensional tables with differing materials is described. The algorithm is tested in a hydrodynamic/radiative transfer code and shown to be of comparable speed to interpolation in regularly spaced tables, which require no table search. The concepts presented are expected to have application in any situation with irregular vector lengths. Also, the procedures that were rejected either because they were too slow or because they involved too much assembly coding are described

An analytical approach for a nodal scheme of two-dimensional neutron transport problems

International Nuclear Information System (INIS)

Barichello, L.B.; Cabrera, L.C.; Prolo Filho, J.F.

2011-01-01

Research highlights: → Nodal equations for a two-dimensional neutron transport problem. → Analytical Discrete Ordinates Method. → Numerical results compared with the literature. - Abstract: In this work, a solution for a two-dimensional neutron transport problem, in cartesian geometry, is proposed, on the basis of nodal schemes. In this context, one-dimensional equations are generated by an integration process of the multidimensional problem. Here, the integration is performed for the whole domain such that no iterative procedure between nodes is needed. The ADO method is used to develop analytical discrete ordinates solution for the one-dimensional integrated equations, such that final solutions are analytical in terms of the spatial variables. The ADO approach along with a level symmetric quadrature scheme, lead to a significant order reduction of the associated eigenvalues problems. Relations between the averaged fluxes and the unknown fluxes at the boundary are introduced as the usually needed, in nodal schemes, auxiliary equations. Numerical results are presented and compared with test problems.

A retrospective and prospective survey of three-dimensional transport calculations

International Nuclear Information System (INIS)

Nakahara, Yasuaki

1985-01-01

A retrospective survey is made on the three-dimensional radiation transport calculations. Introduction is given to computer codes based on the distinctive numerical methods such as the Monte Carlo, Direct Integration, Ssub(n) and Finite Element Methods to solve the three-dimensional transport equations. Prospective discussions are made on pros and cons of these methods. (author)

TRIPOLI-4: Monte Carlo transport code functionalities and applications

International Nuclear Information System (INIS)

Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

2003-01-01

Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)

Two-dimensional QR-coded metamaterial absorber

Science.gov (United States)

Sui, Sai; Ma, Hua; Wang, Jiafu; Pang, Yongqiang; Zhang, Jieqiu; Qu, Shaobo

2016-01-01

In this paper, the design of metamaterial absorbers is proposed based on QR coding and topology optimization. Such absorbers look like QR codes and can be recognized by decoding softwares as well as mobile phones. To verify the design, two lightweight wideband absorbers are designed, which can achieve wideband absorption above 90 % in 6.68-19.30 and 7.00-19.70 GHz, respectively. More importantly, polarization-independent absorption over 90 % can be maintained under incident angle within 55°. The QR code absorber not only can achieve wideband absorption, but also can carry information such as texts and Web sites. They are of important values in applications such identification and electromagnetic protection.

Recent advances in neutral particle transport methods and codes

International Nuclear Information System (INIS)

Azmy, Y.Y.

1996-01-01

An overview of ORNL's three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned

FLAME: A finite element computer code for contaminant transport n variably-saturated media

International Nuclear Information System (INIS)

Baca, R.G.; Magnuson, S.O.

1992-06-01

A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A

FLAME: A finite element computer code for contaminant transport n variably-saturated media

Energy Technology Data Exchange (ETDEWEB)

Baca, R.G.; Magnuson, S.O.

1992-06-01

A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A.

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

FEAST: a two-dimensional non-linear finite element code for calculating stresses

International Nuclear Information System (INIS)

Tayal, M.

1986-06-01

The computer code FEAST calculates stresses, strains, and displacements. The code is two-dimensional. That is, either plane or axisymmetric calculations can be done. The code models elastic, plastic, creep, and thermal strains and stresses. Cracking can also be simulated. The finite element method is used to solve equations describing the following fundamental laws of mechanics: equilibrium; compatibility; constitutive relations; yield criterion; and flow rule. FEAST combines several unique features that permit large time-steps in even severely non-linear situations. The features include a special formulation for permitting many finite elements to simultaneously cross the boundary from elastic to plastic behaviour; accomodation of large drops in yield-strength due to changes in local temperature and a three-step predictor-corrector method for plastic analyses. These features reduce computing costs. Comparisons against twenty analytical solutions and against experimental measurements show that predictions of FEAST are generally accurate to ± 5%

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)

Nonequilibrium Transport and the Bernoulli Effect of Electrons in a Two-Dimensional Electron Gas

Science.gov (United States)

Kaya, Ismet I.

2013-02-01

Nonequilibrium transport of charged carriers in a two-dimensional electron gas is summarized from an experimental point of view. The transport regime in which the electron-electron interactions are enhanced at high bias leads to a range of striking effects in a two-dimensional electron gas. This regime of transport is quite different than the ballistic transport in which particles propagate coherently with no intercarrier energy transfer and the diffusive transport in which the momentum of the electron system is lost with the involvement of the phonons. Quite a few hydrodynamic phenomena observed in classical gasses have the electrical analogs in the current flow. When intercarrier scattering events dominate the transport, the momentum sharing via narrow angle scattering among the hot and cold electrons lead to negative resistance and electron pumping which can be viewed as the analog of the Bernoulli-Venturi effect observed classical gasses. The recent experimental findings and the background work in the field are reviewed.

SALT4: a two-dimensional displacement discontinuity code for thermomechanical analysis in bedded salt deposits

International Nuclear Information System (INIS)

1983-04-01

SALT4 is a two-dimensional analytical/displacement-discontinuity code designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. This code was developed by the University of Minnesota. This documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of the computer code, SALT4. The SALT4 code takes into account: (1) viscoelastic behavior in the pillars adjacent to excavations; (2) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (2) excavation sequence. Major advantages of the SALT4 code are: (1) computational efficiency; (2) the small amount of input data required; and (3) a creep law consistent with laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of SALT4, i.e., temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT4 code can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermal and thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT4 can also be used to verify fully numerical codes. This is similar to the use of analytic solutions for code verification. Although SALT4 was designed for analysis of bedded salt, it is also applicable to crystalline rock if the creep calculation is suppressed. In Section 1.5 of this document the code custodianship and control is described along with the status of verification, validation and peer review of this report

Environmental, Transient, Three-Dimensional, Hydrothermal, Mass Transport Code - FLESCOT

Energy Technology Data Exchange (ETDEWEB)

Onishi, Yasuo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Glass, Kevin A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eyler, L. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Okumura, Masahiko [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-03-28

The purpose of the project was to modify and apply the transient, three-dimensional FLESCOT code to be able to effectively simulate cesium behavior in Fukushima lakes/dam reservoirs, river mouths, and coastal areas. The ultimate objective of the FLESCOT simulation is to predict future changes of cesium accumulation in Fukushima area reservoirs and costal water. These evaluation results will assist ongoing and future environmental remediation activities and policies in a systematic and comprehensive manner.

Multi-dimensional relativistic simulations of core-collapse supernovae with energy-dependent neutrino transport

International Nuclear Information System (INIS)

Mueller, Bernhard

2009-01-01

In this thesis, we have presented the first multi-dimensional models of core-collapse supernovae that combine a detailed, up-to-date treatment of neutrino transport, the equation of state, and - in particular - general relativistic gravity. Building on the well-tested neutrino transport code VERTEX and the GR hydrodynamics code CoCoNuT, we developed and implemented a relativistic generalization of a ray-by-ray-plus method for energy-dependent neutrino transport. The result of these effort, the VERTEX-CoCoNuT code, also incorporates a number of improved numerical techniques that have not been used in the code components VERTEX and CoCoNuT before. In order to validate the VERTEX-CoCoNuT code, we conducted several test simulations in spherical symmetry, most notably a comparison with the one-dimensional relativistic supernova code AGILE-BOLTZTRAN and the Newtonian PROMETHEUSVERTEX code. (orig.)

Multi-dimensional relativistic simulations of core-collapse supernovae with energy-dependent neutrino transport

Energy Technology Data Exchange (ETDEWEB)

Mueller, Bernhard

2009-05-07

In this thesis, we have presented the first multi-dimensional models of core-collapse supernovae that combine a detailed, up-to-date treatment of neutrino transport, the equation of state, and - in particular - general relativistic gravity. Building on the well-tested neutrino transport code VERTEX and the GR hydrodynamics code CoCoNuT, we developed and implemented a relativistic generalization of a ray-by-ray-plus method for energy-dependent neutrino transport. The result of these effort, the VERTEX-CoCoNuT code, also incorporates a number of improved numerical techniques that have not been used in the code components VERTEX and CoCoNuT before. In order to validate the VERTEX-CoCoNuT code, we conducted several test simulations in spherical symmetry, most notably a comparison with the one-dimensional relativistic supernova code AGILE-BOLTZTRAN and the Newtonian PROMETHEUSVERTEX code. (orig.)

Implementation of the kinetics in the transport code AZTRAN; Implementacion de la cinetica en el codigo de transporte AZTRAN

Energy Technology Data Exchange (ETDEWEB)

Duran G, J. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: redfield1290@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2017-09-15

This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S{sub n} for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

Parallelization characteristics of a three-dimensional whole-core code DeCART

International Nuclear Information System (INIS)

Cho, J. Y.; Joo, H.K.; Kim, H. Y.; Lee, J. C.; Jang, M. H.

2003-01-01

Neutron transport calculation for three-dimensional amount of computing time but also huge memory. Therefore, whole-core codes such as DeCART need both also parallel computation and distributed memory capabilities. This paper is to implement such parallel capabilities based on MPI grouping and memory distribution on the DeCART code, and then to evaluate the performance by solving the C5G7 three-dimensional benchmark and a simplified three-dimensional SMART core problem. In C5G7 problem with 24 CPUs, a speedup of maximum 22 is obtained on IBM regatta machine and 21 on a LINUX cluster for the MOC kernel, which indicates good parallel performance of the DeCART code. The simplified SMART problem which need about 11 GBytes memory with one processors requires about 940 MBytes, which means that the DeCART code can now solve large core problems on affordable LINUX clusters

A user's manual for the three-dimensional Monte Carlo transport code SPARTAN

International Nuclear Information System (INIS)

Bending, R.C.; Heffer, P.J.H.

1975-09-01

SPARTAN is a general-purpose Monte Carlo particle transport code intended for neutron or gamma transport problems in reactor physics, health physics, shielding, and safety studies. The code used a very general geometry system enabling a complex layout to be described and allows the user to obtain physics data from a number of different types of source library. Special tracking and scoring techniques are used to improve the quality of the results obtained. To enable users to run SPARTAN, brief descriptions of the facilities available in the code are given and full details of data input and job control language, as well as examples of complete calculations, are included. It is anticipated that changes may be made to SPARTAN from time to time, particularly in those parts of the code which deal with physics data processing. The load module is identified by a version number and implementation date, and updates of sections of this manual will be issued when significant changes are made to the code. (author)

An Implementation of Interfacial Transport Equation into the CUPID code

Energy Technology Data Exchange (ETDEWEB)

Park, Ik Kyu; Cho, Heong Kyu; Yoon, Han Young; Jeong, Jae Jun

2009-11-15

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components for a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted a three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAS, semi-implicit ICE, SIMPLE. The governing equations for a 2-phase flow are composed of mass, momentum, and energy conservation equations for each phase. These equation sets are closed by the interfacial transfer rate of mass, momentum, and energy. The interfacial transfer of mass, momentum, and energy occurs through the interfacial area, and this area plays an important role in the transfer rate. The flow regime based correlations are used for calculating the interracial area in the traditional style 2-phase flow model. This is dependent upon the flow regime and is limited to the fully developed 2-phase flow region. Its application to the multi-dimensional 2-phase flow has some limitation because it adopts the measured results of 2-phase flow in the 1-dimensional tube. The interfacial area concentration transport equation had been suggested in order to calculate the interfacial area without the interfacial area correlations. The source terms to close the interfacial area transport equation should be further developed for a wide ranger usage of it. In this study, the one group interfacial area concentration transport equation has been implemented into the CUPID code. This interfacial area concentration transport equation can be used instead of the interfacial area concentration correlations for the bubbly flow region.

An Implementation of Interfacial Transport Equation into the CUPID code

International Nuclear Information System (INIS)

Park, Ik Kyu; Cho, Heong Kyu; Yoon, Han Young; Jeong, Jae Jun

2009-11-01

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components for a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted a three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAS, semi-implicit ICE, SIMPLE. The governing equations for a 2-phase flow are composed of mass, momentum, and energy conservation equations for each phase. These equation sets are closed by the interfacial transfer rate of mass, momentum, and energy. The interfacial transfer of mass, momentum, and energy occurs through the interfacial area, and this area plays an important role in the transfer rate. The flow regime based correlations are used for calculating the interracial area in the traditional style 2-phase flow model. This is dependent upon the flow regime and is limited to the fully developed 2-phase flow region. Its application to the multi-dimensional 2-phase flow has some limitation because it adopts the measured results of 2-phase flow in the 1-dimensional tube. The interfacial area concentration transport equation had been suggested in order to calculate the interfacial area without the interfacial area correlations. The source terms to close the interfacial area transport equation should be further developed for a wide ranger usage of it. In this study, the one group interfacial area concentration transport equation has been implemented into the CUPID code. This interfacial area concentration transport equation can be used instead of the interfacial area concentration correlations for the bubbly flow region

Effects of non-local electron transport in one-dimensional and two-dimensional simulations of shock-ignited inertial confinement fusion targets

Energy Technology Data Exchange (ETDEWEB)

Marocchino, A.; Atzeni, S.; Schiavi, A. [Dipartimento SBAI, Università di Roma “La Sapienza” and CNISM, Roma 00161 (Italy)

2014-01-15

In some regions of a laser driven inertial fusion target, the electron mean-free path can become comparable to or even longer than the electron temperature gradient scale-length. This can be particularly important in shock-ignited (SI) targets, where the laser-spike heated corona reaches temperatures of several keV. In this case, thermal conduction cannot be described by a simple local conductivity model and a Fick's law. Fluid codes usually employ flux-limited conduction models, which preserve causality, but lose important features of the thermal flow. A more accurate thermal flow modeling requires convolution-like non-local operators. In order to improve the simulation of SI targets, the non-local electron transport operator proposed by Schurtz-Nicolaï-Busquet [G. P. Schurtz et al., Phys. Plasmas 7, 4238 (2000)] has been implemented in the DUED fluid code. Both one-dimensional (1D) and two-dimensional (2D) simulations of SI targets have been performed. 1D simulations of the ablation phase highlight that while the shock profile and timing might be mocked up with a flux-limiter; the electron temperature profiles exhibit a relatively different behavior with no major effects on the final gain. The spike, instead, can only roughly be reproduced with a fixed flux-limiter value. 1D target gain is however unaffected, provided some minor tuning of laser pulses. 2D simulations show that the use of a non-local thermal conduction model does not affect the robustness to mispositioning of targets driven by quasi-uniform laser irradiation. 2D simulations performed with only two final polar intense spikes yield encouraging results and support further studies.

Effects of non-local electron transport in one-dimensional and two-dimensional simulations of shock-ignited inertial confinement fusion targets

International Nuclear Information System (INIS)

Marocchino, A.; Atzeni, S.; Schiavi, A.

2014-01-01

In some regions of a laser driven inertial fusion target, the electron mean-free path can become comparable to or even longer than the electron temperature gradient scale-length. This can be particularly important in shock-ignited (SI) targets, where the laser-spike heated corona reaches temperatures of several keV. In this case, thermal conduction cannot be described by a simple local conductivity model and a Fick's law. Fluid codes usually employ flux-limited conduction models, which preserve causality, but lose important features of the thermal flow. A more accurate thermal flow modeling requires convolution-like non-local operators. In order to improve the simulation of SI targets, the non-local electron transport operator proposed by Schurtz-Nicolaï-Busquet [G. P. Schurtz et al., Phys. Plasmas 7, 4238 (2000)] has been implemented in the DUED fluid code. Both one-dimensional (1D) and two-dimensional (2D) simulations of SI targets have been performed. 1D simulations of the ablation phase highlight that while the shock profile and timing might be mocked up with a flux-limiter; the electron temperature profiles exhibit a relatively different behavior with no major effects on the final gain. The spike, instead, can only roughly be reproduced with a fixed flux-limiter value. 1D target gain is however unaffected, provided some minor tuning of laser pulses. 2D simulations show that the use of a non-local thermal conduction model does not affect the robustness to mispositioning of targets driven by quasi-uniform laser irradiation. 2D simulations performed with only two final polar intense spikes yield encouraging results and support further studies

Effects of non-local electron transport in one-dimensional and two-dimensional simulations of shock-ignited inertial confinement fusion targets

Science.gov (United States)

Marocchino, A.; Atzeni, S.; Schiavi, A.

2014-01-01

In some regions of a laser driven inertial fusion target, the electron mean-free path can become comparable to or even longer than the electron temperature gradient scale-length. This can be particularly important in shock-ignited (SI) targets, where the laser-spike heated corona reaches temperatures of several keV. In this case, thermal conduction cannot be described by a simple local conductivity model and a Fick's law. Fluid codes usually employ flux-limited conduction models, which preserve causality, but lose important features of the thermal flow. A more accurate thermal flow modeling requires convolution-like non-local operators. In order to improve the simulation of SI targets, the non-local electron transport operator proposed by Schurtz-Nicolaï-Busquet [G. P. Schurtz et al., Phys. Plasmas 7, 4238 (2000)] has been implemented in the DUED fluid code. Both one-dimensional (1D) and two-dimensional (2D) simulations of SI targets have been performed. 1D simulations of the ablation phase highlight that while the shock profile and timing might be mocked up with a flux-limiter; the electron temperature profiles exhibit a relatively different behavior with no major effects on the final gain. The spike, instead, can only roughly be reproduced with a fixed flux-limiter value. 1D target gain is however unaffected, provided some minor tuning of laser pulses. 2D simulations show that the use of a non-local thermal conduction model does not affect the robustness to mispositioning of targets driven by quasi-uniform laser irradiation. 2D simulations performed with only two final polar intense spikes yield encouraging results and support further studies.

A two dimensional code (R,Z) for nuclear reactor analysis and its application to the UAR-RI reactor

International Nuclear Information System (INIS)

Bishay, S.T.; Mikhail, I.F.I.; Gaafar, M.A.; Michaiel, M.L.; Nassar, S.F.

1988-01-01

A detailed study is given of fuel consumption in completely reflected cylindrical reactors. A two group, two dimensional (r,z) code is developed to carry out the required procedure. The code is applied to the UAR-RI reactor and the results are found to be in complete agreement with the experimental observations and with the theoretical interpretations. 7 fig., 12 tab

Benchmark testing and independent verification of the VS2DT computer code

International Nuclear Information System (INIS)

McCord, J.T.

1994-11-01

The finite difference flow and transport simulator VS2DT was benchmark tested against several other codes which solve the same equations (Richards equation for flow and the Advection-Dispersion equation for transport). The benchmark problems investigated transient two-dimensional flow in a heterogeneous soil profile with a localized water source at the ground surface. The VS2DT code performed as well as or better than all other codes when considering mass balance characteristics and computational speed. It was also rated highly relative to the other codes with regard to ease-of-use. Following the benchmark study, the code was verified against two analytical solutions, one for two-dimensional flow and one for two-dimensional transport. These independent verifications show reasonable agreement with the analytical solutions, and complement the one-dimensional verification problems published in the code's original documentation

Mesh generation and energy group condensation studies for the jaguar deterministic transport code

International Nuclear Information System (INIS)

Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J.

2012-01-01

The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

Mesh generation and energy group condensation studies for the jaguar deterministic transport code

Energy Technology Data Exchange (ETDEWEB)

Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J. [Knolls Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, P.O. Box 1072, Schenectady, NY 12301-1072 (United States)

2012-07-01

The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

Verification of the network flow and transport/distributed velocity (NWFT/DVM) computer code

International Nuclear Information System (INIS)

Duda, L.E.

1984-05-01

The Network Flow and Transport/Distributed Velocity Method (NWFT/DVM) computer code was developed primarily to fulfill a need for a computationally efficient ground-water flow and contaminant transport capability for use in risk analyses where, quite frequently, large numbers of calculations are required. It is a semi-analytic, quasi-two-dimensional network code that simulates ground-water flow and the transport of dissolved species (radionuclides) in a saturated porous medium. The development of this code was carried out under a program funded by the US Nuclear Regulatory Commission (NRC) to develop a methodology for assessing the risk from disposal of radioactive wastes in deep geologic formations (FIN: A-1192 and A-1266). In support to the methodology development program, the NRC has funded a separate Maintenance of Computer Programs Project (FIN: A-1166) to ensure that the codes developed under A-1192 or A-1266 remain consistent with current operating systems, are as error-free as possible, and have up-to-date documentations for reference by the NRC staff. Part of this effort would include verification and validation tests to assure that a code correctly performs the operations specified and/or is representing the processes or system for which it is intended. This document contains four verification problems for the NWFT/DVM computer code. Two of these problems are analytical verifications of NWFT/DVM where results are compared to analytical solutions. The other two are code-to-code verifications where results from NWFT/DVM are compared to those of another computer code. In all cases NWFT/DVM showed good agreement with both the analytical solutions and the results from the other code

Quantum transport of atomic matter waves in anisotropic two-dimensional and three-dimensional disorder

International Nuclear Information System (INIS)

Piraud, M; Pezzé, L; Sanchez-Palencia, L

2013-01-01

The macroscopic transport properties in a disordered potential, namely diffusion and weak/strong localization, closely depend on the microscopic and statistical properties of the disorder itself. This dependence is rich in counter-intuitive consequences. It can be particularly exploited in matter wave experiments, where the disordered potential can be tailored and controlled, and anisotropies are naturally present. In this work, we apply a perturbative microscopic transport theory and the self-consistent theory of Anderson localization to study the transport properties of ultracold atoms in anisotropic two-dimensional (2D) and three-dimensional (3D) speckle potentials. In particular, we discuss the anisotropy of single-scattering, diffusion and localization. We also calculate disorder-induced shift of the energy states and propose a method to include it, which amounts to renormalizing energies in the standard on-shell approximation. We show that the renormalization of energies strongly affects the prediction for the 3D localization threshold (mobility edge). We illustrate the theoretical findings with examples which are relevant for current matter wave experiments, where the disorder is created with laser speckle. This paper provides a guideline for future experiments aiming at the precise location of the 3D mobility edge and study of anisotropic diffusion and localization effects in 2D and 3D. (paper)

User's manual for the Oak Ridge Tokamak Transport Code

International Nuclear Information System (INIS)

Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.

1977-02-01

A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche

Approximate solutions for the two-dimensional integral transport equation. The critically mixed methods of resolution

International Nuclear Information System (INIS)

Sanchez, Richard.

1980-11-01

This work is divided into two part the first part (note CEA-N-2165) deals with the solution of complex two-dimensional transport problems, the second one treats the critically mixed methods of resolution. These methods are applied for one-dimensional geometries with highly anisotropic scattering. In order to simplify the set of integral equation provided by the integral transport equation, the integro-differential equation is used to obtain relations that allow to lower the number of integral equation to solve; a general mathematical and numerical study is presented [fr

The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark

International Nuclear Information System (INIS)

Kotiluoto, P.

2003-01-01

The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)

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)

SNAP-3D: a three-dimensional neutron diffusion code

International Nuclear Information System (INIS)

McCallien, C.W.J.

1975-10-01

A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)

Biomedical applications of two- and three-dimensional deterministic radiation transport methods

International Nuclear Information System (INIS)

Nigg, D.W.

1992-01-01

Multidimensional deterministic radiation transport methods are routinely used in support of the Boron Neutron Capture Therapy (BNCT) Program at the Idaho National Engineering Laboratory (INEL). Typical applications of two-dimensional discrete-ordinates methods include neutron filter design, as well as phantom dosimetry. The epithermal-neutron filter for BNCT that is currently available at the Brookhaven Medical Research Reactor (BMRR) was designed using such methods. Good agreement between calculated and measured neutron fluxes was observed for this filter. Three-dimensional discrete-ordinates calculations are used routinely for dose-distribution calculations in three-dimensional phantoms placed in the BMRR beam, as well as for treatment planning verification for live canine subjects. Again, good agreement between calculated and measured neutron fluxes and dose levels is obtained

TWO-DIMENSIONAL STELLAR EVOLUTION CODE INCLUDING ARBITRARY MAGNETIC FIELDS. II. PRECISION IMPROVEMENT AND INCLUSION OF TURBULENCE AND ROTATION

International Nuclear Information System (INIS)

Li Linghuai; Sofia, Sabatino; Basu, Sarbani; Demarque, Pierre; Ventura, Paolo; Penza, Valentina; Bi Shaolan

2009-01-01

In the second paper of this series we pursue two objectives. First, in order to make the code more sensitive to small effects, we remove many approximations made in Paper I. Second, we include turbulence and rotation in the two-dimensional framework. The stellar equilibrium is described by means of a set of five differential equations, with the introduction of a new dependent variable, namely the perturbation to the radial gravity, that is found when the nonradial effects are considered in the solution of the Poisson equation. Following the scheme of the first paper, we write the equations in such a way that the two-dimensional effects can be easily disentangled. The key concept introduced in this series is the equipotential surface. We use the underlying cause-effect relation to develop a recurrence relation to calculate the equipotential surface functions for uniform rotation, differential rotation, rotation-like toroidal magnetic fields, and turbulence. We also develop a more precise code to numerically solve the two-dimensional stellar structure and evolution equations based on the equipotential surface calculations. We have shown that with this formulation we can achieve the precision required by observations by appropriately selecting the convergence criterion. Several examples are presented to show that the method works well. Since we are interested in modeling the effects of a dynamo-type field on the detailed envelope structure and global properties of the Sun, the code has been optimized for short timescales phenomena (down to 1 yr). The time dependence of the code has so far been tested exclusively to address such problems.

Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries

Directory of Open Access Journals (Sweden)

Ilić Radovan D.

2002-01-01

Full Text Available This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice.

Upper bound on the capacity of constrained three-dimensional codes

DEFF Research Database (Denmark)

Forchhammer, Søren

2000-01-01

An upper bound on the capacity of constrained three-dimensional codes is presented. The bound for two-dimensional codes of Calkin and Wilf (see SIAM Journal of Discrete Mathematics, vol.11, no.1, p.54-60, 1998) was extended to three dimensions by Nagy and Zeger. Both bounds apply to first order s...

Sub-Transport Layer Coding

DEFF Research Database (Denmark)

Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani

2014-01-01

Packet losses in wireless networks dramatically curbs the performance of TCP. This paper introduces a simple coding shim that aids IP-layer traffic in lossy environments while being transparent to transport layer protocols. The proposed coding approach enables erasure correction while being...... oblivious to the congestion control algorithms of the utilised transport layer protocol. Although our coding shim is indifferent towards the transport layer protocol, we focus on the performance of TCP when ran on top of our proposed coding mechanism due to its widespread use. The coding shim provides gains...

Solution and Study of the Two-Dimensional Nodal Neutron Transport Equation

International Nuclear Information System (INIS)

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

2002-01-01

In the last decade Vilhena and coworkers reported an analytical solution to the two-dimensional nodal discrete-ordinates approximations of the neutron transport equation in a convex domain. The key feature of these works was the application of the combined collocation method of the angular variable and nodal approach in the spatial variables. By nodal approach we mean the transverse integration of the SN equations. This procedure leads to a set of one-dimensional S N equations for the average angular fluxes in the variables x and y. These equations were solved by the old version of the LTS N method, which consists in the application of the Laplace transform to the set of nodal S N equations and solution of the resulting linear system by symbolic computation. It is important to recall that this procedure allow us to increase N the order of S N up to 16. To overcome this drawback we step forward performing a spectral painstaking analysis of the nodal S N equations for N up to 16 and we begin the convergence of the S N nodal equations defining an error for the angular flux and estimating the error in terms of the truncation error of the quadrature approximations of the integral term. Furthermore, we compare numerical results of this approach with those of other techniques used to solve the two-dimensional discrete approximations of the neutron transport equation. (authors)

Integral transport computation of gamma detector response with the CPM2 code

International Nuclear Information System (INIS)

Jones, D.B.

1989-12-01

CPM-2 Version 3 is an enhanced version of the CPM-2 lattice physics computer code which supports the capabilities to (1) perform a two-dimensional gamma flux calculation and (2) perform Restart/Data file maintenance operations. The Gamma Calculation Module implemented in CPM-2 was first developed for EPRI in the CASMO-1 computer code by Studsvik Energiteknik under EPRI Agreement RP2352-01. The gamma transport calculation uses the CPM-HET code module to calculate the transport of gamma rays in two dimensions in a mixed cylindrical-rectangular geometry, where the basic fuel assembly and component regions are maintained in a rectangular geometry, but the fuel pins are represented as cylinders within a square pin cell mesh. Such a capability is needed to represent gamma transport in an essentially transparent medium containing spatially distributed ''black'' cylindrical pins. Under a subcontract to RP2352-01, RPI developed the gamma production and gamma interaction library used for gamma calculation. The CPM-2 gamma calculation was verified against reference results generated by Studsvik using the CASMO-1 program. The CPM-2 Restart/Data file maintenance capabilities provide the user with options to copy files between Restart/Data tapes and to purge files from the Restart/Data tapes

A two-dimensional hydrodynamic code for the interaction of intense heavy ion beams with matter based on the code CONCHAS SPRAY

International Nuclear Information System (INIS)

Schneider, V.; Rentzsch, T.; Maruhn, J.

1988-04-01

In this report we describe a two-dimensional hydrodynamic code applicable to the problems stated. In section II we describe the algorithm solving the hydrodynamic equations. In section III we present test calculations involving the propagation of shocks and contact discontinuities as well as the growth of a Rayleigh-Taylor Instability (RTI). Section IV includes all the modifications and supplements required to use the code to investigate the interaction of intense HI beams with matter. Numcerical simulations of experiments using the RFQ facility and the planned SIS-ESR at GSI are finally discussed in section V. (orig./HSI)

Heavy-ion transport codes for radiotherapy and radioprotection in space

International Nuclear Information System (INIS)

Mancusi, Davide

2006-06-01

Simulation of the transport of heavy ions in matter is a field of nuclear science that has recently received attention in view of its importance for some relevant applications. Accelerated heavy ions can, for example, be used to treat cancers (heavy-ion radiotherapy) and show some superior qualities with respect to more conventional treatment systems, like photons (x-rays) or protons. Furthermore, long-term manned space missions (like a possible future mission to Mars) pose the challenge to protect astronauts and equipment on board against the harmful space radiation environment, where heavy ions can be responsible for a significant share of the exposure risk. The high accuracy expected from a transport algorithm (especially in the case of radiotherapy) and the large amount of semi-empirical knowledge necessary to even state the transport problem properly rule out any analytical approach; the alternative is to resort to numerical simulations in order to build treatment-planning systems for cancer or to aid space engineers in shielding design. This thesis is focused on the description of HIBRAC, a one-dimensional deterministic code optimised for radiotherapy, and PHITS (Particle and Heavy- Ion Transport System), a general-purpose three-dimensional Monte-Carlo code. The structure of both codes is outlined and some relevant results are presented. In the case of PHITS, we also report the first results of an ongoing comprehensive benchmarking program for the main components of the code; we present the comparison of partial charge-changing cross sections for a 400 MeV/n 40 Ar beam impinging on carbon, polyethylene, aluminium, copper, tin and lead targets

Sensitivity analysis using two-dimensional models of the Whiteshell geosphere

Energy Technology Data Exchange (ETDEWEB)

Scheier, N. W.; Chan, T.; Stanchell, F. W.

1992-12-01

As part of the assessment of the environmental impact of disposing of immobilized nuclear fuel waste in a vault deep within plutonic rock, detailed modelling of groundwater flow, heat transport and containment transport through the geosphere is being performed using the MOTIF finite-element computer code. The first geosphere model is being developed using data from the Whiteshell Research Area, with a hypothetical disposal vault at a depth of 500 m. This report briefly describes the conceptual model and then describes in detail the two-dimensional simulations used to help initially define an adequate three-dimensional representation, select a suitable form for the simplified model to be used in the overall systems assessment with the SYVAC computer code, and perform some sensitivity analysis. The sensitivity analysis considers variations in the rock layer properties, variations in fracture zone configurations, the impact of grouting a vault/fracture zone intersection, and variations in boundary conditions. This study shows that the configuration of major fracture zones can have a major influence on groundwater flow patterns. The flows in the major fracture zones can have high velocities and large volumes. The proximity of the radionuclide source to a major fracture zone may strongly influence the time it takes for a radionuclide to be transported to the surface. (auth)

Spin-polarized transport in a two-dimensional electron gas with interdigital-ferromagnetic contacts

DEFF Research Database (Denmark)

Hu, C.-M.; Nitta, Junsaku; Jensen, Ane

2001-01-01

Ferromagnetic contacts on a high-mobility, two-dimensional electron gas (2DEG) in a narrow gap semiconductor with strong spin-orbit interaction are used to investigate spin-polarized electron transport. We demonstrate the use of magnetized contacts to preferentially inject and detect specific spi...

SNAP - a three dimensional neutron diffusion code

International Nuclear Information System (INIS)

McCallien, C.W.J.

1993-02-01

This report describes a one- two- three-dimensional multi-group diffusion code, SNAP, which is primarily intended for neutron diffusion calculations but can also carry out gamma calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP can solve the multi-group neutron diffusion equations using finite difference methods. The one-dimensional slab, cylindrical and spherical geometries and the two-dimensional case are all treated as simple special cases of three-dimensional geometries. Numerous reflective and periodic symmetry options are available and may be used to reduce the number of mesh points necessary to represent the system. Extrapolation lengths can be specified at internal and external boundaries. (Author)

STRUYA a code for two-dimensional fluid flow analysis with and without structure coupling

International Nuclear Information System (INIS)

Katz, F.W.; Schlechtendahl, E.G.; Stoelting, K.

1979-11-01

STRUYA is a code for two-dimensional subsonic and supersonic flow analysis. Both Eulerian and Lagrangian grids are allowed. In the third dimension the flow domain may be bounded by a moving wall. The wall movement may be prescribed in a time-and space varying way or computed by a structural model. STRUYA offers a general scheme for adapting various structural models. As a standard feature it includes a cylindrical shell model (CYLDY2). (orig.) [de

A no-go theorem for a two-dimensional self-correcting quantum memory based on stabilizer codes

International Nuclear Information System (INIS)

Bravyi, Sergey; Terhal, Barbara

2009-01-01

We study properties of stabilizer codes that permit a local description on a regular D-dimensional lattice. Specifically, we assume that the stabilizer group of a code (the gauge group for subsystem codes) can be generated by local Pauli operators such that the support of any generator is bounded by a hypercube of size O(1). Our first result concerns the optimal scaling of the distance d with the linear size of the lattice L. We prove an upper bound d=O(L D-1 ) which is tight for D=1, 2. This bound applies to both subspace and subsystem stabilizer codes. Secondly, we analyze the suitability of stabilizer codes for building a self-correcting quantum memory. Any stabilizer code with geometrically local generators can be naturally transformed to a local Hamiltonian penalizing states that violate the stabilizer condition. A degenerate ground state of this Hamiltonian corresponds to the logical subspace of the code. We prove that for D=1, 2, different logical states can be mapped into each other by a sequence of single-qubit Pauli errors such that the energy of all intermediate states is upper bounded by a constant independent of the lattice size L. The same result holds if there are unused logical qubits that are treated as 'gauge qubits'. It demonstrates that a self-correcting quantum memory cannot be built using stabilizer codes in dimensions D=1, 2. This result is in sharp contrast with the existence of a classical self-correcting memory in the form of a two-dimensional (2D) ferromagnet. Our results leave open the possibility for a self-correcting quantum memory based on 2D subsystem codes or on 3D subspace or subsystem codes.

Comparison of one-, two-, and three-dimensional models for mass transport of radionuclides

International Nuclear Information System (INIS)

Prickett, T.A.; Voorhees, M.L.; Herzog, B.L.

1980-02-01

This technical memorandum compares one-, two-, and three-dimensional models for studying regional mass transport of radionuclides in groundwater associated with deep repository disposal of high-level radioactive wastes. In addition, this report outlines the general conditions for which a one- or two-dimensional model could be used as an alternate to a three-dimensional model analysis. The investigation includes a review of analytical and numerical models in addition to consideration of such conditions as rock and fluid heterogeneity, anisotropy, boundary and initial conditions, and various geometric shapes of repository sources and sinks. Based upon current hydrologic practice, each review is taken separately and discussed to the extent that the researcher can match his problem conditions with the minimum number of model dimensions necessary for an accurate solution

Two dimensional code for modeling of high ione cyclotron harmonic fast wave heating and current drive

International Nuclear Information System (INIS)

Grekov, D.; Kasilov, S.; Kernbichler, W.

2016-01-01

A two dimensional numerical code for computation of the electromagnetic field of a fast magnetosonic wave in a tokamak at high harmonics of the ion cyclotron frequency has been developed. The code computes the finite difference solution of Maxwell equations for separate toroidal harmonics making use of the toroidal symmetry of tokamak plasmas. The proper boundary conditions are prescribed at the realistic tokamak vessel. The currents in the RF antenna are specified externally and then used in Ampere law. The main poloidal tokamak magnetic field and the ''kinetic'' part of the dielectric permeability tensor are treated iteratively. The code has been verified against known analytical solutions and first calculations of current drive in the spherical torus are presented.

System performances of optical space code-division multiple-access-based fiber-optic two-dimensional parallel data link.

Science.gov (United States)

Nakamura, M; Kitayama, K

1998-05-10

Optical space code-division multiple access is a scheme to multiplex and link data between two-dimensional processors such as smart pixels and spatial light modulators or arrays of optical sources like vertical-cavity surface-emitting lasers. We examine the multiplexing characteristics of optical space code-division multiple access by using optical orthogonal signature patterns. The probability density function of interference noise in interfering optical orthogonal signature patterns is calculated. The bit-error rate is derived from the result and plotted as a function of receiver threshold, code length, code weight, and number of users. Furthermore, we propose a prethresholding method to suppress the interference noise, and we experimentally verify that the method works effectively in improving system performance.

Direct-coupled-ray method for design-oriented three-dimensional transport analysis

International Nuclear Information System (INIS)

Bucholz, J.A.; Poncelet, C.G.

1977-01-01

A fast three-dimensional design-oriented transport method has been developed for the solution of both neutron and gamma transport problems. It combines a nodal approach with analytic integral transport to achieve relative speed and accuracy. An analytic solution is obtained for the angular flux in each of the 14 directions defined by the six faces and eight corners of a cubic mesh block. The scheme used to accommodate high-order anisotropic scattering is based on the formulation of ray-to-ray scattering probabilities in an integral sense. A variable mesh approximation has also been introduced to provide greater flexibility. The details of a direct-coupled-ray (DCR) → P 1 conversion technique have been developed but not yet implemented. The DCR method, as implemented in the TRANS3 code, has been used in a number of liquid-metal fast breeder reactor shielding applications. These included a one-dimensional deep penetration configuration and one-, two-, and three dimensional representations of the lower axial shield of the Clinch River Breeder Reactor. Comparisons with ANISN and DOT-III solutions indicated good to excellent agreement in most situations

Heavy-ion transport codes for radiotherapy and radioprotection in space

Energy Technology Data Exchange (ETDEWEB)

Mancusi, Davide

2006-06-15

Simulation of the transport of heavy ions in matter is a field of nuclear science that has recently received attention in view of its importance for some relevant applications. Accelerated heavy ions can, for example, be used to treat cancers (heavy-ion radiotherapy) and show some superior qualities with respect to more conventional treatment systems, like photons (x-rays) or protons. Furthermore, long-term manned space missions (like a possible future mission to Mars) pose the challenge to protect astronauts and equipment on board against the harmful space radiation environment, where heavy ions can be responsible for a significant share of the exposure risk. The high accuracy expected from a transport algorithm (especially in the case of radiotherapy) and the large amount of semi-empirical knowledge necessary to even state the transport problem properly rule out any analytical approach; the alternative is to resort to numerical simulations in order to build treatment-planning systems for cancer or to aid space engineers in shielding design. This thesis is focused on the description of HIBRAC, a one-dimensional deterministic code optimised for radiotherapy, and PHITS (Particle and Heavy- Ion Transport System), a general-purpose three-dimensional Monte-Carlo code. The structure of both codes is outlined and some relevant results are presented. In the case of PHITS, we also report the first results of an ongoing comprehensive benchmarking program for the main components of the code; we present the comparison of partial charge-changing cross sections for a 400 MeV/n {sup 40}Ar beam impinging on carbon, polyethylene, aluminium, copper, tin and lead targets.

Three-dimensional reactor dynamics code for VVER type nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

Three-dimensional reactor dynamics code for VVER type nuclear reactors

International Nuclear Information System (INIS)

Kyrki-Rajamaeki, R.

1995-10-01

A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

Solution of two-dimensional equations of neutron transport in 4P0-approximation of spherical harmonics method

International Nuclear Information System (INIS)

Polivanskij, V.P.

1989-01-01

The method to solve two-dimensional equations of neutron transport using 4P 0 -approximation is presented. Previously such approach was efficiently used for the solution of one-dimensional problems. New an attempt is made to apply the approach to solution of two-dimensional problems. Algorithm of the solution is given, as well as results of test neutron-physical calculations. A considerable as compared with diffusion approximation is shown. 11 refs

New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code

Energy Technology Data Exchange (ETDEWEB)

Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

2016-10-15

The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.

User's manual for the Oak Ridge Tokamak Transport Code

Energy Technology Data Exchange (ETDEWEB)

Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.

1977-02-01

A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche.

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

Jetto a free boundary plasma transport code

International Nuclear Information System (INIS)

Cenacchi, G.; Taroni, A.

1988-01-01

JETTO is a one-and-a-half-dimensional transport code calculating the evolution of plasma parameters in a time dependent axisymmetric MHD equilibrium configuration. A splitting technique gives a consistent solution of coupled equilibrium and transport equations. The plasma boundary is free and defined either by its contact with a limiter (wall) or by a separatrix or by the toroidal magnetic flux. The Grad's approach to the equilibrium problem with adiabatic (or similar) constraints is adopted. This method consists of iterating by alternately solving the Grad-Schluter-Shafranov equation (PDE) and the ODE obtained by averaging the PDE over the magnetic surfaces. The bidimensional equation of the poloidal flux is solved by a finite difference scheme, whereas a Runge-Kutta method is chosen for the averaged equilibrium equation. The 1D transport equations (averaged over the magnetic surfaces) for the electron and ion densities and energies and for the rotational transform are written in terms of a coordinate (ρ) related to the toroidal flux. Impurity transport is also considered, under the hypothesis of coronal equilibrium. The transport equations are solved by an implicit scheme in time and by a finite difference scheme in space. The centering of the source terms and transport coefficients is performed using a Predictor-Corrector scheme. The basic version of the code is described here in detail; input and output parameters are also listed

Modeling A.C. Electronic Transport through a Two-Dimensional Quantum Point Contact

International Nuclear Information System (INIS)

Aronov, I.E.; Beletskii, N.N.; Berman, G.P.; Campbell, D.K.; Doolen, G.D.; Dudiy, S.V.

1998-01-01

We present the results on the a.c. transport of electrons moving through a two-dimensional (2D) semiconductor quantum point contact (QPC). We concentrate our attention on the characteristic properties of the high frequency admittance (ωapproximately0 - 50 GHz), and on the oscillations of the admittance in the vicinity of the separatrix (when a channel opens or closes), in presence of the relaxation effects. The experimental verification of such oscillations in the admittance would be a strong confirmation of the semi-classical approach to the a.c. transport in a QPC, in the separatrix region

The ADO-nodal method for solving two-dimensional discrete ordinates transport problems

International Nuclear Information System (INIS)

Barichello, L.B.; Picoloto, C.B.; Cunha, R.D. da

2017-01-01

Highlights: • Two-dimensional discrete ordinates neutron transport. • Analytical Discrete Ordinates (ADO) nodal method. • Heterogeneous media fixed source problems. • Local solutions. - Abstract: In this work, recent results on the solution of fixed-source two-dimensional transport problems, in Cartesian geometry, are reported. Homogeneous and heterogeneous media problems are considered in order to incorporate the idea of arbitrary number of domain division into regions (nodes) when applying the ADO method, which is a method of analytical features, to those problems. The ADO-nodal formulation is developed, for each node, following previous work devoted to heterogeneous media problem. Here, however, the numerical procedure is extended to higher number of domain divisions. Such extension leads, in some cases, to the use of an iterative method for solving the general linear system which defines the arbitrary constants of the general solution. In addition to solve alternative heterogeneous media configurations than reported in previous works, the present approach allows comparisons with results provided by other metodologies generated with refined meshes. Numerical results indicate the ADO solution may achieve a prescribed accuracy using coarser meshes than other schemes.

Three-dimensional analysis of future groundwater flow conditions and contaminant plume transport in the Hanford Site unconfined aquifer system: FY 1996 and 1997 status report

Energy Technology Data Exchange (ETDEWEB)

Cole, C.R.; Wurstner, S.K.; Williams, M.D.; Thorne, P.D.; Bergeron, M.P.

1997-12-01

A three-dimensional numerical model of groundwater flow and transport, based on the Coupled Fluid Energy, and Solute Transport (CFEST) code, was developed for the Hanford Site to support the Hanford Groundwater Project (HGWP), managed by Pacific Northwest National Laboratory. The model was developed to increase the understanding and better forecast the migration of several contaminant plumes being monitored by the HGWP, and to support the Hanford Site Composite Analysis for low-level waste disposal in the 200-Area Plateau. Recent modeling efforts have focused on continued refinement of an initial version of the three-dimensional model developed in 1995 and its application to simulate future transport of selected contaminant plumes in the aquifer system. This version of the model was updated using a more current version of the CFEST code called CFEST96. Prior to conducting simulations of contaminant transport with the three-dimensional model, a previous steady-state, two-dimensional model of the unconfined aquifer system was recalibrated to 1979 water-table conditions with a statistical inverse method implemented in the CFEST-INV computer code. The results of the recalibration were used to refine the three-dimensional conceptual model and to calibrate it with a conceptualization that preserves the two-dimensional hydraulic properties and knowledge of the aquifer`s three-dimensional properties for the same 1979 water-table conditions. The transient behavior of the three-dimensional flow model was also calibrated by adjusting model storage properties (specific yield) until transient water-table predictions approximated observed water-table elevations between 1979 and 1996.

Three-dimensional analysis of future groundwater flow conditions and contaminant plume transport in the Hanford Site unconfined aquifer system: FY 1996 and 1997 status report

International Nuclear Information System (INIS)

Cole, C.R.; Wurstner, S.K.; Williams, M.D.; Thorne, P.D.; Bergeron, M.P.

1997-12-01

A three-dimensional numerical model of groundwater flow and transport, based on the Coupled Fluid Energy, and Solute Transport (CFEST) code, was developed for the Hanford Site to support the Hanford Groundwater Project (HGWP), managed by Pacific Northwest National Laboratory. The model was developed to increase the understanding and better forecast the migration of several contaminant plumes being monitored by the HGWP, and to support the Hanford Site Composite Analysis for low-level waste disposal in the 200-Area Plateau. Recent modeling efforts have focused on continued refinement of an initial version of the three-dimensional model developed in 1995 and its application to simulate future transport of selected contaminant plumes in the aquifer system. This version of the model was updated using a more current version of the CFEST code called CFEST96. Prior to conducting simulations of contaminant transport with the three-dimensional model, a previous steady-state, two-dimensional model of the unconfined aquifer system was recalibrated to 1979 water-table conditions with a statistical inverse method implemented in the CFEST-INV computer code. The results of the recalibration were used to refine the three-dimensional conceptual model and to calibrate it with a conceptualization that preserves the two-dimensional hydraulic properties and knowledge of the aquifer's three-dimensional properties for the same 1979 water-table conditions. The transient behavior of the three-dimensional flow model was also calibrated by adjusting model storage properties (specific yield) until transient water-table predictions approximated observed water-table elevations between 1979 and 1996

User's manual for ONEDANT: a code package for one-dimensional, diffusion-accelerated, neutral-particle transport

International Nuclear Information System (INIS)

O'Dell, R.D.; Brinkley, F.W. Jr.; Marr, D.R.

1982-02-01

ONEDANT is designed for the CDC-7600, but the program has been implemented and run on the IBM-370/190 and CRAY-I computers. ONEDANT solves the one-dimensional multigroup transport equation in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue search) problems subject to vacuum, reflective, periodic, white, albedo, or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. ONEDANT numerically solves the one-dimensional, multigroup form of the neutral-particle, steady-state form of the Boltzmann transport equation. The discrete-ordinates approximation is used for treating the angular variation of the particle distribution and the diamond-difference scheme is used for phase space discretization. Negative fluxes are eliminated by a local set-to-zero-and-correct algorithm. A standard inner (within-group) iteration, outer (energy-group-dependent source) iteration technique is used. Both inner and outer iterations are accelerated using the diffusion synthetic acceleration method

Two-dimensional charge transport in self-organized, high-mobility conjugated polymers

DEFF Research Database (Denmark)

Sirringhaus, H.; Brown, P.J.; Friend, R.H.

1999-01-01

Self-organization in many solution-processed, semiconducting conjugated polymers results in complex microstructures, in which ordered microcrystalline domains are embedded in an amorphous matrix(I). This has important consequences for electrical properties of these materials: charge transport...... of the ordered microcrystalline domains in the conjugated polymer poly(3-hexylthiophene), P3HT, Self-organization in P3HT results in a lamella structure with two-dimensional conjugated sheets formed by interchain stacking. We find that, depending on processing conditions, the lamellae can adopt two different...... of polymer transistors in logic circuits(5) and active-matrix displays(4,6)....

Numerical modeling of the groundwater contaminant transport for the Lake Karachai Area: The methodological approach and the basic two- dimensional regional model

International Nuclear Information System (INIS)

Petrov, A.V.; Samsonova, L.M.; Vasil'kova, N.A.; Zinin, A.I.; Zinina, G.A.

1994-06-01

Methodological aspects of the numerical modeling of the groundwater contaminant transport for the Lake Karachay area are discussed. Main features of conditions of the task are the high grade of non-uniformity of the aquifer in the fractured rock massif and the high density of the waste solutions, and also the high volume of the input data: both on the part of parameters of the aquifer (number of pump tests) and on the part of observations of functions of processes (long-time observations by the monitoring well grid). The modeling process for constructing the two dimensional regional model is described, and this model is presented as the basic model for subsequent full three-dimensional modeling in sub-areas of interest. Original powerful mathematical apparatus and computer codes for finite-difference numerical modeling are used

Temperature dependent transport of two dimensional electrons in the integral quantum Hall regime

International Nuclear Information System (INIS)

Wi, H.P.

1986-01-01

This thesis is concerned with the temperature dependent electronic transport properties of a two dimensional electron gas subject to background potential fluctuations and a perpendicular magnetic field. The author carried out an extensive temperature dependent study of the transport coefficients, in the region of an integral quantum plateau, in an In/sub x/Ga/sub 1-x/As/InP heterostructure for 4.2K 10 cm -2 meV -1 ) even at the middle between two Landau levels, which is unexpected from model calculations based on short ranged randomness. In addition, the different T dependent behavior of rho/sub xx/ between the states in the tails and those near the center of a Landau level, indicates the existence of different electron states in a Landau level. Additionally, the author reports T-dependent transport measurements in the transition region between two quantum plateaus in several different materials

Spin-orbit coupling, electron transport and pairing instabilities in two-dimensional square structures

Energy Technology Data Exchange (ETDEWEB)

Kocharian, Armen N. [Department of Physics, California State University, Los Angeles, CA 90032 (United States); Fernando, Gayanath W.; Fang, Kun [Department of Physics, University of Connecticut, Storrs, Connecticut 06269 (United States); Palandage, Kalum [Department of Physics, Trinity College, Hartford, Connecticut 06106 (United States); Balatsky, Alexander V. [AlbaNova University Center Nordita, SE-106 91 Stockholm (Sweden)

2016-05-15

Rashba spin-orbit effects and electron correlations in the two-dimensional cylindrical lattices of square geometries are assessed using mesoscopic two-, three- and four-leg ladder structures. Here the electron transport properties are systematically calculated by including the spin-orbit coupling in tight binding and Hubbard models threaded by a magnetic flux. These results highlight important aspects of possible symmetry breaking mechanisms in square ladder geometries driven by the combined effect of a magnetic gauge field spin-orbit interaction and temperature. The observed persistent current, spin and charge polarizations in the presence of spin-orbit coupling are driven by separation of electron and hole charges and opposite spins in real-space. The modeled spin-flip processes on the pairing mechanism induced by the spin-orbit coupling in assembled nanostructures (as arrays of clusters) engineered in various two-dimensional multi-leg structures provide an ideal playground for understanding spatial charge and spin density inhomogeneities leading to electron pairing and spontaneous phase separation instabilities in unconventional superconductors. Such studies also fall under the scope of current challenging problems in superconductivity and magnetism, topological insulators and spin dependent transport associated with numerous interfaces and heterostructures.

Spin-orbit coupling, electron transport and pairing instabilities in two-dimensional square structures

Directory of Open Access Journals (Sweden)

Armen N. Kocharian

2016-05-01

Full Text Available Rashba spin-orbit effects and electron correlations in the two-dimensional cylindrical lattices of square geometries are assessed using mesoscopic two-, three- and four-leg ladder structures. Here the electron transport properties are systematically calculated by including the spin-orbit coupling in tight binding and Hubbard models threaded by a magnetic flux. These results highlight important aspects of possible symmetry breaking mechanisms in square ladder geometries driven by the combined effect of a magnetic gauge field spin-orbit interaction and temperature. The observed persistent current, spin and charge polarizations in the presence of spin-orbit coupling are driven by separation of electron and hole charges and opposite spins in real-space. The modeled spin-flip processes on the pairing mechanism induced by the spin-orbit coupling in assembled nanostructures (as arrays of clusters engineered in various two-dimensional multi-leg structures provide an ideal playground for understanding spatial charge and spin density inhomogeneities leading to electron pairing and spontaneous phase separation instabilities in unconventional superconductors. Such studies also fall under the scope of current challenging problems in superconductivity and magnetism, topological insulators and spin dependent transport associated with numerous interfaces and heterostructures.

Two-dimensional over-all neutronics analysis of the ITER device

Science.gov (United States)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Two-dimensional over-all neutronics analysis of the ITER device

International Nuclear Information System (INIS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi.

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR) and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li 2 O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No.5 and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design. (author)

Gyrokinetic Vlasov code including full three-dimensional geometry of experiments

International Nuclear Information System (INIS)

Nunami, Masanori; Watanabe, Tomohiko; Sugama, Hideo

2010-03-01

A new gyrokinetic Vlasov simulation code, GKV-X, is developed for investigating the turbulent transport in magnetic confinement devices with non-axisymmetric configurations. Effects of the magnetic surface shapes in a three-dimensional equilibrium obtained from the VMEC code are accurately incorporated. Linear simulations of the ion temperature gradient instabilities and the zonal flows in the Large Helical Device (LHD) configuration are carried out by the GKV-X code for a benchmark test against the GKV code. The frequency, the growth rate, and the mode structure of the ion temperature gradient instability are influenced by the VMEC geometrical data such as the metric tensor components of the Boozer coordinates for high poloidal wave numbers, while the difference between the zonal flow responses obtained by the GKV and GKV-X codes is found to be small in the core LHD region. (author)

Numerical studies of unsteady coherent structures and transport in two-dimensional flows

Energy Technology Data Exchange (ETDEWEB)

Hesthaven, J.S.

1995-08-01

The dynamics of unsteady two-dimensional coherent structures in various physical systems is studied through direct numerical solution of the dynamical equations using spectral methods. The relation between the Eulerian and the Lagrangian auto-correlation functions in two-dimensional homogeneous, isotropic turbulence is studied. A simple analytic expression for the Eulerian and Lagrangian auto-correlation function for the fluctuating velocity field is derived solely on the basis of the one-dimensional power spectrum. The long-time evolution of monopolar and dipolar vortices in anisotropic systems relevant for geophysics and plasma physics is studied by direct numerical solution. Transport properties and spatial reorganization of vortical structures are found to depend strongly on the initial conditions. Special attention is given to the dynamics of strong monopoles and the development of unsteady tripolar structures. The development of coherent structures in fluid flows, incompressible as well as compressible, is studied by novel numerical schemes. The emphasis is on the development of spectral methods sufficiently advanced as to allow for detailed and accurate studies of the self-organizing processes. (au) 1 ill., 94 refs.

Transport stochastic multi-dimensional media

International Nuclear Information System (INIS)

Haran, O.; Shvarts, D.

1996-01-01

Many physical phenomena evolve according to known deterministic rules, but in a stochastic media in which the composition changes in space and time. Examples to such phenomena are heat transfer in turbulent atmosphere with non uniform diffraction coefficients, neutron transfer in boiling coolant of a nuclear reactor and radiation transfer through concrete shields. The results of measurements conducted upon such a media are stochastic by nature, and depend on the specific realization of the media. In the last decade there has been a considerable efforts to describe linear particle transport in one dimensional stochastic media composed of several immiscible materials. However, transport in two or three dimensional stochastic media has been rarely addressed. The important effect in multi-dimensional transport that does not appear in one dimension is the ability to bypass obstacles. The current work is an attempt to quantify this effect. (authors)

Transport stochastic multi-dimensional media

Energy Technology Data Exchange (ETDEWEB)

Haran, O; Shvarts, D [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev; Thiberger, R [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

1996-12-01

Many physical phenomena evolve according to known deterministic rules, but in a stochastic media in which the composition changes in space and time. Examples to such phenomena are heat transfer in turbulent atmosphere with non uniform diffraction coefficients, neutron transfer in boiling coolant of a nuclear reactor and radiation transfer through concrete shields. The results of measurements conducted upon such a media are stochastic by nature, and depend on the specific realization of the media. In the last decade there has been a considerable efforts to describe linear particle transport in one dimensional stochastic media composed of several immiscible materials. However, transport in two or three dimensional stochastic media has been rarely addressed. The important effect in multi-dimensional transport that does not appear in one dimension is the ability to bypass obstacles. The current work is an attempt to quantify this effect. (authors).

Computer codes in particle transport physics

International Nuclear Information System (INIS)

Pesic, M.

2004-01-01

Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option

MARG2D code. 1. Eigenvalue problem for two dimensional Newcomb equation

Energy Technology Data Exchange (ETDEWEB)

Tokuda, Shinji [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Watanabe, Tomoko

1997-10-01

A new method and a code MARG2D have been developed to solve the 2-dimensional Newcomb equation which plays an important role in the magnetohydrodynamic (MHD) stability analysis in an axisymmetric toroidal plasma such as a tokamak. In the present formulation, an eigenvalue problem is posed for the 2-D Newcomb equation, where the weight function (the kinetic energy integral) and the boundary conditions at rational surfaces are chosen so that an eigenfunction correctly behaves as the linear combination of the small solution and the analytical solutions around each of the rational surfaces. Thus, the difficulty on solving the 2-D Newcomb equation has been resolved. By using the MARG2D code, the ideal MHD marginally stable state can be identified for a 2-D toroidal plasma. The code is indispensable on computing the outer-region matching data necessary for the resistive MHD stability analysis. Benchmark with ERATOJ, an ideal MHD stability code, has been carried out and the MARG2D code demonstrates that it indeed identifies both stable and marginally stable states against ideal MHD motion. (author)

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B

2003-07-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

Two-dimensional radiation shielding optimization analysis of spent fuel transport container

International Nuclear Information System (INIS)

Tian Yingnan; Chen Yixue; Yang Shouhai

2013-01-01

The intelligent radiation shielding optimization design software platform is a one-dimensional multi-target radiation shielding optimization program which is developed on the basis of the genetic algorithm program and one-dimensional discrete ordinate program-ANISN. This program was applied in the optimization design analysis of the spent fuel transport container radiation shielding. The multi-objective optimization calculation model of the spent fuel transport container radiation shielding was established, and the optimization calculation of the spent fuel transport container weight and radiation dose rate was carried by this program. The calculation results were checked by Monte-Carlo program-MCNP/4C. The results show that the weight of the optimized spent fuel transport container decreases to 81.1% of the origin and the radiation dose rate decreases to below 65.4% of the origin. The maximum deviation between the calculated values from the program and the MCNP is below 5%. The results show that the optimization design scheme is feasible and the calculation result is correct. (authors)

Two-dimensional analysis for a scrapeoff and divertor regions with an MHD model

International Nuclear Information System (INIS)

Ueda, Noriaki; Kasai, Masao; Hitoki, Shigehisa

1987-08-01

With a two-dimensional time dependent fluid code for transport processes in the edge plasma in a tokamak, coupled with Monte-Carlo method for neutral gas behavior, preliminary numerical study has been carried out for the FER divertor. Design base data such as energy flux, particle flux and so on which are essentially important to make an divertor design reliable have been obtained. (author)

Generalized perturbation theory using two-dimensional, discrete ordinates transport theory

International Nuclear Information System (INIS)

Childs, R.L.

1979-01-01

Perturbation theory for changes in linear and bilinear functionals of the forward and adjoint fluxes in a critical reactor has been implemented using two-dimensional discrete ordinates transport theory. The computer program DOT IV was modified to calculate the generalized functions Λ and Λ*. Demonstration calculations were performed for changes in a reaction-rate ratio and a reactivity worth caused by system perturbations. The perturbation theory predictions agreed with direct calculations to within about 2%. A method has been developed for calculating higher lambda eigenvalues and eigenfunctions using techniques similar to those developed for generalized functions. Demonstration calculations have been performed to obtain these eigenfunctions

Effect of beat noise on the performance of two-dimensional time-spreading/wavelength-hopping optical code-division multiple-access systems

Science.gov (United States)

Bazan, T.; Harle, D.; Andonovic, I.; Meenakshi, M.

2005-03-01

The effect of beat noise on optical code-division multiple-access (OCDMA) systems using a range of two-dimensional (2-D) time-spreading/wavelength-hopping (TW) code families is presented. A derivation of a general formula for the error probability of the system is given. The properties of the 2-D codes--namely, the structure, length, and cross-correlation characteristics--are found to have a great influence on system performance. Improved performance can be obtained by use of real-time dynamic thresholding.

Development of 2-D/1-D fusion method for three-dimensional whole-core heterogeneous neutron transport calculations

International Nuclear Information System (INIS)

Lee, Gil Soo

2006-02-01

To describe power distribution and multiplication factor of a reactor core accurately, it is necessary to perform calculations based on neutron transport equation considering heterogeneous geometry and scattering angles. These calculations require very heavy calculations and were nearly impossible with computers of old days. From the limitation of computing power, traditional approach of reactor core design consists of heterogeneous transport calculation in fuel assembly level and whole core diffusion nodal calculation with assembly homogenized properties, resulting from fuel assembly transport calculation. This approach may be effective in computation time, but it gives less accurate results for highly heterogeneous problems. As potential for whole core heterogeneous transport calculation became more feasible owing to rapid development of computing power during last several years, the interests in two and three dimensional whole core heterogeneous transport calculations by deterministic method are increased. For two dimensional calculation, there were several successful approaches using even parity transport equation with triangular meshes, S N method with refined rectangular meshes, the method of characteristics (MOC) with unstructured meshes, and so on. The work in this thesis originally started from the two dimensional whole core heterogeneous transport calculation by using MOC. After successful achievement in two dimensional calculation, there were efforts in three-dimensional whole-core heterogeneous transport calculation using MOC. Since direct extension to three dimensional calculation of MOC requires too much computing power, indirect approach to three dimensional calculation was considered.Thus, 2D/1D fusion method for three dimensional heterogeneous transport calculation was developed and successfully implemented in a computer code. The 2D/1D fusion method is synergistic combination of the MOC for radial 2-D calculation and S N -like methods for axial 1

Metropol: A computer code for the simulation of transport of contaminants with groundwater

International Nuclear Information System (INIS)

Sauter, F.J.; Hassanizadeh, S.M.; Leijnse, A.; Glasbergen, P.; Slot, A.F.M.

1990-01-01

In this report a description is given of the computer code Metropol. This code simulates the three-dimensional flow of groundwater with varying density and the simultaneous transport of contaminants in low concentration and is based on the finite element method. The basic equations for groundwater flow and transport are described as well as the mathematical techniques used to solve these equations. Pre-processing facilities for mesh generation and post-processing facilities such as particle tracking are also discussed. This work was part of the Community Mirage project Second phase, research area Calculation tools

Development and preliminary verification of 2-D transport module of radiation shielding code ARES

International Nuclear Information System (INIS)

Zhang Penghe; Chen Yixue; Zhang Bin; Zang Qiyong; Yuan Longjun; Chen Mengteng

2013-01-01

The 2-D transport module of radiation shielding code ARES is two-dimensional neutron and radiation shielding code. The theory model was based on the first-order steady state neutron transport equation, adopting the discrete ordinates method to disperse direction variables. Then a set of differential equations can be obtained and solved with the source iteration method. The 2-D transport module of ARES was capable of calculating k eff and fixed source problem with isotropic or anisotropic scattering in x-y geometry. The theoretical model was briefly introduced and series of benchmark problems were verified in this paper. Compared with the results given by the benchmark, the maximum relative deviation of k eff is 0.09% and the average relative deviation of flux density is about 0.60% in the BWR cells benchmark problem. As for the fixed source problem with isotropic and anisotropic scattering, the results of the 2-D transport module of ARES conform with DORT very well. These numerical results of benchmark problems preliminarily demonstrate that the development process of the 2-D transport module of ARES is right and it is able to provide high precision result. (authors)

Finite element solution of two dimensional time dependent heat equation

International Nuclear Information System (INIS)

Maaz

1999-01-01

A Microsoft Windows based computer code, named FHEAT, has been developed for solving two dimensional heat problems in Cartesian and Cylindrical geometries. The programming language is Microsoft Visual Basic 3.0. The code makes use of Finite element formulation for spatial domain and Finite difference formulation for time domain. Presently the code is capable of solving two dimensional steady state and transient problems in xy- and rz-geometries. The code is capable excepting both triangular and rectangular elements. Validation and benchmarking was done against hand calculations and published results. (author)

TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticality studies

International Nuclear Information System (INIS)

Ermumcu, G.; Gonnord, J.; Nimal, J.C.

1980-01-01

TRIMARAN is developed for safety analysis of nuclear components containing fissionable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method, in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies

International Nuclear Information System (INIS)

Ermuncu, G.; Gonnord, J.; Nimal, J.C.

1980-04-01

TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

Explicit finite-difference solution of two-dimensional solute transport with periodic flow in homogenous porous media

Directory of Open Access Journals (Sweden)

Djordjevich Alexandar

2017-12-01

Full Text Available The two-dimensional advection-diffusion equation with variable coefficients is solved by the explicit finitedifference method for the transport of solutes through a homogenous two-dimensional domain that is finite and porous. Retardation by adsorption, periodic seepage velocity, and a dispersion coefficient proportional to this velocity are permitted. The transport is from a pulse-type point source (that ceases after a period of activity. Included are the firstorder decay and zero-order production parameters proportional to the seepage velocity, and periodic boundary conditions at the origin and at the end of the domain. Results agree well with analytical solutions that were reported in the literature for special cases. It is shown that the solute concentration profile is influenced strongly by periodic velocity fluctuations. Solutions for a variety of combinations of unsteadiness of the coefficients in the advection-diffusion equation are obtainable as particular cases of the one demonstrated here. This further attests to the effectiveness of the explicit finite difference method for solving two-dimensional advection-diffusion equation with variable coefficients in finite media, which is especially important when arbitrary initial and boundary conditions are required.

An analytical approach for a nodal formulation of a two-dimensional fixed-source neutron transport problem in heterogeneous medium

Energy Technology Data Exchange (ETDEWEB)

Basso Barichello, Liliane; Dias da Cunha, Rudnei [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Inst. de Matematica; Becker Picoloto, Camila [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Tres, Anderson [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Matematica Aplicada

2015-05-15

A nodal formulation of a fixed-source two-dimensional neutron transport problem, in Cartesian geometry, defined in a heterogeneous medium, is solved by an analytical approach. Explicit expressions, in terms of the spatial variables, are derived for averaged fluxes in each region in which the domain is subdivided. The procedure is an extension of an analytical discrete ordinates method, the ADO method, for the solution of the two-dimensional homogeneous medium case. The scheme is developed from the discrete ordinates version of the two-dimensional transport equation along with the level symmetric quadrature scheme. As usual for nodal schemes, relations between the averaged fluxes and the unknown angular fluxes at the contours are introduced as auxiliary equations. Numerical results are in agreement with results available in the literature.

Motivation for Using Generalized Geometry in the Time Dependent Transport Code TDKENO

Energy Technology Data Exchange (ETDEWEB)

Dustin Popp; Zander Mausolff; Sedat Goluoglu

2016-04-01

We are proposing to use the code, TDKENO, to model TREAT. TDKENO solves the time dependent, three dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the neutron flux is factored into two components – a rapidly varying amplitude equation and a slowly varying shape equation and each is solved separately on different time scales. The shape equation is solved using the 3D Monte Carlo transport code KENO, from Oak Ridge National Laboratory’s SCALE code package. Using the Monte Carlo method to solve the shape equation is still computationally intensive, but the operation is only performed when needed. The amplitude equation is solved deterministically and frequently, so the solution gives an accurate time-dependent solution without having to repeatedly We have modified TDKENO to incorporate KENO-VI so that we may accurately represent the geometries within TREAT. This paper explains the motivation behind using generalized geometry, and provides the results of our modifications. TDKENO uses the Improved Quasi-Static method to accomplish this. In this method, the neutron flux is factored into two components. One component is a purely time-dependent and rapidly varying amplitude function, which is solved deterministically and very frequently (small time steps). The other is a slowly varying flux shape function that weakly depends on time and is only solved when needed (significantly larger time steps).

Global vertical mass transport by clouds - A two-dimensional model study

International Nuclear Information System (INIS)

Olofsson, Mats

1988-05-01

A two-dimensional global dispersion model, where vertical transport in the troposphere carried out by convective as well as by frontal cloud systems is explicitly treated, is developed from an existing diffusion model. A parameterization scheme for the cloud transport, based on global cloud statistics, is presented. The model has been tested by using Kr-85, Rn-222 and SO 2 as tracers. Comparisons have been made with observed distributions of these tracers, but also with model results without the cloud transport, using eddy diffusion as the primary means of vertical transport. The model results indicate that for trace species with a turnover time of days to weeks, the introduction of cloud-transport gives much more realistic simulations of their vertical distribution. Layers of increased mixing ratio with height, which can be found in real atmosphere, are reproduced in our cloud-transport model profiles, but can never be simulated with a pure eddy diffusion model. The horizontal transport in the model, by advection and eddy diffusion, gives a realistic distribution between the hemispheres of the more long-lived tracers (Kr-85). A combination of vertical transport by convective and frontal cloud systems is shown to improve the model simulations, compared to limiting it to convective transport only. The importance of including cumulus clouds in the convective transport scheme, in addition to the efficient transport by cumulonimbus clouds, is discussed. The model results are shown to be more sensitive to the vertical detrainment distribution profile than to the absolute magnitude of the vertical mass transport. The scavenging processes for SO 2 are parameterized without the introduction of detailed chemistry. An enhanced removal, due to the increased contact with droplets in the in-cloud lifting process, is introduced in the model. (author)

Influence of Dzyaloshinskii-Moriya interaction and ballistic spin transport in the two and three-dimensional Heisenberg model

Science.gov (United States)

Lima, L. S.

2018-06-01

We study the effect of Dzyaloshisnkii-Moriya interaction on spin transport in the two and three-dimensional Heisenberg antiferromagnetic models in the square lattice and cubic lattice respectively. For the three-dimensional model, we obtain a large peak for the spin conductivity and therefore a finite AC conductivity. For the two-dimensional model, we have gotten the AC spin conductivity tending to the infinity at ω → 0 limit and a suave decreasing in the spin conductivity with increase of ω. We obtain a small influence of the Dzyaloshinskii-Moriya interaction on the spin conductivity in all cases analyzed.

Comparative study of boron transport models in NRC Thermal-Hydraulic Code Trace

Energy Technology Data Exchange (ETDEWEB)

Olmo-Juan, Nicolás; Barrachina, Teresa; Miró, Rafael; Verdú, Gumersindo; Pereira, Claubia, E-mail: nioljua@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es, E-mail: claubia@nuclear.ufmg.br [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM). Universitat Politècnica de València (Spain); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes, as TRACE. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models have been created using the different options and configurations that could have influence in boron transport. These models allow to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately. (author)

Simulation of unsaturated flow and solute transport at the Las Cruces trench site using the PORFLO-3 computer code

International Nuclear Information System (INIS)

Rockhold, M.L.; Wurstner, S.K.

1991-03-01

The objective of this work was to test the ability of the PORFLO-3 computer code to simulate water infiltration and solute transport in dry soils. Data from a field-scale unsaturated zone flow and transport experiment, conducted near Las Cruces, New Mexico, were used for model validation. A spatial moment analysis was used to provide a quantitative basis for comparing the mean simulated and observed flow behavior. The scope of this work was limited to two-dimensional simulations of the second experiment at the Las Cruces trench site. Three simulation cases are presented. The first case represents a uniform soil profile, with homogeneous, isotropic hydraulic and transport properties. The second and third cases represent single stochastic realizations of randomly heterogeneous hydraulic conductivity fields, generated from the cumulative probability distribution of the measured data. Two-dimensional simulations produced water content changes that matched the observed data reasonably well. Models that explicitly incorporated heterogeneous hydraulic conductivity fields reproduced the characteristics of the observed data somewhat better than a uniform, homogeneous model. Improved predictions of water content changes at specific spatial locations were obtained by adjusting the soil hydraulic properties. The results of this study should only be considered a qualitative validation of the PORFLO-3 code. However, the results of this study demonstrate the importance of site-specific data for model calibration. Applications of the code for waste management and remediation activities will require site-specific data for model calibration before defensible predictions of unsaturated flow and containment transport can be made. 23 refs., 16 figs., 3 tabs

Two-dimensional beam profiles and one-dimensional projections

Science.gov (United States)

Findlay, D. J. S.; Jones, B.; Adams, D. J.

2018-05-01

One-dimensional projections of improved two-dimensional representations of transverse profiles of particle beams are proposed for fitting to data from harp-type monitors measuring beam profiles on particle accelerators. Composite distributions, with tails smoothly matched on to a central (inverted) parabola, are shown to give noticeably better fits than single gaussian and single parabolic distributions to data from harp-type beam profile monitors all along the proton beam transport lines to the two target stations on the ISIS Spallation Neutron Source. Some implications for inferring beam current densities on the beam axis are noted.

BALTORO a general purpose code for coupling discrete ordinates and Monte-Carlo radiation transport calculations

International Nuclear Information System (INIS)

Zazula, J.M.

1983-01-01

The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)

Modeling a TRIGA Mark II reactor using the Attila three-dimensional deterministic transport code

International Nuclear Information System (INIS)

Keller, S.T.; Palmer, T.S.; Wareing, T.A.

2005-01-01

A benchmark model of a TRIGA reactor constructed using materials and dimensions similar to existing TRIGA reactors was analyzed using MCNP and the recently developed deterministic transport code Attila TM . The benchmark reactor requires no MCNP modeling approximations, yet is sufficiently complex to validate the new modeling techniques. Geometric properties of the benchmark reactor are specified for use by Attila TM with CAD software. Materials are treated individually in MCNP. Materials used in Attila TM that are clad are homogenized. Attila TM uses multigroup energy discretization. Two cross section libraries were constructed for comparison. A 16 group library collapsed from the SCALE 4.4.a 238 group library provided better results than a seven group library calculated with WIMS-ANL. Values of the k-effective eigenvalue and scalar flux as a function of location and energy were calculated by the two codes. The calculated values for k-effective and spatially averaged neutron flux were found to be in good agreement. Flux distribution by space and energy also agreed well. Attila TM results could be improved with increased spatial and angular resolution and revised energy group structure. (authors)

Two-phase flow characteristics analysis code: MINCS

International Nuclear Information System (INIS)

Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.

1992-03-01

Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)

Hydrogen recycle modeling in transport codes

International Nuclear Information System (INIS)

Howe, H.C.

1979-01-01

The hydrogen recycling models now used in Tokamak transport codes are reviewed and the method by which realistic recycling models are being added is discussed. Present models use arbitrary recycle coefficients and therefore do not model the actual recycling processes at the wall. A model for the hydrogen concentration in the wall serves two purposes: (1) it allows a better understanding of the density behavior in present gas puff, pellet, and neutral beam heating experiments; and (2) it allows one to extrapolate to long pulse devices such as EBT, ISX-C and reactors where the walls are observed or expected to saturate. Several wall models are presently being studied for inclusion in transport codes

APPLE-3: improvement of APPLE for neutron and gamma-ray flux, spectrum and reaction rate plotting code, and of its code manual

International Nuclear Information System (INIS)

Kawasaki, Hiromitu; Maki, Koichi; Seki, Yasushi.

1991-03-01

A code APPLE was produced in 1976 for calculating and plotting tritium breeding ratio and tritium production rate distributions. That code was improved as 'APPLE-2' in 1982, to calculate and plot not only tritium breeding ratio but also distributions of neutron and gamma-ray fluxes, their spectra, nuclear heating rates and other reaction rates, and dose rate distributions during operation and after shutdown in 1982. The code APPLE-2 can calculate and plot these nuclear properties derived from neutron and gamma-ray fluxes by ANISN (one dimensional transport code), DOT3.5 (two dimensional transport code) and MORSE (three dimensional Monte Carlo code). We revised the code APPLE-2 as 'APPLE-3' by adding many functions to the APPLE-2 code in accordance with users' requirements proposed in recent progress of fusion reaction nuclear design. With minor modification of APPLE-2, a number of inconsistencies have been found between the code manual and the input data in the code. In the present report, the new functions added to APPLE-2 and improved users' manual are explained. (author)

C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

Energy Technology Data Exchange (ETDEWEB)

Yesilyurt, Gokhan [ORNL; Clarno, Kevin T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL; Fox, Patricia B [ORNL

2011-01-01

The C5 benchmark problem proposed by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency was modeled to examine the capabilities of Denovo, a three-dimensional (3-D) parallel discrete ordinates (S{sub N}) radiation transport code, for problems with no spatial homogenization. Denovo uses state-of-the-art numerical methods to obtain accurate solutions to the Boltzmann transport equation. Problems were run in parallel on Jaguar, a high-performance supercomputer located at Oak Ridge National Laboratory. Both the two-dimensional (2-D) and 3-D configurations were analyzed, and the results were compared with the reference MCNP Monte Carlo calculations. For an additional comparison, SCALE/KENO-V.a Monte Carlo solutions were also included. In addition, a sensitivity analysis was performed for the optimal angular quadrature and mesh resolution for both the 2-D and 3-D infinite lattices of UO{sub 2} fuel pin cells. Denovo was verified with the C5 problem. The effective multiplication factors, pin powers, and assembly powers were found to be in good agreement with the reference MCNP and SCALE/KENO-V.a Monte Carlo calculations.

Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

International Nuclear Information System (INIS)

Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.; Soldevila, M.

2003-01-01

In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented

Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

Energy Technology Data Exchange (ETDEWEB)

Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B; Soldevila, M [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S/SERMA/LEPP), 91 - Gif sur Yvette (France)

2003-07-01

In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented.

Contaminant transport at a waste residue deposit

DEFF Research Database (Denmark)

Engesgaard, Peter Knudegaard; Traberg, Rikke

1996-01-01

Contaminant transport in an aquifer at an incinerator waste residue deposit in Denmark is simulated. A two-dimensional, geochemical transport code is developed for this purpose and tested by comparison to results from another code, The code is applied to a column experiment and to the field site...

RTk/SN Solutions of the Two-Dimensional Multigroup Transport Equations in Hexagonal Geometry

International Nuclear Information System (INIS)

Valle, Edmundo del; Mund, Ernest H.

2004-01-01

This paper describes an extension to the hexagonal geometry of some weakly discontinuous nodal finite element schemes developed by Hennart and del Valle for the two-dimensional discrete ordinates transport equation in quadrangular geometry. The extension is carried out in a way similar to the extension to the hexagonal geometry of nodal element schemes for the diffusion equation using a composite mapping technique suggested by Hennart, Mund, and del Valle. The combination of the weakly discontinuous nodal transport scheme and the composite mapping is new and is detailed in the main section of the paper. The algorithm efficiency is shown numerically through some benchmark calculations on classical problems widely referred to in the literature

Vectorized Matlab Codes for Linear Two-Dimensional Elasticity

Directory of Open Access Journals (Sweden)

Jonas Koko

2007-01-01

Full Text Available A vectorized Matlab implementation for the linear finite element is provided for the two-dimensional linear elasticity with mixed boundary conditions. Vectorization means that there is no loop over triangles. Numerical experiments show that our implementation is more efficient than the standard implementation with a loop over all triangles.

ANISN-L, a CDC-7600 code which solves the one-dimensional, multigroup, time dependent transport equation by the method of discrete ordinates

Energy Technology Data Exchange (ETDEWEB)

Wilcox, T. P.

1973-09-20

The code ANISN-L solves the one-dimensional, multigroup, time-independent Boltzmann transport equation by the method of discrete ordinates. In problems involving a fissionable system, it can calculate the system multiplication or alpha. In such cases, it is also capable of determining isotopic concentrations, radii, zone widths, or buckling in order to achieve a given multiplication or alpha. The code may also calculate fluxes caused by a specified fixed source. Neutron, gamma, and coupled neutron--gamma problems may be solved in either the forward or adjoint (backward) modes. Cross sections describing upscatter, as well as the usual downscatter, may be employed. This report describes the use of ANISN-L; this is a revised version of ANISN which handles both large and small problems efficiently on CDC-7600 computers. (RWR)

NCEL: two dimensional finite element code for steady-state temperature distribution in seven rod-bundle

International Nuclear Information System (INIS)

Hrehor, M.

1979-01-01

The paper deals with an application of the finite element method to the heat transfer study in seven-pin models of LMFBR fuel subassembly. The developed code NCEL solves two-dimensional steady state heat conduction equation in the whole subassembly model cross-section and enebles to perform the analysis of thermal behaviour in both normal and accidental operational conditions as eccentricity of the central rod or full or partial (porous) blockage of some part of the cross-flow area. The heat removal is simulated by heat sinks in coolant under conditions of subchannels slug flow approximation

Design and implementation in VHDL code of the two-dimensional fast Fourier transform for frequency filtering, convolution and correlation operations

Science.gov (United States)

Vilardy, Juan M.; Giacometto, F.; Torres, C. O.; Mattos, L.

2011-01-01

The two-dimensional Fast Fourier Transform (FFT 2D) is an essential tool in the two-dimensional discrete signals analysis and processing, which allows developing a large number of applications. This article shows the description and synthesis in VHDL code of the FFT 2D with fixed point binary representation using the programming tool Simulink HDL Coder of Matlab; showing a quick and easy way to handle overflow, underflow and the creation registers, adders and multipliers of complex data in VHDL and as well as the generation of test bench for verification of the codes generated in the ModelSim tool. The main objective of development of the hardware architecture of the FFT 2D focuses on the subsequent completion of the following operations applied to images: frequency filtering, convolution and correlation. The description and synthesis of the hardware architecture uses the XC3S1200E family Spartan 3E FPGA from Xilinx Manufacturer.

Development of a two-dimensional simulation code (koad) including atomic processes for beam direct energy conversion

International Nuclear Information System (INIS)

Yamamoto, Y.; Yoshikawa, K.; Hattori, Y.

1987-01-01

A two-dimensional simulation code for the beam direct energy conversion called KVAD (Kyoto University Advanced DART) including various loss mechanisms has been developed, and shown excellent agreement with the authors' experiments using the He + beams. The beam direct energy converter (BDC) is the device to recover the kinetic energy of unneutralized ions in the neutral beam injection (NBI) system directly into electricity. The BDC is very important and essential not only to the improvements of NBI system efficiency, but also to the relaxation of high heat flux problems on the beam dump with increase of injection energies. So far no simulation code could have successfully predicted BDC experimental results. The KUAD code applies, an optimized algorithm for vector processing, the finite element method (FEM) for potential calculation, and a semi-automatic method for spatial segmentations. Since particle trajectories in the KVAD code are analytically solved, very high speed tracings of the particle could be achieved by introducing an adjacent element matrix to identify the neighboring triangle elements and electrodes. Ion space charges are also analytically calculated by the Cloud in Cell (CIC) method, as well as electron space charges. Power losses due to atomic processes can be also evaluated in the KUAD code

FINEDAN - an explicit finite-element calculation code for two-dimensional analyses of fast dynamic transients in nuclear reactor technology

International Nuclear Information System (INIS)

Adamik, V.; Matejovic, P.

1989-01-01

The problems are discussed of nonstationary, nonlinear dynamics of the continuum. A survey is presented of calculation methods in the given area with emphasis on the area of impact problems. A description is presented of the explicit finite elements method and its application to two-dimensional Cartesian and cylindrical configurations. Using the method the explicit calculation code FINEDAN was written which was tested in a series of verification calculations for different configurations and different types of continuum. The main characteristics are presented of the code and of some, of its practical applications. Envisaged trends of the development of the code and its possible applications in the technology of nuclear reactors are given. (author). 9 figs., 4 tabs., 10 refs

Rn3D: A finite element code for simulating gas flow and radon transport in variably saturated, nonisothermal porous media

International Nuclear Information System (INIS)

Holford, D.J.

1994-01-01

This document is a user's manual for the Rn3D finite element code. Rn3D was developed to simulate gas flow and radon transport in variably saturated, nonisothermal porous media. The Rn3D model is applicable to a wide range of problems involving radon transport in soil because it can simulate either steady-state or transient flow and transport in one-, two- or three-dimensions (including radially symmetric two-dimensional problems). The porous materials may be heterogeneous and anisotropic. This manual describes all pertinent mathematics related to the governing, boundary, and constitutive equations of the model, as well as the development of the finite element equations used in the code. Instructions are given for constructing Rn3D input files and executing the code, as well as a description of all output files generated by the code. Five verification problems are given that test various aspects of code operation, complete with example input files, FORTRAN programs for the respective analytical solutions, and plots of model results. An example simulation is presented to illustrate the type of problem Rn3D is designed to solve. Finally, instructions are given on how to convert Rn3D to simulate systems other than radon, air, and water

Integral Transport Theory in One-dimensional Geometries

Energy Technology Data Exchange (ETDEWEB)

Carlvik, I

1966-06-15

A method called DIT (Discrete Integral Transport) has been developed for the numerical solution of the transport equation in one-dimensional systems. The characteristic features of the method are Gaussian integration over the coordinate as described by Kobayashi and Nishihara, and a particular scheme for the calculation of matrix elements in annular and spherical geometry that has been used for collision probabilities in earlier Flurig programmes. The paper gives a general theory including such things as anisotropic scattering and multi-pole fluxes, and it gives a brief description of the Flurig scheme. Annular geometry is treated in some detail, and corresponding formulae are given for spherical and plane geometry. There are many similarities between DIT and the method of collision probabilities. DIT is in many cases faster, because for a certain accuracy in the fluxes DIT often needs fewer space points than the method of collision probabilities needs regions. Several computer codes using DIT, both one-group and multigroup, have been written. It is anticipated that experience gained in calculations with these codes will be reported in another paper.

FLUST-2D - A computer code for the calculation of the two-dimensional flow of a compressible medium in coupled retangular areas

International Nuclear Information System (INIS)

Enderle, G.

1979-01-01

The computer-code FLUST-2D is able to calculate the two-dimensional flow of a compressible fluid in arbitrary coupled rectangular areas. In a finite-difference scheme the program computes pressure, density, internal energy and velocity. Starting with a basic set of equations, the difference equations in a rectangular grid are developed. The computational cycle for coupled fluid areas is described. Results of test calculations are compared to analytical solutions and the influence of time step and mesh size are investigated. The program was used to precalculate the blowdown experiments of the HDR experimental program. Downcomer, plena, internal vessel region, blowdown pipe and a containment area have been modelled two-dimensionally. The major results of the precalculations are presented. This report also contains a description of the code structure and user information. (orig.) [de

Revealing origin of quasi-one dimensional current transport in defect rich two dimensional materials

DEFF Research Database (Denmark)

Lotz, Mikkel Rønne; Boll, Mads; Hansen, Ole

2014-01-01

to a non-uniform current flow characteristic of lower dimensionality. In this work, simulations based on a finite element method together with a Monte Carlo approach are used to establish the transition from 2D to quasi-1D current transport, when applying a micro four-point probe to measure on 2D...... conductors with an increasing amount of line-shaped defects. Clear 2D and 1D signatures are observed at low and high defect densities, respectively, and current density plots reveal the presence of current channels or branches in defect configurations yielding 1D current transport. A strong correlation...

Application of a parallel 3-dimensional hydrogeochemistry HPF code to a proposed waste disposal site at the Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Gwo, Jin-Ping; Yeh, Gour-Tsyh

1997-01-01

The objectives of this study are (1) to parallelize a 3-dimensional hydrogeochemistry code and (2) to apply the parallel code to a proposed waste disposal site at the Oak Ridge National Laboratory (ORNL). The 2-dimensional hydrogeochemistry code HYDROGEOCHEM, developed at the Pennsylvania State University for coupled subsurface solute transport and chemical equilibrium processes, was first modified to accommodate 3-dimensional problem domains. A bi-conjugate gradient stabilized linear matrix solver was then incorporated to solve the matrix equation. We chose to parallelize the 3-dimensional code on the Intel Paragons at ORNL by using an HPF (high performance FORTRAN) compiler developed at PGI. The data- and task-parallel algorithms available in the HPF compiler proved to be highly efficient for the geochemistry calculation. This calculation can be easily implemented in HPF formats and is perfectly parallel because the chemical speciation on one finite-element node is virtually independent of those on the others. The parallel code was applied to a subwatershed of the Melton Branch at ORNL. Chemical heterogeneity, in addition to physical heterogeneities of the geological formations, has been identified as one of the major factors that affect the fate and transport of contaminants at ORNL. This study demonstrated an application of the 3-dimensional hydrogeochemistry code on the Melton Branch site. A uranium tailing problem that involved in aqueous complexation and precipitation-dissolution was tested. Performance statistics was collected on the Intel Paragons at ORNL. Implications of these results on the further optimization of the code were discussed

The discrete cones method for two-dimensional neutron transport calculations

International Nuclear Information System (INIS)

Watanabe, Y.; Maynard, C.W.

1986-01-01

A novel method, the discrete cones method (DC/sub N/), is proposed as an alternative to the discrete ordinates method (S/sub N/) for solutions of the two-dimensional neutron transport equation. The new method utilizes a new concept, discrete cones, which are made by partitioning a unit spherical surface that the direction vector of particles covers. In this method particles in a cone are simultaneously traced instead of those in discrete directions so that an anomaly of the S/sub N/ method, the ray effects, can be eliminated. The DC/sub N/ method has been formulated for X-Y geometry and a program has been creaed by modifying the standard S/sub N/ program TWOTRAN-II. Our sample calculations demonstrate a strong mitigation of the ray effects without a computing cost penalty

SWAT4.0 - The integrated burnup code system driving continuous energy Monte Carlo codes MVP, MCNP and deterministic calculation code SRAC

International Nuclear Information System (INIS)

Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki

2015-03-01

There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)

A comparison of two three-dimensional shell-element transient electromagnetics codes

International Nuclear Information System (INIS)

Yugo, J.J.; Williamson, D.E.

1992-01-01

Electromagnetic forces due to eddy currents strongly influence the design of components for the next generation of fusion devices. An effort has been made to benchmark two computer programs used to generate transient electromagnetic loads: SPARK and EddyCuFF. Two simple transient field problems were analyzed, both of which had been previously analyzed by the SPARK code with results recorded in the literature. A third problem that uses an ITER inboard blanket benchmark model was analyzed as well. This problem was driven with a self-consistent, distributed multifilament plasma model generated by an axisymmetric physics code. The benchmark problems showed good agreement between the two shell-element codes. Variations in calculated eddy currents of 1--3% have been found for similar, finely meshed models. A difference of 8% was found in induced current and 20% in force for a coarse mesh and complex, multifilament field driver. Because comparisons were made to results obtained from literature, model preparation and code execution times were not evaluated

Tuning spin transport across two-dimensional organometallic junctions

Science.gov (United States)

Liu, Shuanglong; Wang, Yun-Peng; Li, Xiangguo; Fry, James N.; Cheng, Hai-Ping

2018-01-01

We study via first-principles modeling and simulation two-dimensional spintronic junctions made of metal-organic frameworks consisting of two Mn-phthalocyanine ferromagnetic metal leads and semiconducting Ni-phthalocyanine channels of various lengths. These systems exhibit a large tunneling magnetoresistance ratio; the transmission functions of such junctions can be tuned using gate voltage by three orders of magnitude. We find that the origin of this drastic change lies in the orbital alignment and hybridization between the leads and the center electronic states. With physical insight into the observed on-off phenomenon, we predict a gate-controlled spin current switch based on two-dimensional crystallines and offer general guidelines for designing spin junctions using 2D materials.

Parallel thermal radiation transport in two dimensions

International Nuclear Information System (INIS)

Smedley-Stevenson, R.P.; Ball, S.R.

2003-01-01

This paper describes the distributed memory parallel implementation of a deterministic thermal radiation transport algorithm in a 2-dimensional ALE hydrodynamics code. The parallel algorithm consists of a variety of components which are combined in order to produce a state of the art computational capability, capable of solving large thermal radiation transport problems using Blue-Oak, the 3 Tera-Flop MPP (massive parallel processors) computing facility at AWE (United Kingdom). Particular aspects of the parallel algorithm are described together with examples of the performance on some challenging applications. (author)

Parallel thermal radiation transport in two dimensions

Energy Technology Data Exchange (ETDEWEB)

Smedley-Stevenson, R.P.; Ball, S.R. [AWE Aldermaston (United Kingdom)

2003-07-01

This paper describes the distributed memory parallel implementation of a deterministic thermal radiation transport algorithm in a 2-dimensional ALE hydrodynamics code. The parallel algorithm consists of a variety of components which are combined in order to produce a state of the art computational capability, capable of solving large thermal radiation transport problems using Blue-Oak, the 3 Tera-Flop MPP (massive parallel processors) computing facility at AWE (United Kingdom). Particular aspects of the parallel algorithm are described together with examples of the performance on some challenging applications. (author)

Converting Panax ginseng DNA and chemical fingerprints into two-dimensional barcode.

Science.gov (United States)

Cai, Yong; Li, Peng; Li, Xi-Wen; Zhao, Jing; Chen, Hai; Yang, Qing; Hu, Hao

2017-07-01

In this study, we investigated how to convert the Panax ginseng DNA sequence code and chemical fingerprints into a two-dimensional code. In order to improve the compression efficiency, GATC2Bytes and digital merger compression algorithms are proposed. HPLC chemical fingerprint data of 10 groups of P. ginseng from Northeast China and the internal transcribed spacer 2 (ITS2) sequence code as the DNA sequence code were ready for conversion. In order to convert such data into a two-dimensional code, the following six steps were performed: First, the chemical fingerprint characteristic data sets were obtained through the inflection filtering algorithm. Second, precompression processing of such data sets is undertaken. Third, precompression processing was undertaken with the P. ginseng DNA (ITS2) sequence codes. Fourth, the precompressed chemical fingerprint data and the DNA (ITS2) sequence code were combined in accordance with the set data format. Such combined data can be compressed by Zlib, an open source data compression algorithm. Finally, the compressed data generated a two-dimensional code called a quick response code (QR code). Through the abovementioned converting process, it can be found that the number of bytes needed for storing P. ginseng chemical fingerprints and its DNA (ITS2) sequence code can be greatly reduced. After GTCA2Bytes algorithm processing, the ITS2 compression rate reaches 75% and the chemical fingerprint compression rate exceeds 99.65% via filtration and digital merger compression algorithm processing. Therefore, the overall compression ratio even exceeds 99.36%. The capacity of the formed QR code is around 0.5k, which can easily and successfully be read and identified by any smartphone. P. ginseng chemical fingerprints and its DNA (ITS2) sequence code can form a QR code after data processing, and therefore the QR code can be a perfect carrier of the authenticity and quality of P. ginseng information. This study provides a theoretical

The simulation of a two-dimensional (2D) transport problem in a rectangular region with Lattice Boltzmann method with two-relaxation-time

Science.gov (United States)

Sugiyanto, S.; Hardyanto, W.; Marwoto, P.

2018-03-01

Transport phenomena are found in many problems in many engineering and industrial sectors. We analyzed a Lattice Boltzmann method with Two-Relaxation Time (LTRT) collision operators for simulation of pollutant moving through the medium as a two-dimensional (2D) transport problem in a rectangular region model. This model consists of a 2D rectangular region with 54 length (x), 27 width (y), and it has isotropic homogeneous medium. Initially, the concentration is zero and is distributed evenly throughout the region of interest. A concentration of 1 is maintained at 9 < y < 18, whereas the concentration of zero is maintained at 0 < y < 9 and 18 < y < 27. A specific discharge (Darcy velocity) of 1.006 is assumed. A diffusion coefficient of 0.8333 is distributed uniformly with a uniform porosity of 0.35. A computer program is written in MATLAB to compute the concentration of pollutant at any specified place and time. The program shows that LTRT solution with quadratic equilibrium distribution functions (EDFs) and relaxation time τa=1.0 are in good agreement result with other numerical solutions methods such as 3DLEWASTE (Hybrid Three-dimensional Lagrangian-Eulerian Finite Element Model of Waste Transport Through Saturated-Unsaturated Media) obtained by Yeh and 3DFEMWATER-LHS (Three-dimensional Finite Element Model of Water Flow Through Saturated-Unsaturated Media with Latin Hypercube Sampling) obtained by Hardyanto.

EFDC1D - A ONE DIMENSIONAL HYDRODYNAMIC AND SEDIMENT TRANSPORT MODEL FOR RIVER AND STREAM NETWORKS: MODEL THEORY AND USERS GUIDE

Science.gov (United States)

This technical report describes the new one-dimensional (1D) hydrodynamic and sediment transport model EFDC1D. This model that can be applied to stream networks. The model code and two sample data sets are included on the distribution CD. EFDC1D can simulate bi-directional unstea...

Transport properties of diazonium functionalized graphene: chiral two-dimensional hole gases

International Nuclear Information System (INIS)

Huang Ping; Jing Long; Zhu Huarui; Gao Xueyun

2012-01-01

The electric transport properties of diazonium functionalized graphene (DFG) were investigated. The temperature dependence of the resistivity (ρ-T) and the Shubnikov-de Haas oscillation of the DFG revealed two-dimensional hole gas (2DHG) behaviors. The DFGs exhibited unusual weak localization behaviors in which both inelastic and chirality-breaking elastic scattering processes should be taken into account, meaning that graphene chirality was maintained. Because of the giant decrease in the diffusion coefficient, the scattering rates remained relatively low in the presence of suppression of the scattering lengths. The decreases of both the mean free path and the Fermi velocity were responsible for the suppression of the diffusion coefficient and hence the charge mobility. (paper)

FLICA-4 (version 1) a computer code for three dimensional thermal analysis of nuclear reactor cores

International Nuclear Information System (INIS)

Raymond, P.; Allaire, G.; Boudsocq, G.

1995-01-01

FLICA-4 is a thermal-hydraulic computer code developed at the French Energy Atomic Commission (CEA) for three dimensional steady state or transient two phase flow for design and safety thermal analysis of nuclear reactor cores. The two phase flow model of FLICA-4 is based on four balance equations for the fluid which includes: three balance equations for the mixture and a mass balance equation for the less concentrated phase which permits the calculation of non-equilibrium flows as sub cooled boiling and superheated steam. A drift velocity model takes into account the velocity disequilibrium between phases. The thermal behaviour of fuel elements can be computed by a one dimensional heat conduction equation in plane, cylindrical or spherical geometries and coupled to the fluid flow calculation. Convection and diffusion of solution products which are transported either by the liquid or by the gas, can be evaluated by solving specific mass conservation equations. A one dimensional two phase flow model can also be used to compute 1-D flow in pipes, guide tubes, BWR assemblies or RBMK channels. The FLICA-4 computer code uses fast running time steam-water functions. Phasic and saturation physical properties are computed by using bi-cubic spline functions. Polynomial coefficients are tabulated from 0.1 to 22 MPa and 0 to 800 degrees C. Specific modules can be utilised in order to generate the spline coefficients for any other fluid properties

Three-dimensional Boltzmann-Hydro Code for Core-collapse in Massive Stars. II. The Implementation of Moving-mesh for Neutron Star Kicks

Energy Technology Data Exchange (ETDEWEB)

Nagakura, Hiroki [TAPIR, Walter Burke Institute for Theoretical Physics, Mailcode 350-17, California Institute of Technology, Pasadena, CA 91125 (United States); Iwakami, Wakana [Yukawa Institute for Theoretical Physics, Kyoto University, Oiwake-cho, Kitashirakawa, Sakyo-ku, Kyoto, 606-8502 (Japan); Furusawa, Shun [Center for Computational Astrophysics, National Astronimical Observatory of Japan, Mitaka, Tokyo 181-8588 (Japan); Sumiyoshi, Kohsuke [Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Yamada, Shoichi [Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Matsufuru, Hideo [High Energy Accelerator Research Organization, 1-1 Oho, Tsukuba, Ibaraki 308-0801 (Japan); Imakura, Akira [University of Tsukuba, 1-1-1, Tennodai Tsukuba, Ibaraki 305-8577 (Japan)

2017-04-01

We present a newly developed moving-mesh technique for the multi-dimensional Boltzmann-Hydro code for the simulation of core-collapse supernovae (CCSNe). What makes this technique different from others is the fact that it treats not only hydrodynamics but also neutrino transfer in the language of the 3 + 1 formalism of general relativity (GR), making use of the shift vector to specify the time evolution of the coordinate system. This means that the transport part of our code is essentially general relativistic, although in this paper it is applied only to the moving curvilinear coordinates in the flat Minknowski spacetime, since the gravity part is still Newtonian. The numerical aspect of the implementation is also described in detail. Employing the axisymmetric two-dimensional version of the code, we conduct two test computations: oscillations and runaways of proto-neutron star (PNS). We show that our new method works fine, tracking the motions of PNS correctly. We believe that this is a major advancement toward the realistic simulation of CCSNe.

Three-dimensional Boltzmann-Hydro Code for Core-collapse in Massive Stars. II. The Implementation of Moving-mesh for Neutron Star Kicks

International Nuclear Information System (INIS)

Nagakura, Hiroki; Iwakami, Wakana; Furusawa, Shun; Sumiyoshi, Kohsuke; Yamada, Shoichi; Matsufuru, Hideo; Imakura, Akira

2017-01-01

We present a newly developed moving-mesh technique for the multi-dimensional Boltzmann-Hydro code for the simulation of core-collapse supernovae (CCSNe). What makes this technique different from others is the fact that it treats not only hydrodynamics but also neutrino transfer in the language of the 3 + 1 formalism of general relativity (GR), making use of the shift vector to specify the time evolution of the coordinate system. This means that the transport part of our code is essentially general relativistic, although in this paper it is applied only to the moving curvilinear coordinates in the flat Minknowski spacetime, since the gravity part is still Newtonian. The numerical aspect of the implementation is also described in detail. Employing the axisymmetric two-dimensional version of the code, we conduct two test computations: oscillations and runaways of proto-neutron star (PNS). We show that our new method works fine, tracking the motions of PNS correctly. We believe that this is a major advancement toward the realistic simulation of CCSNe.

Regional Atmospheric Transport Code for Hanford Emission Tracking, Version 2 (RATCHET2)

International Nuclear Information System (INIS)

Ramsdell, James V.; Rishel, Jeremy P.

2006-01-01

This manual describes the atmospheric model and computer code for the Atmospheric Transport Module within SAC. The Atmospheric Transport Module, called RATCHET2, calculates the time-integrated air concentration and surface deposition of airborne contaminants to the soil. The RATCHET2 code is an adaptation of the Regional Atmospheric Transport Code for Hanford Emissions Tracking (RATCHET). The original RATCHET code was developed to perform the atmospheric transport for the Hanford Environmental Dose Reconstruction Project. Fundamentally, the two sets of codes are identical; no capabilities have been deleted from the original version of RATCHET. Most modifications are generally limited to revision of the run-specification file to streamline the simulation process for SAC.

Regional Atmospheric Transport Code for Hanford Emission Tracking, Version 2(RATCHET2)

Energy Technology Data Exchange (ETDEWEB)

Ramsdell, James V.; Rishel, Jeremy P.

2006-07-01

This manual describes the atmospheric model and computer code for the Atmospheric Transport Module within SAC. The Atmospheric Transport Module, called RATCHET2, calculates the time-integrated air concentration and surface deposition of airborne contaminants to the soil. The RATCHET2 code is an adaptation of the Regional Atmospheric Transport Code for Hanford Emissions Tracking (RATCHET). The original RATCHET code was developed to perform the atmospheric transport for the Hanford Environmental Dose Reconstruction Project. Fundamentally, the two sets of codes are identical; no capabilities have been deleted from the original version of RATCHET. Most modifications are generally limited to revision of the run-specification file to streamline the simulation process for SAC.

Proton and hydrogen transport through two-dimensional monolayers

International Nuclear Information System (INIS)

Seel, Max; Pandey, Ravindra

2016-01-01

Diffusion of protons and hydrogen atoms in representative two-dimensional materials is investigated. Specifically, density functional calculations were performed on graphene, hexagonal boron nitride (h-BN), phosphorene, silicene, and molybdenum disulfide (MoS 2 ) monolayers to study the surface interaction and penetration barriers for protons and hydrogen atoms employing finite cluster models. The calculated barrier heights correlate approximately with the size of the opening formed by the three-fold open sites in the monolayers considered. They range from 1.56 eV (proton) and 4.61 eV (H) for graphene to 0.12 eV (proton) and 0.20 eV (H) for silicene. The results indicate that only graphene and h-BN monolayers have the potential for membranes with high selective permeability. The MoS 2 monolayer behaves differently: protons and H atoms become trapped between the outer S layers in the Mo plane in a well with a depth of 1.56 eV (proton) and 1.5 eV (H atom), possibly explaining why no proton transport was detected, suggesting MoS 2 as a hydrogen storage material instead. For graphene and h-BN, off-center proton penetration reduces the barrier to 1.38 eV for graphene and 0.11 eV for h-BN. Furthermore, Pt acting as a substrate was found to have a negligible effect on the barrier height. In defective graphene, the smallest barrier for proton diffusion (1.05 eV) is found for an oxygen-terminated defect. Therefore, it seems more likely that thermal protons can penetrate a monolayer of h-BN but not graphene and defects are necessary to facilitate the proton transport in graphene. (paper)

Proton and hydrogen transport through two-dimensional monolayers

Science.gov (United States)

Seel, Max; Pandey, Ravindra

2016-06-01

Diffusion of protons and hydrogen atoms in representative two-dimensional materials is investigated. Specifically, density functional calculations were performed on graphene, hexagonal boron nitride (h-BN), phosphorene, silicene, and molybdenum disulfide (MoS2) monolayers to study the surface interaction and penetration barriers for protons and hydrogen atoms employing finite cluster models. The calculated barrier heights correlate approximately with the size of the opening formed by the three-fold open sites in the monolayers considered. They range from 1.56 eV (proton) and 4.61 eV (H) for graphene to 0.12 eV (proton) and 0.20 eV (H) for silicene. The results indicate that only graphene and h-BN monolayers have the potential for membranes with high selective permeability. The MoS2 monolayer behaves differently: protons and H atoms become trapped between the outer S layers in the Mo plane in a well with a depth of 1.56 eV (proton) and 1.5 eV (H atom), possibly explaining why no proton transport was detected, suggesting MoS2 as a hydrogen storage material instead. For graphene and h-BN, off-center proton penetration reduces the barrier to 1.38 eV for graphene and 0.11 eV for h-BN. Furthermore, Pt acting as a substrate was found to have a negligible effect on the barrier height. In defective graphene, the smallest barrier for proton diffusion (1.05 eV) is found for an oxygen-terminated defect. Therefore, it seems more likely that thermal protons can penetrate a monolayer of h-BN but not graphene and defects are necessary to facilitate the proton transport in graphene.

VAMPIR - A two-group two-dimensional diffusion computer code for burnup calculation

International Nuclear Information System (INIS)

Zmijarevic, I.; Petrovic, I.

1985-01-01

VAMPIR is a computer code which simulates the burnup within a reactor coe. It computes the neutron flux, power distribution and burnup taking into account spatial variations of temperature and xenon poisoning. Its overall reactor calculation uses diffusion theory with finite differences approximation in X-Y or R-Z geometry. Two-group macroscopic cross section data are prepared by the lattice cell code WIMS-D4 and stored in the library form of multi entry tabulation against the various parameters that significantly affect the physical conditions in the reactor core. herein, the main features of the program are presented. (author)

A zero-dimensional EXTRAP computer code

International Nuclear Information System (INIS)

Karlsson, P.

1982-10-01

A zero-dimensional computer code has been designed for the EXTRAP experiment to predict the density and the temperature and their dependence upon paramenters such as the plasma current and the filling pressure of neutral gas. EXTRAP is a Z-pinch immersed in a vacuum octupole field and could be either linear or toroidal. In this code the density and temperature are assumed to be constant from the axis up to a breaking point from where they decrease linearly in the radial direction out to the plasma radius. All quantities, however, are averaged over the plasma volume thus giving the zero-dimensional character of the code. The particle, momentum and energy one-fluid equations are solved including the effects of the surrounding neutral gas and oxygen impurities. The code shows that the temperature and density are very sensitive to the shape of the plasma, flatter profiles giving higher temperatures and densities. The temperature, however, is not strongly affected for oxygen concentration less than 2% and is well above the radiation barrier even for higher concentrations. (Author)

Code Coupling for Multi-Dimensional Core Transient Analysis

International Nuclear Information System (INIS)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

2015-01-01

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

Code Coupling for Multi-Dimensional Core Transient Analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

2015-05-15

After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

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)

Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

International Nuclear Information System (INIS)

Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

2009-01-01

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

The discrete cones methods for two-dimensional neutral particle transport problems with voids

International Nuclear Information System (INIS)

Watanabe, Y.; Maynard, C.W.

1983-01-01

One of the most widely applied deterministic methods for time-independent, two-dimensional neutron transport calculations is the discrete ordinates method (DSN). The DSN solution, however, fails to be accurate in a void due to the ray effect. In order to circumvent this drawback, the authors have been developing a novel approximation: the discrete cones method (DCN), where a group of particles in a cone are simultaneously traced instead of particles in discrete directions for the DSN method. Programs, which apply to the DSN method in a non-vacuum region and the DCN method in a void, have been written for transport calculations in X-Y coordinates. The solutions for test problems demonstrate mitigation of the ray effect in voids without loosing the computational efficiency of the DSN method

Depletion methodology in the 3-D whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun

2005-02-01

Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.

A control volume scheme for three-dimensional transport: buffer and matrix effect on a decay chain transport in the repository

International Nuclear Information System (INIS)

Lee, Y. M.; Hwang, Y. S.; Kim, S. G.; Kang, C. H.

2002-01-01

Using a three-dimensional numerical code, B3R developed for nuclide transport of an arbitrary length of decay chain in the buffer between the canister and adjacent rock in a high-level radioactive waste repository by adopting a finite difference method utilizing the control-volume scheme, some illustrative calculations have been done. A linear sorption isotherm, nuclide transport due to diffusion in the buffer and the rock matrix, and advection and dispersion along thin rigid parallel fractures existing in a saturated porous rock matrix as well as diffusion through the fracture wall into the matrix is assumed. In such kind of repository, buffer and rock matrix are known to be important physico-chemical barriers in nuclide retardation. To show effects of buffer and rock matrix on nuclide transport in HLW repository and also to demonstrate usefulness of B3R, several cases of breakthrough curves as well as three-dimensional plots of concentration isopleths associated with these two barriers are introduced for a typical case of decay chain of 234 U→ 230 Th→ 226 Ra, which is the most important chain as far as the human environment is concerned

The nodal discrete-ordinate transport calculation of anisotropy scattering problem in three-dimensional cartesian geometry

International Nuclear Information System (INIS)

Wu Hongchun; Xie Zhongsheng; Zhu Xuehua

1994-01-01

The nodal discrete-ordinate transport calculating model of anisotropy scattering problem in three-dimensional cartesian geometry is given. The computing code NOTRAN/3D has been encoded and the satisfied conclusion is gained

Improvements in practical applicability of NSHEX: nodal transport calculation code for three-dimensional hexagonal-Z geometry

International Nuclear Information System (INIS)

Sugino, Kazuteru

1998-07-01

As a tool to perform a fast reactor core calculations with high accuracy, NSHEX the nodal transport calculation code for three-dimensional hexagonal-Z geometry is under development. To improve the practical applicability of NSHEX, for instance, in its application to safety analysis and commercial reactor core design studies, we investigated the basic theory used in it, improved the program performance, and evaluated its applicability to the analysis of commercial reactor cores. The current studies show the following: (1) An improvement in the treatment of radial leakage in the radial nodal coupling equation bettered calculational convergence for safety analysis calculation, so the applicability of NSHEX to safety analysis was improved. (2) As a result of comparison of results from NSHEX and the standard core calculation code, it was confirmed that there was consistency between them. (3) According to the evaluation of the effect due to the difference of calculational condition, it was found that the calculation under appropriate nodal expansion orders and Sn orders correspond to the one under most detailed condition. However further investigation is required to reduce the uncertainty in calculational results due to the treatment of high order flux moments. (4) A whole core version of NSHEX enabling calculation for any FBR core geometry has been developed, this improved general applicability of NSHEX. (5) An investigation of the applicability of the rebalance method to acceleration clarified that this improved calculational convergence and it was effective. (J.P.N.)

Velocity and Dispersion for a Two-Dimensional Random Walk

International Nuclear Information System (INIS)

Li Jinghui

2009-01-01

In the paper, we consider the transport of a two-dimensional random walk. The velocity and the dispersion of this two-dimensional random walk are derived. It mainly show that: (i) by controlling the values of the transition rates, the direction of the random walk can be reversed; (ii) for some suitably selected transition rates, our two-dimensional random walk can be efficient in comparison with the one-dimensional random walk. Our work is motivated in part by the challenge to explain the unidirectional transport of motor proteins. When the motor proteins move at the turn points of their tracks (i.e., the cytoskeleton filaments and the DNA molecular tubes), some of our results in this paper can be used to deal with the problem. (general)

Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

International Nuclear Information System (INIS)

Chiu, C.

1981-01-01

Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nuclear Boiling Ratio (DNBR) which ultimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithm. First, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for this channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum. Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties. Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code. (orig.)

Probing Carrier Transport and Structure-Property Relationship of Highly Ordered Organic Semiconductors at the Two-Dimensional Limit.

Science.gov (United States)

Zhang, Yuhan; Qiao, Jingsi; Gao, Si; Hu, Fengrui; He, Daowei; Wu, Bing; Yang, Ziyi; Xu, Bingchen; Li, Yun; Shi, Yi; Ji, Wei; Wang, Peng; Wang, Xiaoyong; Xiao, Min; Xu, Hangxun; Xu, Jian-Bin; Wang, Xinran

2016-01-08

One of the basic assumptions in organic field-effect transistors, the most fundamental device unit in organic electronics, is that charge transport occurs two dimensionally in the first few molecular layers near the dielectric interface. Although the mobility of bulk organic semiconductors has increased dramatically, direct probing of intrinsic charge transport in the two-dimensional limit has not been possible due to excessive disorders and traps in ultrathin organic thin films. Here, highly ordered single-crystalline mono- to tetralayer pentacene crystals are realized by van der Waals (vdW) epitaxy on hexagonal BN. We find that the charge transport is dominated by hopping in the first conductive layer, but transforms to bandlike in subsequent layers. Such an abrupt phase transition is attributed to strong modulation of the molecular packing by interfacial vdW interactions, as corroborated by quantitative structural characterization and density functional theory calculations. The structural modulation becomes negligible beyond the second conductive layer, leading to a mobility saturation thickness of only ∼3  nm. Highly ordered organic ultrathin films provide a platform for new physics and device structures (such as heterostructures and quantum wells) that are not possible in conventional bulk crystals.

Applications of Transport/Reaction Codes to Problems in Cell Modeling; TOPICAL

International Nuclear Information System (INIS)

MEANS, SHAWN A.; RINTOUL, MARK DANIEL; SHADID, JOHN N.

2001-01-01

We demonstrate two specific examples that show how our exiting capabilities in solving large systems of partial differential equations associated with transport/reaction systems can be easily applied to outstanding problems in computational biology. First, we examine a three-dimensional model for calcium wave propagation in a Xenopus Laevis frog egg and verify that a proposed model for the distribution of calcium release sites agrees with experimental results as a function of both space and time. Next, we create a model of the neuron's terminus based on experimental observations and show that the sodium-calcium exchanger is not the route of sodium's modulation of neurotransmitter release. These state-of-the-art simulations were performed on massively parallel platforms and required almost no modification of existing Sandia codes

Theoretical background and user's manual for the computer code on groundwater flow and radionuclide transport calculation in porous rock

International Nuclear Information System (INIS)

Shirakawa, Toshihiko; Hatanaka, Koichiro

2001-11-01

In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostatistical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report. (author)

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

Parallel Implementation of the Multi-Dimensional Spectral Code SPECT3D on large 3D grids.

Science.gov (United States)

Golovkin, Igor E.; Macfarlane, Joseph J.; Woodruff, Pamela R.; Pereyra, Nicolas A.

2006-10-01

The multi-dimensional collisional-radiative, spectral analysis code SPECT3D can be used to study radiation from complex plasmas. SPECT3D can generate instantaneous and time-gated images and spectra, space-resolved and streaked spectra, which makes it a valuable tool for post-processing hydrodynamics calculations and direct comparison between simulations and experimental data. On large three dimensional grids, transporting radiation along lines of sight (LOS) requires substantial memory and CPU resources. Currently, the parallel option in SPECT3D is based on parallelization over photon frequencies and allows for a nearly linear speed-up for a variety of problems. In addition, we are introducing a new parallel mechanism that will greatly reduce memory requirements. In the new implementation, spatial domain decomposition will be utilized allowing transport along a LOS to be performed only on the mesh cells the LOS crosses. The ability to operate on a fraction of the grid is crucial for post-processing the results of large-scale three-dimensional hydrodynamics simulations. We will present a parallel implementation of the code and provide a scalability study performed on a Linux cluster.

Three-dimensional thermal hydraulic best estimate code BAGIRA: new results of verification

International Nuclear Information System (INIS)

Peter Kohut; Sergey D Kalinichenko; Alexander E Kroshilin; Vladimir E Kroshilin; Alexander V Smirnov

2005-01-01

Full text of publication follows: BAGIRA is a three-dimensional inhomogeneous two-velocity two-temperature thermal hydraulic code of best estimate, elaborated in VNIIAES for modeling two-phase flows in the primary circuit and steam generators of VVER-type nuclear reactors under various accident, transient or normal operation conditions. In this talk we present verification results of the BAGIRA code, obtained on the basis of different experiments performed on special and integral thermohydraulic experimental facilities as well as on real NPPs. Special attention is paid to the verification of three-dimensional flow models. Besides that we expose new results of the code benchmark analysis made on the basis of two recent LOCA-type experiments - 'Leak 2 x 25% from the hot leg double-side rupture' and 'Leak 3% from the cold leg' - performed on the PSB-VVER integral test facility (Electrogorsk Research and Engineering Center, Electrogorsk, Russia) - the most up-to-date Russian large-scale four-loop unit which has been designed for modelling the primary circuit of VVER-1000 type reactors. (authors)

Simplified two and three dimensional HTTR benchmark problems

International Nuclear Information System (INIS)

Zhang Zhan; Rahnema, Farzad; Zhang Dingkang; Pounders, Justin M.; Ougouag, Abderrafi M.

2011-01-01

To assess the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of whole core configurations. In this paper we have created two and three dimensional numerical benchmark problems typical of high temperature gas cooled prismatic cores. Additionally, a single cell and single block benchmark problems are also included. These problems were derived from the HTTR start-up experiment. Since the primary utility of the benchmark problems is in code-to-code verification, minor details regarding geometry and material specification of the original experiment have been simplified while retaining the heterogeneity and the major physics properties of the core from a neutronics viewpoint. A six-group material (macroscopic) cross section library has been generated for the benchmark problems using the lattice depletion code HELIOS. Using this library, Monte Carlo solutions are presented for three configurations (all-rods-in, partially-controlled and all-rods-out) for both the 2D and 3D problems. These solutions include the core eigenvalues, the block (assembly) averaged fission densities, local peaking factors, the absorption densities in the burnable poison and control rods, and pin fission density distribution for selected blocks. Also included are the solutions for the single cell and single block problems.

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

International Nuclear Information System (INIS)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN

Energy Technology Data Exchange (ETDEWEB)

Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki

1996-12-01

The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)

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

International Nuclear Information System (INIS)

Warner, Paul; Oliveira, R.E. de

2000-01-01

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

Transport and equilibrium in field-reversed mirrors

International Nuclear Information System (INIS)

Boyd, J.K.

1982-09-01

Two plasma models relevant to compact torus research have been developed to study transport and equilibrium in field reversed mirrors. In the first model for small Larmor radius and large collision frequency, the plasma is described as an adiabatic hydromagnetic fluid. In the second model for large Larmor radius and small collision frequency, a kinetic theory description has been developed. Various aspects of the two models have been studied in five computer codes ADB, AV, NEO, OHK, RES. The ADB code computes two dimensional equilibrium and one dimensional transport in a flux coordinate. The AV code calculates orbit average integrals in a harmonic oscillator potential. The NEO code follows particle trajectories in a Hill's vortex magnetic field to study stochasticity, invariants of the motion, and orbit average formulas. The OHK code displays analytic psi(r), B/sub Z/(r), phi(r), E/sub r/(r) formulas developed for the kinetic theory description. The RES code calculates resonance curves to consider overlap regions relevant to stochastic orbit behavior

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

Recent improvements and new features in the Westinghouse lattice physics codes

International Nuclear Information System (INIS)

Huria, H.C.; Buechel, R.J.

1995-01-01

Westinghouse has been using the ANC three-dimensional, two-energy-group nodal model for nuclear analysis and fuel management calculations for standard pressurized water reactor (PWR) reload design analysis since 1988. The cross sections are obtained from PHOENIX-P, a modified version of the PHOENIX lattice physics code for all square-assembly PWR cores. The PHOENIX-H code was developed for modeling both the VVER-1000 and VVER-440 fuel lattice configurations. The PHOENIX-H code has evolved from PHOENIX-P, the primary difference being in the neutronic solution modules. The PHOENIX-P code determines the assembly flux distribution using integral transport theory-based pin-cell nodal coupling followed by two-dimensional discrete ordinates solution in x-y geometry. The PHOENIX-H code uses the two-dimensional heterogeneous response method. The other infrastructure is identical in both the codes, and they share the same 42-group cross-section library

Theory and application of a three-dimensional code SHAPS to complex piping systems

International Nuclear Information System (INIS)

Wang, C.Y.

1983-01-01

This paper describes the theory and application of a three-dimensional computer code SHAPS to the complex piping systems. The code utilizes a two-dimensional implicit Eulerian method for the hydrodynamic analysis together with a three-dimensional elastic-plastic finite-element program for the structural calculation. A three-dimensional pipe element with eight degrees of freedom is employed to account for the hoop, flexural, axial, and the torsional mode of the piping system. In the SHAPS analysis the hydrodynamic equations are modified to include the global piping motion. Coupling between fluid and structure is achieved by enforcing the free-slip boundary conditions. Also, the response of the piping network generated by the seismic excitation can be included. A thermal transient capability is also provided in SHAPS. To illustrate the methodology, many sample problems dealing with the hydrodynamic, structural, and thermal analyses of reactor-piping systems are given. Validation of the SHAPS code with experimental data is also presented

Collisional plasma transport: two-dimensional scalar formulation of the initial boundary value problem and quasi one-dimensional models

International Nuclear Information System (INIS)

Mugge, J.W.

1979-10-01

The collisional plasma transport problem is formulated as an initial boundary value problem for general characteristic boundary conditions. Starting from the full set of hydrodynamic and electrodynamic equations an expansion in the electron-ion mass ratio together with a multiple timescale method yields simplified equations on each timescale. On timescales where many collisions have taken place for the simplified equations the initial boundary value problem is formulated. Through the introduction of potentials a two-dimensional scalar formulation in terms of quasi-linear integro-differential equations of second order for a domain consisting of plasma and vacuum sub-domains is obtained. (Auth.)

Analytical simulation of two dimensional advection dispersion ...

African Journals Online (AJOL)

The study was designed to investigate the analytical simulation of two dimensional advection dispersion equation of contaminant transport. The steady state flow condition of the contaminant transport where inorganic contaminants in aqueous waste solutions are disposed of at the land surface where it would migrate ...

Analytical Simulation of Two Dimensional Advection Dispersion ...

African Journals Online (AJOL)

ADOWIE PERE

ABSTRACT: The study was designed to investigate the analytical simulation of two dimensional advection dispersion equation of contaminant transport. The steady state flow condition of the contaminant transport where inorganic contaminants in aqueous waste solutions are disposed of at the land surface where it would ...

User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

International Nuclear Information System (INIS)

Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

2003-01-01

This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

High Energy Transport Code HETC

International Nuclear Information System (INIS)

Gabriel, T.A.

1985-09-01

The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs

ALGE3D: A Three-Dimensional Transport Model

Science.gov (United States)

Maze, G. M.

2017-12-01

Of the top 10 most populated US cities from a 2015 US Census Bureau estimate, 7 of the cities are situated near the ocean, a bay, or on one of the Great Lakes. A contamination of the water ways in the United States could be devastating to the economy (through tourism and industries such as fishing), public health (from direct contact, or contaminated drinking water), and in some cases even infrastructure (water treatment plants). Current national response models employed by emergency response agencies have well developed models to simulate the effects of hazardous contaminants in riverine systems that are primarily driven by one-dimensional flows; however in more complex systems, such as tidal estuaries, bays, or lakes, a more complex model is needed. While many models exist, none are capable of quick deployment in emergency situations that could contain a variety of release situations including a mixture of both particulate and dissolved chemicals in a complex flow area. ALGE3D, developed at the Department of Energy's (DOE) Savannah River National Laboratory (SRNL), is a three-dimensional hydrodynamic code which solves the momentum, mass, and energy conservation equations to predict the movement and dissipation of thermal or dissolved chemical plumes discharged into cooling lakes, rivers, and estuaries. ALGE3D is capable of modeling very complex flows, including areas with tidal flows which include wetting and drying of land. Recent upgrades have increased the capabilities including the transport of particulate tracers, allowing for more complete modeling of the transport of pollutants. In addition the model is capable of coupling with a one-dimension riverine transport model or a two-dimension atmospheric deposition model in the event that a contamination event occurs upstream or upwind of the water body.

Continuous Energy, Multi-Dimensional Transport Calculations for Problem Dependent Resonance Self-Shielding

International Nuclear Information System (INIS)

Downar, T.

2009-01-01

The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.

Two-Dimensional Spatial Imaging of Charge Transport in Germanium Crystals at Cryogenic Temperatures

Energy Technology Data Exchange (ETDEWEB)

Moffatt, Robert [Stanford Univ., CA (United States)

2016-03-01

In this dissertation, I describe a novel apparatus for studying the transport of charge in semiconductors at cryogenic temperatures. The motivation to conduct this experiment originated from an asymmetry observed between the behavior of electrons and holes in the germanium detector crystals used by the Cryogenic Dark Matter Search (CDMS). This asymmetry is a consequence of the anisotropic propagation of electrons in germanium at cryogenic temperatures. To better model our detectors, we incorporated this effect into our Monte Carlo simulations of charge transport. The purpose of the experiment described in this dissertation is to test those models in detail. Our measurements have allowed us to discover a shortcoming in our most recent Monte Carlo simulations of electrons in germanium. This discovery would not have been possible without the measurement of the full, two-dimensional charge distribution, which our experimental apparatus has allowed for the first time at cryogenic temperatures.

Transport Methods Conquering the Seven-Dimensional Mountain

International Nuclear Information System (INIS)

Graziani, F; Olson, G

2003-01-01

In a wide variety of applications, a significant fraction of the momentum and energy present in a physical problem is carried by the transport of particles. Depending on the circumstances, the types of particles might involve some or all of photons, neutrinos, charged particles, or neutrons. In application areas that use transport, the computational time is usually dominated by the transport calculation. Therefore, there is a potential for great synergy; progress in transport algorithms could help quicken the time to solution for many applications. The complexity, and hence expense, involved in solving the transport problem can be understood by realizing that the general solution to the Boltzmann transport equation is seven dimensional: 3 spatial coordinates, 2 angles, 1 time, and 1 for speed or energy. Low-order approximations to the transport equation are frequently used due in part to physical justification but many times simply because a solution to the full transport problem is too computationally expensive. An example is the diffusion equation, which effectively drops the two angles in phase space by assuming that a linear representation in angle is adequate. Another approximation is the grey approximation, which drops the energy variable by averaging over it. If the grey approximation is applied to the diffusion equation, the expense of solving what amounts to the simplest possible description of transport is roughly equal to the cost of implicit computational fluid dynamics. It is clear therefore, that for those application areas needing some form of transport, fast, accurate and robust transport algorithms can lead to an increase in overall code performance and a decrease in time to solution. The seven-dimensional nature of transport means that factors of 100 or 1000 improvement in computer speed or memory are quickly absorbed in slightly higher resolution in space, angle, and energy. Therefore, the biggest advances in the last few years and in the next

Comparison of preconditioned generalized conjugate gradient methods to two-dimensional neutron and photon transport equation

International Nuclear Information System (INIS)

Chen, G.S.; Yang, D.Y.

1998-01-01

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 (conjugate gradient square algorithm), Bi-CGSTAB (bi-conjugate gradient stabilized algorithm) and QMRCGSTAB (quasi-minimal residual variant of bi-conjugate gradient stabilized algorithm). These subroutines are connected to computer program DORT. Several problems are tested on a personal computer with Intel Pentium CPU. The reasons to choose the generalized conjugate gradient methods are that the methods have better residual (equivalent to error) control procedures in the computation and have better convergent rate. The pointwise incomplete LU factorization ILU, modified pointwise incomplete LU factorization MILU, block incomplete factorization BILU and modified blockwise incomplete LU factorization MBILU are the preconditioning techniques used in the several testing problems. In Bi-CGSTAB, CGS, TFQMR and QMRCGSTAB method, we find that either CGS or Bi-CGSTAB method combined with preconditioner MBILU is the most efficient algorithm in these methods in the several testing problems. The numerical solution of flux by preconditioned CGS and Bi-CGSTAB methods has the same result as those from Cray computer, obtained by either the point successive relaxation method or the line successive relaxation method combined with Gaussian elimination

A structural modification of the two dimensional fuel behaviour analysis code FEMAXI-III with high-speed vectorized operation

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Ishiguro, Misako; Yamazaki, Takashi; Tokunaga, Yasuo.

1985-02-01

Though the two-dimensional fuel behaviour analysis code FEMAXI-III has been developed by JAERI in form of optimized scalar computer code, the call for more efficient code usage generally arized from the recent trends like high burn-up and load follow operation asks the code into further modification stage. A principal aim of the modification is to transform the already implemented scalar type subroutines into vectorized forms to make the programme structure efficiently run on high-speed vector computers. The effort of such structural modification has been finished on a fair way to success. The benchmarking two tests subsequently performed to examine the effect of the modification led us the following concluding remarks: (1) In the first benchmark test, comparatively high-burned three fuel rods that have been irradiated in HBWR, BWR, and PWR condition are prepared. With respect to all cases, a net computing time consumed in the vectorized FEMAXI is approximately 50 % less than that consumed in the original one. (2) In the second benchmark test, a total of 26 PWR fuel rods that have been irradiated in the burn-up ranges of 13-30 MWd/kgU and subsequently power ramped in R2 reactor, Sweden is prepared. In this case the code is purposed to be used for making an envelop of PCI-failure threshold through 26 times code runs. Before coming to the same conclusion, the vectorized FEMAXI-III consumed a net computing time 18 min., while the original FEMAXI-III consumed a computing time 36 min. respectively. (3) The effects obtained from such structural modification are found to be significantly attributed to saving a net computing time in a mechanical calculation in the vectorized FEMAXI-III code. (author)

Vectorization, parallelization and porting of nuclear codes. 2001

International Nuclear Information System (INIS)

Akiyama, Mitsunaga; Katakura, Fumishige; Kume, Etsuo; Nemoto, Toshiyuki; Tsuruoka, Takuya; Adachi, Masaaki

2003-07-01

Several computer codes in the nuclear field have been vectorized, parallelized and transported on the super computer system at Center for Promotion of Computational Science and Engineering in Japan Atomic Energy Research Institute. We dealt with 10 codes in fiscal 2001. In this report, the parallelization of Neutron Radiography for 3 Dimensional CT code NR3DCT, the vectorization of unsteady-state heat conduction code THERMO3D, the porting of initial program of MHD simulation, the tuning of Heat And Mass Balance Analysis Code HAMBAC, the porting and parallelization of Monte Carlo N-Particle transport code MCNP4C3, the porting and parallelization of Monte Carlo N-Particle transport code system MCNPX2.1.5, the porting of induced activity calculation code CINAC-V4, the use of VisLink library in multidimensional two-fluid model code ACD3D and the porting of experiment data processing code from GS8500 to SR8000 are described. (author)

Multi-dimensional two-fluid flow computation. An overview

International Nuclear Information System (INIS)

Carver, M.B.

1992-01-01

This paper discusses a repertoire of three-dimensional computer programs developed to perform critical analysis of single-phase, two-phase and multi-fluid flow in reactor components. The basic numerical approach to solving the governing equations common to all the codes is presented and the additional constitutive relationships required for closure are discussed. Particular applications are presented for a number of computer codes. (author). 12 refs

Fractional calculus phenomenology in two-dimensional plasma models

Science.gov (United States)

Gustafson, Kyle; Del Castillo Negrete, Diego; Dorland, Bill

2006-10-01

Transport processes in confined plasmas for fusion experiments, such as ITER, are not well-understood at the basic level of fully nonlinear, three-dimensional kinetic physics. Turbulent transport is invoked to describe the observed levels in tokamaks, which are orders of magnitude greater than the theoretical predictions. Recent results show the ability of a non-diffusive transport model to describe numerical observations of turbulent transport. For example, resistive MHD modeling of tracer particle transport in pressure-gradient driven turbulence for a three-dimensional plasma reveals that the superdiffusive (2̂˜t^α where α> 1) radial transport in this system is described quantitatively by a fractional diffusion equation Fractional calculus is a generalization involving integro-differential operators, which naturally describe non-local behaviors. Our previous work showed the quantitative agreement of special fractional diffusion equation solutions with numerical tracer particle flows in time-dependent linearized dynamics of the Hasegawa-Mima equation (for poloidal transport in a two-dimensional cold-ion plasma). In pursuit of a fractional diffusion model for transport in a gyrokinetic plasma, we now present numerical results from tracer particle transport in the nonlinear Hasegawa-Mima equation and a planar gyrokinetic model. Finite Larmor radius effects will be discussed. D. del Castillo Negrete, et al, Phys. Rev. Lett. 94, 065003 (2005).

Kinetic instabilities of thin current sheets: Results of two-and-one-half-dimensional Vlasov code simulations

International Nuclear Information System (INIS)

Silin, I.; Buechner, J.

2003-01-01

Nonlinear triggering of the instability of thin current sheets is investigated by two-and-one-half- dimensional Vlasov code simulations. A global drift-resonant instability (DRI) is found, which results from the lower-hybrid-drift waves penetrating from the current sheet edges to the center where they resonantly interact with unmagnetized ions. This resonant nonlinear instability grows faster than a Kelvin-Helmholtz instability obtained in previous studies. The DRI is either asymmetric or symmetric mode or a combination of the two, depending on the relative phase of the lower-hybrid-drift waves at the edges of the current sheet. With increasing particle mass ratio the wavenumber of the fastest-growing mode increases as kL z ∼(m i /m e ) 1/2 /2 and the growth rate of the DRI saturates at a finite level

Two-dimensional study of shock breakout at the rear face of laser irradiated metallic targets

Energy Technology Data Exchange (ETDEWEB)

Cottet, F.; Marty, L.; Hallouin, M.; Romain, J.P.; Virmont, J.; Fabbro, R.; Faral, B.

1988-11-01

The two-dimensional propagation dynamics of laser-driven shock waves in solids is studied through the analysis of the shock breakout at the rear face of the target for a set of materials and laser intensities. The laser shock simulations were carried out by means of a two-dimensional hydrodynamics code in which the laser-ablation pressure is replaced by an equivalent pressure pulse. It is shown that the two-dimensional code is a very useful tool to analyze laser-shock experiments where two-dimensional effects arise from a finite laser-spot size or a heterogeneous energy deposition.

Two-dimensional study of shock breakout at the rear face of laser irradiated metallic targets

International Nuclear Information System (INIS)

Cottet, F.; Marty, L.; Hallouin, M.; Romain, J.P.; Virmont, J.; Fabbro, R.; Faral, B.

1988-01-01

The two-dimensional propagation dynamics of laser-driven shock waves in solids is studied through the analysis of the shock breakout at the rear face of the target for a set of materials and laser intensities. The laser shock simulations were carried out by means of a two-dimensional hydrodynamics code in which the laser-ablation pressure is replaced by an equivalent pressure pulse. It is shown that the two-dimensional code is a very useful tool to analyze laser-shock experiments where two-dimensional effects arise from a finite laser-spot size or a heterogeneous energy deposition

Predictive capabilities of a two-dimensional model in the ground water transport of radionuclides

International Nuclear Information System (INIS)

Gureghian, A.B.; Beskid, N.J.; Marmer, G.J.

1978-01-01

The discharge of low-level radioactive waste into tailings ponds is a potential source of ground water contamination. The estimation of the radiological hazards related to the ground water transport of radionuclides from tailings retention systems depends on reasonably accurate estimates of the movement of both water and solute. A two-dimensional mathematical model having predictive capability for ground water flow and solute transport has been developed. The flow equation has been solved under steady-state conditions and the mass transport equation under transient conditions. The simultaneous solution of both equations is achieved through the finite element technique using isoparametric elements, based on the Galerkin formulation. However, in contrast to the flow equation solution, the weighting functions used in the solution of the mass transport equation have a non-symmetric form. The predictive capability of the model is demonstrated using an idealized case based on analyses of field data obtained from the sites of operating uranium mills. The pH of the solution, which regulates the variation of the distribution coefficient (K/sub d/) in a particular site, appears to be the most important factor in the assessment of the rate of migration of the elements considered herein

FISH: A THREE-DIMENSIONAL PARALLEL MAGNETOHYDRODYNAMICS CODE FOR ASTROPHYSICAL APPLICATIONS

International Nuclear Information System (INIS)

Kaeppeli, R.; Whitehouse, S. C.; Scheidegger, S.; Liebendoerfer, M.; Pen, U.-L.

2011-01-01

FISH is a fast and simple ideal magnetohydrodynamics code that scales to ∼10,000 processes for a Cartesian computational domain of ∼1000 3 cells. The simplicity of FISH has been achieved by the rigorous application of the operator splitting technique, while second-order accuracy is maintained by the symmetric ordering of the operators. Between directional sweeps, the three-dimensional data are rotated in memory so that the sweep is always performed in a cache-efficient way along the direction of contiguous memory. Hence, the code only requires a one-dimensional description of the conservation equations to be solved. This approach also enables an elegant novel parallelization of the code that is based on persistent communications with MPI for cubic domain decomposition on machines with distributed memory. This scheme is then combined with an additional OpenMP parallelization of different sweeps that can take advantage of clusters of shared memory. We document the detailed implementation of a second-order total variation diminishing advection scheme based on flux reconstruction. The magnetic fields are evolved by a constrained transport scheme. We show that the subtraction of a simple estimate of the hydrostatic gradient from the total gradients can significantly reduce the dissipation of the advection scheme in simulations of gravitationally bound hydrostatic objects. Through its simplicity and efficiency, FISH is as well suited for hydrodynamics classes as for large-scale astrophysical simulations on high-performance computer clusters. In preparation for the release of a public version, we demonstrate the performance of FISH in a suite of astrophysically orientated test cases.

Two dimensional electron transport in disordered and ordered distributions of magnetic flux vortices

International Nuclear Information System (INIS)

Nielsen, M.; Hedegaard, P.

1994-04-01

We have considered the conductivity properties of a two dimensional electron gas (2DEG) in two different kinds of inhomogeneous magnetic fields, i.e. a disordered distribution of magnetic flux vortices, and a periodic array of magnetic flux vortices. The work falls in two parts. In the first part we show how the phase shifts for an electron scattering on an isolated vortex, can be calculated analytically, and related to the transport properties through the differential cross section. In the second part we present numerical results for the Hall conductivity of the 2DEG in a periodic array of flux vortices found by exact diagonalization. We find characteristic spikes in the Hall conductance, when it is plotted against the filling fraction. It is argued that the spikes can be interpreted in terms of ''topological charge'' piling up across local and global gaps in the energy spectrum. (au) (23 refs.)

Two-dimensional simulations of magnetically-driven instabilities

International Nuclear Information System (INIS)

Peterson, D.; Bowers, R.; Greene, A.E.; Brownell, J.

1986-01-01

A two-dimensional Eulerian MHD code is used to study the evolution of magnetically-driven instabilities in cylindrical geometry. The code incorporates an equation of state, resistivity, and radiative cooling model appropriate for an aluminum plasma. The simulations explore the effects of initial perturbations, electrical resistivity, and radiative cooling on the growth and saturation of the instabilities. Comparisons are made between the 2-D simulations, previous 1-D simulations, and results from the Pioneer experiments of the Los Alamos foil implosion program

Particle and heavy ion transport code system; PHITS

International Nuclear Information System (INIS)

Niita, Koji

2004-01-01

Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

Fourier analysis of parallel block-Jacobi splitting with transport synthetic acceleration in two-dimensional geometry

International Nuclear Information System (INIS)

Rosa, M.; Warsa, J. S.; Chang, J. H.

2007-01-01

A Fourier analysis is conducted in two-dimensional (2D) Cartesian geometry for the discrete-ordinates (SN) approximation of the neutron transport problem solved with Richardson iteration (Source Iteration) and Richardson iteration preconditioned with Transport Synthetic Acceleration (TSA), using the Parallel Block-Jacobi (PBJ) algorithm. The results for the un-accelerated algorithm show that convergence of PBJ can degrade, leading in particular to stagnation of GMRES(m) in problems containing optically thin sub-domains. The results for the accelerated algorithm indicate that TSA can be used to efficiently precondition an iterative method in the optically thin case when implemented in the 'modified' version MTSA, in which only the scattering in the low order equations is reduced by some non-negative factor β<1. (authors)

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

Application of diffusion theory to the transport of neutral particles in fusion plasmas

International Nuclear Information System (INIS)

Hasan, M.Z.

1985-01-01

It is shown that the widely held view that diffusion theory can not provide good accuracy for the transport of neutral particles in fusion plasmas is misplaced. In fact, it is shown that multigroup diffusion theory gives quite good accuracy as compared to the transport theory. The reasons for this are elaborated and some of the physical and theoretical reasons which make the multigroup diffusion theory provide good accuracy are explained. Energy dependence must be taken into consideration to obtain a realistic neutral atom distribution in fusion plasmas. There are two reasons for this; presence of either is enough to necessitate an energy dependent treatment. First, the plasma temperature varies spatially, and second, the ratio of charge-exchange to total plasma-neutral interaction cross section (c) is not close to one. A computer code to solve the one-dimensional multigroup diffusion theory in general geometry (slab, cylindrical and spherical) has been written for use on Cray computers, and its results are compared with those from the one-dimensional transport code ANISN to support the above finding. A fast, compact and versatile two-dimensional finite element multigroup diffusion theory code, FINAT, in X-Y and R-Z cylindrical/toroidal geometries has been written for use on CRAY computers. This code has been compared with the two dimensional transport code DOT-4.3. The accuracy is very good, and FENAT runs much faster compared even to DOT-4.3 which is a finite difference code

Noncoherent Spectral Optical CDMA System Using 1D Active Weight Two-Code Keying Codes

Directory of Open Access Journals (Sweden)

Bih-Chyun Yeh

2016-01-01

Full Text Available We propose a new family of one-dimensional (1D active weight two-code keying (TCK in spectral amplitude coding (SAC optical code division multiple access (OCDMA networks. We use encoding and decoding transfer functions to operate the 1D active weight TCK. The proposed structure includes an optical line terminal (OLT and optical network units (ONUs to produce the encoding and decoding codes of the proposed OLT and ONUs, respectively. The proposed ONU uses the modified cross-correlation to remove interferences from other simultaneous users, that is, the multiuser interference (MUI. When the phase-induced intensity noise (PIIN is the most important noise, the modified cross-correlation suppresses the PIIN. In the numerical results, we find that the bit error rate (BER for the proposed system using the 1D active weight TCK codes outperforms that for two other systems using the 1D M-Seq codes and 1D balanced incomplete block design (BIBD codes. The effective source power for the proposed system can achieve −10 dBm, which has less power than that for the other systems.

Current status of high energy nucleon-meson transport code

Energy Technology Data Exchange (ETDEWEB)

Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)

NASA space radiation transport code development consortium

International Nuclear Information System (INIS)

Townsend, L. W.

2005-01-01

Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)

PHITS-a particle and heavy ion transport code system

International Nuclear Information System (INIS)

Niita, Koji; Sato, Tatsuhiko; Iwase, Hiroshi; Nose, Hiroyuki; Nakashima, Hiroshi; Sihver, Lembit

2006-01-01

The paper presents a summary of the recent development of the multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS. In particular, we discuss in detail the development of two new models, JAM and JQMD, for high energy particle interactions, incorporated in PHITS, and show comparisons between model calculations and experiments for the validations of these models. The paper presents three applications of the code including spallation neutron source, heavy ion therapy and space radiation. The results and examples shown indicate PHITS has great ability of carrying out the radiation transport analysis of almost all particles including heavy ions within a wide energy range

RADTRAN 5: A computer code for transportation risk analysis

International Nuclear Information System (INIS)

Neuhauser, K.S.; Kanipe, F.L.

1991-01-01

RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air

Moderator feedback effects in two-dimensional nodal methods for pressurized water reactor analysis

International Nuclear Information System (INIS)

Downar, T.J.

1987-01-01

A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback

Multi-dimensional rheology-based two-phase model for sediment transport and applications to sheet flow and pipeline scour

International Nuclear Information System (INIS)

Lee, Cheng-Hsien; Low, Ying Min; Chiew, Yee-Meng

2016-01-01

Sediment transport is fundamentally a two-phase phenomenon involving fluid and sediments; however, many existing numerical models are one-phase approaches, which are unable to capture the complex fluid-particle and inter-particle interactions. In the last decade, two-phase models have gained traction; however, there are still many limitations in these models. For example, several existing two-phase models are confined to one-dimensional problems; in addition, the existing two-dimensional models simulate only the region outside the sand bed. This paper develops a new three-dimensional two-phase model for simulating sediment transport in the sheet flow condition, incorporating recently published rheological characteristics of sediments. The enduring-contact, inertial, and fluid viscosity effects are considered in determining sediment pressure and stresses, enabling the model to be applicable to a wide range of particle Reynolds number. A k − ε turbulence model is adopted to compute the Reynolds stresses. In addition, a novel numerical scheme is proposed, thus avoiding numerical instability caused by high sediment concentration and allowing the sediment dynamics to be computed both within and outside the sand bed. The present model is applied to two classical problems, namely, sheet flow and scour under a pipeline with favorable results. For sheet flow, the computed velocity is consistent with measured data reported in the literature. For pipeline scour, the computed scour rate beneath the pipeline agrees with previous experimental observations. However, the present model is unable to capture vortex shedding; consequently, the sediment deposition behind the pipeline is overestimated. Sensitivity analyses reveal that model parameters associated with turbulence have strong influence on the computed results.

The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL

International Nuclear Information System (INIS)

Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.

1983-03-01

This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)

Electrical detection of spin transport in Si two-dimensional electron gas systems

Science.gov (United States)

Chang, Li-Te; Fischer, Inga Anita; Tang, Jianshi; Wang, Chiu-Yen; Yu, Guoqiang; Fan, Yabin; Murata, Koichi; Nie, Tianxiao; Oehme, Michael; Schulze, Jörg; Wang, Kang L.

2016-09-01

Spin transport in a semiconductor-based two-dimensional electron gas (2DEG) system has been attractive in spintronics for more than ten years. The inherent advantages of high-mobility channel and enhanced spin-orbital interaction promise a long spin diffusion length and efficient spin manipulation, which are essential for the application of spintronics devices. However, the difficulty of making high-quality ferromagnetic (FM) contacts to the buried 2DEG channel in the heterostructure systems limits the potential developments in functional devices. In this paper, we experimentally demonstrate electrical detection of spin transport in a high-mobility 2DEG system using FM Mn-germanosilicide (Mn(Si0.7Ge0.3)x) end contacts, which is the first report of spin injection and detection in a 2DEG confined in a Si/SiGe modulation doped quantum well structure (MODQW). The extracted spin diffusion length and lifetime are l sf = 4.5 μm and {τ }{{s}}=16 {{ns}} at 1.9 K respectively. Our results provide a promising approach for spin injection into 2DEG system in the Si-based MODQW, which may lead to innovative spintronic applications such as spin-based transistor, logic, and memory devices.

Mechanical transport in two-dimensional networks of fractures

International Nuclear Information System (INIS)

Endo, H.K.

1984-04-01

The objectives of this research are to evaluate directional mechanical transport parameters for anisotropic fracture systems, and to determine if fracture systems behave like equivalent porous media. The tracer experiments used to measure directional tortuosity, longitudinal geometric dispersivity, and hydraulic effective porosity are conducted with a uniform flow field and measurements are made from the fluid flowing within a test section where linear length of travel is constant. Since fluid flow and mechanical transport are coupled processes, the directional variations of specific discharge and hydraulic effective porosity are measured in regions with constant hydraulic gradients to evaluate porous medium equivalence for the two processes, respectively. If the fracture region behaves like an equivalent porous medium, the system has the following stable properties: (1) specific discharge is uniform in any direction and can be predicted from a permeability tensor; and (2) hydraulic effective porosity is directionally stable. Fracture systems with two parallel sets of continuous fractures satisfy criterion 1. However, in these systems hydraulic effective porosity is directionally dependent, and thus, criterion 2 is violated. Thus, for some fracture systems, fluid flow can be predicted using porous media assumptions, but it may not be possible to predict transport using porous media assumptions. Two discontinuous fracture systems were studied which satisfied both criteria. Hydraulic effective porosity for both systems has a value between rock effective porosity and total porosity. A length-density analysis (LDS) of Canadian fracture data shows that porous media equivalence for fluid flow and transport is likely when systems have narrow aperture distributions. 54 references, 90 figures, 7 tables

Two-dimensional Haar wavelet Collocation Method for the solution of Stationary Neutron Transport Equation in a homogeneous isotropic medium

International Nuclear Information System (INIS)

Patra, A.; Saha Ray, S.

2014-01-01

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: This paper emphasizes on finding the solution for a stationary transport equation using the technique of Haar wavelet Collocation Method (HWCM). Haar wavelet Collocation Method is efficient and powerful in solving wide class of linear and nonlinear differential equations. Recently Haar wavelet transform has gained the reputation of being a very effective tool for many practical applications. 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 for solution of the stationary Neutron Transport Equation in homogeneous isotropic medium. The proposed method is mathematically very simple, easy and fast. To demonstrate about the efficiency of the method, one test problem is discussed. It can be observed from the computational simulation that the numerical approximate solution is much closer to the exact solution

One dimensional neutron kinetics in the TRAC-BF1 code

International Nuclear Information System (INIS)

Weaver, W.L. III; Wagner, K.C.

1987-01-01

The TRAC-BWR code development program at the Idaho National Engineering Laboratory is developing a version of the TRAC code for the U.S. Nuclear Regulatory Commission (USNRC) to provide a best-estimate analysis capability for the simulation of postulated accidents in boiling water reactor (BWR) power systems and related experimental facilities. Recent development efforts in the TRAC-BWR program have focused on improving the computational efficiency through the incorporation of a hybrid Courant- limit-violating numerical solution scheme in the one-dimensional component models and on improving code accuracy through the development of a one-dimensional neutron kinetics model. Many other improvements have been incorporated into TRAC-BWR to improve code portability, accuracy, efficiency, and maintainability. This paper will describe the one- dimensional neutron kinetics model, the generation of the required input data for this model, and present results of the first calculations using the model

The use of non-dimensional representation of the solute transport equations

International Nuclear Information System (INIS)

Laurens, J.-M.

1988-07-01

This report presents the results obtained in a pilot investigation into the use of non-dimensional representations of the solute transport equations, so as to improve the efficiency of the PRA codes used by the DoE and its contractors. A reduced set of parameters was obtained for a single layer transport model. As expected, the response was shown to be highly sensitive on the new parameters. A faster convergence of the system was observed when the sampling technique used was changed to take into account the properties of the new parameters, such that uniform coverage of the reduced parameter hyperspace was achieved. (author)

Multi-dimensional Code Development for Safety Analysis of LMR

International Nuclear Information System (INIS)

Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

2006-08-01

A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

BOT3P: a mesh generation software package for the transport analysis codes Dort, Tort, Twodant, Threedant and MCNP

International Nuclear Information System (INIS)

Orsi, R.

2003-01-01

Bot3p consists of a set of standard Fortran 77 language programs that gives the users of the deterministic transport codes Dort and Tort some useful diagnostic tools to prepare and check the geometry of their input data files for both Cartesian and cylindrical geometries including graphical display modules. Bot3p produces at the same time the geometrical and material distribution data for the deterministic transport codes Twodant and Threedant and, only in three-dimensional (3D) Cartesian geometry, for the Monte Carlo Transport Code MCNP. This makes it possible to compare directly for the same geometry the effects stemming from the use of different data libraries and solution approaches on transport analysis results. Through the use of Bot3p, radiation transport problems with complex 3D geometrical structures can be modelled easily, as a relatively small amount of engineer-time is required and refinement is achieved by changing few parameters. This tool is useful for solving very large challenging problems. (author)

Development of a multi-dimensional realistic thermal-hydraulic system analysis code, MARS 1.3 and its verification

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-06-01

A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other

Simplified model for radioactive contaminant transport: the TRANSS code

International Nuclear Information System (INIS)

Simmons, C.S.; Kincaid, C.T.; Reisenauer, A.E.

1986-09-01

A simplified ground-water transport model called TRANSS was devised to estimate the rate of migration of a decaying radionuclide that is subject to sorption governed by a linear isotherm. Transport is modeled as a contaminant mass transmitted along a collection of streamlines constituting a streamtube, which connects a source release zone with an environmental arrival zone. The probability-weighted contaminant arrival distribution along each streamline is represented by an analytical solution of the one-dimensional advection-dispersion equation with constant velocity and dispersion coefficient. The appropriate effective constant velocity for each streamline is based on the exact travel time required to traverse a streamline with a known length. An assumption used in the model to facilitate the mathematical simplification is that transverse dispersion within a streamtube is negligible. Release of contaminant from a source is described in terms of a fraction-remaining curve provided as input information. However, an option included in the code is the calculation of a fraction-remaining curve based on four specialized release models: (1) constant release rate, (2) solubility-controlled release, (3) adsorption-controlled release, and (4) diffusion-controlled release from beneath an infiltration barrier. To apply the code, a user supplies only a certain minimal number of parameters: a probability-weighted list of travel times for streamlines, a local-scale dispersion coefficient, a sorption distribution coefficient, total initial radionuclide inventory, radioactive half-life, a release model choice, and size dimensions of the source. The code is intended to provide scoping estimates of contaminant transport and does not predict the evolution of a concentration distribution in a ground-water flow field. Moreover, the required travel times along streamlines must be obtained from a prior ground-water flow simulation

Gray and multigroup radiation transport models for two-dimensional binary stochastic media using effective opacities

International Nuclear Information System (INIS)

Olson, Gordon L.

2016-01-01

One-dimensional models for the transport of radiation through binary stochastic media do not work in multi-dimensions. Authors have attempted to modify or extend the 1D models to work in multidimensions without success. Analytic one-dimensional models are successful in 1D only when assuming greatly simplified physics. State of the art theories for stochastic media radiation transport do not address multi-dimensions and temperature-dependent physics coefficients. Here, the concept of effective opacities and effective heat capacities is found to well represent the ensemble averaged transport solutions in cases with gray or multigroup temperature-dependent opacities and constant or temperature-dependent heat capacities. In every case analyzed here, effective physics coefficients fit the transport solutions over a useful range of parameter space. The transport equation is solved with the spherical harmonics method with angle orders of n=1 and 5. Although the details depend on what order of solution is used, the general results are similar, independent of angular order. - Highlights: • Gray and multigroup radiation transport is done through 2D stochastic media. • Approximate models for the mean radiation field are found for all test problems. • Effective opacities are adjusted to fit the means of stochastic media transport. • Test problems include temperature dependent opacities and heat capacities • Transport solutions are done with angle orders n=1 and 5.

Computation of the bounce-average code

International Nuclear Information System (INIS)

Cutler, T.A.; Pearlstein, L.D.; Rensink, M.E.

1977-01-01

The bounce-average computer code simulates the two-dimensional velocity transport of ions in a mirror machine. The code evaluates and bounce-averages the collision operator and sources along the field line. A self-consistent equilibrium magnetic field is also computed using the long-thin approximation. Optionally included are terms that maintain μ, J invariance as the magnetic field changes in time. The assumptions and analysis that form the foundation of the bounce-average code are described. When references can be cited, the required results are merely stated and explained briefly. A listing of the code is appended

Development of the three dimensional flow model in the SPACE code

International Nuclear Information System (INIS)

Oh, Myung Taek; Park, Chan Eok; Kim, Shin Whan

2014-01-01

SPACE (Safety and Performance Analysis CodE) is a nuclear plant safety analysis code, which has been developed in the Republic of Korea through a joint research between the Korean nuclear industry and research institutes. The SPACE code has been developed with multi-dimensional capabilities as a requirement of the next generation safety code. It allows users to more accurately model the multi-dimensional flow behavior that can be exhibited in components such as the core, lower plenum, upper plenum and downcomer region. Based on generalized models, the code can model any configuration or type of fluid system. All the geometric quantities of mesh are described in terms of cell volume, centroid, face area, and face center, so that it can naturally represent not only the one dimensional (1D) or three dimensional (3D) Cartesian system, but also the cylindrical mesh system. It is possible to simulate large and complex domains by modelling the complex parts with a 3D approach and the rest of the system with a 1D approach. By 1D/3D co-simulation, more realistic conditions and component models can be obtained, providing a deeper understanding of complex systems, and it is expected to overcome the shortcomings of 1D system codes. (author)

Modification of the finite element heat and mass transfer code (FEHM) to model multicomponent reactive transport

International Nuclear Information System (INIS)

Viswanathan, H.S.

1996-08-01

The finite element code FEHMN, developed by scientists at Los Alamos National Laboratory (LANL), is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developing hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent Kd model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The new chemical capabilities of FEHMN are illustrated by using Los Alamos National Laboratory's site scale model of Yucca Mountain to model two-dimensional, vadose zone 14 C transport. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also prove that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies

Los Alamos radiation transport code system on desktop computing platforms

International Nuclear Information System (INIS)

Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.

1990-01-01

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines

GANTRAS - a system of codes for the solution of the multigroup transport equation with a rigorous treatment of anisotropic neutron scattering

International Nuclear Information System (INIS)

Schwenk-Ferrero, A.

1986-11-01

GANTRAS is a system of codes for neutron transport calculations in which the anisotropy of elastic and inelastic (including (n,n'x)-reactions) scattering is fully taken into account. This is achieved by employing a rigorous method, so-called I * -method, to represent the scattering term of the transport equation and with the use of double-differential cross-sections for the description of the emission of secondary neutrons. The I * -method was incorporated into the conventional transport code ONETRAN. The ONETRAN subroutines were modified for the new purpose. An implementation of the updated version ANTRA1 was accomplished for plane and spherical geometry. ANTRA1 was included in GANTRAS and linked to another modules which prepare angle-dependent transfer matrices. The GANTRAS code consists of three modules: 1. The CROMIX code which calculates the macroscopic transfer matrices for mixtures on the base of microscopic nuclide-dependent data. 2. The ATP code which generates discretized angular transfer probabilities (i.e. discretizes the I * -function). 3. The ANTRA1 code to perform S N transport calculations in one-dimensional plane and spherical geometries. This structure of GANTRAS allows to accommodate the system to various transport problems. (orig.) [de

Two-dimensional carbon crystals. Electrical transport in single- and double-layer graphene; Zweidimensionale Kohlenstoffkristalle. Elektrischer Transport in Einzel- und Doppellagen-Graphen

Energy Technology Data Exchange (ETDEWEB)

Schmidt, Hennrik

2012-02-03

In his work atomically thin layers of carbon, socalled graphene, are investigated. These systems exhibit outstanding electronic properties which are analysed using magnetotransport measurements. For this purpose, different types of samples are prepared, analysed and discussed. In addition to conventional single layer and single crystal bilayer systems, folded flakes with twisted planes are examined. Since monolayer graphene is a two dimensional crystal in which every atom sits at the surface, it is very sensitive to any type of perturbation. Three different cases are investigated: Firstly, dopants are removed from the surface and the change in transport properties is monitored. Secondly, the regime of small carrier concentrations is used to observe field induced recharging of inhomogeneities. Thirdly, an atomic force microscope is used to alter the graphene itself in a defined region. The implications of this modification are again investigated using magnetotransport measurements. The influence of one layer on another one is studied in decoupled two layer samples. A folded sample with separatly contacted layers is used to show transport through the folded region. For jointly contacted layers parallel transport measurements are performed to analyse screening effects of an applied electric field and substrate influence. The interaction of the two layers is shown by a significant reduction of the Fermivelocity.

An improved version of the MICROX-2 code

Energy Technology Data Exchange (ETDEWEB)

Mathews, D. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1997-11-01

The MICROX-2 code prepares broad group neutron cross sections for use in diffusion- and/or transport-theory codes from an input library of fine group and pointwise cross sections. The neutron weighting spectrum is obtained by solving the B{sub 1} neutron balance equations at about 10000 energies in a one-dimensional (planar, spherical or cylindrical), two-region unit cell. The regions are coupled by collision probabilities based upon spatially flat neutron emission. Energy dependent Dancoff factors and bucklings correct the one-dimensional calculations for multi-dimensional lattice effects. A critical buckling search option is also included. The inner region may include two different types of fuel particles (grains). This report describes the present PSI FORTRAN 90 version of the MICROX-2 code which operates on CRAY computers and IBM PC`s. The equations which are solved in the various energy ranges are given along with descriptions of various changes that have been made in the present PSI version of the code. A completely re-written description of the user input is also included. (author) 7 figs., 4 tabs., 59 refs.

Modified Three-Dimensional Multicarrier Optical Prime Codes

Directory of Open Access Journals (Sweden)

Rajesh Yadav

2016-01-01

Full Text Available We propose a mathematical model for novel three-dimensional multicarrier optical codes in terms of wavelength/time/space based on the prime sequence algorithm. The proposed model has been extensively simulated on MATLAB for prime numbers (P to analyze the performance of code in terms of autocorrelation and cross-correlation. The simulated outcome resembles the mathematical model and gives better results over other methods available in the literature as far as autocorrelation and cross-correlation are concerned. The proposed 3D optical codes are more efficient in terms of cardinality, improved security, and providing quality of services.

Two-Dimensional Tellurene as Excellent Thermoelectric Material

KAUST Repository

Sharma, Sitansh; Singh, Nirpendra; Schwingenschlö gl, Udo

2018-01-01

We study the thermoelectric properties of two-dimensional tellurene by first-principles calculations and semiclassical Boltzmann transport theory. The HSE06 hybrid functional results in a moderate direct band gap of 1.48 eV at the Γ point. A high

Two-dimensional electron flow in pulsed power transmission lines and plasma opening switches

International Nuclear Information System (INIS)

Church, B.W.; Longcope, D.W.; Ng, C.K.; Sudan, R.N.

1991-01-01

The operation of magnetically insulated transmission lines (MITL) and the interruption of current in a plasma opening switch (POS) are determined by the physics of the electrons emitted by the cathode surface. A mathematical model describes the self-consistent two-dimensional flow of an electron fluid. A finite element code, FERUS, has been developed to solve the two equations which describe Poisson's and Ampere's law in two dimensions. The solutions from this code are obtained for parameters where the electron orbits are considerably modified by the self-magnetic field of the current. Next, the self-insulated electron flow in a MITL with a step change in cross-section is studied using a conventional two-dimensional fully electromagnetic particle-in-cell code, MASK. The equations governing two-dimensional quasi-static electron flow are solved numerically by a third technique which is suitable for predicting current interruption in a POS. The object of the study is to determine the critical load impedance, Z CL , required for current interruption for a given applied voltage, cathode voltage and plasma length. (author). 9 refs, 5 figs

SSS: A code for computing one dimensional shock and detonation wave propagation

International Nuclear Information System (INIS)

Sun Chengwei

1986-01-01

The one-dimensional hydrodynamic code SSS for shock and detonation wave propagation in inert and reactive media is described. The elastic-plastic-hydrodynamic model and four burn techniques (the Arrhenius law, C-J volume, sharp shock and Forest Fire) are used. There are HOM and JWL options for the state equation of detonation products. Comparing with the SIN code published by LANL, the SSS code has several new options: laser effects, blast waves, diverging and instantaneous detonation waves with arbitrary initiation positions. Two examples are given to compare the SSS and SIN calculations with the experimental data

A description of the two-dimensional combustion code FLARE

International Nuclear Information System (INIS)

Martin, D.

1986-07-01

This report gives details of the computer code FLARE. The model used for the turbulent combustion of premixed gases is described. Details of the numerical scheme used to solve the resulting equations are discussed. The input and output for the code are also described. Details of the coding are given in the Appendices together with sample input and output. (author)

Modelling of the spent fuel heat-up in the spent fuel pools using one-dimensional system codes and CFD codes

Energy Technology Data Exchange (ETDEWEB)

Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety

2017-07-15

The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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

Space applications of the MITS electron-photon Monte Carlo transport code system

International Nuclear Information System (INIS)

Kensek, R.P.; Lorence, L.J.; Halbleib, J.A.; Morel, J.E.

1996-01-01

The MITS multigroup/continuous-energy electron-photon Monte Carlo transport code system has matured to the point that it is capable of addressing more realistic three-dimensional adjoint applications. It is first employed to efficiently predict point doses as a function of source energy for simple three-dimensional experimental geometries exposed to simulated uniform isotropic planar sources of monoenergetic electrons up to 4.0 MeV. Results are in very good agreement with experimental data. It is then used to efficiently simulate dose to a detector in a subsystem of a GPS satellite due to its natural electron environment, employing a relatively complex model of the satellite. The capability for survivability analysis of space systems is demonstrated, and results are obtained with and without variance reduction

A Two-Temperature Open-Source CFD Model for Hypersonic Reacting Flows, Part Two: Multi-Dimensional Analysis †

OpenAIRE

Vincent Casseau; Daniel E. R. Espinoza; Thomas J. Scanlon; Richard E. Brown

2016-01-01

hy2Foam is a newly-coded open-source two-temperature computational fluid dynamics (CFD) solver that has previously been validated for zero-dimensional test cases. It aims at (1) giving open-source access to a state-of-the-art hypersonic CFD solver to students and researchers; and (2) providing a foundation for a future hybrid CFD-DSMC (direct simulation Monte Carlo) code within the OpenFOAM framework. This paper focuses on the multi-dimensional verification of hy2Foam and firstly describes th...

Development of a new EMP code at LANL

Science.gov (United States)

Colman, J. J.; Roussel-Dupré, R. A.; Symbalisty, E. M.; Triplett, L. A.; Travis, B. J.

2006-05-01

A new code for modeling the generation of an electromagnetic pulse (EMP) by a nuclear explosion in the atmosphere is being developed. The source of the EMP is the Compton current produced by the prompt radiation (γ-rays, X-rays, and neutrons) of the detonation. As a first step in building a multi- dimensional EMP code we have written three kinetic codes, Plume, Swarm, and Rad. Plume models the transport of energetic electrons in air. The Plume code solves the relativistic Fokker-Planck equation over a specified energy range that can include ~ 3 keV to 50 MeV and computes the resulting electron distribution function at each cell in a two dimensional spatial grid. The energetic electrons are allowed to transport, scatter, and experience Coulombic drag. Swarm models the transport of lower energy electrons in air, spanning 0.005 eV to 30 keV. The swarm code performs a full 2-D solution to the Boltzmann equation for electrons in the presence of an applied electric field. Over this energy range the relevant processes to be tracked are elastic scattering, three body attachment, two body attachment, rotational excitation, vibrational excitation, electronic excitation, and ionization. All of these occur due to collisions between the electrons and neutral bodies in air. The Rad code solves the full radiation transfer equation in the energy range of 1 keV to 100 MeV. It includes effects of photo-absorption, Compton scattering, and pair-production. All of these codes employ a spherical coordinate system in momentum space and a cylindrical coordinate system in configuration space. The "z" axis of the momentum and configuration spaces is assumed to be parallel and we are currently also assuming complete spatial symmetry around the "z" axis. Benchmarking for each of these codes will be discussed as well as the way forward towards an integrated modern EMP code.

The CNCSN-2: One, two-and three-dimensional coupled neutral and charged particle discrete ordinates code system

International Nuclear Information System (INIS)

Voloschenko, A. M.; Gukov, S. V.; Russkov, A. A.; Gurevich, M. I.; Shkarovsky, D. A.; Kryuchkov, V. P.; Sumaneev, O. V.; Dubinin, A. A.

2009-01-01

KATRIN, KASKAD-Sand ROZ-6 codes solve the multigroup transport equation for neutrons, photons and charged particles in 3D. BOT3P-5., ConDat can be used as preprocessor. ARVES-2.5, a cross-section preprocessor (the package of utilities for operating with the cross section file in FMAC-M format) is included. Auxiliary codes MIXERM, CEPXS-BFP, CEPXS-BFP, SADCO-2.4 and CNCSN-2 are used

Equatorial spread F studies using SAMI3 with two-dimensional and three-dimensional electrostatics

Directory of Open Access Journals (Sweden)

H. C. Aveiro

2013-12-01

Full Text Available This letter presents a study of equatorial F region irregularities using the NRL SAMI3/ESF model, comparing results using a two-dimensional (2-D and a three-dimensional (3-D electrostatic potential solution. For the 3-D potential solution, two cases are considered for parallel plasma transport: (1 transport based on the parallel ambipolar field, and (2 transport based on the parallel electric field. The results show that the growth rate of the generalized Rayleigh–Taylor instability is not affected by the choice of the potential solution. However, differences are observed in the structures of the irregularities between the 2-D and 3-D solutions. Additionally, the plasma velocity along the geomagnetic field computed using the full 3-D solution shows complex structures that are not captured by the simplified model. This points out that only the full 3-D model is able to fully capture the complex physics of the equatorial F region.

Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET)

International Nuclear Information System (INIS)

Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.

1994-02-01

The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses that individuals may have received from operations at the Hanford Site since 1944. This report deals specifically with the atmospheric transport model, Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). RATCHET is a major rework of the MESOILT2 model used in the first phase of the HEDR Project; only the bookkeeping framework escaped major changes. Changes to the code include (1) significant changes in the representation of atmospheric processes and (2) incorporation of Monte Carlo methods for representing uncertainty in input data, model parameters, and coefficients. To a large extent, the revisions to the model are based on recommendations of a peer working group that met in March 1991. Technical bases for other portions of the atmospheric transport model are addressed in two other documents. This report has three major sections: a description of the model, a user's guide, and a programmer's guide. These sections discuss RATCHET from three different perspectives. The first provides a technical description of the code with emphasis on details such as the representation of the model domain, the data required by the model, and the equations used to make the model calculations. The technical description is followed by a user's guide to the model with emphasis on running the code. The user's guide contains information about the model input and output. The third section is a programmer's guide to the code. It discusses the hardware and software required to run the code. The programmer's guide also discusses program structure and each of the program elements

Two-dimensional ion effects in relativistic diodes

International Nuclear Information System (INIS)

Poukey, J.W.

1975-01-01

In relativistic diodes, ions are emitted from the anode plasma. The effects and properties of these ions are studied via a two-dimensional particle simulation code. The space charge of these ions enhances the electron emission, and this additional current (including that of the ions, themselves) aids in obtaining superpinched electron beams for use in pellet fusion studies. (U.S.)

Development of One Dimensional Hyperbolic Coupled Solver for Two-Phase Flows

International Nuclear Information System (INIS)

Kim, Eoi Jin; Kim, Jong Tae; Jeong, Jae June

2008-08-01

The purpose of this study is a code development for one dimensional two-phase two-fluid flows. In this study, the computations of two-phase flow were performed by using the Roe scheme which is one of the upwind schemes. The upwind scheme is widely used in the computational fluid dynamics because it can capture discontinuities clearly such as a shock. And this scheme is applicable to multi-phase flows by the extension methods which were developed by Toumi, Stadtke, etc. In this study, the extended Roe upwind scheme by Toumi for two-phase flow was implemented in the one-dimensional code. The scheme was applied to a shock tube problem and a water faucet problem. This numerical method seems efficient for non oscillating solutions of two phase flow problems, and also capable for capturing discontinuities

Development of One Dimensional Hyperbolic Coupled Solver for Two-Phase Flows

Energy Technology Data Exchange (ETDEWEB)

Kim, Eoi Jin; Kim, Jong Tae; Jeong, Jae June

2008-08-15

The purpose of this study is a code development for one dimensional two-phase two-fluid flows. In this study, the computations of two-phase flow were performed by using the Roe scheme which is one of the upwind schemes. The upwind scheme is widely used in the computational fluid dynamics because it can capture discontinuities clearly such as a shock. And this scheme is applicable to multi-phase flows by the extension methods which were developed by Toumi, Stadtke, etc. In this study, the extended Roe upwind scheme by Toumi for two-phase flow was implemented in the one-dimensional code. The scheme was applied to a shock tube problem and a water faucet problem. This numerical method seems efficient for non oscillating solutions of two phase flow problems, and also capable for capturing discontinuities.

Huemul: a two dimensional multigroup collision probability code for general geometries

International Nuclear Information System (INIS)

Calabrese, C.R.; Grant, C.R.

1990-01-01

The control rod calculation and the necessity of having a 2-D transport code able to calculate geometries as different as pool reactor or power reactor control rods resulted in the development of a new tool according to these requirements. This new tool permits a 2-D spatial representation, and the calculation mesh is formed by circumferential arcs segments not necessarily parallel to the coordinate axis. It includes the possibility of considering boundary conditions in the form of an albedo matrix (J + /J - ) as well as different external currents for each face of the model. It also allows an arbitrary number of energy groups compatible with computer limitations. These possibilities make HUEMUL a useful tool for a great variety of control rod or supercell type calculations. HUEMUL has been tested with copper activity measurements performed in the Canadian D2O facility ZED-2 with stainless-steel adjuster rods obtaining a very good agreement (better than 2%). Also manganese activity measurements in RA-2 pool reactor were used to compare calculated and measured values inside a MTR fuel element calculations, (better than 1%). Comparisons with results from the WIMS code for a light water cell with 3% enriched UO 2 has also shown a very good agreement in fluxes and multiplication constant (better than 1.5% in fluxes and 50 pcm in k-infinity). (Author) [es

Benchmarking NNWSI flow and transport codes: COVE 1 results

International Nuclear Information System (INIS)

Hayden, N.K.

1985-06-01

The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs

Transmission probability method for solving neutron transport equation in three-dimensional triangular-z geometry

Energy Technology Data Exchange (ETDEWEB)

Liu Guoming [Department of Nuclear Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi 710049 (China)], E-mail: gmliusy@gmail.com; Wu Hongchun; Cao Liangzhi [Department of Nuclear Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi 710049 (China)

2008-09-15

This paper presents a transmission probability method (TPM) to solve the neutron transport equation in three-dimensional triangular-z geometry. The source within the mesh is assumed to be spatially uniform and isotropic. At the mesh surface, the constant and the simplified P{sub 1} approximation are invoked for the anisotropic angular flux distribution. Based on this model, a code TPMTDT is encoded. It was verified by three 3D Takeda benchmark problems, in which the first two problems are in XYZ geometry and the last one is in hexagonal-z geometry, and an unstructured geometry problem. The results of the present method agree well with those of Monte-Carlo calculation method and Spherical Harmonics (P{sub N}) method.

Progress in nuclear well logging modeling using deterministic transport codes

International Nuclear Information System (INIS)

Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.

2002-01-01

Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-01

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

Geometry system used in the General Monte Carlo transport code SPARTAN

International Nuclear Information System (INIS)

Bending, R.C.; Easter, P.G.

1974-01-01

The geometry routines used in the general-purpose, three-dimensional particle transport code SPARTAN are described. The code is designed to deal with the very complex geometries encountered in lattice cell and fuel handling calculations, health physics, and shielding problems. Regions of the system being studied may be represented by simple shapes (spheres, cylinders, and so on) or by multinomial surfaces of any order, and many simple shapes may be combined to make up a complex layout. The geometry routines are designed to allow the program to carry out a number of tasks (such as sampling for a random point or tracking a path through several regions) in any order, so that the use of the routines is not restricted to a particular tracking or scoring method. Routines for reading, checking, and printing the data are included. (U.S.)

Code-To-Code Benchmarking Of The Porflow And GoldSim Contaminant Transport Models Using A Simple 1-D Domain - 11191

International Nuclear Information System (INIS)

Hiergesell, R.; Taylor, G.

2010-01-01

An investigation was conducted to compare and evaluate contaminant transport results of two model codes, GoldSim and Porflow, using a simple 1-D string of elements in each code. Model domains were constructed to be identical with respect to cell numbers and dimensions, matrix material, flow boundary and saturation conditions. One of the codes, GoldSim, does not simulate advective movement of water; therefore the water flux term was specified as a boundary condition. In the other code, Porflow, a steady-state flow field was computed and contaminant transport was simulated within that flow-field. The comparisons were made solely in terms of the ability of each code to perform contaminant transport. The purpose of the investigation was to establish a basis for, and to validate follow-on work that was conducted in which a 1-D GoldSim model developed by abstracting information from Porflow 2-D and 3-D unsaturated and saturated zone models and then benchmarked to produce equivalent contaminant transport results. A handful of contaminants were selected for the code-to-code comparison simulations, including a non-sorbing tracer and several long- and short-lived radionuclides exhibiting both non-sorbing to strongly-sorbing characteristics with respect to the matrix material, including several requiring the simulation of in-growth of daughter radionuclides. The same diffusion and partitioning coefficients associated with each contaminant and the half-lives associated with each radionuclide were incorporated into each model. A string of 10-elements, having identical spatial dimensions and properties, were constructed within each code. GoldSim's basic contaminant transport elements, Mixing cells, were utilized in this construction. Sand was established as the matrix material and was assigned identical properties (e.g. bulk density, porosity, saturated hydraulic conductivity) in both codes. Boundary conditions applied included an influx of water at the rate of 40 cm/yr at one

Diverse anisotropy of phonon transport in two-dimensional group IV-VI compounds: A comparative study

Science.gov (United States)

Qin, Guangzhao; Qin, Zhenzhen; Fang, Wu-Zhang; Zhang, Li-Chuan; Yue, Sheng-Ying; Yan, Qing-Bo; Hu, Ming; Su, Gang

2016-05-01

New classes of two-dimensional (2D) materials beyond graphene, including layered and non-layered, and their heterostructures, are currently attracting increasing interest due to their promising applications in nanoelectronics, optoelectronics and clean energy, where thermal transport is a fundamental physical parameter. In this paper, we systematically investigated the phonon transport properties of the 2D orthorhombic group IV-VI compounds of GeS, GeSe, SnS and SnSe by solving the Boltzmann transport equation (BTE) based on first-principles calculations. Despite their similar puckered (hinge-like) structure along the armchair direction as phosphorene, the four monolayer compounds possess diverse anisotropic properties in many aspects, such as phonon group velocity, Young's modulus and lattice thermal conductivity (κ), etc. Especially, the κ along the zigzag and armchair directions of monolayer GeS shows the strongest anisotropy while monolayer SnS and SnSe show almost isotropy in phonon transport. The origin of the diverse anisotropy is fully studied and the underlying mechanism is discussed in details. With limited size, the κ could be effectively lowered, and the anisotropy could be effectively modulated by nanostructuring, which would extend the applications to nanoscale thermoelectrics and thermal management. Our study offers fundamental understanding of the anisotropic phonon transport properties of 2D materials, and would be of significance for further study, modulation and applications in emerging technologies.

Performance testing of the sediment-contaminant transport model, SERATRA, at different rivers

International Nuclear Information System (INIS)

Onishi, Y.; Yabusaki, S.B.; Kincaid, C.T.

1982-04-01

Mathematical models of sediment-contaminant migration in surface water must account for transport, intermedia transfer, decay and degradation, and transformation processes. The unsteady, two dimensional, sediment-contaminant transport code, SERATRA (Onishi, Schreiber and Codell 1980) includes these mechanisms. To assess the accuracy of SERATRA to simulate the sediment-contaminant transport and fate processes, the code was tested against one-dimensional analytical solutions, checked for its mass balance, and applied to field sites. The field application cases ranged from relatively simple, steady conditions to unsteady, nonuniform conditions for large, intermediate, and small rivers. It was found that SERATRA is capable of simulating sediment-contaminant transport under a wide range of conditions

Intrinsic two-dimensional states on the pristine surface of tellurium

Science.gov (United States)

Li, Pengke; Appelbaum, Ian

2018-05-01

Atomic chains configured in a helical geometry have fascinating properties, including phases hosting localized bound states in their electronic structure. We show how the zero-dimensional state—bound to the edge of a single one-dimensional helical chain of tellurium atoms—evolves into two-dimensional bands on the c -axis surface of the three-dimensional trigonal bulk. We give an effective Hamiltonian description of its dispersion in k space by exploiting confinement to a virtual bilayer, and elaborate on the diminished role of spin-orbit coupling. These intrinsic gap-penetrating surface bands were neglected in the interpretation of seminal experiments, where two-dimensional transport was otherwise attributed to extrinsic accumulation layers.

DNA barcode goes two-dimensions: DNA QR code web server.

Science.gov (United States)

Liu, Chang; Shi, Linchun; Xu, Xiaolan; Li, Huan; Xing, Hang; Liang, Dong; Jiang, Kun; Pang, Xiaohui; Song, Jingyuan; Chen, Shilin

2012-01-01

The DNA barcoding technology uses a standard region of DNA sequence for species identification and discovery. At present, "DNA barcode" actually refers to DNA sequences, which are not amenable to information storage, recognition, and retrieval. Our aim is to identify the best symbology that can represent DNA barcode sequences in practical applications. A comprehensive set of sequences for five DNA barcode markers ITS2, rbcL, matK, psbA-trnH, and CO1 was used as the test data. Fifty-three different types of one-dimensional and ten two-dimensional barcode symbologies were compared based on different criteria, such as coding capacity, compression efficiency, and error detection ability. The quick response (QR) code was found to have the largest coding capacity and relatively high compression ratio. To facilitate the further usage of QR code-based DNA barcodes, a web server was developed and is accessible at http://qrfordna.dnsalias.org. The web server allows users to retrieve the QR code for a species of interests, convert a DNA sequence to and from a QR code, and perform species identification based on local and global sequence similarities. In summary, the first comprehensive evaluation of various barcode symbologies has been carried out. The QR code has been found to be the most appropriate symbology for DNA barcode sequences. A web server has also been constructed to allow biologists to utilize QR codes in practical DNA barcoding applications.

DNA barcode goes two-dimensions: DNA QR code web server.

Directory of Open Access Journals (Sweden)

Chang Liu

Full Text Available The DNA barcoding technology uses a standard region of DNA sequence for species identification and discovery. At present, "DNA barcode" actually refers to DNA sequences, which are not amenable to information storage, recognition, and retrieval. Our aim is to identify the best symbology that can represent DNA barcode sequences in practical applications. A comprehensive set of sequences for five DNA barcode markers ITS2, rbcL, matK, psbA-trnH, and CO1 was used as the test data. Fifty-three different types of one-dimensional and ten two-dimensional barcode symbologies were compared based on different criteria, such as coding capacity, compression efficiency, and error detection ability. The quick response (QR code was found to have the largest coding capacity and relatively high compression ratio. To facilitate the further usage of QR code-based DNA barcodes, a web server was developed and is accessible at http://qrfordna.dnsalias.org. The web server allows users to retrieve the QR code for a species of interests, convert a DNA sequence to and from a QR code, and perform species identification based on local and global sequence similarities. In summary, the first comprehensive evaluation of various barcode symbologies has been carried out. The QR code has been found to be the most appropriate symbology for DNA barcode sequences. A web server has also been constructed to allow biologists to utilize QR codes in practical DNA barcoding applications.

Magneto-transport studies on curved two-dimensional electron gases in InGaAs-microscrolls

International Nuclear Information System (INIS)

Schumacher, O.

2007-01-01

In this thesis magneto-resistance studies on evenly curved two-dimensional electron systems in cylindric geometry are presented and discussed. A principle first introduced by Prinz and co-workers in 1998 enables us to roll up thin semiconductor layer systems by taking advantage of internal elastic strain. The radius of such a semiconductor tube can be adjusted ranging from a few nanometers up to several micrometers. The tubes' shape and place on the substrate can be defined by lithographic methods which are presented in this work. Furthermore, we show rolled-up structures containing a two-dimensional electron system in the tube wall. With a special lithographic procedure we are able to structure, to contact and to roll up these 2D-electron-gases in Hall geometry. As a result, a cylindric two-dimensional electron system is produced, which experiences a modulation of the perpendicular magnetic field component. The radius of curvature of our structures is about 10 μm, the carrier mobility is optimized to values up to 125,000 cm 2 /Vs. In transport experiments on curved Hall bars containing two dimensional electron systems two Hall bar orientations, with respect to the curvature, may be distinguished. In this work both orientations, i.e. with a Hall bar along the tube curvature as well as a Hall bar along the tube axis, are presented and discussed. Measurements on Hall bars along the curvature show signatures in the longitudinal resistance, which can be understood with the help of the Landauer-Buettiker-formalism and the model of magnetic barriers. For Hall bars oriented along the tube axis the perpendicular magnetic field component averaged over the width of the bar defines the minimum position of the Shubnikov-de Haas-oscillations as well as the slope of the Hall resistance. Furthermore, measurements on so-called van the Pauw-lamellas are presented. In this geometry the magneto-resistance shows a slope which refers to highly mobile conditions at the zero crossing of

A comparison of two nodal codes : Advanced nodal code (ANC) and analytic function expansion nodal (AFEN) code

International Nuclear Information System (INIS)

Chung, S.K.; Hah, C.J.; Lee, H.C.; Kim, Y.H.; Cho, N.Z.

1996-01-01

Modern nodal methods usually employs the transverse integration technique in order to reduce a multi-dimensional diffusion equation to one-dimensional diffusion equations. The use of the transverse integration technique requires two major approximations such as a transverse leakage approximation and a one-dimensional flux approximation. Both the transverse leakage and the one-dimensional flux are approximated by polynomials. ANC (Advanced Nodal Code) developed by Westinghouse employs a modern nodal expansion method for the flux calculation, the equivalence theory for the homogenization error reduction and a group theory for pin power recovery. Unlike the conventional modern nodal methods, AFEN (Analytic Function Expansion Nodal) method expands homogeneous flux distributions within a node into non-separable analytic basis functions, which eliminate two major approximations of the modern nodal methods. A comparison study of AFEN with ANC has been performed to see the applicability of AFEN to commercial PWR and different types of reactors such as MOX fueled reactor. The qualification comparison results demonstrate that AFEN methodology is accurate enough to apply for commercial PWR analysis. The results show that AFEN provides very accurate results (core multiplication factor and assembly power distribution) for cores that exhibit strong flux gradients as in a MOX loaded core. (author)

Revealing origin of quasi-one dimensional current transport in defect rich two dimensional materials

International Nuclear Information System (INIS)

Lotz, Mikkel R.; Boll, Mads; Bøggild, Peter; Petersen, Dirch H.; Hansen, Ole; Kjær, Daniel

2014-01-01

The presence of defects in graphene have for a long time been recognized as a bottleneck for its utilization in electronic and mechanical devices. We recently showed that micro four-point probes may be used to evaluate if a graphene film is truly 2D or if defects in proximity of the probe will lead to a non-uniform current flow characteristic of lower dimensionality. In this work, simulations based on a finite element method together with a Monte Carlo approach are used to establish the transition from 2D to quasi-1D current transport, when applying a micro four-point probe to measure on 2D conductors with an increasing amount of line-shaped defects. Clear 2D and 1D signatures are observed at low and high defect densities, respectively, and current density plots reveal the presence of current channels or branches in defect configurations yielding 1D current transport. A strong correlation is found between the density filling factor and the simulation yield, the fraction of cases with 1D transport and the mean sheet conductance. The upper transition limit is shown to agree with the percolation threshold for sticks. Finally, the conductance of a square sample evaluated with macroscopic edge contacts is compared to the micro four-point probe conductance measurements and we find that the micro four-point probe tends to measure a slightly higher conductance in samples containing defects

One-dimensional radionuclide transport under time-varying conditions

International Nuclear Information System (INIS)

Gelbard, F.; Olague, N.E.; Longsine, D.E.

1990-01-01

This paper discusses new analytical and numerical solutions presented for one-dimensional radionuclide transport under time-varying fluid-flow conditions including radioactive decay. The analytical solution assumes that all radionuclides have identical retardation factors, and is limited to instantaneous releases. The numerical solution does not have these limitations, but is tested against the limiting case given for the analytical solution. Reasonable agreement between the two solutions was found. Examples are given for the transport of a three-member radionuclide chain transported over distances and flow rates comparable to those reported for Yucca Mountain, the proposed disposal site for high-level nuclear waste

Interfacial and Wall Transport Models for SPACE-CAP Code

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun

2009-01-01

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code

Interfacial and Wall Transport Models for SPACE-CAP Code

Energy Technology Data Exchange (ETDEWEB)

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

2009-10-15

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

Assessment of RELAP5-3D copyright using data from two-dimensional RPI flow tests

International Nuclear Information System (INIS)

Davis, C.B.

1998-01-01

The capability of the RELAP5-3D copyright computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code's logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved

Two-terminal video coding.

Science.gov (United States)

Yang, Yang; Stanković, Vladimir; Xiong, Zixiang; Zhao, Wei

2009-03-01

Following recent works on the rate region of the quadratic Gaussian two-terminal source coding problem and limit-approaching code designs, this paper examines multiterminal source coding of two correlated, i.e., stereo, video sequences to save the sum rate over independent coding of both sequences. Two multiterminal video coding schemes are proposed. In the first scheme, the left sequence of the stereo pair is coded by H.264/AVC and used at the joint decoder to facilitate Wyner-Ziv coding of the right video sequence. The first I-frame of the right sequence is successively coded by H.264/AVC Intracoding and Wyner-Ziv coding. An efficient stereo matching algorithm based on loopy belief propagation is then adopted at the decoder to produce pixel-level disparity maps between the corresponding frames of the two decoded video sequences on the fly. Based on the disparity maps, side information for both motion vectors and motion-compensated residual frames of the right sequence are generated at the decoder before Wyner-Ziv encoding. In the second scheme, source splitting is employed on top of classic and Wyner-Ziv coding for compression of both I-frames to allow flexible rate allocation between the two sequences. Experiments with both schemes on stereo video sequences using H.264/AVC, LDPC codes for Slepian-Wolf coding of the motion vectors, and scalar quantization in conjunction with LDPC codes for Wyner-Ziv coding of the residual coefficients give a slightly lower sum rate than separate H.264/AVC coding of both sequences at the same video quality.

Optix: A Monte Carlo scintillation light transport code

Energy Technology Data Exchange (ETDEWEB)

Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)

2014-02-11

The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.

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)

Development of general-purpose particle and heavy ion transport monte carlo code

International Nuclear Information System (INIS)

Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji

2002-01-01

The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)

Development of three dimensional transient analysis code STTA for SCWR core

International Nuclear Information System (INIS)

Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping

2015-01-01

Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation

Two-point model for divertor transport

International Nuclear Information System (INIS)

Galambos, J.D.; Peng, Y.K.M.

1984-04-01

Plasma transport along divertor field lines was investigated using a two-point model. This treatment requires considerably less effort to find solutions to the transport equations than previously used one-dimensional (1-D) models and is useful for studying general trends. It also can be a valuable tool for benchmarking more sophisticated models. The model was used to investigate the possibility of operating in the so-called high density, low temperature regime

THIDA: code system for calculation of the exposure dose rate around a fusion device

International Nuclear Information System (INIS)

Iida, Hiromasa; Igarashi, Masahito.

1978-12-01

A code system THIDA has been developed for calculation of the exposure dose rates around a fusion device. It consists of the following: one- and two-dimensional discrete ordinate transport codes; induced activity calculation code; activation chain, activation cross section, radionuclide gamma-ray energy/intensity and gamma-ray group constant files; and gamma ray flux to exposure dose rate conversion coefficients. (author)

Two-dimensional model of coupled heat and moisture transport in frost-heaving soils

International Nuclear Information System (INIS)

Guymon, G.L.; Berg, R.L.; Hromadka, T.V.

1984-01-01

A two-dimensional model of coupled heat and moisture flow in frost-heaving soils is developed based upon well known equations of heat and moisture flow in soils. Numerical solution is by the nodal domain integration method which includes the integrated finite difference and the Galerkin finite element methods. Solution of the phase change process is approximated by an isothermal approach and phenomenological equations are assumed for processes occurring in freezing or thawing zones. The model has been verified against experimental one-dimensional freezing soil column data and experimental two-dimensional soil thawing tank data as well as two-dimensional soil seepage data. The model has been applied to several simple but useful field problems such as roadway embankment freezing and frost heaving

TART 2000: A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

International Nuclear Information System (INIS)

Cullen, D.E

2000-01-01

TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files

TART 2000 A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

CERN Document Server

Cullen, D

2000-01-01

TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.

Updated version of the DOT 4 one- and two-dimensional neutron/photon transport code

International Nuclear Information System (INIS)

Rhoades, W.A.; Childs, R.L.

1982-07-01

DOT 4 is designed to allow very large transport problems to be solved on a wide range of computers and memory arrangements. Unusual flexibilty in both space-mesh and directional-quadrature specification is allowed. For example, the radial mesh in an R-Z problem can vary with axial position. The directional quadrature can vary with both space and energy group. Several features improve performance on both deep penetration and criticality problems. The program has been checked and used extensively

BLT [Breach, Leach, and Transport]: A source term computer code for low-level waste shallow land burial

International Nuclear Information System (INIS)

Suen, C.J.; Sullivan, T.M.

1990-01-01

This paper discusses the development of a source term model for low-level waste shallow land burial facilities and separates the problem into four individual compartments. These are water flow, corrosion and subsequent breaching of containers, leaching of the waste forms, and solute transport. For the first and the last compartments, we adopted the existing codes, FEMWATER and FEMWASTE, respectively. We wrote two new modules for the other two compartments in the form of two separate Fortran subroutines -- BREACH and LEACH. They were incorporated into a modified version of the transport code FEMWASTE. The resultant code, which contains all three modules of container breaching, waste form leaching, and solute transport, was renamed BLT (for Breach, Leach, and Transport). This paper summarizes the overall program structure and logistics, and presents two examples from the results of verification and sensitivity tests. 6 refs., 7 figs., 1 tab

Vectorization of three-dimensional neutron diffusion code CITATION

International Nuclear Information System (INIS)

Harada, Hiroo; Ishiguro, Misako

1985-01-01

Three-dimensional multi-group neutron diffusion code CITATION has been widely used for reactor criticality calculations. The code is expected to be run at a high speed by using recent vector supercomputers, when it is appropriately vectorized. In this paper, vectorization methods and their effects are described for the CITATION code. Especially, calculation algorithms suited for vectorization of the inner-outer iterative calculations which spend most of the computing time are discussed. The SLOR method, which is used in the original CITATION code, and the SOR method, which is adopted in the revised code, are vectorized by odd-even mesh ordering. The vectorized CITATION code is executed on the FACOM VP-100 and VP-200 computers, and is found to run over six times faster than the original code for a practical-scale problem. The initial value of the relaxation factor and the number of inner-iterations given as input data are also investigated since the computing time depends on these values. (author)

Numerical simulation of complex multi-dimensional two-phase flows in nuclear power plant coolant circuits by means of the best-estimate thermal-hydraulic code BAGIRA

International Nuclear Information System (INIS)

Kalinichenko, S.D.; Kroshilin, A.E.; Kroshilin, V.E.; Smirnov, A.V.

2009-01-01

Recent results are exposed, obtained by applying the best-estimate thermal hydraulic code BAGIRA for three-dimensional modeling complex two-phase flow dynamics inside the vessel of the horizontal steam generator PGV-1000 used in reactor units with VVER-1000. Spatial volumetric void fraction and velocity distributions are calculated and compared with available experimental data. (author)

SERPENT Monte Carlo reactor physics code

International Nuclear Information System (INIS)

Leppaenen, J.

2010-01-01

SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

Three-Dimensional Color Code Thresholds via Statistical-Mechanical Mapping

Science.gov (United States)

Kubica, Aleksander; Beverland, Michael E.; Brandão, Fernando; Preskill, John; Svore, Krysta M.

2018-05-01

Three-dimensional (3D) color codes have advantages for fault-tolerant quantum computing, such as protected quantum gates with relatively low overhead and robustness against imperfect measurement of error syndromes. Here we investigate the storage threshold error rates for bit-flip and phase-flip noise in the 3D color code (3DCC) on the body-centered cubic lattice, assuming perfect syndrome measurements. In particular, by exploiting a connection between error correction and statistical mechanics, we estimate the threshold for 1D stringlike and 2D sheetlike logical operators to be p3DCC (1 )≃1.9 % and p3DCC (2 )≃27.6 % . We obtain these results by using parallel tempering Monte Carlo simulations to study the disorder-temperature phase diagrams of two new 3D statistical-mechanical models: the four- and six-body random coupling Ising models.

Coding for Two Dimensional Constrained Fields