Three-dimensional multigroup diffusion code ANDEX based on nodal method for cartesian geometry
International Nuclear Information System (INIS)
An analytic polynomial nodal method using partial currents has been derived for the solution of multigroup neutron diffusion equations in three-dimensional (3-D) cartesian geometry. This method is characterized by expressing the source and leakage terms in an auxiliary 1-D diffusion equation by quadratic polynomials and solving it analytically. Based on this method, we have developed a 3-D multigroup diffusion code ANDEX, and applied to 2-D LWR and 3-D FBR models. The results of keff, power distributions and computing time have been compared with those of finite difference method calculations. (author)
Development and validation of Apros multigroup nodal diffusion model
Rintala, Antti
2015-01-01
The development of a steady state and transient multigroup nodal diffusion model for process simulation software Apros was continued and the models were validated. The initial implementation of the model was performed in 2009 and it has not been under continuous development afterwards. Some errors in the steady state model were corrected. The transient model was found to be incorrect. The solution method of the transient model was derived, and the program code not common with the steady s...
FINELM: a multigroup finite element diffusion code
International Nuclear Information System (INIS)
FINELM is a FORTRAN IV program to solve the Neutron Diffusion Equation in X-Y, R-Z, R-theta, X-Y-Z and R-theta-Z geometries using the method of Finite Elements. Lagrangian elements of linear or higher degree to approximate the spacial flux distribution have been provided. The method of dissections, coarse mesh rebalancing and Chebyshev acceleration techniques are available. Simple user defined input is achieved through extensive input subroutines. The input preparation is described followed by a program structure description. Sample test cases are provided. (Auth.)
1-D EQUILIBRIUM DISCRETE DIFFUSION MONTE CARLO
Energy Technology Data Exchange (ETDEWEB)
T. EVANS; ET AL
2000-08-01
We present a new hybrid Monte Carlo method for 1-D equilibrium diffusion problems in which the radiation field coexists with matter in local thermodynamic equilibrium. This method, the Equilibrium Discrete Diffusion Monte Carlo (EqDDMC) method, combines Monte Carlo particles with spatially discrete diffusion solutions. We verify the EqDDMC method with computational results from three slab problems. The EqDDMC method represents an incremental step toward applying this hybrid methodology to non-equilibrium diffusion, where it could be simultaneously coupled to Monte Carlo transport.
Converged accelerated finite difference scheme for the multigroup neutron diffusion equation
International Nuclear Information System (INIS)
Computer codes involving neutron transport theory for nuclear engineering applications always require verification to assess improvement. Generally, analytical and semi-analytical benchmarks are desirable, since they are capable of high precision solutions to provide accurate standards of comparison. However, these benchmarks often involve relatively simple problems, usually assuming a certain degree of abstract modeling. In the present work, we show how semi-analytical equivalent benchmarks can be numerically generated using convergence acceleration. Specifically, we investigate the error behavior of a 1D spatial finite difference scheme for the multigroup (MG) steady-state neutron diffusion equation in plane geometry. Since solutions depending on subsequent discretization can be envisioned as terms of an infinite sequence converging to the true solution, extrapolation methods can accelerate an iterative process to obtain the limit before numerical instability sets in. The obtained results have been compared to the analytical solution to the 1D multigroup diffusion equation when available, using FORTRAN as the computational language. Finally, a slowing down problem has been solved using a cascading source update, showing how a finite difference scheme performs for ultra-fine groups (104 groups) in a reasonable computational time using convergence acceleration. (authors)
Multigroup finite element-boundary element method for neutron diffusion
International Nuclear Information System (INIS)
Full text: The finite element method (FEM) is an efficient method used for the solution of partial differential equations (PDE's) of engineering physics due to its symmetric, sparse and positive-definite coefficient matrix. FEM has been successfully applied for the solution of multigroup neutron transport and diffusion equations since 1970's. The boundary element method (BEM), on the other hand, is a newer method and is unique among the numerical methods used for the solution of PDE's with its property of confining the unknowns only to the boundaries of homogeneous regions, thus, greatly reducing matrix dimensions. The first application of BEM to the neutron diffusion equation (NDE) dates back to 1985 and many researchers are currently working in this area. Although BEM is known to have the desirable property of being an internal-mesh free method, this advantage is lost in some of its application to the NDE due to the existence of fission source volume integrals in fissionable regions unless domain-decomposition methods are used. To exploit the favorable properties of both FEM and BEM, a hybrid FE/BE method has been recently proposed for reflected systems treated by one or two-group diffusion theories in a recent paper co-authored by the first author. In this work, the hybrid FE/BE method for reflected systems is generalized to multigroup diffusion theory. The core is treated by FEM to preserve the high accuracy of FEM in such neutron-producing regions. Using a boundary integral equation formerly proposed by the second author, BEM, is utilized for the discretization of the reflector, thus, eliminating the internal mesh completely for this nonfissionable region. The multigroup FE/BE method has been implemented in our recently developed FORTRAN program. The program is validated by comparison of the calculated effective multiplication factor and the group fluxes with their analytical counterparts for a two-group reflected system. Comparison of these results and
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A code called COMESH based on corner mesh finite difference scheme has been developed to solve multigroup diffusion theory equations. One can solve 1-D, 2-D or 3-D problems in Cartesian geometry and 1-D (r) or 2-D (r-z) problem in cylindrical geometry. On external boundary one can use either homogeneous Dirichlet (θ-specified) or Neumann (∇θ specified) type boundary conditions or a linear combination of the two. Internal boundaries for control absorber simulations are also tackled by COMESH. Many an acceleration schemes like successive line over-relaxation, two parameter Chebyschev acceleration for fission source, generalised coarse mesh rebalancing etc., render the code COMESH a very fast one for estimating eigenvalue and flux/power profiles in any type of reactor core configuration. 6 refs. (author)
International Nuclear Information System (INIS)
The subject is divided in two parts: In the first part a nodal method has been worked out to solve the steady state multigroup diffusion equation. This method belongs to the same set of nodal methods currently used to calculate the exact fission powers and neutron fluxes in a very short computing time. It has been tested on a two dimensional idealized reactors. The effective multiplication factor and the fission powers for each fuel element have been calculated. The second part consists in studying and mastering the multigroup diffusion code DAHRA - a reduced version of DIANE - a two dimensional code using finite difference method
FINELM: a multigroup finite element diffusion code. Part I
International Nuclear Information System (INIS)
The author presents a two dimensional code for multigroup diffusion using the finite element method. It was realized that the extensive connectivity which contributes significantly to the accuracy, results in a matrix which, although symmetric and positive definite, is wide band and possesses an irregular profile. Hence, it was decided to introduce sparsity techniques into the code. The introduction of the R-Z geometry lead to a great deal of changes in the code since the rotational invariance of the removal matrices in X-Y geometry did not carry over in R-Z geometry. Rectangular elements were introduced to remedy the inability of the triangles to model essentially one dimensional problems such as slab geometry. The matter is discussed briefly in the text in the section on benchmark problems. This report is restricted to the general theory of the triangular elements and to the sparsity techniques viz. incomplete disections. The latter makes the size of the problem that can be handled independent of core memory and dependent only on disc storage capacity which is virtually unlimited. (Auth.)
International Nuclear Information System (INIS)
1 - Description of program or function: PARTISN (Parallel, Time-Dependent SN) is the evolutionary successor to CCC-0547/DANTSYS. User input and cross section formats are very similar to that of DANTSYS. The linear Boltzmann transport equation is solved for neutral particles using the deterministic (SN) method. Both the static (fixed source or eigenvalue) and time-dependent forms of the transport equation are solved in forward or adjoint mode. Vacuum, reflective, periodic, white, or inhomogeneous boundary conditions are solved. General anisotropic scattering and inhomogeneous sources are permitted. PARTISN solves the transport equation on orthogonal (single level or block-structured AMR) grids in 1-D (slab, two-angle slab, cylindrical, or spherical), 2-D (X-Y, R-Z, or R-T) and 3-D (X-Y-Z or R-Z-T) geometries. 2 - Methods:PARTISN numerically solves the multigroup form of the neutral-particle Boltzmann transport equation. The discrete-ordinates form of approximation is used for treating the angular variation of the particle distribution. For curvilinear geometries, diamond differencing is used for angular discretization. The spatial discretizations may be either low-order (diamond difference or Adaptive Weighted Diamond Difference (AWDD)) or higher-order (linear discontinuous or exponential discontinuous). Negative fluxes are eliminated by a local set-to-zero-and-correct algorithm for the diamond case (DD/STZ). Time differencing is Crank-Nicholson (diamond), also with a set-to-zero fix-up scheme. Both inner and outer iterations can be accelerated using the diffusion synthetic acceleration method, or transport synthetic acceleration can be used to accelerate the inner iterations. The diffusion solver uses either the conjugate gradient or multigrid method. Chebyshev acceleration of the fission source is used. The angular source terms may be treated either via standard PN expansions or Galerkin scattering. An option is provided for strictly positive scattering sources
International Nuclear Information System (INIS)
Highlights: → Coupled neutron and gamma transport is considered in the multigroup diffusion approximation. → The model accommodates fission, up- and down-scattering and common neutron-gamma interactions. → The exact solution to the diffusion equation in a heterogeneous media of any number of regions is found. → The solution is shown to parallel the one-group case in a homogeneous medium. → The discussion concludes with a heterogeneous, 2 fuel-plate 93.2% enriched reactor fuel benchmark demonstration. - Abstract: The angular flux for the 'rod model' describing coupled neutron/gamma (n, γ) diffusion has a particularly straightforward analytical representation when viewed from the perspective of a one-group homogeneous medium. Cast in the form of matrix functions of a diagonalizable matrix, the solution to the multigroup equations in heterogeneous media is greatly simplified. We shall show exactly how the one-group homogeneous medium solution leads to the multigroup solution.
Multigroup neutron transport equation in the diffusion and P1 approximation
International Nuclear Information System (INIS)
Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P1 approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P1 approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)
APPLE, Plot of 1-D Multigroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN
International Nuclear Information System (INIS)
A - Description of problem or function: The APPLE-2 code has the following functions: (1) It plots multi-group energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT-3.5, and MORSE. (2) It gives an overview plot of multi-group neutron fluxes calculated by ANISN and DOT-3.5. The scalar neutron flux phi(r,E) is plotted with the spatial parameter r linear along the Y-axis, logE along the X-axis and log phi(r,E) in the Z direction. (3) It calculates the spatial distribution and region volume integrated values of reaction rates using the scalar flux calculated with ANISN and DOT-3.5. (4) Reaction rate distribution along the R or Z direction may be plotted. (5) An overview plot of reaction rates or scalar fluxes summed over specified groups may be plotted. R(ri,zi) or phi(ri,zi) is plotted with spatial parameters r and z along the X- and Y-axes in an orthogonal coordinate system. (6) Angular flux calculated by ANISN is rearranged and a shell source at any specified spatial mesh point may be punched out in FIDO format. The shell source obtained may be employed in solving deep penetration problems with ANISN, when the entire reactor system is divided into two or more parts and the neutron fluxes in two adjoining parts are connected by using the shell source. B - Method of solution: (a) The input data specification is made as simple as possible by making use of the input data required in the radiation transport code. For example, geometry related data in ANISN and DOT are transmitted to APPLE-2 along with scalar flux data so as to reduce duplicity and errors in reproducing these data. (b) Most the input data follow the free form FIDO format developed at Oak Ridge National Laboratory and used in the ANISN code. Furthermore, the mixture specifying method used in ANISN is also employed by APPLE-2. (c) Libraries for some standard response functions required in fusion reactor design have been prepared and are made available to users of the 42-group neutron
SIXTUS-2. A two dimensional multigroup diffusion theory code in hexagonal geometry. Pt. 1
International Nuclear Information System (INIS)
A new algorithm for solving the 2-dimensional multigroup diffusion equations in hexagonal geometry is described. It is based on three novel ideas: analytic intranodal solutions, use of the group irreducible representations and an explicit scheme for solving the response matrix equations. The resulting computer code SIXTUS-2 has been found to be very accurate and effective. (Auth.)
Second order time evolution of the multigroup diffusion and P1 equations for radiation transport
International Nuclear Information System (INIS)
Highlights: → An existing multigroup transport algorithm is extended to be second-order in time. → A new algorithm is presented that does not require a grey acceleration solution. → The two algorithms are tested with 2D, multi-material problems. → The two algorithms have comparable computational requirements. - Abstract: An existing solution method for solving the multigroup radiation equations, linear multifrequency-grey acceleration, is here extended to be second order in time. This method works for simple diffusion and for flux-limited diffusion, with or without material conduction. A new method is developed that does not require the solution of an averaged grey transport equation. It is effective solving both the diffusion and P1 forms of the transport equation. Two dimensional, multi-material test problems are used to compare the solution methods.
Tredit A 3-D multigroup diffusion theory simulator for hexagonal fuel assembly cores
International Nuclear Information System (INIS)
A multigroup 3-D reactor core simulator based on neutron diffusion theory, called TREDIT has been developed for Light Water Reactors (LWRs). It considers triangle shaped meshes in X-Y plane and variable mesh spacing in Z-direction. Thus it is especially suited for designing and analysing LWR cores with hexagonal fuel assemblied like the Russian WWER reactors. When fuel assembly cross-sections in multigroup form are input as fitted constants, the computer code TREDIT can build up core burnup distribution with power distribution computed for initial reactor conditions. The results of this code have been compared with another diffusion theory based code and found satisfactory. Xenon feedback effects on core power distribution are demonstrated. (author)
Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis
International Nuclear Information System (INIS)
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)
FINELM: a multigroup finite element diffusion code. III
International Nuclear Information System (INIS)
The authors describe the formalism of the finite element for r-theta geometry. The treatment of a four sided element with two circular arcs and two radial sides is presented. At the centre of the disc this element is collapsed into a three sided element by reducing the inner arc length to zero. The singularity of the leakage term of such an element at r=0 is described. The modelling of control and safety rods by introducing boundary conditions interior to the reactor is presented. This has the effect of converting the diffusion region from a simply connected to a multiply connected region. Dirichlet conditions with prescribed flux values zero are imposed on nodes in the interior of the non diffusion zones in order to decouple these nodes from the rest. The method is tested by calculating the effectiveness of control and safety rods for a 500 MW Th high temperature reactor. The results are compared with those obtained with DIFGEN using x-y geometry. The DIFGEN results and the problem data are by courtsey of Hochtemperatur Reaktor Bau AG, Mannheim. (Auth.)
GIS-BASED 1-D DIFFUSIVE WAVE OVERLAND FLOW MODEL
Energy Technology Data Exchange (ETDEWEB)
KALYANAPU, ALFRED [Los Alamos National Laboratory; MCPHERSON, TIMOTHY N. [Los Alamos National Laboratory; BURIAN, STEVEN J. [NON LANL
2007-01-17
This paper presents a GIS-based 1-d distributed overland flow model and summarizes an application to simulate a flood event. The model estimates infiltration using the Green-Ampt approach and routes excess rainfall using the 1-d diffusive wave approximation. The model was designed to use readily available topographic, soils, and land use/land cover data and rainfall predictions from a meteorological model. An assessment of model performance was performed for a small catchment and a large watershed, both in urban environments. Simulated runoff hydrographs were compared to observations for a selected set of validation events. Results confirmed the model provides reasonable predictions in a short period of time.
Simulate-HEX - The multi-group diffusion equation in hexagonal-z geometry
International Nuclear Information System (INIS)
The multigroup diffusion equation is solved for the hexagonal-z geometry by dividing each hexagon into 6 triangles. In each triangle, the Fourier solution of the wave equation is approximated by 8 plane waves to describe the intra-nodal flux accurately. In the end an efficient Finite Difference like equation is obtained. The coefficients of this equation depend on the flux solution itself and they are updated once per power/void iteration. A numerical example demonstrates the high accuracy of the method. (authors)
FEM-BABEL, 3-D Multigroup Neutron Diffusion by Galerkin Method
International Nuclear Information System (INIS)
1 - Nature of the physical problem solved: This program computes the three-dimensional multigroup neutron diffusion equation using the finite element method. 2 - Method of solution: The equation is solved using a solution algorithm based on a Galerkin-type scheme. Prism and box-shaped finite elements are used. The resulting equation system is solved using the successive over-relaxation method and the inner iterations are accelerated by a coarse mesh re-balancing technique. 3 - Restrictions on the complexity of the problem: Any down-scattering of neutrons is allowed but up-scattering and region-dependent fission spectra are not permitted
Development of a three-dimensional multigroup nodal diffusion code for the LMR
International Nuclear Information System (INIS)
STEP is a three-dimensional multigroup nodal diffusion code for the neutronics analysis of the LMR core and accepts microscopic cross section data. Material cross sections are obtained by summing the product of atom densities and microscopic cross sections over all isotopes comprising the material. STEP contains a thermal-hydraulics module which enables feedback effects from both fuel temperature and coolant temperature changes. Numerical results of the STEP code over the KALIMER core (392 MWt) agree well with those of DIF-3D. And it has been observed that the thermal-hydraulics module is working properly
An effective method of solving the multigroup diffusion problem in hexagonal geometry. Part I
International Nuclear Information System (INIS)
An effective method of solving two-dimensional multigroup diffusion equations in hexagonal geometry is described. The method is based on the following two ideas: nodal approach, and expansion of one-dimensional neutron fluxes inside the node into polynomials up to the third order. The resulting relations for the interface-averaged partial currents, node-averaged fluxes and flux moments are used in computer code NEHEX. The code was found to be an accurate and effective computational tool. Its description and validation against reference benchmark problems will be published as Part II of this report. (author) 1 fig., 1 tab., 9 refs
International Nuclear Information System (INIS)
The good features of the Analytic Function Expansion Nodal (AFEN) method are utilized to develop a practical scheme for the multigroup diffusion problems, in combination with the polynomial expansion nodal (PEN) method. The thermal group fluxes exhibiting strong gradients are solved by the AFEN method, while the fast group fluxes that are smoother than the thermal group fluxes are solved by the PEN method. The scheme is developed for cores of rectangular and hexagonal geometries. In particular, to model the fast group fluxes in the hexagonal geometry by the PEN method, a polynomial function set which shows good performance in accuracy and numerical stability is derived, in premiere. (author)
International Nuclear Information System (INIS)
Nonlinear diffusion acceleration (NDA) can improve the performance of a neutron transport solver significantly especially for the multigroup eigenvalue problems. The high-order transport equation and the transport-corrected low-order diffusion equation form a nonlinear system in NDA, which can be solved via a Picard iteration. The consistency of the correction of the low-order equation is important to ensure the stabilization and effectiveness of the iteration. It also makes the low-order equation preserve the scalar flux of the high-order equation. In this paper, the consistent correction for a particular discretization scheme, self-adjoint angular flux (SAAF) formulation with discrete ordinates method (SN) and continuous finite element method (CFEM) is proposed for the multigroup neutron transport equation. Equations with the anisotropic scatterings and a void treatment are included. The Picard iteration with this scheme has been implemented and tested with RattleSNake, a MOOSE-based application at INL. Convergence results are presented. (authors)
International Nuclear Information System (INIS)
Most of the neutron diffusion codes use numerical methods giving accurate results in structured meshes. However, the application of these methods in unstructured meshes to deal with complex geometries is not straightforward and it may cause problems of stability and convergence of the solution. By contrast, the Finite Volume Method (FVM) is easily applied to unstructured meshes and is typically used in the transport equations due to the conservation of the transported quantity within the volume. In this paper, the FVM algorithm implemented in the ARB Partial Differential Equations Solver has been used to discretize the multigroup neutron diffusion equation to obtain the matrices of the generalized eigenvalue problem, which has been solved by means of the SLEPc library. Nevertheless, these matrices could be large for fine meshes and the eigenvalue problem resolution could require a high calculation time. Therefore, a transformation of the generalized eigenvalue problem into a standard one is performed in order to reduce the calculation time. (author)
Solution of the 1D kinetic diffusion equations using a reduced nodal cubic scheme
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In this work it is described a novel method to solve the multi-group time-dependent diffusion equations based on a nodal cubic space interpolation in addition to the application of quadrature rules simplifying the stiffness and mass matrices arising in a finite element procedure. Numerical results for a well known benchmark problem are also provided. (authors)
International Nuclear Information System (INIS)
In this work we developed a software to model and generate results in tables and graphs of one-dimensional neutron transport problems in multi-group formulation of energy. The numerical method we use to solve the problem of neutron diffusion is analytic, thus eliminating the truncation errors that appear in classical numerical methods, e.g., the method of finite differences. This numerical analytical method increases the computational efficiency, since they are not refined spatial discretization necessary because for any spatial discretization grids used, the numerical result generated for the same point of the domain remains unchanged unless the rounding errors of computational finite arithmetic. We chose to develop a computational application in MatLab platform for numerical computation and program interface is simple and easy with knobs. We consider important to model this neutron transport problem with a fixed source in the context of shielding calculations of radiation that protects the biosphere, and could be sensitive to ionizing radiation
Lozano Montero, Juan Andrés; García Herranz, Nuria; Ahnert Iglesias, Carolina; Aragonés Beltrán, José María
2008-01-01
In this work we address the development and implementation of the analytic coarse-mesh finite-difference (ACMFD) method in a nodal neutron diffusion solver called ANDES. The first version of the solver is implemented in any number of neutron energy groups, and in 3D Cartesian geometries; thus it mainly addresses PWR and BWR core simulations. The details about the generalization to multigroups and 3D, as well as the implementation of the method are given. The transverse integration procedure i...
International Nuclear Information System (INIS)
An extended version of Hassitt's one-dimension Multi-group diffusion programme has been prepared which allows for a maximum of twenty-two energy groups rather than sixteen. It also permits the use of drum storage by programme and data up to the full machine capacity of 1024 sectors, rather than 478 sectors. Some minor corrections have been made, A binary tape of an 830-sector version has been prepared (Winfrith P.5272). (author)
FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: FEM-2D solves the two-dimensional diffusion equation in x-y geometry. This is done by the finite elements method. 2 - Method of solution: FEM-2D uses triangular elements with first and second order Lagrange approximations. The systems equations are formulated in multigroup form and solved by Cholesky procedure which operates only on nonzero elements. Various acceleration techniques are available for the outer iteration. Fluxes along various lines and rates in arbitrary zones may be output. 3 - Restrictions on the complexity of the problem: The code uses variable dimensioning. Thus, the problem size is restricted by the largest array which usually is the systems matrix. Fluxes of all groups are kept in memory. This might become another restrictive data set for a large number of groups. The validity of the results is restricted by the approximations used. FEM-2D requires a finite element net which allows the approximation of fluxes by at most parabolas. The node distribution should be more dense in areas of heavy flux changes (near absorbers or the reflector)
International Nuclear Information System (INIS)
The fine mesh diffusion formulation is extended to deal with multigroup 3-D problems in rectangular geometries. The formulation includes interface discontinuity factors per cell type, pre-calculated from transport solutions. The iterative scheme, aiming to an efficient parallel implementation in memory distributed multi-processors, is based on domain decomposition in the 4 possible sets of 4 neighbor quarters of assemblies. The alternate dissections achieve convergence to the exact boundary conditions, while attenuating high frequency noise. Whole core convergence is accelerated in the long wavelength effects by a consistent high-order analytical nodal solution performed by the ANDES solver. A neutronics - thermal-hydraulics iterative scheme is also developed to compute best estimate results, by coupling at the detailed cell-subchannel scale the COBAYA3 code with several TH subchannel codes. The numerical performance and convergence rates are verified by computing pin-cell scale solutions for the OECD/NEA/USNRC PWR MOX/UO2 Core Transient Benchmark in 8 energy groups and heterogeneous assemblies. The cell-subchannel scale neutronics and thermal-hydraulics coupling, allows the verification of the effects of the detailed TH feedbacks on cross-sections and, thus, on fuel pin powers, calculated here for a 3D color-set of two different fuel types of the previous benchmark, using COBAYA3 and COBRA-3C. (authors)
Multigroup radiation hydrodynamics with flux-limited diffusion and adaptive mesh refinement
González, Matthias; Commerçon, Benoît; Masson, Jacques
2015-01-01
Radiative transfer plays a key role in the star formation process. Due to a high computational cost, radiation-hydrodynamics simulations performed up to now have mainly been carried out in the grey approximation. In recent years, multi-frequency radiation-hydrodynamics models have started to emerge, in an attempt to better account for the large variations of opacities as a function of frequency. We wish to develop an efficient multigroup algorithm for the adaptive mesh refinement code RAMSES which is suited to heavy proto-stellar collapse calculations. Due to prohibitive timestep constraints of an explicit radiative transfer method, we constructed a time-implicit solver based on a stabilised bi-conjugate gradient algorithm, and implemented it in RAMSES under the flux-limited diffusion approximation. We present a series of tests which demonstrate the high performance of our scheme in dealing with frequency-dependent radiation-hydrodynamic flows. We also present a preliminary simulation of a three-dimensional p...
Three-dimensional h-adaptivity for the multigroup neutron diffusion equations
Wang, Yaqi
2009-04-01
Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library\\'s extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6. © 2008 Elsevier Ltd. All rights reserved.
International Nuclear Information System (INIS)
The Analytic Coarse-Mesh Finite-Difference method is developed in detail for multi-group and multi-dimensional diffusion calculations, including the general and particular modal solutions in the complex space for any number of groups. For rectangular multidimensional geometries, the Chao's generalized relations with transverse integration provide a high-order approximation of the ACMFD method, where all energy groups are coupled by matrix-vector FD relations and the errors are limited to the ones incurred by the interpolation of the transverse interface currents, in a non-linear iterative scheme. The implementation of the method in a multigroup 3D rectangular geometry nodal solver called ANDES is discussed, pointing out the encapsulation achieved for integration of the solver as an optional module within larger code systems. The performance of the ANDES solver in 3D rectangular (X-Y-Z) geometry and multi-groups is verified by its application to several 2D-3D model and international benchmarks (NEA-OECD), with given diffusion cross section sets in few-groups (2 to 8). The extensive verification, always required for new methods and codes, shows a quite fast convergence of ANDES in both the eigenvalue and transverse leakage iteration loops and with the nodal coarse-mesh size, allowing to reach the conclusion that quite high accuracy is achieved with rather large nodes, one node or four nodes per PWR fuel assembly, as compared with reference solutions obtained with fine-mesh finite-difference diffusion calculations using mesh sizes 64 to 128 times smaller than the ANDES nodes. (authors)
International Nuclear Information System (INIS)
Highlights: ► We develop a 2-D, multigroup neutron/adjoint diffusion computer code based on GFEM. ► The spatial discretization is performed using unstructured triangle elements. ► Multiplication factor, flux/adjoint and power distribution are outputs of the code. ► Sensitivity analysis to the number, arrangement and size of elements is performed. ► We proved that the developed code is a reliable tool to solve diffusion equation. -- Abstract: Various methods for solving the forward/adjoint equation in hexagonal and rectangular geometries are known in the literatures. In this paper, the solution of multigroup forward/adjoint equation using Finite Element Method (FEM) for hexagonal and rectangular reactor cores is reported. The spatial discretization of equations is based on Galerkin FEM (GFEM) using unstructured triangle elements. Calculations are performed for both linear and quadratic approximations of the shape function; based on which results are compared. Using power iteration method for the forward and adjoint calculations, the forward and adjoint fluxes with the corresponding eigenvalues are obtained. The results are then benchmarked against the valid results for IAEA-2D, BIBLIS-2D and IAEA-PWR benchmark problems. Convergence rate of GFEM in linear and quadratic approximations of the shape function are calculated and results are quantitatively compared. A sensitivity analysis of the calculations to the number and arrangement of elements has been performed.
Energy Technology Data Exchange (ETDEWEB)
Shestakov, A I; Offner, S R
2007-03-02
We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory
VARI-QUIR-3, 2-D Multigroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: The steady-state, multigroup, two-dimensional neutron diffusion equations are solved in x-y, r-z, and r-theta geometry. 2 - Method of solution: A Gauss-Seidel type of solution with inner and outer iterations is used. The source is held constant during the inner iterations
International Nuclear Information System (INIS)
A multigroup diffusion theory code, TRIHEX-3D, has been developed for hexagonal lattice core analyses. For 2-D problems one can use hexagonal or triangular centre-mesh finite difference (FD) schemes. The geometrical description of the problem is for hexagonal geometry only. Subdivision of each hexagon into uniform triangles is facilitated by a built-in auto-triangularisati on procedure. One can analyse any symmetric part of the core or the whole core as well. Reflective (30deg, 60deg, 90deg, 120deg and 180deg) and rotational (60deg, 120deg and 180deg) symmetry boundary conditions are allowed. For 3-D problems one can use a direct 3-D FDM or an axial flux synthesis method. TRIHEX-3D can be used for the core design problems of VVER type of hexagonal lattice cores. The code has been validated against a LMFBR SNR-300 benchmark problem. (author). 8 tabs., 9 figs., 9 refs., 5 appendixes
EQUIVA, Few-Group Diffusion Parameter for PWR Reflector Region by 1-D Transport Calculation
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1 - Description of program or function: EQUIVA-1 generates few-group equivalent diffusion theory parameters for pressurized water (PWR) reflector regions from the results of simple one-dimensional (slab) multigroup transport calculations. The one-dimensional 'normalized generalized equivalence theory' (NGET) method is used for this purpose. Equivalent parameters can be generated for any number of condensed energy groups and for different material regions, such as for the explicit baffle and water reflector of the radial reflector of a PWR, or for the homogenized baffle/reflector. EQUIVA-2 generates environment-insensitive equivalent diffusion theory parameters for pressurized water (PWR) reflector regions from the results of simple one-dimensional (slab) multigroup transport calculations. The one-dimensional reflector models have been implemented, namely the few-group NGET-RM method and the two-group KOEBKE-RM method. Equivalent parameters can be generated for one homogenized region only, although this region itself may be constituted by any number of component regions. 2 - Method of solution: EQUIVA-1: Analytic functions of non-symmetric real matrices are computed by means of a spectral analysis method. EQUIVA-2: Analytic functions of non-symmetric real matrices are computed by means of a spectral analysis method. Component region slab response matrices are combined using rigorous addition rules to obtain a global response matrix for a single region/node. 3 - Restrictions on the complexity of the problem: EQUIVA-1.1 is a variably dimensioned code and there is no restriction on the number of energy groups, etc. The size of the problem is restricted only by the computer core storage available. EQUIVA-2.0 is a variably dimensioned code and there is no restriction on the number of energy groups, etc. The size of the problem is restricted only by the computer core storage available
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The objective of this work is to obtain an analytical solution of the neutron diffusion kinetic equation in one-dimensional cartesian geometry, to monoenergetic and multigroup problems. These equations are of the type stiff, due to large differences in the orders of magnitude of the time scales of the physical phenomena involved, which make them difficult to solve. The basic idea of the proposed method is applying the spectral expansion in the scalar flux and in the precursor concentration, taking moments and solving the resulting matrix problem by the Laplace transform technique. Bearing in mind that the equation for the precursor concentration is a first order linear differential equation in the time variable, to enable the application of the spectral method we introduce a fictitious diffusion term multiplied by a positive value which tends to zero. This procedure opened the possibility to find an analytical solution to the problem studied. We report numerical simulations and analysis of the results obtained with the precision controlled by the truncation order of the series. (author)
Prinsloo, Rian Hendrik
2006-01-01
Nodal diffusion methods have been used extensively in nuclear reactor calculations specifically for their performance advantage, but also their superior accuracy. In this work a nodal diffusion method is developed for three-dimensional cylindrical geometry. Recent developments in the Pebble Bed Modular Reactor (PBMR) project have sparked renewed interest in the application of different modelling methods to its design, and naturally included in this effort is a nodal method for ...
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Today, we can use a computer cluster consist of a few hundreds CPUs with reasonable budget. Such computer system enables us to do detailed modeling of reactor core. The detailed modeling will improve the safety and the economics of a nuclear reactor by eliminating un-necessary conservatism or missing consideration. To take advantage of such a cluster computer, efficient parallel algorithms must be developed. Mechanical structure analysis community has studied the domain decomposition method to solve the stress-strain equation using the finite element methods. One of the most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multi-group diffusion problem in previous study. In this study, we report the result of recent modification to handle the three-dimensional subdomain partitioning, and the sub-domain multi-group problem. Modified FETI-DP algorithm has been successfully applied for the eigenvalue problem of multi-group neutron diffusion equation. The overall CPU time is decreasing as number of sub-domains (partitions) is increasing. However, there may be a limit in decrement due to increment of the number of primal points will increase the CPU time spent by the solution of the global equation. Even distribution of computational load (criterion a) is important to achieve fast computation. The subdomain partition can be effectively performed using suitable graph theory partition package such as MeTIS
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Highlights: ► The multi-group IDE-NDK was solved numerically in 2D-Cartesian geometry. ► The progressive basic polynomial (BPn) methods showed no numerical oscillations. ► The BP2 algorithm showed good accuracy and efficiency. -- Abstract: The multi-group time-integro-differential equations of the neutron diffusion kinetics (IDE-NDK) was solved numerically in 2D Cartesian geometry with the use of the basic-progressive polynomial approximation (BPn). Two applications were computed: a ramp, and an instantaneous change of the thermal removal macroscopic cross sections of the driver material of the 2D-TWGL benchmark problems. The BP2 algorithm showed good accuracy when compared with the results of other codes. BPn did not show the undesirable oscillations that appeared in other codes.
Benchmarking with the multigroup diffusion high-order response matrix method
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The benchmarking capabilities of the high-order response matrix eigenvalue method, which was developed more than a decade ago, are demonstrated by means of the numerical analysis of a variety of two-dimensional Cartesian geometry light-water reactor test problems. These problems are typical of those generally used for the benchmarking of coarse-mesh (nodal) diffusion methods and the numerical results show that the high-order response matrix eigenvalue method is well suited to be used as an alternative to fine-mesh finite-difference and refined mesh nodal methods for the purpose of generating reference solutions to such problems. (author)
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In this work we address the development and implementation of the analytic coarse-mesh finite-difference (ACMFD) method in a nodal neutron diffusion solver called ANDES. The first version of the solver is implemented in any number of neutron energy groups, and in 3D Cartesian geometries; thus it mainly addresses PWR and BWR core simulations. The details about the generalization to multigroups and 3D, as well as the implementation of the method are given. The transverse integration procedure is the scheme chosen to extend the ACMFD formulation to multidimensional problems. The role of the transverse leakage treatment in the accuracy of the nodal solutions is analyzed in detail: the involved assumptions, the limitations of the method in terms of nodal width, the alternative approaches to implement the transverse leakage terms in nodal methods - implicit or explicit -, and the error assessment due to transverse integration. A new approach for solving the control rod 'cusping' problem, based on the direct application of the ACMFD method, is also developed and implemented in ANDES. The solver architecture turns ANDES into an user-friendly, modular and easily linkable tool, as required to be integrated into common software platforms for multi-scale and multi-physics simulations. ANDES can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. The verification and performance of the solver are demonstrated using both proof-of-principle test cases and well-referenced international benchmarks
Development of 3D multi-group neutron diffusion code for hexagonal geometry
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Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)
Improvement of the axial diffusion solver of DeCART employing 1-D transport solution
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Instead of the 3-D transport equation, DeCART solves a transverse leakage coupled radial transport and axial diffusion equations to obtain an approximate 3-D transport solution. In this paper, two of the approximations contained in DeCART related with diffusion constants and cell homogenization are exmained, and practical improvement schemes are suggested. To overcome the diffusion approximation used in the axial direction, a current conservation scheme based on the axial 1-D transport solution is introduced. To overcome the cell homogenization effect, a plane height refinement scheme is employed near the axial core/reflector boundary where homogenization constants vary significantly in the axial direction within the plane. These schemes are evaluated by solving the 3-D VENUS-2 MOX core benchmark. The current conservation and plane height refinement schemes bring about 280 pcm and 100 pcm improvement in k-eff, respectively, and about 390 pcm in total, but trivial effects in the power distribution
A variational nodal expansion method for the solution of multigroup neutron diffusion equations
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An accurate neutronics analysis method is needed for light water reactor core monitoring systems to efficiently operate the core with a smaller margin to limiting parameters. It is also required in in-core fuel management systems to optimize the core loading patterns, and the fuel designs with a higher reliability. When mixed oxide fuel or much higher burnup fuel is used, a new higher order nodal method seems necessary to introduce. Based on these considerations, a new nodal diffusion method for the neutronics analysis of light water reactor cores has been developed. The method is based on an approximation of neutron fluxes by expanding them with a set of functions defined within a node. The expansion coefficients are determined in such a way that the solution becomes the most accurate approximation to the exact solution by utilizing the variational principle. The expansion functions are obtained only from single assembly diffusion calculations. The present method includes no homogenization procedure, and the assembly heterogeneity effect on neutron fluxes is taken into account in a consistent way. The intra-nodal pin-power distribution can also be determined in a consistent way with high accuracy. The present method was implemented in a two-dimensional nodal code, and tested for benchmark cases. The results proved that the accuracy of the present method was excellent. The root mean square errors of both nodal powers and nodal maximum pin powers were observed to be less than 1%. The computing time of the code was measured to be about 3% of the reference, fine-mesh calculation. A three-dimensional version is currently being developed, and since the heterogeneity effect is of less importance in axial direction, a more efficient calculation method can be adopted for the axial solution of the neutron flux. The new method can be used as a ''plug-in'' module to existing core simulators to increase the accuracy of the neutronics part of existing core models, including the
HEXPEDITE: A net current multigroup nodal diffusion method for hexagonal-z geometry
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The feasibility of a nodal diffusion algorithm for hexagonal cores was first demonstrated by Duracz and by Lawrence. They implemented a polynomial method with partial currents for internode coupling. Following them, several authors introduced variants of the expansion technique. Wagner developed an analytical method; however, like all previous authors, he still used partial currents for internode coupling and a response matrix solution approach. Very recently, another polynomial model with net currents expressed in terms of transverse-integrated fluxes and a nodal integral method based on coordinate transformations were presented. A transformation-group method was also introduced. In this paper, a hexagonal-z method similar in approach to that of the Cartesian geometry ILLICO is presented. The new method uses an analytical solution of the transverse-integrated equations, net currents for internode coupling, and a global coupling solution scheme different from that of the methods discussed earlier. An extension that treats explicitly the in-node spatial dependence of cross sections is also introduced
Al-Chalabi, Rifat M. Khalil
1997-09-01
Development of an improvement to the computational efficiency of the existing nested iterative solution strategy of the Nodal Exapansion Method (NEM) nodal based neutron diffusion code NESTLE is presented. The improvement in the solution strategy is the result of developing a multilevel acceleration scheme that does not suffer from the numerical stalling associated with a number of iterative solution methods. The acceleration scheme is based on the multigrid method, which is specifically adapted for incorporation into the NEM nonlinear iterative strategy. This scheme optimizes the computational interplay between the spatial discretization and the NEM nonlinear iterative solution process through the use of the multigrid method. The combination of the NEM nodal method, calculation of the homogenized, neutron nodal balance coefficients (i.e. restriction operator), efficient underlying smoothing algorithm (power method of NESTLE), and the finer mesh reconstruction algorithm (i.e. prolongation operator), all operating on a sequence of coarser spatial nodes, constitutes the multilevel acceleration scheme employed in this research. Two implementations of the multigrid method into the NESTLE code were examined; the Imbedded NEM Strategy and the Imbedded CMFD Strategy. The main difference in implementation between the two methods is that in the Imbedded NEM Strategy, the NEM solution is required at every MG level. Numerical tests have shown that the Imbedded NEM Strategy suffers from divergence at coarse- grid levels, hence all the results for the different benchmarks presented here were obtained using the Imbedded CMFD Strategy. The novelties in the developed MG method are as follows: the formulation of the restriction and prolongation operators, and the selection of the relaxation method. The restriction operator utilizes a variation of the reactor physics, consistent homogenization technique. The prolongation operator is based upon a variant of the pin power
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The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P1) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level
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In the present contribution we discuss the solution of the two-dimensional multi-group neutron kinetic equation in cylindrical geometry. The solution is obtained in analytical representation. To this end the scalar flux is extended in terms of the eigenfunctions associated to the respective problem in Cartesian geometry. Taking moments and using orthogonality properties of the eigenfunctions we get a matrix differential equation for the expansion coefficients which has a known solution. We apply this methodology for the neutron kinetic diffusion equation and show numerical results for two-energy groups. (author)
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Oliveira, F.R.; Vilhena, Marco T.; Bodmann, B.E.J., E-mail: fernando.rodrigues@ufrgs.br, E-mail: bardo.bodmann@ufrgs.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil); Carvalho, F., E-mail: fernando@nuclear.ufrj.br [Coordenacao dos Cursos de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Institute Alberto Luiz Coimbra
2015-07-01
In the present contribution we discuss the solution of the two-dimensional multi-group neutron kinetic equation in cylindrical geometry. The solution is obtained in analytical representation. To this end the scalar flux is extended in terms of the eigenfunctions associated to the respective problem in Cartesian geometry. Taking moments and using orthogonality properties of the eigenfunctions we get a matrix differential equation for the expansion coefficients which has a known solution. We apply this methodology for the neutron kinetic diffusion equation and show numerical results for two-energy groups. (author)
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The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P1) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently
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This report presents the HEXAGA-III-programme solving multi-group time-independent real and/or adjoint neutron diffusion equations for three-dimensional-triangular-z-geometry. The method of solution is based on the AGA two-sweep iterative method belonging to the family of factorization techniques. An arbitrary neutron scattering model is permitted. The report written for users provides the description of the programme input and output and the use of HEXAGA-III is illustrated by a sample reactor problem. (orig.)
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Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.
1977-11-01
The report documents the computer code block VENTURE designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P/sub 1/) in up to three-dimensional geometry. It uses and generates interface data files adopted in the cooperative effort sponsored by the Reactor Physics Branch of the Division of Reactor Research and Development of the Energy Research and Development Administration. Several different data handling procedures have been incorporated to provide considerable flexibility; it is possible to solve a wide variety of problems on a variety of computer configurations relatively efficiently.
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The multi-group integro-differential equations of the neutron diffusion kinetics (IDE-NDK) was presented and solved numerically in multi-slab geometry with the use of the progressive polynomial approximation. Four applications were computed: a positive ramp, a negative ramp, a sinusoidal and an instantaneous change of thermal macroscopic cross-sections in an 120 slab-nuclear reactor for a 2 prompt-group model. The results showed good accuracy for the developed non-iterative algorithms. It was shown the advantage of using the IDE-NDK over the traditional partial differential equations of the neutron diffusion kinetics from an accuracy point of view. Finite difference algorithms were also developed to obtain initial conditions and to make desired comparisons.
Quantum Diffusion on Molecular Tubes: Universal Scaling of the 1D to 2D Transition
Chuang, Chern; Lee, Chee Kong; Moix, Jeremy M.; Knoester, Jasper; Cao, Jianshu
2016-05-01
The transport properties of disordered systems are known to depend critically on dimensionality. We study the diffusion coefficient of a quantum particle confined to a lattice on the surface of a tube, where it scales between the 1D and 2D limits. It is found that the scaling relation is universal and independent of the temperature, disorder, and noise parameters, and the essential order parameter is the ratio between the localization length in 2D and the circumference of the tube. Phenomenological and quantitative expressions for transport properties as functions of disorder and noise are obtained and applied to real systems: In the natural chlorosomes found in light-harvesting bacteria the exciton transfer dynamics is predicted to be in the 2D limit, whereas a family of synthetic molecular aggregates is found to be in the homogeneous limit and is independent of dimensionality.
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Methodology for 3-D calculation analysis of nuclear reactor cell with axial symmetry and finite mesh step is described. This methodology is based on the axial leakage calculation analysis method that has been developed for nuclear reactor with closed lattice like VVER-type. The trial functions that are used at full core level of nuclear reactor calculation analysis are defined. Connection between core reactor equation and the definition of trial functions is given. Importance of different trial functions from the point of view the full reactor core calculation is analyzed. If we deal with the case when reactor has strong neutron flux gradients caused with regularization rods it is important to take into account the influence of neutron spectrum into axial leakage. So this paper focuses upon just multi-group approach to obtain matrixes that are defined with trial functions values and with boundary conditions. Previous numerical results of comparison of the matrixes elements analytically obtained and matrix elements obtained with described methodology are given. Analytical expressions for two-group matrix elements are considered as verification results for multi-group numerical scheme. (authors)
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The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)
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A modular computer code system called FEMSYN has been developed to solve the multigroup diffusion theory equations. The various methods that are incorporated in FEMSYN are (i) finite difference method (FDM) (ii) finite element method (FEM) and (iii) single channel flux synthesis method (SCFS). These methods are described in detail in parts II, III and IV of the present report. In this report, a comparison of the accuracy and the speed of different methods of solution for some benchmark problems are reported. The input preparation and listing of sample input and output are included in the Appendices. The code FEMSYN has been used to solve a wide variety of reactor core problems. It can be used for both LWR and PHWR applications. (author)
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Ceolin, C.; Schramm, M.; Vilhena, M.T.; Bodmann, B.E.J., E-mail: celina.ceolin@gmail.com, E-mail: marceloschramm@hotmail.com, E-mail: vilhena@pq.cnpq.br, E-mail: bardo.bodmann@ufrgs.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica
2013-07-01
In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on a expansion in Taylor Series, which was proven to be useful in [1] [2] [3]. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method [4]. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations. (author)
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Ceolin, Celina; Schramm, Marcelo; Bodmann, Bardo Ernst Josef; Vilhena, Marco Tullio Mena Barreto de [Universidade Federal do Rio Grande do Sul, Porto Alegre (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Bogado Leite, Sergio de Queiroz [Comissao Nacional de Energia Nuclear, Rio de Janeiro (Brazil)
2014-11-15
In this work the authors solved the steady state neutron diffusion equation for a multi-layer slab assuming the multi-group energy model. The method to solve the equation system is based on an expansion in Taylor Series resulting in an analytical expression. The results obtained can be used as initial condition for neutron space kinetics problems. The neutron scalar flux was expanded in a power series, and the coefficients were found by using the ordinary differential equation and the boundary and interface conditions. The effective multiplication factor k was evaluated using the power method. We divided the domain into several slabs to guarantee the convergence with a low truncation order. We present the formalism together with some numerical simulations.
International Nuclear Information System (INIS)
1 - Description of program or function: DIF3D solves multigroup diffusion theory eigenvalue, adjoint, fixed source and criticality (concentration search) problems in 1-, 2- and 3-space dimensions for orthogonal (rectangular or cylindrical), triangular and hexagonal geometries. Anisotropic diffusion coefficients are permitted. Flux and power density maps by mesh cell and region-wise balance integrals are provided. Although primarily designed for fast reactor problems, up-scattering and internal black boundary conditions are also treated. The DIF3D8.0/VARIANT8.0 release differs from the previous DIF3D7.0 release in that it includes a significantly expanded set of solution techniques using variational nodal methods. DIF3D's nodal option solves the multigroup steady state neutron diffusion equation in two- and three-dimensional hexagonal and cartesian geometries and solves the transport equation in two-and three-dimensional cartesian geometries. Eigenvalue, adjoint, fixed source and criticality (concentration) search problems are permitted as are anisotropic diffusion coefficients. Flux and power density maps by mesh cell and region-wise balance integrals are provided. Although primarily designed for fast reactor problems, up-scattering and for finite difference option only internal black boundary conditions are also treated. VARIANT solves the multigroup steady-state neutron diffusion and transport equations in two- and three-dimensional Cartesian and hexagonal geometries using variational nodal methods. The transport approximations involve complete spherical harmonic expansions up to order P5. Eigenvalue, adjoint, fixed source, gamma heating, and criticality (concentration) search problems are permitted. Anisotropic scattering is treated, and although primarily designed for fast reactor problems, up-scattering options are also included. Related and Auxiliary Programs: DIF3D reads and writes the standard interface files specified by the Committee on Computer Code
Retroviral intasomes search for a target DNA by 1D diffusion which rarely results in integration.
Jones, Nathan D; Lopez, Miguel A; Hanne, Jeungphill; Peake, Mitchell B; Lee, Jong-Bong; Fishel, Richard; Yoder, Kristine E
2016-01-01
Retroviruses must integrate their linear viral cDNA into the host genome for a productive infection. Integration is catalysed by the retrovirus-encoded integrase (IN), which forms a tetramer or octamer complex with the viral cDNA long terminal repeat (LTR) ends termed an intasome. IN removes two 3'-nucleotides from both LTR ends and catalyses strand transfer of the recessed 3'-hydroxyls into the target DNA separated by 4-6 bp. Host DNA repair restores the resulting 5'-Flap and single-stranded DNA (ssDNA) gap. Here we have used multiple single molecule imaging tools to determine that the prototype foamy virus (PFV) retroviral intasome searches for an integration site by one-dimensional (1D) rotation-coupled diffusion along DNA. Once a target site is identified, the time between PFV strand transfer events is 470 ms. The majority of PFV intasome search events were non-productive. These observations identify new dynamic IN functions and suggest that target site-selection limits retroviral integration. PMID:27108531
International Nuclear Information System (INIS)
More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be integrated in common platforms. These premises are in the origin of the NURESIM Integrated Project within the 6. European Framework Program, which is envisaged to provide the initial step towards a Common European Standard Software Platform for nuclear reactors simulations. In the frame of this project and to reach the above-mentioned goals, a 3-D multigroup nodal solver for neutron diffusion calculations called ANDES (Analytic Nodal Diffusion Equation Solver) has been developed and tested in-depth in this Thesis. ANDES solves the steady-state and time-dependent neutron diffusion equation in three-dimensions and any number of energy groups, utilizing the Analytic Coarse-Mesh Finite-Difference (ACMFD) scheme to yield the nodal coupling equations. It can be applied to both Cartesian and triangular-Z geometries, so that simulations of LWR as well as VVER, HTR and fast reactors can be performed. The solver has been implemented in a fully encapsulated way, enabling it as a module to be readily integrated in other codes and platforms. In fact, it can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. Verification of performance has shown that ANDES is a code with high order definition for whole core realistic nodal simulations. In this paper, the methodology developed and involved in ANDES is presented. (authors)
Hernandez-Charpak, J.; Hoogeboom-Pot, K.; Anderson, E.; Murnane, M.; Kapteyn, H.; Nardi, D.
2014-03-01
How is thermal transport affected by spatial confinement in nanoscale systems? In past work we and others demonstrated that the Fourier Law of heat diffusion fails for length scales smaller than the mean free path of the energy carriers in a material. Here we probe how the transition from macroscopic diffusive behavior of phonons through the quasi-ballistic regime is different for 1D and 2D nano-confined hot spots. We study a series of periodic nickel lines (1D) and dots (2D) with linewidths varying from 750 to 30 nm deposited on both sapphire and silicon substrates. The thermal relaxation of these femtosecond-laser-excited nanostructures is monitored by the diffraction of extreme ultraviolet (EUV) light obtained from tabletop high harmonic generation. The short wavelength of EUV light, combined with the coherence and ultrashort pulses of high harmonic sources, provides a unique and powerful probe for nanostructured materials on their intrinsic length and time scales. The relaxation dynamics are linked to an effective thermal boundary resistivity with the assistance of multi-physics finite element analysis to quantify the stronger deviation from macroscopic diffusive behavior as a function of nanostructure linewidth in 2D hot spots compared to 1D. This work was supported by SRC Contract 2012-OJ-2304, by NSF Award No.: DGE 1144083, and used facilities provided by the NSF Engineering Research Center in EUV Science and Technology.
Iterative 2-D/1-D methods for the 3-D neutron diffusion calculation
International Nuclear Information System (INIS)
To remedy the problems arising from assembly homogenization and de-homogenization, several efforts have been made to solve directly the heterogeneous problem with a fine mesh and to reduce the computational burden by coupling 2-D planar with 1-D axial solutions using a Transverse Leakage (TL) coupling. However, the potential for a numerical instability at a small axial mesh size has been observed. Lee et al. showed that one of the two existing methods, method A, is mathematically unstable at a small mesh size while the other, method B, is always stable. They also proposed a new method for a 2-D/1-D coupling, method C, and they showed that it is always stable and it provides the best performance in terms of the convergence rate. In this paper another algorithm, method D, is proposed and its stability is also investigated
Very high-order finite volume scheme for 1D convection diffusion problem: the implicit case
Machado, Gaspar; Clain, Stephane; Pereira, Rui
2012-01-01
We present a fourth-order in space, second-order in time finite volume scheme for transient convection diffusion problem based on the Polynomial Reconstruction Operator and a Crank-Nicholson method. A detailed description of the scheme is provided and we perform numerical tests to highlight the performance of the method in comparison with the classical Patankar method.
Ferizi, Uran; Schneider, Torben; Alipoor, Mohammad; Eufracio, Odin; Fick, Rutger H J; Deriche, Rachid; Nilsson, Markus; Loya-Olivas, Ana K; Rivera, Mariano; Poot, Dirk H J; Ramirez-Manzanares, Alonso; Marroquin, Jose L; Rokem, Ariel; Pötter, Christian; Dougherty, Robert F; Sakaie, Ken; Wheeler-Kingshott, Claudia; Warfield, Simon K; Witzel, Thomas; Wald, Lawrence L; Raya, José G; Alexander, Daniel C
2016-01-01
A large number of mathematical models have been proposed to describe the measured signal in diffusion-weighted (DW) magnetic resonance imaging (MRI) and infer properties about the white matter microstructure. However, a head-to-head comparison of DW-MRI models is critically missing in the field. To address this deficiency, we organized the "White Matter Modeling Challenge" during the International Symposium on Biomedical Imaging (ISBI) 2015 conference. This competition aimed at identifying the DW-MRI models that best predict unseen DW data. in vivo DW-MRI data was acquired on the Connectom scanner at the A.A.Martinos Center (Massachusetts General Hospital) using gradients strength of up to 300 mT/m and a broad set of diffusion times. We focused on assessing the DW signal prediction in two regions: the genu in the corpus callosum, where the fibres are relatively straight and parallel, and the fornix, where the configuration of fibres is more complex. The challenge participants had access to three-quarters of t...
A New 2D-Transport, 1D-Diffusion Approximation of the Boltzmann Transport equation
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Larsen, Edward
2013-06-17
The work performed in this project consisted of the derivation, implementation, and testing of a new, computationally advantageous approximation to the 3D Boltz- mann transport equation. The solution of the Boltzmann equation is the neutron flux in nuclear reactor cores and shields, but solving this equation is difficult and costly. The new “2D/1D” approximation takes advantage of a special geometric feature of typical 3D reactors to approximate the neutron transport physics in a specific (ax- ial) direction, but not in the other two (radial) directions. The resulting equation is much less expensive to solve computationally, and its solutions are expected to be sufficiently accurate for many practical problems. In this project we formulated the new equation, discretized it using standard methods, developed a stable itera- tion scheme for solving the equation, implemented the new numerical scheme in the MPACT code, and tested the method on several realistic problems. All the hoped- for features of this new approximation were seen. For large, difficult problems, the resulting 2D/1D solution is highly accurate, and is calculated about 100 times faster than a 3D discrete ordinates simulation.
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Selected neutron reaction nuclear data libraries and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into MATXSR format using the NJOY system on the VAX4000 computer of the IAEA. This document lists the resulting multigroup data libraries. All the multigroup data generated are available cost-free upon request from the IAEA Nuclear Data Section. (author). 9 refs
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Highlights: • New type of multi-level rebalancing approach for nodal transport. • Generally improved and more mesh-independent convergence behavior. • Importance for intended regime of 3D pin-by-pin core computations. - Abstract: A new multi-level surface rebalancing (MLSR) approach has been developed, aimed at enabling an improved non-linear acceleration of nodal flux iteration convergence in 3D steady-state and transient reactor simulation. This development is meant specifically for anticipating computational needs for solving envisaged multi-group diffusion-like SPN calculations with enhanced mesh resolution (i.e. 3D multi-box up to 3D pin-by-pin grid). For the latter grid refinement regime, the previously available multi-level coarse mesh rebalancing (MLCMR) strategy has been observed to become increasingly inefficient with increasing 3D mesh resolution. Furthermore, for very fine 3D grids that feature a very fine axial mesh as well, non-convergence phenomena have been observed to emerge. In the verifications pursued up to now, these problems have been resolved by the new approach. The novelty arises from taking the interface current balance equations defined over all Cartesian box edges, instead of the nodal volume-integrated process-rate balance equation, as an appropriate restriction basis for setting up multi-level acceleration of fine grid interface current iterations. The new restriction strategy calls for the use of a newly derived set of adjoint spectral equations that are needed for computing a limited set of spectral response vectors per node. This enables a straightforward determination of group-condensed interface current spectral coupling operators that are of crucial relevance in the new rebalancing setup. Another novelty in the approach is a new variational method for computing the neutronic eigenvalue. Within this context, the latter is treated as a control parameter for driving another, newly defined and numerically more fundamental
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D3D and D3E, branches of a computer program, solve two- and three-dimensional real and ajoint stationary multigroup neutron diffusion equations by approximating the differential equations by finite difference equations. The discrete grid is a mesh edged one, so that the neutron fluxes are calculated on surfaces separating zones to which different physical conditions apply. Different options allow to treat homogeneous, i.e. eigenvalue problems as well as inhomogeneous, i.e. external source driven problems. The linear algebraic system of the difference equations is solved by the outer and inner iterations method. An outer iteration of the homogeneous problem is the power iteration with the fission source, whereas the outer iteration of the inhomogeneous problem is an iteration with the fission source. Within the process of an outer iteration the group fluxes are determined by inner iterations, either via block overrelaxation or a method of conjugate gradients. (orig./HP)
The Suppression of Energy Discretization Errors in Multigroup Transport Calculations
Energy Technology Data Exchange (ETDEWEB)
Larsen, Edward
2013-06-17
The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to "coarsen" the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial and energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.
The Suppression of Energy Discretization Errors in Multigroup Transport Calculations
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The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.
AMPX-77, Modular System for Coupled Neutron-Gamma Multigroup Cross-Sections from ENDF/B-5
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1 - Description of program or function: The AMPX system is a system of computer programs (modules) capable of producing coupled multigroup neutron-gamma-ray cross section sets. The system is one of the standards for producing multigroup neutron, gamma-ray production, gamma-ray interaction, and coupled neutron-gamma cross-section sets from ENDF data. AMPX-produced cross sections can be used directly with a variety of diffusion theory, discrete ordinates, and Monte Carlo radiation transport computer codes. A one-dimensional Sn calculation capability is provided for general use and for cross section collapsing. Treatments are included for resonance self-shielding effects. 2 - Method of solution: The system includes a full range of features needed to: (1) produce multigroup neutron, gamma-ray production, and/or gamma-ray interaction cross-section data, (2) resonance self-shield, (3) spectrally collapse, (4) convert cross-section libraries from one format to another format, (5) execute a one- dimensional (1-D) discrete-ordinates calculation, and (6) perform miscellaneous cross section-operations. 3 - Restrictions on the complexity of the problem: The principal restriction is the availability of adequate core storage. All large modules are variably dimensioned. Certain modules will automatically use external storage (disk,tape), if in-core storage is inadequate. While these procedures are of little consequence on today's large computers with 'virtual memory' capabilities, they can be important when small-core PC's or workstations are used
Coupling of Nod1D and HOTCHANNEL: static case
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In this work the joining of the programs Nod1D and HOTCHANNEL, developed in the National Polytechnic Institute (IPN) and in the Electrical Research Institute (IIE) respectively is described. The first one allows to study the neutronic of a nuclear reactor and the second one allows to carry out the analysis of hot channel of a Boiling Water Reactor (BWR). Nod1 D is a program that it solves by nodal methods type finite element those diffusion equations in multigroup, and it is the static part of Nod Kin that it solves the diffusion equation in their time dependent part. For another side HOTCHANNEL is based on a mathematical model constituted by four conservation equations (two of mass conservation, one of motion quantity and one of energy), which are solved applying one discretization in implicit finite differences. Both programs have been verified in independent form using diverse test problems. In this work the modifications that were necessary to carry out to both for obtaining a coupled program that it provides the axial distribution of the neutron flux, the power, the burnup and the void fraction, among others parameters as much as neutronic as thermal hydraulics are described. Those are also mentioned limitations, advantages and disadvantages of the final product to which has been designated Nod1 D-HotChn. Diverse results for the Cycle 1 of the Laguna Verde Unit 1 reactor of the Nucleo electric central comparing them with those obtained directly with the CoreMasterPresto code are provided. (Author)
Coupling of Nod1D and HOTCHANNEL: static case; Acoplamiento de Nod1D y HOTCHANNEL: caso estatico
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Gomez T, A.M. [IPN-ESFM, 07738 Mexico D.F. (Mexico); Ovando C, R. [IIE-Gcia. de Energia Nuclear, Cuernavaca, Morelos (Mexico)]. e-mail: rovando@iie.org.mx
2003-07-01
In this work the joining of the programs Nod1D and HOTCHANNEL, developed in the National Polytechnic Institute (IPN) and in the Electrical Research Institute (IIE) respectively is described. The first one allows to study the neutronic of a nuclear reactor and the second one allows to carry out the analysis of hot channel of a Boiling Water Reactor (BWR). Nod1 D is a program that it solves by nodal methods type finite element those diffusion equations in multigroup, and it is the static part of Nod Kin that it solves the diffusion equation in their time dependent part. For another side HOTCHANNEL is based on a mathematical model constituted by four conservation equations (two of mass conservation, one of motion quantity and one of energy), which are solved applying one discretization in implicit finite differences. Both programs have been verified in independent form using diverse test problems. In this work the modifications that were necessary to carry out to both for obtaining a coupled program that it provides the axial distribution of the neutron flux, the power, the burnup and the void fraction, among others parameters as much as neutronic as thermal hydraulics are described. Those are also mentioned limitations, advantages and disadvantages of the final product to which has been designated Nod1 D-HotChn. Diverse results for the Cycle 1 of the Laguna Verde Unit 1 reactor of the Nucleo electric central comparing them with those obtained directly with the CoreMasterPresto code are provided. (Author)
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We study the physical-vapor-deposition of 1D bismuth nanostructures. Bi nanowire elongating along [012] and/or [110] direction as well as anisotropic Bi nano-columns are physical-vapor-deposited successfully. The coexistence and competition of surface diffusion and geometric shielding are critical to their formation as well as growth mode transition among them. Since physical-vapor-deposition is a vacuum process, we make use of it to fabricate the ohmic contact to prevent the damage to the bismuth nanostructures brought by the etching to their thick surface oxide layer. (paper)
The multigroup neutronics model of NuStar's 3D core code EGRET
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As a key component of NuStar's core analysis system for PWR application, EGRET is designed to perform steady-state coupled neutronic/hydraulic analysis of PWRs. This paper presents EGRET's unique 3D nodal diffusion model and 2D pin power reconstruction (PPR) model. Unlike the practice in most of today's production codes that iteratively solves the global 3D coarse-mesh problem and the local axially 1D fine-mesh problem to handle the axial heterogeneity within a node caused by fuel grid and partially-inserted control rod, EGRET resolves the issue by inventing a new nodal technology and introducing the adaptive meshing technique to follow the movement of control rod tip. The new nodal method employs fine-mesh heterogeneous calculation with coarse-mesh transverse coupling such that the axial heterogeneous nodes can be explicitly modeled in exact geometry and directly incorporated into the scheme of transversely coupled coarse-mesh nodal methods. Each axial channel can have its own fine-mesh division without the need of dividing the whole core into radially coupled fine-meshes. There is no need to do 1D fine-mesh and 3D coarse-mesh iteration either. While for the PPR model, EGRET adopts a group-decoupled direct fitting method, which avoids both the complication of constructing 2D analytic multigroup flux solution and any group-coupled iteration. Another unique feature of the PPR model is that it fully utilizes all the information available from 3D core calculation into the downstream PPR process. Particularly, for the first time, the 1D profiles of transversely-integrated fluxes are utilized as the additional conditions to reconstruct pin power. Numerical results of series of benchmark problems verify the good performance of EGRET's unique multi-group neutronics model. (author)
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Petersen, Claudio Z. [Universidade Federal de Pelotas, Capao do Leao (Brazil). Programa de Pos Graduacao em Modelagem Matematica; Bodmann, Bardo E.J.; Vilhena, Marco T. [Universidade Federal do Rio Grande do Sul, Porto Alegre, RS (Brazil). Programa de Pos-graduacao em Engenharia Mecanica; Barros, Ricardo C. [Universidade do Estado do Rio de Janeiro, Nova Friburgo, RJ (Brazil). Inst. Politecnico
2014-12-15
In the present work we solve in analytical representation the three dimensional neutron kinetic diffusion problem in rectangular Cartesian geometry for homogeneous and bounded domains for any number of energy groups and precursor concentrations. The solution in analytical representation is constructed using a hierarchical procedure, i.e. the original problem is reduced to a problem previously solved by the authors making use of a combination of the spectral method and a recursive decomposition approach. Time dependent absorption cross sections of the thermal energy group are considered with step, ramp and Chebyshev polynomial variations. For these three cases, we present numerical results and discuss convergence properties and compare our results to those available in the literature.
Hinata, Hirofumi; Kataoka, Tomoya
2016-08-15
We propose a belt transect setting strategy for mark-recapture experiments (MREs) to evaluate the time-independent 1D diffusion coefficient (〈Dp0〉) of marine litter in the cross-shore direction that determines the backwashing flux of the litter, based on two-year MREs for plastic floats (PFs) on Wadahama Beach, Nii-jima Island, Japan. When the alongshore width of the belt transect (Lt) was of the order of, or longer than, the length scale of wave-induced nearshore current circulation (Lc), the PFs were rarely transported alongshore across the selected transects prior to being backwashed offshore. Thus, the transect residence time became longer and showed a much weaker dependence on the transect position, in contrast to when Lt was even shorter than Lc. We therefore obtained the diffusion coefficients close to the value of (〈Dp0〉) when we set Lt to the order of, or longer than, Lc. PMID:27263978
A multigroup treatment of radiation transport
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A multi-group radiation package is outlined which will accurately handle radiation transfer problems in laser-produced plasmas. Bremsstrahlung, recombination and line radiation are included as well as fast electron Bremsstrahlung radiation. The entire radiation field is divided into a large number of groups (typically 20), which diffuse radiation energy in real space as well as in energy space, the latter occurring via electron-radiation interaction. Using this model a radiation transport code will be developed to be incorporated into MEDUSA. This modified version of MEDUSA will be used to study radiative preheat effects in laser-compression experiments at the Central Laser Facility, Rutherford Laboratory. The model is also relevant to heavy ion fusion studies. (author)
Procedure to Generate the MPACT Multigroup Library
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Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2015-12-17
The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.
Procedure to Generate the MPACT Multigroup Library
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The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.
Laginha Silva, Patricia; Martins, Flávio A.; Boski, Tomász; Sampath, Dissanayake M. R.
2010-05-01
processes. In this viewpoint the system is broken down into its fundamental components and processes and the model is build up by selecting the important processes regardless of its time and space scale. This viewpoint was only possible to pursue in the recent years due to improvement in system knowledge and computer power (Paola, 2000). The primary aim of this paper is to demonstrate that it is possible to simulate the evolution of the sediment river bed, traditionally studied with synthetic models, with a process-based hydrodynamic, sediment transport and morphodynamic model, solving explicitly the mass and momentum conservation equations. With this objective, a comparison between two mathematical models for alluvial rivers is made to simulate the evolution of the sediment river bed of a conceptual 1D embayment for periods in the order of a thousand years: the traditional synthetic basin infilling aggregate diffusive type model based on the diffusion equation (Paola, 2000), used in the "synthesist" viewpoint and the process-based model MOHID (Miranda et al., 2000). The simulation of the sediment river bed evolution achieved by the process-based model MOHID is very similar to those obtained by the diffusive type model, but more complete due to the complexity of the process-based model. In the MOHID results it is possible to observe a more comprehensive and realistic results because this type of model include processes that is impossible to a synthetic model to describe. At last the combined effect of tide, sea level rise and river discharges was investigated in the process based model. These effects cannot be simulated using the diffusive type model. The results demonstrate the feasibility of using process based models to perform studies in scales of 10000 years. This is an advance relative to the use of synthetic models, enabling the use of variable forcing. REFERENCES • Briggs, L.I. and Pollack, H.N., 1967. Digital model of evaporate sedimentation. Science, 155, 453
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The spatial eigenfunction expansion method is used to solve the multigroup time-dependent diffusion equation when the absorption cross-section in the thermal group is a function of time. An expression for the multi region reactor transfer function is obtained. Some numerical results for two energy groups are also presented. (author)
A discretization of the multigroup PN radiative transfer equation on general meshes
Hermeline, F.
2016-05-01
We propose and study a finite volume method of discrete duality type for discretizing the multigroup PN approximation of radiative transfer equation on general meshes. This method is second order-accurate on a very large variety of meshes, stable under a Courant-Friedrichs-Lewy condition and it preserves naturally the diffusion asymptotic limit.
A 3D multigroup transport kinetics code in hexagonal geometry for fast reactor transient analysis
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A description of the 3D multigroup diffusion/transport kinetics code HEXNODYN is given and numerical results are reported. HEXNODYN couples time integration by the quasi-static method with space integration by HEXNOD's analytic (diffusion option) or discrete ordinates (transport option) nodal method. An equivalent hexagonal version of the KfK rod ejection problem has been set up to validate the diffusion option by comparison with available 2D diffusion codes. The transport option has been validated by comparison with the diffusion option. Numerical results indicate that the diffusion option may be considered as fully validated while the transport version is at least internally consistent
A new method for the calculation of diffusion coefficients with Monte Carlo
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This paper presents a new Monte Carlo-based method for the calculation of diffusion coefficients. One distinctive feature of this method is that it does not resort to the computation of transport cross sections directly, although their functional form is retained. Instead, a special type of tally derived from a deterministic estimate of Fick's Law is used for tallying the total cross section, which is then combined with a set of other standard Monte Carlo tallies. Some properties of this method are presented by means of numerical examples for a multi-group 1-D implementation. Calculated diffusion coefficients are in general good agreement with values obtained by other methods. (author)
A New Method for the Calculation of Diffusion Coefficients with Monte Carlo
Dorval, Eric
2014-06-01
This paper presents a new Monte Carlo-based method for the calculation of diffusion coefficients. One distinctive feature of this method is that it does not resort to the computation of transport cross sections directly, although their functional form is retained. Instead, a special type of tally derived from a deterministic estimate of Fick's Law is used for tallying the total cross section, which is then combined with a set of other standard Monte Carlo tallies. Some properties of this method are presented by means of numerical examples for a multi-group 1-D implementation. Calculated diffusion coefficients are in general good agreement with values obtained by other methods.
A numerical model for multigroup radiation hydrodynamics
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We present in this paper a multigroup model for radiation hydrodynamics to account for variations of the gas opacity as a function of frequency. The entropy closure model (M1) is applied to multigroup radiation transfer in a radiation hydrodynamics code. In difference from the previous grey model, we are able to reproduce the crucial effects of frequency-variable gas opacities, a situation omnipresent in physics and astrophysics. We also account for the energy exchange between neighbouring groups which is important in flows with strong velocity divergence. These terms were computed using a finite volume method in the frequency domain. The radiative transfer aspect of the method was first tested separately for global consistency (reversion to grey model) and against a well-established kinetic model through Marshak wave tests with frequency-dependent opacities. Very good agreement between the multigroup M1 and kinetic models was observed in all tests. The successful coupling of the multigroup radiative transfer to the hydrodynamics was then confirmed through a second series of tests. Finally, the model was linked to a database of opacities for a Xe gas in order to simulate realistic multigroup radiative shocks in Xe. The differences with the previous grey models are discussed.
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In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the Sn solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice
Chacón Rebollo, Tomás
2015-03-01
This paper introduces a variational multi-scale method where the sub-grid scales are computed by spectral approximations. It is based upon an extension of the spectral theorem to non necessarily self-adjoint elliptic operators that have an associated base of eigenfunctions which are orthonormal in weighted L2 spaces. This allows to element-wise calculate the sub-grid scales by means of the associated spectral expansion. We propose a feasible VMS-spectral method by truncation of this spectral expansion to a finite number of modes. We apply this general framework to the convection-diffusion equation, by analytically computing the family of eigenfunctions. We perform a convergence and error analysis. We also present some numerical tests that show the stability of the method for an odd number of spectral modes, and an improvement of accuracy in the large resolved scales, due to the adding of the sub-grid spectral scales.
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1 - Description of problem or function: KENO is a multigroup, Monte Carlo criticality code containing a special geometry package which allows easy description of systems composed of cylinders, spheres, and cuboids (rectangular parallelepipeds) arranged in any order with only one restriction. They cannot be rotated or translated. Each geometrical region must be described as completely enclosing all regions interior to it. For systems not describable using this special geometry package, the program can use the generalized geometry package (GEOM) developed for the O5R Monte Carlo code. It allows any system that can be described by a collection of planes and/or quadratic surfaces, arbitrarily oriented and intersecting in arbitrary fashion. The entire problem can be mocked up in generalized geometry, or one generalized geometry unit or box type can be used alone or in combination with standard KENO units or box types. Rectangular arrays of fissile units are allowed with or without external reflector regions. Output from KENO consists of keff for the system plus an estimate of its standard deviation and the leakage, absorption, and fissions for each energy group plus the totals for all groups. Flux as a function of energy group and region and fission densities as a function of region are optional output. KENO-4: Added features include a neutron balance edit, PICTURE routines to check the input geometry, and a random number sequencing subroutine written in FORTRAN-4. 2 - Method of solution: The scattering treatment used in KENO assumes that the differential neutron scattering cross section can be represented by a P1 Legendre polynomial. Absorption of neutrons in KENO is not allowed. Instead, at each collision point of a neutron tracking history the weight of the neutron is reduced by the absorption probability. When the neutron weight has been reduced below a specified point for the region in which the collision occurs, Russian roulette is played to determine if the
Multigroup albedo method applied to gamma radiation shielding
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The Albedo method, when applied to shielding calculations, is characterized by following the radiation through the materials, determining the reflected, absorbed and transmitted fractions of the incident current, independently of flux calculations. The excellent results obtained to neutron shielding cases in which the diffusion approximation could be applied motivated this work, where the method was applied in order to develop a multigroup and multilayered algorithm. A gamma radiation shielding simulation was carried out to a system constituted by three infinite slabs of varied materials and six energy groups. The results obtained by Albedo Method were the same generated by ANISN, a consecrated deterministic nuclear code. Concludingly, this work demonstrates the validity of Albedo Method to gamma radiation shielding analysis through its agreement with the full Transport Equation. (author)
Parallel computation of multigroup reactivity coefficient using iterative method
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One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium
Establishment of multi-groups atomic parametric database
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A method is given to establish multi-groups atomic parametric database for multi-groups radiation transport equation. The equation can be used in calculating the X-ray radiation from plasma. Several methods to check the calculation of the multi-groups database is also given. A 20 groups atomic parametric database of Au element with grid of 20 (plasma density) x 20 (electron temperature) x 20 (photon temperature) is given too
Cross section probability tables in multi-group transport calculations
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The use of cross section probability tables in multigroup transport calculations is presented. Emphasis is placed on how probability table parameters are generated in a multigroup cross section processor and how existing transport codes must be modifed to use them. In order to illustrate the accuracy obtained by using probability tables, results are presented for a variety of neutron and photon transport problems
A Note on Multigroup Comparisons Using SAS PROC CALIS
Jones-Farmer, L. Allison; Pitts, Jennifer P.; Rainer, R. Kelly
2008-01-01
Although SAS PROC CALIS is not designed to perform multigroup comparisons, it is believed that SAS can be "tricked" into doing so for groups of equal size. At present, there are no comprehensive examples of the steps involved in performing a multigroup comparison in SAS. The purpose of this article is to illustrate these steps. We demonstrate…
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In multigroup calculations of reactivity and sensitivity coefficients, methodical errors can appear if the interdependence of multigroup constants is not taken into account. For this effect to be taken into account, so-called implicit components of the aforementioned values are introduced. A simple technique for computing these values is proposed. It is based on the use of subgroup parameters.
Multigroup neutron dose calculations for proton therapy
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We have developed tools for the preparation of coupled multigroup proton/neutron cross section libraries. Our method is to use NJOY to process evaluated nuclear data files for incident particles below 150 MeV and MCNPX to produce data for higher energies. We modified the XSEX3 program of the MCNPX code system to produce Legendre expansions of scattering matrices generated by sampling the physics models that are comparable to the output of the GROUPR routine of NJOY. Our code combines the low and high energy scattering data with user input stopping powers and energy deposition cross sections that we also calculated using MCNPX. Our code also calculates momentum transfer coefficients for the library and optionally applies an energy straggling model to the scattering cross sections and stopping powers. The motivation was initially for deterministic solution of space radiation shielding calculations using Attila, but noting that proton therapy treatment planning may neglect secondary neutron dose assessments because of difficulty and expense, we have also investigated the feasibility of multi group methods for this application. We have shown that multigroup MCNPX solutions for secondary neutron dose compare well with continuous energy solutions and are obtainable with less than half computational cost. This efficiency comparison neglects the cost of preparing the library data, but this becomes negligible when distributed over many multi group calculations. Our deterministic calculations illustrate recognized obstacles that may have to be overcome before discrete ordinates methods can be efficient alternatives for proton therapy neutron dose calculations
Multigroup neutron dose calculations for proton therapy
Energy Technology Data Exchange (ETDEWEB)
Kelsey Iv, Charles T [Los Alamos National Laboratory; Prinja, Anil K [Los Alamos National Laboratory
2009-01-01
We have developed tools for the preparation of coupled multigroup proton/neutron cross section libraries. Our method is to use NJOY to process evaluated nuclear data files for incident particles below 150 MeV and MCNPX to produce data for higher energies. We modified the XSEX3 program of the MCNPX code system to produce Legendre expansions of scattering matrices generated by sampling the physics models that are comparable to the output of the GROUPR routine of NJOY. Our code combines the low and high energy scattering data with user input stopping powers and energy deposition cross sections that we also calculated using MCNPX. Our code also calculates momentum transfer coefficients for the library and optionally applies an energy straggling model to the scattering cross sections and stopping powers. The motivation was initially for deterministic solution of space radiation shielding calculations using Attila, but noting that proton therapy treatment planning may neglect secondary neutron dose assessments because of difficulty and expense, we have also investigated the feasibility of multi group methods for this application. We have shown that multigroup MCNPX solutions for secondary neutron dose compare well with continuous energy solutions and are obtainable with less than half computational cost. This efficiency comparison neglects the cost of preparing the library data, but this becomes negligible when distributed over many multi group calculations. Our deterministic calculations illustrate recognized obstacles that may have to be overcome before discrete ordinates methods can be efficient alternatives for proton therapy neutron dose calculations.
Multigroup Free-atom Doppler-broadening Approximation. Theory
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Gray, Mark Girard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-11-06
Multigroup cross sections at a one target temperature can be Doppler-broadened to multigroup cross sections at a higher target temperature by matrix multiplication if the group structure suf- ficiently resolves the original temperature continuous energy cross section. Matrix elements are the higher temperature group weighted averages of the integral over the lower temperature group boundaries of the free-atom Doppler-broadening kernel. The results match theory for constant and 1/v multigroup cross sections at 618 lanl group structure resolution.
Modelling and simulations of macroscopic multi-group pedestrian flow
Mahato, Naveen K; Tiwari, Sudarshan
2016-01-01
We consider a multi-group microscopic model for pedestrian flow describing the behaviour of large groups. It is based on an interacting particle system coupled to an eikonal equation. Hydrodynamic multi-group models are derived from the underlying particle system as well as scalar multi-group models. The eikonal equation is used to compute optimal paths for the pedestrians. Particle methods are used to solve the macroscopic equations. Numerical test cases are investigated and the models and, in particular, the resulting evacuation times are compared for a wide range of different parameters.
Multigroup fast fission factor treatment in a thermal reactor lattice
International Nuclear Information System (INIS)
A multigroup procedure for the studies of the fast fission effects in the thermal reactor lattice and the calculation of the fast fission factor was developed. The Monte Carlo method and the multigroup procedure were combined to calculate the fast neutron interaction and backscattering effects in a reactor lattice. A set of probabilities calculated by the Monte Carlo method gives a multigroup spectrum of neutrons coming from the moderator and entering the fuel element. Thus, the assumptions adopted so far in defining and calculating the fast fission factor has been avoided, and a new definition including the backscattering and interaction effects in a reactor lattice have been given. (author)
Multigroup cross section library; WIMS library
International Nuclear Information System (INIS)
The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings
Multi-group neutron transport theory
International Nuclear Information System (INIS)
Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author)
Application of equivalence methods on Monte Carlo method based homogenization multi-group constants
International Nuclear Information System (INIS)
The multi-group constants generated via continuous energy Monte Carlo method do not satisfy the equivalence between reference calculation and diffusion calculation applied in reactor core analysis. To the satisfaction of the equivalence theory, general equivalence theory (GET) and super homogenization method (SPH) were applied to the Monte Carlo method based group constants, and a simplified reactor core and C5G7 benchmark were examined with the Monte Carlo constants. The results show that the calculating precision of group constants is improved, and GET and SPH are good candidates for the equivalence treatment of Monte Carlo homogenization. (authors)
A code to calculate multigroup constants for fast neutron reactor
International Nuclear Information System (INIS)
KQCS-2 code is a new improved version of KQCS code, which was designed to calculate multigroup constants for fast neutron reactor. The changes and improvements on KQCS are described in this paper. (author)
Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen IV Reactors
Energy Technology Data Exchange (ETDEWEB)
Rahnema, Farzad; Haghighat, Alireza; Ougouag, Abderrafi
2013-11-29
The objective of this proposal is the development of a consistent multi-group theory that accurately accounts for the energy-angle coupling associated with collapsed-group cross sections. This will allow for coarse-group transport and diffusion theory calculations that exhibit continuous energy accuracy and implicitly treat cross- section resonances. This is of particular importance when considering the highly heterogeneous and optically thin reactor designs within the Next Generation Nuclear Plant (NGNP) framework. In such reactors, ignoring the influence of anisotropy in the angular flux on the collapsed cross section, especially at the interface between core and reflector near which control rods are located, results in inaccurate estimates of the rod worth, a serious safety concern. The scope of this project will include the development and verification of a new multi-group theory enabling high-fidelity transport and diffusion calculations in coarse groups, as well as a methodology for the implementation of this method in existing codes. This will allow for a higher accuracy solution of reactor problems while using fewer groups and will reduce the computational expense. The proposed research represents a fundamental advancement in the understanding and improvement of multi- group theory for reactor analysis.
Radiation Transport for Explosive Outflows: A Multigroup Hybrid Monte Carlo Method
Wollaeger, Ryan T; Graziani, Carlo; Couch, Sean M; Jordan, George C; Lamb, Donald Q; Moses, Gregory A
2013-01-01
We explore the application of Implicit Monte Carlo (IMC) and Discrete Diffusion Monte Carlo (DDMC) to radiation transport in strong fluid outflows with structured opacity. The IMC method of Fleck & Cummings is a stochastic computational technique for nonlinear radiation transport. IMC is partially implicit in time and may suffer in efficiency when tracking Monte Carlo particles through optically thick materials. The DDMC method of Densmore accelerates an IMC computation where the domain is diffusive. Recently, Abdikamalov extended IMC and DDMC to multigroup, velocity-dependent neutrino transport with the intent of modeling neutrino dynamics in core-collapse supernovae. Densmore has also formulated a multifrequency extension to the originally grey DDMC method. In this article we rigorously formulate IMC and DDMC over a high-velocity Lagrangian grid for possible application to photon transport in the post-explosion phase of Type Ia supernovae. The method described is suitable for a large variety of non-mono...
A general multigroup formulation of the analytic nodal method
International Nuclear Information System (INIS)
In this paper the theoretical description of an alternative approach to the Analytic Nodal Method is given, in which a full multigroup formulations is developed. This approach differs from the well known QUANDRY approach in three aspects. Firstly, a notation which is more widely used in Quantum Mechanics has been adopted to enable a clear and concise presentation of this multigroup approach. A basis transformation is then used to reduce the directional equations to a scalar form and finally, Green's secondary identity is used to rewrite each of the resulting scalar equations in a form which eventually leads to a response matrix, as opposed to using classical methods to actually solve the coupled multigroup directional equations
Scattering approach to classical quasi-1D transport
Kogan, Eugene
1996-01-01
General dynamical transport of classical particles in disordered quasi-1D samples is viewed in the framework of scattering approach. Simple equation for the transfer-matrix is obtained within this unified picture. In the case of diffusive transport the solution of this equation exactly coincides with the solution of diffusion equation.
Energy Technology Data Exchange (ETDEWEB)
Smith, L.A.; Gallmeier, F.X. [Oak Ridge Institute for Science and Energy, TN (United States); Gehin, J.C. [Oak Ridge National Lab., TN (United States)] [and others
1995-05-01
The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are {approx} 13%, while the average differences are < 8%.
Multigroup Confirmatory Factor Analysis: Locating the Invariant Referent Sets
French, Brian F.; Finch, W. Holmes
2008-01-01
Multigroup confirmatory factor analysis (MCFA) is a popular method for the examination of measurement invariance and specifically, factor invariance. Recent research has begun to focus on using MCFA to detect invariance for test items. MCFA requires certain parameters (e.g., factor loadings) to be constrained for model identification, which are…
International Nuclear Information System (INIS)
We witnessed an initial hyped period and enthusiasm on carbon nano tubes in the 1990s later went through a significant expansion into nano tubes of other materials (metal di chalcogenides, boron nitride, etc.) as well as various nano wires and nano rods. While much of the hype might have gone, the research on one-dimensional (1D) nano materials has matured as one of the most active research areas within the nano science and nano technology community, flourishing with ample, exciting, and new research opportunities. Just like any other research frontier, researchers working in the 1D nano materials field are constantly striving to develop new fundamental science as well as potential applications. It remains a common belief that versatility and tunability of 1D nano materials would challenge many new rising tasks coming from our resource and energy demanding modern society. The traditional semiconductor industry has produced so many devices and systems from transistors, sensors, lasers, and LEDs to more sophisticated solar panels, which are now part of our daily lives. By down sizing the core components or parts to 1D form, one might wonder how fundamentally the dimensionality and morphology would impact the device performance, this is, as always, requiring us to fully understand the structure-property relationship in 1D nano materials. It may be equally crucial in connecting discovery-driven fundamental science to market-driven technology industry concerning potentially relevant findings derived from these novel materials. The importance of a platform that allows active researchers in this field to present their new development in a timely and efficient manner is therefore self-evident. Following the success of two early special issues devoted to 1D nano materials, this is the third one in a row organized by the same group of guest editors, attesting that such a platform has been well received by the readers
Application de la methode des sous-groupes au calcul Monte-Carlo multigroupe
Martin, Nicolas
effects of the scattering reaction consistent with the subgroup method. In this study, we generalize the Discrete Angle Technique, already proposed for homogeneous, multigroup cross sections, to isotopic cross sections on the form of probability tables. In this technique, the angular density is discretized into probability tables. Similarly to the cross-section case, a moment approach is used to compute the probability tables for the scattering cosine. (4) The introduction of a leakage model based on the B1 fundamental mode approximation. Unlike deterministic lattice packages, most Monte Carlo-based lattice physics codes do not include leakage models. However the generation of homogenized and condensed group constants (cross sections, diffusion coefficients) require the critical flux. This project has involved the development of a program into the DRAGON framework, written in Fortran 2003 and wrapped with a driver in C, the GANLIB 5. Choosing Fortran 2003 has permitted the use of some modern features, such as the definition of objects and methods, data encapsulation and polymorphism. The validation of the proposed code has been performed by comparison with other numerical methods: (1) The continuous-energy Monte Carlo method of the SERPENT code. (2) The Collision Probability (CP) method and the discrete ordinates (SN) method of the DRAGON lattice code. (3) The multigroup Monte Carlo code MORET, coupled with the DRAGON code. Benchmarks used in this work are representative of some industrial configurations encountered in reactor and criticality-safety calculations: (1)Pressurized Water Reactors (PWR) cells and assemblies. (2) Canada-Deuterium Uranium Reactors (CANDU-6) clusters. (3) Critical experiments from the ICSBEP handbook (International Criticality Safety Benchmark Evaluation Program).
Multi-group calculations for fast reactors
International Nuclear Information System (INIS)
The paper deals with various causes of error in calculations. The first part sets out the mathematical approximations (diffusion approximation, Sn method, etc.), the numerical resolution methods (effect of integration step), the models used, and the implications of these various factors in the determination of the principal characteristics of a fast neutron reactor. The second part studies the effect on reactivity of variations of element cross-sections, using various fuels, in a reactor of rather hard spectrum. (author)
International Nuclear Information System (INIS)
SCALE 6 computes problem-dependent multigroup (MG) cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic pointwise (PW) transport calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The PW calculation is performed by the CENTRM code using a 1-D cylindrical Wigner-Seitz model with the white boundary condition instead of the real rectangular cell shape to represent a lattice unit cell. The pointwise fluxes computed by CENTRM are not exact because a 1-D model is used for the transport calculation, which introduces discrepancies in the MG self-shielded cross sections, resulting in some deviation in the eigenvalue. In order to solve this problem, the method of characteristics (MOC) has been applied to enable the CENTRM PW transport calculation for a 2-D square pin cell. The computation results show that the new BONAMI/CENTRM-MOC procedure produces very precise self-shielded cross sections compared to MCNP reaction rates.
Multigroup cross sections of resonant nuclei considering moderator mass differences
International Nuclear Information System (INIS)
The multigroup constants library MGCL in the nuclear criticality safety evaluation code system JACS has been produced by the Bondarenko method to treat self-shielding effects. For estimating errors of this treatment, the multigroup cross sections of MGCL are compared with those obtained by precise treatment, i.e. with the weighted cross sections by ultra-fine spectra of neutron. The precise calculations are made for homogeneous mixtures of a resonant nucleus (235U, 238U, 239Pu, 240Pu, 242Pu or 56Fe) and a fictitious moderator nucleus with mass number 1, 12 or 200. The ultra-fine spectrum is calculated by the RABBLE code. Distinct differences are found in the self-shielding factors by comparisons between both treatments. Moreover, as the mass number increases, depressions of the self-shielding factor at the resonance peaks and its enhancements at the window of resonances are observed. (author)
Cyclotron radiation by a multi-group method
International Nuclear Information System (INIS)
A multi-energy group technique is developed to study conditions under which cyclotron radiation emission can shift a Maxwellian electron distribution into a non-Maxwellian; and if the electron distribution is non-Maxwellian, to study the rate of cyclotron radiation emission as compared to that emitted by a Maxwellian having the same mean electron density and energy. The assumptions in this study are: the electrons should be in an isotropic medium and the magnetic field should be uniform. The multi-group technique is coupled into a multi-group Fokker-Planck computer code to study electron behavior under the influence of cyclotron radiation emission in a self-consistent fashion. Several non-Maxwellian distributions were simulated to compare their cyclotron emissions with the corresponding energy and number density equivalent Maxwellian distribtions
Optimal calculational schemes for solving multigroup photon transport problem
International Nuclear Information System (INIS)
A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems
Multigroup-multiwaves Lisrel modeling in tourist satisfaction analysis
Cristina Bernini; Silvia Cagnone
2013-01-01
The paper analyzes the influence of tourist heterogeneity on the Tourist Local System Overall Satisfaction and its changes over time. We investigate two aspects: if different tourists segmented according to their trip motivation (seaside, conference and sport) show the same pattern of evaluation toward some relevant features of the TLS and if the evaluation scheme is dynamic. At this aim, a Multigroup-Multiwaves Lisrel model is estimated on a data set from the Tourist Satisfaction Survey, con...
Nuclear data and multigroup methods in fast reactor calculations
International Nuclear Information System (INIS)
The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)
A nodal expansion method for solving the multigroup SP3 equations in the reactor code DYN3D
International Nuclear Information System (INIS)
The core model DYN3D which has been developed for three-dimensional analyses of steady states and transients in thermal reactors with quadratic or hexagonal fuel assemblies is based on nodal methods for the solution of the two-group neutron diffusion equation. Loading cores with higher content of MOX fuel, the increase of the fuel cycle length and new types of reactors are challenging for these standard methods. A nodal expansion method for solving the equations of the simplified P3 approximation (SP3) of the multigroup transport equation was developed to improve the accuracy of the DYN3D code. In this paper, the method used in DYN3D-SP3 is described. It is applied for the pin-wise calculation of a steady state of the OECD/NEA and U.S. NRC PWR MOX/UO2 Core Transient Benchmark. The eigenvalue keff, assembly powers and the pin powers are computed. The results calculated with different approaches including diffusion theory are compared with the reference solution obtained from a heterogeneous transport calculation with the code DeCART. Different approaches of the diffusion coefficient used in the SP3 equations are investigated. The SP3 results obtained with the transport cross section of multigroup diffusion theory show the smallest deviations from the reference solution. These deviations are in the same order as the results of the code DORT, whereas the DORT and DYN3D calculations were carried out with the same library of group constants for homogenized pin cells. (authors)
SERKON program for compiling a multigroup library to be used in BETTY calculation
International Nuclear Information System (INIS)
A SERKON-type program was written to compile data sets generated by FEDGROUP-3 into a multigroup library for BETTY calculation. A multigroup library was generated from the ENDF/B-IV data file and tested against the TRX-1 and TRX-2 lattices with good results. (author)
MUXS: a code to generate multigroup cross sections for sputtering calculations
International Nuclear Information System (INIS)
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc
Nonparametric Multi-group Membership Model for Dynamic Networks
Kim, Myunghwan; Leskovec, Jure
2013-01-01
Relational data-like graphs, networks, and matrices-is often dynamic, where the relational structure evolves over time. A fundamental problem in the analysis of time-varying network data is to extract a summary of the common structure and the dynamics of the underlying relations between the entities. Here we build on the intuition that changes in the network structure are driven by the dynamics at the level of groups of nodes. We propose a nonparametric multi-group membership model for dynami...
Status of multigroup cross-section data for shielding applications
International Nuclear Information System (INIS)
Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V
Multigroup Free-atom Doppler-broadening Approximation. Experiment
Energy Technology Data Exchange (ETDEWEB)
Gray, Mark Girard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-11-06
The multigroup energy Doppler-broadening approximation agrees with continuous energy Dopplerbroadening generally to within ten percent for the total cross sections of ^{1}H,^{ 56}Fe, and ^{235}U at 250 lanl. Although this is probably not good enough for broadening from room temperature through the entire temperature range in production use, it is better than any interpolation scheme between temperatures proposed to date, and may be good enough for extrapolation from high temperatures. The method deserves further study since additional improvements are possible.
Variational nodal solution algorithms for multigroup criticality problems
International Nuclear Information System (INIS)
Variational nodal transport methods are generalized for the treatment of multigroup criticality problems. The generation of variational response matrices is streamlined and automated through the use of symbolic manipulation. A new red-black partitioned matrix algorithm for the solution of the within-group equations is formulated and shown to be at once both a regular matrix splitting and a synthetic acceleration method. The methods are implemented in X- Y geometry as a module of the Argonne National Laboratory code DIF3D. For few group problems highly accurate P3 eigenvalues are obtained with computing times comparable to those of an existing interface-current nodal transport method
SNAP - a three dimensional neutron diffusion code
International Nuclear Information System (INIS)
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)
International Nuclear Information System (INIS)
Highlights: • A new AFEN code, MGANSP3, is developed for simplified P3 (SP3) calculations. • Surface averaged partial currents are used for coupling the nodes. • Coarse group rebalancing method is applied to increase the speed of calculations. • Four benchmark problems are used to examine the accuracy of the MGANSP3 code. - Abstract: In this study, a new analytic function expansion nodal (AFEN) method was developed to solve multi-group and three dimensional neutron simplified P3 equations (SP3) in reactor cores with rectangular fuel assemblies. In this method, the intranodal fluxes are expanded into a set of analytic basis functions for each group and moment. The nodes are coupled through the surface averaged partial currents at each nodal interface. Thus, six boundary conditions at each group and Legendre moments have been considered. Coarse group rebalancing (CGR) method was applied to increase the speed of code calculations. The code takes few-groups cross sections produced by a lattice code such as WIMS and calculates the effective multiplication factor, zeroth and second moments of the flux in multi-group energy, reactivity, and the relative power density at each fuel assembly. The numerical results for different benchmark problems demonstrate that solution of SP3 equations by our AFEN method improves both effective multiplication factor (keff) and power distribution compared to our AFEN diffusion method, especially in heterogeneous geometry and mixed-oxide (MOX) fuel problems
BETA-S, Multi-Group Beta-Ray Spectra
International Nuclear Information System (INIS)
1 - Description of program or function: BETA-S calculates beta-decay source terms and energy spectra in multigroup format for time-dependent radionuclide inventories of actinides, fission products, and activation products. Multigroup spectra may be calculated in any arbitrary energy-group structure. The code also calculates the total beta energy release rate from the sum of the average beta-ray energies as determined from the spectral distributions. BETA-S also provides users with an option to determine principal beta-decaying radionuclides contributing to each energy group. The CCC-545/SCALE 4.3 (or SCALE4.2) code system must be installed on the computer before installing BETA-S, which requires the SCALE subroutine library and nuclide-inventory generation from the ORIGEN-S code. 2 - Methods:Well-established models for beta-energy distributions are used to explicitly represent allowed, and 1., 2. - and 3. -forbidden transition types. Forbidden non-unique transitions are assumed to have a spectral shape of allowed transitions. The multigroup energy spectra are calculated by numerically integrating the energy distribution functions using an adaptive Simpson's Rule algorithm. Nuclide inventories are obtained from a binary interface produced by the ORIGEN-S code. BETA-S calculates the spectra for all isotopes on the binary interface that have associated beta-decay transition data in the ENSDF-95 library, developed for the BETA-S code. This library was generated from ENSDF data and contains 715 materials, representing approximately 8500 individual beta transition branches. 3 - Restrictions on the complexity of the problem: The algorithms do not treat positron decay transitions or internal conversion electrons. The neglect of positron transitions in inconsequential for most applications involving aggregate fission products, since most of the decay modes are via electrons. The neglect of internal conversion electrons may impact on the accuracy of the spectrum in the low
Generation of subgroup parameters from JENDL-2 based multigroup data set for FBR core materials
International Nuclear Information System (INIS)
Subgroup method gives a more accurate treatment to the resonance absorption in nuclear reactors, especially when it is heterogeneous, than the usual multigroup method. An algorithm has been developed based on a modified form of Roth's procedure, to calculate subgroup parameters, from the multigroup table of self-shielding factors given against a set of temperatures and dilution cross sections. A program SPART has been written with this algorithm, and it has been used to generate subgroup parameters for some important fast reactor core materials from the JENDL-2 based multigroup set, recently created and validated at IGCAR. In this report, the algorithm is discussed, and the subgroup parameters generated are presented. (author)
International Nuclear Information System (INIS)
Adaptive matrix formation (AMF) method has been developed for the numerical solution of the transient multigroup neutron diffusion and delayed precursor equations in two- and three-dimensional geometry. The method is applied to a general class of two- and three- dimensional problems. The results of numerical experiments, as well as comparison with space-time experimental results indicate that the method is accurate and that the two- and three-dimensional calculations can be performed at 'reasonable' computer costs. Moreover, the AMF method offers the flexibility of using smaller time steps between flux shape calculations to achieve a specified accuracy and capability, without encountering numerical problems that occur in the other conventional methods. There is a large considerable saving in computer time and costs due to the partitioning of the matrix adopted in the presented AMF method. The two- and three-dimensional problems were analyzed with the present calculations model to illustrate the accuracy and stability of the method. Furthermore, the stability of the investigated method has been tested for sinusoidal, ramp, and step-change reactivity insertions. The results are in a good agreement with those of the other less approximate methods, including the problems in which the reflector zone is perturbed
Reflector modelling of small high leakage cores making use of multi-group nodal equivalence theory
International Nuclear Information System (INIS)
This research focuses on modelling reflectors in typical material testing reactors (MTRs). Equivalence theory is used to homogenise and collapse detailed transport solutions to generate equivalent nodal parameters and albedo boundary conditions for reflectors, for subsequent use in full core nodal diffusion codes. This approach to reflector modelling has been shown to be accurate for two-group large commercial light water reactor (LWR) analysis, but has not been investigated for MTRs. MTRs are smaller, with much larger leakage, environment sensitivity and multi-group spectrum dependencies than LWRs. This study aims to determine if this approach to reflector modelling is an accurate and plausible homogenisation technique for the modelling of small MTR cores. The successful implementation will result in simplified core models, better accuracy and improved efficiency of computer simulations. Codes used in this study include SCALE 6.1, OSCAR-4 and EQUIVA (the last two codes are developed and used at Necsa). The results show a five times reduction in calculational time for the proposed reduced reactor model compared to the traditional explicit model. The calculated equivalent parameters however show some sensitivity to the environment used to generate them. Differences in the results compared to the current explicit model, require more careful investigation including comparisons with a reference result, before its implementation can be recommended. (authors)
Development and verification of a nodal approach for solving the multigroup SP{sub 3} equations
Energy Technology Data Exchange (ETDEWEB)
Beckert, C. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O. Box 51 01 19, D-01314 Dresden (Germany); Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O. Box 51 01 19, D-01314 Dresden (Germany)], E-mail: U.Grundmann@fzd.de
2008-01-15
The core model DYN3D which has been developed for three-dimensional analyses of steady states and transients in thermal reactors with quadratic or hexagonal fuel assemblies is based on nodal methods for the solution of the two-group neutron diffusion equation. Loading cores with higher content of MOX fuel, the increase of the fuel cycle length, and the consideration of new reactor types are challenging for these standard methods. A nodal expansion method for solving the equations of the simplified P{sub 3} (SP{sub 3}) approximation of the multigroup transport equation was developed to improve the accuracy of the DYN3D code. The method described in the paper is verified with pinwise calculations of a steady state of the OECD/NEA and US NRC PWR MOX/UO{sub 2} Core Transient Benchmark. The used 16-group cross section library was generated for DORT calculations with homogenized pin cells. Two different approximations of the diffusion coefficient which occurs in the within-group form of the SP{sub 3} equations are investigated. Using the transport cross section for the calculation of the diffusion coefficient gives much better results than those obtained with the removal cross section. The improvement of the results in comparison to a pinwise diffusion calculation is shown. The results are compared with the DORT and the heterogeneous reference solution of the code DeCART. Concerning the SP{sub 3} calculation using the diffusion coefficient based on the transport cross section (DYN3D-SP3-TR) the deviations of the eigenvalue k{sub eff} and the assembly powers from the transport solutions of DORT and DeCART are in the same order as those between the two transport solutions themselves. The improvement of the DYN3D-SP3-TR results in comparison to the diffusion calculation is presented. As the DYN3D-SP3-TR and DORT calculations are performed with homogenized pin cells, the pin powers of the two calculations are closer to each other than to the pin powers of the DeCART solution
Development and verification of a nodal approach for solving the multigroup SP3 equations
International Nuclear Information System (INIS)
The core model DYN3D which has been developed for three-dimensional analyses of steady states and transients in thermal reactors with quadratic or hexagonal fuel assemblies is based on nodal methods for the solution of the two-group neutron diffusion equation. Loading cores with higher content of MOX fuel, the increase of the fuel cycle length, and the consideration of new reactor types are challenging for these standard methods. A nodal expansion method for solving the equations of the simplified P3 (SP3) approximation of the multigroup transport equation was developed to improve the accuracy of the DYN3D code. The method described in the paper is verified with pinwise calculations of a steady state of the OECD/NEA and US NRC PWR MOX/UO2 Core Transient Benchmark. The used 16-group cross section library was generated for DORT calculations with homogenized pin cells. Two different approximations of the diffusion coefficient which occurs in the within-group form of the SP3 equations are investigated. Using the transport cross section for the calculation of the diffusion coefficient gives much better results than those obtained with the removal cross section. The improvement of the results in comparison to a pinwise diffusion calculation is shown. The results are compared with the DORT and the heterogeneous reference solution of the code DeCART. Concerning the SP3 calculation using the diffusion coefficient based on the transport cross section (DYN3D-SP3-TR) the deviations of the eigenvalue keff and the assembly powers from the transport solutions of DORT and DeCART are in the same order as those between the two transport solutions themselves. The improvement of the DYN3D-SP3-TR results in comparison to the diffusion calculation is presented. As the DYN3D-SP3-TR and DORT calculations are performed with homogenized pin cells, the pin powers of the two calculations are closer to each other than to the pin powers of the DeCART solution. To estimate the contribution of
Intragroup Socialization for Adult Korean Adoptees: A Multigroup Analysis
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Kimberly J. Langrehr
2014-06-01
Full Text Available The purpose of the current study was to test a model of socialization among a sample of adult Korean adoptees. Based on the tenants of homophily and social identity theory, it was hypothesized that participants’ early racial and ethnic socialization experiences would account for their current intragroup friendships as adults, and that this relationship would be mediated by early intragroup contact and moderated by early ethnic identity status. The two ethnic and racial socialization variables (i.e., ethnic heritage activities and racial in-exposure significantly accounted for participants’ relationships with other Korean adoptees and nonadopted Koreans, and the effects were partially explained by early intragroup contact. Results of multigroup testing indicated the proposed socialization model was non-invariant across groups, such that the effects of ethnic heritage activities on intragroup contact and the effect of racial in-exposure on friendships with Korean adoptees were significantly different based on early ethnic identity status.
MORET: Version 4.B. A multigroup Monte Carlo criticality code
International Nuclear Information System (INIS)
MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)
Multigroup representation of fusion product orbits in a plasma column
International Nuclear Information System (INIS)
A method is derived for describing the time-depending behavior of α particles produced in a radially nonuniform slender plasma column as a distribution function among the possible orbits. A multigroup numerical approximation is introduced to analyze the development of the distribution function and its moments. Results are presented of calculations of the time-dependent α-particle energy spectrum and radial density, energy, and electron heating profiles in plasma columns with radii comparable to the α Larmor radius. This technique allows calculation of the α particle history at much more rapid rates than allowed by Monte Carlo technuques: The characteristic time scale is the α-electron slowing-down time rather than the cyclotron period
Multigroup-multiwaves Lisrel modeling in tourist satisfaction analysis
Directory of Open Access Journals (Sweden)
Cristina Bernini
2013-05-01
Full Text Available The paper analyzes the influence of tourist heterogeneity on the Tourist Local System Overall Satisfaction and its changes over time. We investigate two aspects: if different tourists segmented according to their trip motivation (seaside, conference and sport show the same pattern of evaluation toward some relevant features of the TLS and if the evaluation scheme is dynamic. At this aim, a Multigroup-Multiwaves Lisrel model is estimated on a data set from the Tourist Satisfaction Survey, conducted in Rimini from 2004 to 2006 by the Faculty of Statistics – University of Bologna. The analysis shows that tourist evaluation scheme toward Rimini is quite similar among groups and over time, suggesting that differences among tourists do not affect TLS satisfaction.
Multigroup covariance matrices for fast-reactor studies
International Nuclear Information System (INIS)
This report presents the multigroup covariance matrices based on the ENDF/B-V nuclear data evaluations. The materials and reactions have been chosen according to the specifications of ORNL-5517. Several cross section covariances, other than those specified by that report, are included due to the derived nature of the uncertainty files in ENDF/B-V. The materials represented are Ni, Cr, 16O, 12C, Fe, Na, 235U, 238U, 239Pu, 240Pu, 241Pu, and 10B (present due to its correlation to 238U). The data have been originally processed into a 52-group energy structure by PUFF-II and subsequently collapsed to smaller subgroup strutures. The results are illustrated in 52-group correlation matrix plots and tabulated into thirteen groups for convenience
International Nuclear Information System (INIS)
Starting from the radiation transport equation for homogeneous, refractive lossy media, we derive the corresponding time-dependent multifrequency diffusion equations. Zeroth and first moments of the transport equation couple the energy density, flux and pressure tensor. The system is closed by neglecting the temporal derivative of the flux and replacing the pressure tensor by its diagonal analogue. The radiation equations are coupled to a diffusion equation for the matter temperature. We are interested in modeling heating and cooling of silica (SiO2), at possibly rapid rates. Hence, in contrast to related work, we retain the temporal derivative of the radiation field. We derive boundary conditions at a planar air-silica interface taking account of reflectivities obtained from the Fresnel relations that include absorption. The spectral dimension is discretized into a finite number of intervals leading to a system of multigroup diffusion equations. Three simulations are presented. One models cooling of a silica slab, initially at 2500 K, for 10 s. The other two are 1D and 2D simulations of irradiating silica with a CO2 laser, λ = 10.59 μm. In 2D, a laser beam (Gaussian profile, r0 = 0.5 mm for 1/e decay) shines on a disk (radius = 0.4, thickness = 0.4 cm).
Kalpakkam multigroup cross section set for fast reactor applications - status and performance
International Nuclear Information System (INIS)
This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs
MCMG: a 3-D multigroup P3 Monte Carlo code and its benchmarks
International Nuclear Information System (INIS)
In this paper a 3-D Monte Carlo multigroup neutron transport code MCMG has been developed from a coupled neutron and photon transport Monte Carlo code MCNP. The continuous-energy cross section library of the MCNP code is replaced by the multigroup cross section data generated by the transport lattice code, such as the WIMS code. It maintains the strong abilities of MCNP for geometry treatment, counting, variance reduction techniques and plotting. The multigroup neutron scattering cross sections adopt the Pn (n ≤ 3) approximation. The test results are in good agreement with the results of other methods and experiments. The number of energy groups can be varied from few groups to multigroup, and either macroscopic or microscopic cross section can be used. (author)
One-Dimensional (1-D) Nanoscale Heterostructures
Institute of Scientific and Technical Information of China (English)
Guozhen SHEN; Di CHEN; Yoshio BANDO; Dmitri GOLBERG
2008-01-01
One-dimensional (1-D) nanostructures have been attracted much attention as a result of their exceptional properties, which are different from bulk materials. Among 1-D nanostructures, 1-D heterostructures with modulated compositions and interfaces have recently become of particular interest with respect to potential applications in nanoscale building blocks of future optoelectronic devices and systems. Many kinds of methods have been developed for the synthesis of 1-D nanoscale heterostructures. This article reviews the most recent development, with an emphasize on our own recent efforts, on 1-D nanoscale heterostructures, especially those synthesized from the vapor deposition methods, in which all the reactive precursors are mixed together in the reaction chamber. Three types of 1-D nanoscale heterostructures, defined from their morphologies characteristics, are discussed in detail, which include 1-D co-axial core-shell heterostructures, 1-D segmented heterostructures and hierarchical heterostructures. This article begins with a brief survey of various methods that have been developed for synthesizing 1-D nanoscale heterostructures and then mainly focuses on the synthesis, structures and properties of the above three types of nanoscale heterostructures. Finally, this review concludes with personal views towards the topic of 1-D nanoscale heterostructures.
Energy Technology Data Exchange (ETDEWEB)
Coste-Delclaux, M
2006-03-15
This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)
International Nuclear Information System (INIS)
A variational finite element-spherical harmonics method is presented for the solution of the even-parity multigroup equations with anisotropic scattering and sources. It is shown that by using a simple and natural formulation the numerical implementation of the method for any desired geometry is greatly eased and the anisotropy of scatter treated without any difficulty. Numerical examples demonstrate the ability of the resulting code to solve geometrically complex multigroup problems. (Author)
Radiation Transport for Explosive Outflows: A Multigroup Hybrid Monte Carlo Method
Wollaeger, Ryan T.; van Rossum, Daniel R.; Graziani, Carlo; Couch, Sean M.; Jordan, George C., IV; Lamb, Donald Q.; Moses, Gregory A.
2013-12-01
We explore Implicit Monte Carlo (IMC) and discrete diffusion Monte Carlo (DDMC) for radiation transport in high-velocity outflows with structured opacity. The IMC method is a stochastic computational technique for nonlinear radiation transport. IMC is partially implicit in time and may suffer in efficiency when tracking MC particles through optically thick materials. DDMC accelerates IMC in diffusive domains. Abdikamalov extended IMC and DDMC to multigroup, velocity-dependent transport with the intent of modeling neutrino dynamics in core-collapse supernovae. Densmore has also formulated a multifrequency extension to the originally gray DDMC method. We rigorously formulate IMC and DDMC over a high-velocity Lagrangian grid for possible application to photon transport in the post-explosion phase of Type Ia supernovae. This formulation includes an analysis that yields an additional factor in the standard IMC-to-DDMC spatial interface condition. To our knowledge the new boundary condition is distinct from others presented in prior DDMC literature. The method is suitable for a variety of opacity distributions and may be applied to semi-relativistic radiation transport in simple fluids and geometries. Additionally, we test the code, called SuperNu, using an analytic solution having static material, as well as with a manufactured solution for moving material with structured opacities. Finally, we demonstrate with a simple source and 10 group logarithmic wavelength grid that IMC-DDMC performs better than pure IMC in terms of accuracy and speed when there are large disparities between the magnitudes of opacities in adjacent groups. We also present and test our implementation of the new boundary condition.
Variational methods in steady state diffusion problems
International Nuclear Information System (INIS)
Classical variational techniques are used to obtain accurate solutions to the multigroup multiregion one dimensional steady state neutron diffusion equation. Analytic solutions are constructed for benchmark verification. Functionals with cubic trial functions and conservational lagrangian constraints are exhibited and compared with nonconservational functionals with respect to neutron balance and to relative flux and current at interfaces. Excellent agreement of the conservational functionals using cubic trial functions is obtained in comparison with analytic solutions
Macroscopic multigroup constants for accelerator driven system core calculation
International Nuclear Information System (INIS)
The high-level wastes stored in facilities above ground or shallow repositories, in close connection with its nuclear power plant, can take almost 106 years before the radiotoxicity became of the order of the background. While the disposal issue is not urgent from a technical viewpoint, it is recognized that extended storage in the facilities is not acceptable since these ones cannot provide sufficient isolation in the long term and neither is it ethical to leave the waste problem to future generations. A technique to diminish this time is to transmute these long-lived elements into short-lived elements. The approach is to use an Accelerator Driven System (ADS), a sub-critical arrangement which uses a Spallation Neutron Source (SNS), after separation the minor actinides and the long-lived fission products (LLFP), to convert them to short-lived isotopes. As an advanced reactor fuel, still today, there is a few data around these type of core systems. In this paper we generate macroscopic multigroup constants for use in calculations of a typical ADS fuel, take into consideration, the ENDF/BVI data file. Four energy groups are chosen to collapse the data from ENDF/B-VI data file by PREPRO code. A typical MOX fuel cell is used to validate the methodology. The results are used to calculate one typical subcritical ADS core. (author)
MCNP - transport calculations in ducts using multigroup albedo coefficients
International Nuclear Information System (INIS)
In this work, the use of multigroup albedo coefficients in Monte Carlo calculations of particle reflection and transmission by ducts is investigated. The procedure consists in modifying the MCNP code so that an albedo matrix computed previously by deterministic methods or Monte Carlo is introduced into the program to describe particle reflection by a surface. This way it becomes possible to avoid the need of considering particle transport in the duct wall explicitly, changing the problem to a problem of transport in the duct interior only and reducing significantly the difficulty of the real problem. The probability of particle reflection at the duct wall is given, for each group, as the sum of the albedo coefficients over the final groups. The calculation is started by sampling a source particle and simulating its reflection on the duct wall by sampling a group for the emerging particle. The particle weight is then reduced by the reflection probability. Next, a new direction and trajectory for the particle is selected. Numerical results obtained for the model are compared with results from a discrete ordinates code and results from Monte Carlo simulations that take particle transport in the wall into account. (author)
Multi-group SP3 approximation for simulation of a three-dimensional PWR rod ejection accident
International Nuclear Information System (INIS)
Highlights: • The multi-group SP3 method developed and implemented in PARCS for the MOX analysis. • The verifications were performed in 2D and 3D, 2G and MG, diffusion and transport, with and without feedback. • All results show consistency with the reference results obtained from the ANL PN transport code VARIANT for steady-state and transport calculations. • It was found that the SP3 angular approximation captures sufficient transport effects for both steady-state and transient, and provides essentially the same results as the VARIANT P5 method. • From the transient results of the full-core problem, it was noted that MG is more conservative than 2G, and P1 is more conservative than SP3. - Abstract: Previous researchers have shown that the simplified P3 (SP3) approximation is capable of providing sufficiently high accuracy for both static and transient simulations for reactor core analysis with considerably less computational expense than higher order transport methods such as the discrete ordinate or the full spherical harmonics methods. The objective of this paper is to provide a consistent comparison of two-group (2G) and multi-group (MG) diffusion and SP3 transport for rod ejection accident (REA) in a practical light water reactor (LWR) problem. The analysis is performed on two numerical benchmarks, a 3 × 3 assembly mini-core and a full pressurized water reactor (PWR) core. The calculations were performed using pin homogenized and assembly homogenized cross sections for a series of benchmarks of increasing difficulty, in two-dimensional (2D) and three-dimensional (3D), 2G and MG, diffusion and transport, as well as with and without feedback. All results show consistency with the reference results obtained from higher-order methods. It is demonstrated that the analyzed problems show small group-homogenization effects, but relatively significant transport effects which are satisfactorily addressed by the SP3 transport method. The sensitivity tests
Dorval, Eric
2016-01-01
Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient model...
International Nuclear Information System (INIS)
Current theories for approximating the effects of stochastic media on radiation transport assume very limited physics such as one dimension, constant grey opacities, and no material energy balance equation. When applied to more complex physical problems, the standard theory fails to match the results from direct numerical simulations. This work presents the first direct numerical simulations of multigroup radiation transport coupled to a material temperature equation in a 2D stochastic medium that are compared to closures proposed by various authors. After extending it from grey to multigroup physics, one closure that is not commonly used successfully models the results in dilute systems where one material comprises less than 5% of the total. This closure is more accurate for related grey transport problems than it is for the multigroup problem. When the specific heats are material- and temperature-dependent, it is much more difficult to fit the direct numerical solutions with an approximate closure.
Optimal control in multi-group coupled within-host and between-host models
Directory of Open Access Journals (Sweden)
Eric Numfor
2016-03-01
Full Text Available We formulate and then analyze a multi-group coupled within-host model of ODEs and between-host model of ODE and first-order PDEs, using the Human Immunodeficiency Virus (HIV for illustration. The basic reproduction number of the multi-group coupled epidemiological model is derived, steady states solutions are calculated and stability analysis of equilbria is investigated. An optimal control problem for our model with drug treatment on the multi-group within-host system is formulated and analyzed. Ekeland's principle is used in proving existence and uniqueness of an optimal control pair. Numerical simulations based on the semi-implicit finite difference schemes and the forward-backward sweep iterative method are obtained.
Development of a multi-group SN transport calculation code with unstructured tetrahedral meshes
International Nuclear Information System (INIS)
This paper reviews the computational methods used in the MUST (Multi-group Unstructured geometry SN Transport) code for solving the multi-group Sn transport equation in general geometries and describes the status of development of MUST. MUST solves the multi-group transport equation with unstructured tetrahedral meshes for modeling complicated geometrical problems. For tetrahedral mesh generation, input generation, and output visualization, we developed a management program where the mesh generation is based on Gmsh and TetGen that are open softwares. The geometrical modeling is done with the commercial CAD softwares such as CATIA. MUST uses the discontinuous finite element method (DFEM) and two-sub cell balance methods with linear discontinuous expansion (LDEM-SCB) to spatially discretize the transport equation. We applied MUST to three neutron and gamma coupled test problems for testing MUST. (author)
Development of a Multi-Group Neutron Cross Section Library Generation System for PWR
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog; Hong, Ser Gi; Song, Jae Seung; Lee, Kyung Hoon; Cho, Jin Young; Kim, Ha Yong; Koo, Bon Seung; Shim, Hyung Jin; Park, Sang Yoon
2008-10-15
This report describes a generation system of multi-group cross section library which is used in the KARMA lattice calculation code. In particular, the theoretical methodologies, program structures, and input preparations for the constituent programs of the system are described in detail. The library generation system consists of the following five programs : ANJOY, GREDIT, MERIT, SUBDATA, and LIBGEN. ANJOY generates automatically the NJOY input files and two batch files for automatic NJOY run for all the nuclides considered. The automatic NJOY run gives TAPE 23 (PENDF output file of BROADR module of NJOY) and TAPE24 (GENDF output file of GROUPR module of NJOY) files for each nuclide. GREDIT prepares a formatted multi-group cross section file in which the cross sections are tabulated versus temperature and background cross section after reading the TAPE24 file. MERIT generates the hydrogen equivalence factors and the resonance integral tables by solving the slowing down equation with ultra-fine group cross sections which are prepared with the TAPE 23 file. SUBDATA generates the subgroup data including subgroup levels and weights after reading the MERIT output file. Finally, LIBGEN generates the final multi-group library file by assembling the data prepared in the previous steps and by reading the other data such as fission product yield data and decay data.The multi-group cross section library includes general multi-group cross sections, resonance data, subgroup data, fission product yield data, kappa-values (energy release per fission), and all the data which are required in the depletion calculation. The addition or elimination of the cross sections for some nuclides can be easily done by changing the LIBGEN input file if the general multi-group cross section and the subgroup data files are prepared.
Development of a Multi-Group Neutron Cross Section Library Generation System for PWR
International Nuclear Information System (INIS)
This report describes a generation system of multi-group cross section library which is used in the KARMA lattice calculation code. In particular, the theoretical methodologies, program structures, and input preparations for the constituent programs of the system are described in detail. The library generation system consists of the following five programs : ANJOY, GREDIT, MERIT, SUBDATA, and LIBGEN. ANJOY generates automatically the NJOY input files and two batch files for automatic NJOY run for all the nuclides considered. The automatic NJOY run gives TAPE 23 (PENDF output file of BROADR module of NJOY) and TAPE24 (GENDF output file of GROUPR module of NJOY) files for each nuclide. GREDIT prepares a formatted multi-group cross section file in which the cross sections are tabulated versus temperature and background cross section after reading the TAPE24 file. MERIT generates the hydrogen equivalence factors and the resonance integral tables by solving the slowing down equation with ultra-fine group cross sections which are prepared with the TAPE 23 file. SUBDATA generates the subgroup data including subgroup levels and weights after reading the MERIT output file. Finally, LIBGEN generates the final multi-group library file by assembling the data prepared in the previous steps and by reading the other data such as fission product yield data and decay data.The multi-group cross section library includes general multi-group cross sections, resonance data, subgroup data, fission product yield data, kappa-values (energy release per fission), and all the data which are required in the depletion calculation. The addition or elimination of the cross sections for some nuclides can be easily done by changing the LIBGEN input file if the general multi-group cross section and the subgroup data files are prepared
Application of diffusion theory to the transport of neutral particles in fusion plasmas
International Nuclear Information System (INIS)
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
Social exploration of 1D games
DEFF Research Database (Denmark)
Valente, Andrea; Marchetti, Emanuela
2013-01-01
In this paper the apparently meaningless concept of a 1 dimensional computer game is explored, via netnography. A small number of games was designed and implemented, in close contact with online communities of players and developers, providing evidence that 1 dimension is enough to produce intere...... interesting gameplay, to allow for level design and even to leave room for artistic considerations on 1D rendering. General techniques to re-design classic 2D games into 1D are also emerging from this exploration....
Two-dimensional multigroup finite element calculation of fast reactor in diffusion approximation
International Nuclear Information System (INIS)
When a linear element of triangular shape is used the actual finite element calculation is relatively simple. Extensive programs for mesh generation were written for easy inputting the configuration of reactors. A number of other programs were written for plotting neutron flux fields in individual groups, the power distribution, spatial plotting of fields, etc. The operation of selected programs, data preparation and operating instructions are described and examples given of data and results. All programs are written in GIER ALGOL. The used method and the developed programs have demonstrated that they are a useful instrument for the calculation of criticality and the distribution of neutron flux and power of both fast and thermal reactors. (J.B.)
Energy Technology Data Exchange (ETDEWEB)
Ceolin, Celina; Vilhena, Marco T.; Bodmann, Bardo E.J., E-mail: vilhena@pq.cnpq.b, E-mail: bardo.bodmann@ufrgs.b [Universidade Federal do Rio Grande do Sul (PROMEC/UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Alvim, Antonio Carlos Marques, E-mail: alvim@nuclear.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Energia Nuclear
2011-07-01
The authors solved analytically the neutron kinetic equations in a homogeneous slab, assuming the multi group energy model and six delayed neutron precursor groups by the Generalized Integral Laplace Transform Technique (GILTT) for a multi-layered slab. To this end, averaged values for the nuclear parameters in the multi-layered slab are used and the solution is constructed following the idea of Adomian's decomposition method upon reducing the heterogeneous problem to a set of recursive problems with constant parameters in the multi-layered slab. More specifically, the corrections that render the initially homogeneous problem into a heterogeneous one are plugged into the equation as successive source terms. To the best of our knowledge this sort of solution is novel and not found in literature. We further present some numerical simulations. (author)
Adoption as a Social Marker: The Diffusion of Products in a Multigroup Environment
Smaldino, Paul E; Hillis, Vicken; Bednar, Jenna
2015-01-01
Social identities are among the key factors driving social behavior in complex societies. Recent attention to social identity in consumer behavior indicates that individuals avoid products that might signal membership in an outgroup. Yet the population-level consequences of adoption as identity signaling are largely unknown. Whereas previous work has focused on asymmetric attraction and repulsion among groups with different social identities, here we consider the spread of innovations in a structured population in which there are multiple groups who don't want to be mistaken for one another, using both analytical and agent-based modeling. This formal analysis, the first to take the spatial structure of a population into account, allows us to consider empirically important factors, including demographic skew and communication scale, that likely influence overall patterns of adoption. We find that as products become emergent social markers, aversion to outgroup-associated products can decrease overall patterns ...
Intracellular facilitated diffusion: searchers, crowders and blockers
Brackley, C A; Marenduzzo, D
2013-01-01
In bacteria, regulatory proteins search for a specific DNA binding target via "facilitated diffusion": a series of rounds of 3D diffusion in the cytoplasm, and 1D linear diffusion along the DNA contour. Using large scale Brownian dynamics simulations we find that each of these steps is affected differently by crowding proteins, which can either be bound to the DNA acting as a road block to the 1D diffusion, or freely diffusing in the cytoplasm. Macromolecular crowding can strongly affect mechanistic features such as the balance between 3D and 1D diffusion, but leads to surprising robustness of the total search time.
DIAMANT2 - A multigroup neutron transport program for triangular and hexagonal geometry
International Nuclear Information System (INIS)
DIAMANT2 evolved out of the DIAMANT-code. DIAMANT2 solves the multigroup neutron transport equation in planar geometry using the Ssub(N) method. Spatial discretization is accomplished by taking finite differences on a meshgrid composed of equilateral triangles. This report contains a detailed documentation of the program and the input description. (orig./HJ)
Jones, K.; Johnston, R.; Manley, D.J.; Owen, D.; Charlton, C.
2015-01-01
We develop and apply a multilevel modeling approach that is simultaneously capable of assessing multigroup and multiscale segregation in the presence of substantial stochastic variation that accompanies ethnicity rates based on small absolute counts. Bayesian MCMC estimation of a log-normal Poisson
A finite element option for the MARC transport/ diffusion theory computer code
International Nuclear Information System (INIS)
The MARC multigroup transport/diffusion theory computer code has been extended to include a finite element option. The facility is available in two-dimensional geometry and has a novel feature in allowing high order polynomial approximations to the flux using an automated computer procedure. (U.K.)
YORP torques with 1D thermal model
Breiter, Slawomir; Czekaj, Maria
2010-01-01
A numerical model of the Yarkovsky-O'Keefe-Radzievskii-Paddack (YORP) effect for objects defined in terms of a triangular mesh is described. The algorithm requires that each surface triangle can be handled independently, which implies the use of a 1D thermal model. Insolation of each triangle is determined by an optimized ray-triangle intersection search. Surface temperature is modeled with a spectral approach; imposing a quasi-periodic solution we replace heat conduction equation by the Helmholtz equation. Nonlinear boundary conditions are handled by an iterative, FFT based solver. The results resolve the question of the YORP effect in rotation rate independence on conductivity within the nonlinear 1D thermal model regardless of the accuracy issues and homogeneity assumptions. A seasonal YORP effect in attitude is revealed for objects moving on elliptic orbits when a nonlinear thermal model is used.
1D ferrimagnetism in homometallic chains
Coronado Miralles, Eugenio; Gómez García, Carlos José; Borrás Almenar, Juan José
1990-01-01
The magnetic properties of the cobalt zigzag chain Co(bpy)(NCS)2 (bpy=2,2′‐bipyridine) are discussed on the basis of an Ising‐chain model that takes into account alternating Landé factors. It is emphasized, for the first time, that a homometallic chain containing only one type of site can give rise to a 1D ferrimagneticlike behavior. ,
Mezzacappa, A; Bruenn, S W; Blondin, J M; Guidry, M W; Strayer, M R; Umar, A S
1996-01-01
We investigate neutrino-driven convection in core collapse supernovae and its ramifications for the explosion mechanism. We begin with an ``optimistic'' 15 solar mass precollapse model, which is representative of the class of stars with compact iron cores. This model is evolved through core collapse and bounce in one dimension using multigroup (neutrino-energy--dependent) flux-limited diffusion (MGFLD) neutrino transport and Lagrangian hydrodynamics, providing realistic initial conditions for the postbounce convection and evolution. Our two-dimensional simulation begins at 106 ms after bounce at a time when there is a well-developed gain region, and proceeds for 400 ms. We couple two-dimensional (PPM) hydrodynamics to one-dimensional MGFLD neutrino transport. At 225 ms after bounce we see large-scale convection behind the shock, characterized by high-entropy, mushroom-like, expanding upflows and dense, low-entropy, finger-like downflows. The upflows reach the shock and distort it from sphericity. The radial c...
International Nuclear Information System (INIS)
A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the Keff, neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.
Energy Technology Data Exchange (ETDEWEB)
Zou Jun, E-mail: jzou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)
2010-12-15
A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K{sub eff}, neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.
International Nuclear Information System (INIS)
A hybrid multigroup/continuous-energy Monte Carlo algorithm is developed for solving the Boltzmann-Fokker-Planck equation. This algorithm differs significantly from previous charged-particle Monte Carlo algorithms. Most importantly, it can be used to perform both forward and adjoint transport calculations, using the same basic multigroup cross-section data. The new algorithm is fully described, computationally tested, and compared with a standard condensed history algorithm for coupled electron-photon transport calculations
FAYEZ MOUSTAFA MOAWAD, RAGAB
2016-01-01
[EN] The neutron diffusion equation is an approximation of the neutron transport equation that describes the neutron population in a nuclear reactor core. In particular, we will consider here VVER-type reactors which use the neutron diffusion equation discretized on hexagonal meshes. Most of the simulation codes of a nuclear power reactor use the multigroup neutron diffusion equation to describe the neutron distribution inside the reactor core.To study the stationary state of a reactor, the r...
MC2-2: a code to calculate fast neutron spectra and multigroup cross sections
International Nuclear Information System (INIS)
MC2-2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC2-2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC2-2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC2-2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC2-2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers
A Multigroup Method for the Calculation of Neutron Fluence with a Source Term
Heinbockel, J. H.; Clowdsley, M. S.
1998-01-01
Current research on the Grant involves the development of a multigroup method for the calculation of low energy evaporation neutron fluences associated with the Boltzmann equation. This research will enable one to predict radiation exposure under a variety of circumstances. Knowledge of radiation exposure in a free-space environment is a necessity for space travel, high altitude space planes and satellite design. This is because certain radiation environments can cause damage to biological and electronic systems involving both short term and long term effects. By having apriori knowledge of the environment one can use prediction techniques to estimate radiation damage to such systems. Appropriate shielding can be designed to protect both humans and electronic systems that are exposed to a known radiation environment. This is the goal of the current research efforts involving the multi-group method and the Green's function approach.
Processing ENDF/B-V uncertainty data into multigroup covariance matrices
International Nuclear Information System (INIS)
The purpose of this work is to develop and demonstrate the capability of processing Evaluated Nuclear Data File, system B, version five (ENDF/B-V) uncertainty data into multigroup covariance matrices. These covariances may then be folded with sensitivity coefficients to obtain uncertainties in selected integral parameters such as K-effective and breeding ratio. The project consisted of separating the previous uncertainty processor (PUFF) from the basic nuclear data cross section processor (MINX), updating the uncertanty processor to theENDF/B-V format, programming the processor for new uncertainty data, and demonstrating the processor capabilities by producing a multigroup covariance library. These capabilities were verified in various ways including hand calculations and comparisons with other known results. A computer code named PUFF-II was written to perform the task described above
A conservative multi-group approach to the Boltzmann equations for reactive gas mixtures
Bisi, M.; Rossani, A.; Spiga, G.
2015-11-01
Starting from a simple kinetic model for a quaternary mixture of gases undergoing a bimolecular chemical reaction, multi-group integro-differential equations are derived for the particle distribution functions of all species. The procedure takes advantage of a suitable probabilistic formulation, based on the underlying collision frequencies and transition probabilities, of the relevant reactive kinetic equations of Boltzmann type. Owing to an appropriate choice of a sufficiently large number of weight functions, it is shown that the proposed multi-group equations are able to fulfil exactly, at any order of approximation, the correct conservation laws that must be inherited from the original kinetic equations, where speed was a continuous variable. Future developments are also discussed.
Updated multi-group cross sections of minor actinides with improved resonance treatment
International Nuclear Information System (INIS)
The study of minor actinide in transmutation reactors and other future applications makes resonance self-shielding treatment a significant issue for criticality and isotope depletion. Resonance treatment for minor actinides has been carried out by subgroup method with improved interference effect through interference correction. Subgroup data was generated using RMET21 and GENP codes along with multi-group cross section data by NJOY nuclear data processing system. Updated multi-group cross section data library for a neutron transport code nTRACER was compared with solutions from MCNPX. The resonance interaction of uranium with minor actinides has been included by modified interference treatment of interference correction in subgroup methodology. The comparison of cross sections and multiplication factor in pin and assembly problems showed significant improvement from systematic resonance treatment especially for 237Np and 243Am. (author)
Correction of multigroup cross sections for resolved resonance interference in mixed absorbers
International Nuclear Information System (INIS)
The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed
PHISICS multi-group transport neutronic capabilities for RELAP5
Energy Technology Data Exchange (ETDEWEB)
Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G. [Idaho National Laboratory (INL), 2525 N. Fremont Ave., Idaho Falls, ID 83402 (United States)
2012-07-01
PHISICS is a neutronic code system currently under development at INL. Its goal is to provide state of the art simulation capability to reactor designers. This paper reports on the effort of coupling this package to the thermal hydraulic system code RELAP5. This will enable full prismatic core and system modeling and the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5 (NESTLE). The paper describes the capabilities of the coupling and illustrates them with a set of sample problems. (authors)
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies
International Nuclear Information System (INIS)
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
Energy Technology Data Exchange (ETDEWEB)
Ford, W.E. III; Roussin, R.W.; Petrie, L.M.; Diggs, B.R.; Comolander, H.E.
1979-01-01
Contents of the IBM version of the APMX system distributed by the Radiation Shielding Information Center (APMX-II) are described. Sample problems which demonstrate the procedure for implementing AMPX-II modules to generate point cross sections; generate multigroup neutron, photon production, and photon interaction cross sections for various transport codes; collapse multigroup cross sections; check, edit, and punch multigroup cross sections; and execute a one-dimensional discrete ordinates transport calculation are detailed. 25 figures, 9 tables.
2D/1D approximations to the 3D neutron transport equation. I: Theory
International Nuclear Information System (INIS)
A new class of '2D/1D' approximations is proposed for the 3D linear Boltzmann equation. These approximate equations preserve the exact transport physics in the radial directions x and y and diffusion physics in the axial direction z. Thus, the 2D/1D equations are more accurate approximations of the 3D Boltzmann equation than the conventional 3D diffusion equation. The 2D/1D equations can be systematically discretized, to yield accurate simulation methods for 3D reactor core problems. The resulting solutions will be more accurate than 3D diffusion solutions, and less expensive to generate than standard 3D transport solutions. In this paper, we (i) show that the simplest 2D/1D equation has certain desirable properties, (ii) systematically discretize this equation, and (iii) derive a stable iteration scheme for solving the discrete system of equations. In a companion paper [1], we give numerical results that confirm the theoretical predictions of accuracy and iterative stability. (authors)
International Nuclear Information System (INIS)
For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIAR next generation core simulation software package. ARCADIAR will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP3 nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIAR. For enabling a high computational performance, the SPN calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP3 is about 1.4 times as expensive computationally as SP1 (diffusion). The developed SP3 solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIAR. With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP3 instead of SP1 has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SPN results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO2/MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP3 computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)
Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh
International Nuclear Information System (INIS)
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
Global analysis on a class of multi-group SEIR model with latency and relapse.
Wang, Jinliang; Shu, Hongying
2016-02-01
In this paper, we investigate the global dynamics of a multi-group SEIR epidemic model, allowing heterogeneity of the host population, delay in latency and delay due to relapse distribution for the human population. Our results indicate that when certain restrictions on nonlinear growth rate and incidence are fulfilled, the basic reproduction number R0 plays the key role of a global threshold parameter in the sense that the long-time behaviors of the model depend only on R0. The proofs of the main results utilize the persistence theory in dynamical systems, Lyapunov functionals guided by graph-theoretical approach. PMID:26776266
Modification of the resonance treatment in multigroup neutron slowing-down codes
International Nuclear Information System (INIS)
The previously reported computer codes GRACE and BETTY for resonance treatment in the multigroup neutron slowing-down processes have been improved, employing the new results of resonance absorption calculations. The total resonance integral formulae were changed, 239Pu resonance integral data were included in the library of group constants and the selection of partial resonance integral distribution functions was automatized. The users of the GRACE and BETTY codes are provided with a more credible and more comfortable resonance treatment. Explicit description of modification of user's manuals is given. (D.P.)
Multi-group pin power reconstruction method based on colorset form functions
International Nuclear Information System (INIS)
A multi-group pin power reconstruction method that fully exploits nodal information obtained from global coarse mesh solution has been developed. It expands the intra-nodal flux distributions into nonseparable semi-analytic basis functions, and a colorset based form function generating method is proposed, which can accurately model the spectral interaction occurring at assembly interface. To demonstrate its accuracy and applicability to realistic problems, the new method is tested against two benchmark problems, including a mixed-oxide fuel problem. The results show that the new methods is comparable in accuracy to fine-mesh methods. (authors)
Specifications for a two-dimensional multi-group scattering code: ALCI
International Nuclear Information System (INIS)
This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, RΘ. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors)
Directory of Open Access Journals (Sweden)
Xiaoming Fan
2014-01-01
Full Text Available We discuss multigroup SIRS (susceptible, infectious, and recovered epidemic models with random perturbations. We carry out a detailed analysis on the asymptotic behavior of the stochastic model; when reproduction number ℛ0>1, we deduce the globally asymptotic stability of the endemic equilibrium by measuring the difference between the solution and the endemic equilibrium of the deterministic model in time average. Numerical methods are employed to illustrate the dynamic behavior of the model and simulate the system of equations developed. The effect of the rate of immunity loss on susceptible and recovered individuals is also analyzed in the deterministic model.
On the dynamics of a class of multi-group models for vector-borne diseases
Iggidr, Abderrahman; Sallet, Gauthier; Souza, Max O.
2016-01-01
The resurgence of vector-borne diseases is an increasing public health concern, and there is a need for a better understanding of their dynamics. For a number of diseases, e.g. dengue and chikungunya, this resurgence occurs mostly in urban environments, which are naturally very heterogeneous, particularly due to population circulation. In this scenario, there is an increasing interest in both multi-patch and multi-group models for such diseases. In this work, we study the dynamics of a vector...
On the completeness of the multigroup eigenfunctions set of a reactor system Boltzmann operator
International Nuclear Information System (INIS)
An example is given, which illustrates how the set of the eigenfunctions shifts from incompleteness to completeness when a coupling relationship is established between the spectrum of the neutrons produced by fission and the energy of the neutrons which generate the fissions. The proposed method allows one to complete the set of eigenfunctions of the Boltzmann operator in the multigroup case. That, in principle, enlarges the possibility to apply the SM, Standard Method, and the GSM, Generalized Standard Method, to any problem in reactor physics, regardless of the number of energy groups. (author)
Hybrid method of deterministic and probabilistic approaches for multigroup neutron transport problem
International Nuclear Information System (INIS)
A hybrid method of deterministic and probabilistic methods is proposed to solve Boltzmann transport equation. The new method uses a deterministic method, Method of Characteristics (MOC), for the fast and thermal neutron energy ranges and a probabilistic method, Monte Carlo (MC), for the intermediate resonance energy range. The hybrid method, in case of continuous energy problem, will be able to take advantage of fast MOC calculation and accurate resonance self shielding treatment of MC method. As a proof of principle, this paper presents the hybrid methodology applied to a multigroup form of Boltzmann transport equation and confirms that the hybrid method can produce consistent results with MC and MOC methods. (authors)
REX1-87, Multigroup Neutron Cross-Sections from ENDF/B
International Nuclear Information System (INIS)
1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations
Geospatial Data Fusion and Multigroup Decision Support for Surface Water Quality Management
Sun, A. Y.; Osidele, O.; Green, R. T.; Xie, H.
2010-12-01
Social networking and social media have gained significant popularity and brought fundamental changes to many facets of our everyday life. With the ever-increasing adoption of GPS-enabled gadgets and technology, location-based content is likely to play a central role in social networking sites. While location-based content is not new to the geoscience community, where geographic information systems (GIS) are extensively used, the delivery of useful geospatial data to targeted user groups for decision support is new. Decision makers and modelers ought to make more effective use of the new web-based tools to expand the scope of environmental awareness education, public outreach, and stakeholder interaction. Environmental decision processes are often rife with uncertainty and controversy, requiring integration of multiple sources of information and compromises between diverse interests. Fusing of multisource, multiscale environmental data for multigroup decision support is a challenging task. Toward this goal, a multigroup decision support platform should strive to achieve transparency, impartiality, and timely synthesis of information. The latter criterion often constitutes a major technical bottleneck to traditional GIS-based media, featuring large file or image sizes and requiring special processing before web deployment. Many tools and design patterns have appeared in recent years to ease the situation somewhat. In this project, we explore the use of Web 2.0 technologies for “pushing” location-based content to multigroups involved in surface water quality management and decision making. In particular, our granular bottom-up approach facilitates effective delivery of information to most relevant user groups. Our location-based content includes in-situ and remotely sensed data disseminated by NASA and other national and local agencies. Our project is demonstrated for managing the total maximum daily load (TMDL) program in the Arroyo Colorado coastal river basin
Energy Technology Data Exchange (ETDEWEB)
Hursin, Mathieu [School of Nuclear Engineering, Purdue University, 400 Central Drive, IN 47907 (United States); Xiao Shanjie [School of Nuclear Engineering, Purdue University, 400 Central Drive, IN 47907 (United States); Jevremovic, Tatjana [School of Nuclear Engineering, Purdue University, 400 Central Drive, IN 47907 (United States)]. E-mail: tatjanaj@purdue.edu
2006-09-15
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments.
International Nuclear Information System (INIS)
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments
Analysis list: Nr1d2 [Chip-atlas[Archive
Lifescience Database Archive (English)
Full Text Available Nr1d2 Liver + mm9 http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/target/Nr1d2.1.tsv... http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/target/Nr1d2.5.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/target/Nr1d2....10.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/colo/Nr1d2.Liver.tsv http://dbarchive.biosciencedbc.jp/kyushu-u/mm9/colo/Liver.gml ...
A new derivation of Akcasu's 'MLP' equations for 1-D particle transport in stochastic media
International Nuclear Information System (INIS)
This paper presents a new derivation of Akcasu's modified Levermore-Pomraning (MLP) model, which estimates the ensemble-averaged angular flux for particle transport problems in 1-D geometrically random media. The significant new feature of the MLP equations is that, unlike the earlier Levermore-Pomraning (LP) model, the MLP equations are exact for certain classes of problems with scattering. We also show, via asymptotic analyses, that the MLP equations are accurate in the atomic mix and diffusion limits
Multi-group unified nodal method with two-group coarse-mesh finite difference formulation
International Nuclear Information System (INIS)
The one-node kernels of the unified nodal method (UNM) which were originally developed for two-group (2G) problems are extended to solve multi-group (MG) problems within the framework of the 2G coarse-mesh finite difference (CMFD) formulation. The analytic nodal method (ANM) kernel of UNM is reformulated for the MG application by adopting the Pade approximation to avoid the similarity transform required to diagonalize the G x G buckling matrix. In addition, a one-node semi-analytic nodal method (SANM) kernel which is considered adequate for multi-group calculations is also integrated into the UNM formulation by expressing it in the form consistent with the other UNM kernels. As an efficient global solution framework, the 2G CMFD formulation with dynamic group condensation and prolongation is established and the performance of the various MG kernels is examined using various static and transient benchmark problems. It turns out that the SANM kernel is the best one for MG problems not only because it retains accuracy comparable to MGANM with a shorter computing time but also because its accuracy or its convergence does not depend on the eigenvalue range of the buckling matrix of the system. The 2G CMFD formulation with MG one-node UNM kernels turns out to be very effective in that it conveniently accelerates the MG source iteration
International Nuclear Information System (INIS)
Comparative calculations of the experimental benchmark of iron sphere with Cf source have been performed in order to assess the sensibility of the calculations of neutron transmission through iron media to different multigroup libraries generated on the base of ENDF/B-6 and ENDF/B-4. Similar calculations and comparison of the neutron flux passed through media typical as geometry and material compositions for the WWER-1000 and WWER-440 vessels have been carried out. Except the already well-known problem dependent libraries, the new libraries BGL-440 and BGL-1000 generated on the base of ENDF/B-6 for the WWER-440 and WWER-1000 RPV neutron fluence calculations have been applied. The solving of neutron transport through iron media using ENDF/B-6 data gives better consistency with the experiment than using ENDF/B-4. The latter underestimate the experimental fluxes more substantially in the energy range above 2 MeV and the evaluations of the neutron flux responses for the WWER vessel surveillance is preferably to be carried out by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries
Verification of a Multi-group Cross Section Library for Burnup Calculation
Energy Technology Data Exchange (ETDEWEB)
Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of); Joo, Hang Yu [Seoul National Univ., Seoul (Korea, Republic of)
2013-05-15
Despite satisfying the estimation of the neutronic parameters without depletion to some extent, it still requires detailed investigation of the behavior of a fuel with strong neutron absorber over its operating life time by nTRACER, the direct whole core calculation code with the conventional semi Predictor-Corrector method. This study is mainly focused on the verification of the newly generated multi-group library for burnup calculation by nTRACER through the analysis of its performance of depletion calculation of UO{sub 2} fuel with strong neutron absorbers such as Gadolinium. Firstly, the depletion calculation results of nTRACER are presented by comparing the evolution of k-inf and the inventories of commonly found important isotopes as a function of burnup in the cases of gadolinia(GAD)-bearing fuel pin and fuel assembly (FA) with those of MCNPX-version.2.6.0. The newly generated multi-group library for burnup calculation by nTRACER was verified through GAD-bearing fuel after the new approach of resonance treatment had been employed. Though very good agreement in the overall effect reflected on the multiplication factor of FA at BOC, the evolution of k-inf along fuel irradiation history was systematically well underestimated by nTRACER when compared to Monte Carlo results.
The group-level consequences of sexual conflict in multigroup populations.
Directory of Open Access Journals (Sweden)
Omar Tonsi Eldakar
Full Text Available In typical sexual conflict scenarios, males best equipped to exploit females are favored locally over more prudent males, despite reducing female fitness. However, local advantage is not the only relevant form of selection. In multigroup populations, groups with less sexual conflict will contribute more offspring to the next generation than higher conflict groups, countering the local advantage of harmful males. Here, we varied male aggression within- and between-groups in a laboratory population of water striders and measured resulting differences in local population growth over a period of three weeks. The overall pool fitness (i.e., adults produced of less aggressive pools exceeded that of high aggression pools by a factor of three, with the high aggression pools essentially experiencing no population growth over the course of the study. When comparing the fitness of individuals across groups, aggression appeared to be under stabilizing selection in the multigroup population. The use of contextual analysis revealed that overall stabilizing selection was a product of selection favoring aggression within groups, but selected against it at the group-level. Therefore, this report provides further evidence to show that what evolves in the total population is not merely an extension of within-group dynamics.
International Nuclear Information System (INIS)
The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures
Historical review of group diffusion computation
International Nuclear Information System (INIS)
Diffusion theory neutron flux computations are the backbone of reactor physics design studies. Early one-space-dimension computations of simple configurations with Marchand and Friden desk computers were swept with the computer revolution into detailed three-space-dimension calculations of complex reactor configurations on today's parallel and vector machines. Methods for modeling reactors have evolved during the past 45 yr. The multigroup diffusion theory first described by Ehrlich and Hurwitz has been embellished with transport theory corrections and accounting for mesh effects in discrete modeling. Spectral variations within a few-group framework were accounted for by space-dependent group constants based on estimated hardness of spectrum. A more sophisticated procedure used overlapping energy groups
P1 adaptation of TRIPOLI-4 code for the use of 3D realistic core multigroup cross section generation
International Nuclear Information System (INIS)
In this paper, we discuss some improvements we recently implemented in the Monte-Carlo code TRIPOLI-4 associated with the homogenization and collapsing of subassemblies cross sections. The improvement offered us another approach to get critical multigroup cross sections with Monte-Carlo method. The new calculation method in TRIPOLI-4 tries to ensure the neutronic balances, the multiplicative factors and the critical flux spectra for some realistic geometries. We make it by at first improving the treatment of the energy transfer probability, the neutron excess weight and the neutron fission spectrum. This step is necessary for infinite geometries. The second step which will be enlarged in this paper is aimed at better dealing with the multigroup anisotropy distribution law for finite geometries. Usually, Monte-Carlo homogenized multi-group cross sections are validated within a core calculation by a deterministic code. Here, the validation of multigroup constants will also be carried out by Monte-Carlo core calculation code. Different subassemblies are tested with the new collapsing method, especially for the fast neutron reactors subassemblies. (authors)
International Nuclear Information System (INIS)
The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values
International Nuclear Information System (INIS)
Liquid Salt Cooled Reactors (LSCRs) are high temperature reactors, cooled by liquid salt, with a TRISO-particle based fuel in a solid form (stationary fuel elements or circulating fuel pebbles); this paper is focusing on the former. In either case, due to the double heterogeneity, core physics analysis require different considerations with more complex approaches than LWRs core physics calculations. Additional challenges appear when using the multi-group approach. In this paper we examine the use of SCALE6.1.1. Double heterogeneity may be accounted for through the Dancoff factor, however, SCALE6.1.1 does not provide an automated method to calculate Dancoff Factors for fuel planks with TRISO fuel particles. Therefore, depletion with continuous energy Monte Carlo Transport (CE depletion) in SCALE6.2 beta was used to generate MC Dancoff factors for multi-group calculations. MCDancoff corrected multi-group depletion agrees with the results for CE depletion within ±100 pcm, and within ±2σ. Producing MCDancoff factors for multi-group (MG) depletion calculations is necessary to LSCR analysis because CE depletion runtime and memory requirements are prohibitive for routine use. MG depletion with MCDancoff provides significantly shorter runtime and lower memory requirements while providing results of acceptable accuracy. (author)
Cai, Li; Pénéliau, Yannick; Diop, Cheikh M.; Malvagi, Fausto
2014-06-01
In this paper, we discuss some improvements we recently implemented in the Monte-Carlo code TRIPOLI-4® associated with the homogenization and collapsing of subassemblies cross sections. The improvement offered us another approach to get critical multigroup cross sections with Monte-Carlo method. The new calculation method in TRIPOLI-4® tries to ensure the neutronic balances, the multiplicative factors and the critical flux spectra for some realistic geometries. We make it by at first improving the treatment of the energy transfer probability, the neutron excess weight and the neutron fission spectrum. This step is necessary for infinite geometries. The second step which will be enlarged in this paper is aimed at better dealing with the multigroup anisotropy distribution law for finite geometries. Usually, Monte-Carlo homogenized multi-group cross sections are validated within a core calculation by a deterministic code. Here, the validation of multigroup constants will also be carried out by Monte-Carlo core calculation code. Different subassemblies are tested with the new collapsing method, especially for the fast neutron reactors subassemblies.
International Nuclear Information System (INIS)
A - Nature of physical problem solved: The ANISN system treats neutron and gamma transport in one- dimensional plane, spherical and cylinder geometry. The multigroup cross sections prepared by the programs LIANE and SUPERTOG are processed by the program RETTOG, which produces a binary library with Legendre expansions. The binary library can be updated and edited with the program LGR/B. The photon multigroup cross sections are created with the program GAMLEG/A. If the bulk of the data is too large, the program TAPEMA produces a special group-by-group library. The volume sources are calculated from a reduced set of input data and punched in a format suitable for input to ANISN, using the program PRESOU. ANISN calculates fluxes by groups, space intervals, angle and any number of reaction rates. The energy and space dependent fluxes are stored on tape and can be reprocessed, edited and plotted with the program ANISEX, which also permits to calculate supplementary reaction rates. The program ANISN can condense cross sections into a reduced number of groups. The ANISN system is used as a reference system for the evaluation of approximation methods (space-diffusion or point kernel) or for the preparation of multigroup libraries for two-dimensional transport codes (DOT). In particular it is used for shielding problems with high attenuation in water reactors and fast reactors. ANISN-E solves the same problems as the original ANISN code. Some modifications concern weighted cross sections output and fixed distributed sources input/output. ANISN-E (CCC-0082/09): The CYBER 175 version of ANISN-E also contains the free-format input capability. ANISN-JR extends the applicability of the original ANISN code for shielding analyses by adding options of calculating the reaction rates distributions from detector response, generating the volume- flux weighted cross sections in arbitrary regions or zones and plotting the neutron or gamma-ray spectra and the reaction rates distributions
Nelson, Adam
Multi-group scattering moment matrices are critical to the solution of the multi-group form of the neutron transport equation, as they are responsible for describing the change in direction and energy of neutrons. These matrices, however, are difficult to correctly calculate from the measured nuclear data with both deterministic and stochastic methods. Calculating these parameters when using deterministic methods requires a set of assumptions which do not hold true in all conditions. These quantities can be calculated accurately with stochastic methods, however doing so is computationally expensive due to the poor efficiency of tallying scattering moment matrices. This work presents an improved method of obtaining multi-group scattering moment matrices from a Monte Carlo neutron transport code. This improved method of tallying the scattering moment matrices is based on recognizing that all of the outgoing particle information is known a priori and can be taken advantage of to increase the tallying efficiency (therefore reducing the uncertainty) of the stochastically integrated tallies. In this scheme, the complete outgoing probability distribution is tallied, supplying every one of the scattering moment matrices elements with its share of data. In addition to reducing the uncertainty, this method allows for the use of a track-length estimation process potentially offering even further improvement to the tallying efficiency. Unfortunately, to produce the needed distributions, the probability functions themselves must undergo an integration over the outgoing energy and scattering angle dimensions. This integration is too costly to perform during the Monte Carlo simulation itself and therefore must be performed in advance by way of a pre-processing code. The new method increases the information obtained from tally events and therefore has a significantly higher efficiency than the currently used techniques. The improved method has been implemented in a code system
MPI version of NJOY and its application to multigroup cross-section generation
International Nuclear Information System (INIS)
Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures
International Nuclear Information System (INIS)
Highlights: • Multigroup nuclear data are sampled based on multivariate normal distributions. • Multigroup perturbation factors are applied to pointwise-ACE nuclear data. • Samples of perturbed pointwise-ACE nuclear data are generated by NUSS for MCNPX. • Variances in MCNPX outputs due to perturbed samples of ACE data are quantified. • NUSS is verified with TSUNAMI and MCNPX PERT CARD sensitivity/uncertainty methods. - Abstract: Stochastic sampling (SS) method for quantifying nuclear data uncertainties is accomplished by using perturbed nuclear data in routine neutronics calculations and determining the variance of output parameters due to the input nuclear data uncertainties. Existing SS-based methods have demonstrated the feasibility and efficiency of propagating uncertainties in multigroup nuclear data. However, in fields such as criticality safety assessment, pointwise representation of nuclear data is more appropriate in order to corroborate the increasing safety demand and best-estimate modeling capabilities. In this work, an SS-based tool, called NUSS is implemented which perturbs pointwise ACE-formatted nuclear data using multigroup nuclear data covariance. The use of pointwise ACE-formatted nuclear data in NUSS can accommodate flexible multigroup covariance structures and allows for nuclear data uncertainty propagation through the continuous/pointwise-energy transport code MCNPX. As a first step of the NUSS development and verification, uncertainty contributions from 239Pu and 235U nuclear data were assessed for Jezebel (Pu-fueled) and Godiva (U-fueled) fast-spectrum criticality benchmarks. NUSS results are compared to those by other uncertainty quantification methods such as TSUNAMI and MCNPX PERT CARD. Next, Light Water Reactor (LWR) pin cell models from the OECD/NEA UAM Phase-1 benchmark were analyzed. Results of cross section and kinf uncertainties in consideration of different nuclear data covariance libraries are presented
Development of advanced nodal diffusion methods for modern computer architectures
International Nuclear Information System (INIS)
A family of highly efficient multidimensional multigroup advanced neutron-diffusion nodal methods, ILLICO, were implemented on sequential, vector, and vector-concurrent computers. Three-dimensional realistic benchmark problems can be solved in vectorized mode in less than 0.73 s (33.86 Mflops) on a Cray X-MP/48. Vector-concurrent implementations yield speedups as high as 9.19 on an Alliant FX/8. These results show that the ILLICO method preserves essentially all of its speed advantage over finite-difference methods. A self-consistent higher-order nodal diffusion method was developed and implemented. Nodal methods for global nuclear reactor multigroup diffusion calculations which account explicitly for heterogeneities in the assembly nuclear properties were developed and evaluated. A systematic analysis of the zero-order variable cross section nodal method was conducted. Analyzing the KWU PWR depletion benchmark problem, it is shown that when burnup heterogeneities arise, ordinary nodal methods, which do not explicitly treat the heterogeneities, suffer a significant systematic error that accumulates. A nodal method that treats explicitly the space dependence of diffusion coefficients was developed and implemented. A consistent burnup-correction method for nodal microscopic depletion analysis was developed
Facilitated diffusion buffers noise in gene expression
Schoech, Armin; Zabet, Nicolae Radu
2014-01-01
Transcription factors perform facilitated diffusion (3D diffusion in the cytosol and 1D diffusion on the DNA) when binding to their target sites to regulate gene expression. Here, we investigated the influence of this binding mechanism on the noise in gene expression. Our results showed that, for biologically relevant parameters, the binding process can be represented by a two-state Markov model and that the accelerated target finding due to facilitated diffusion leads to a reduction in both ...
AIRDIF, Neutron and Gamma Doses from Nuclear Explosion by 2-D Air Diffusion
International Nuclear Information System (INIS)
1 - Description of problem or function: AIRDIF is a two-dimensional atmospheric radiation diffusion code designed to calculate neutron and gamma doses in the environment of a nuclear explosion. It calculates radiation fluxes in one-dimensional homogeneous air, or two-dimensional variable density air. The results are limited by the assumptions inherent in diffusion theory: the region of interest must be large compared to the radiation mean free path, the spatial flux gradients must not be steep, flux varies linearly with the cosine of the direction angle. The code requires as input data neutron and gamma source spectra, coupled neutron-gamma multigroup cross sections, and, for two- dimensional problems, a set of mass integral scaling (MIS) coefficients. These latter are calculated from an AIRDIF output flux file for a one-dimensional problem by the auxiliary program MISFIT, using a least squares fitting technique to Murphy's radiation transmission equation. MISFIT can also be used to calculate one- dimensional MIS doses. The MIS coefficients and doses can be input to AIRDIF, in two- dimensional mode to calculate 2-D fluxes, doses and K-factors (the ratio of 2-D to 1-D dose). Alternatively the 2-D doses and K-factors may be computed using the output 2-D flux file of a previous AIRDIF run using the auxiliary program DOSCOMP. 2 - Method of solution: Un-collided particle flux is determined from an analytic expression describing exponential attenuation with distance. Diffusion theory is used for the flux, using un-collided flux as a source term. A central collided differencing technique is used to reduce the diffusion equation to a matrix equation, which is solved by the Successive Line Over-relaxation (SLOR) method. Total flux is calculated as the sum of collided and un-collided components. To maintain a mesh interval which has the same relationship to mean free path at all heights, an expanding non-orthogonal coordinate system is used. In homogeneous air this system
Energy Technology Data Exchange (ETDEWEB)
Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.
1992-10-01
AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.
International Nuclear Information System (INIS)
AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available
Global dynamics of a novel multi-group model for computer worms
International Nuclear Information System (INIS)
In this paper, we study worm dynamics in computer networks composed of many autonomous systems. A novel multi-group SIQR (susceptible-infected-quarantined-removed) model is proposed for computer worms by explicitly considering anti-virus measures and the network infrastructure. Then, the basic reproduction number of worm R0 is derived and the global dynamics of the model are established. It is shown that if R0 is less than or equal to 1, the disease-free equilibrium is globally asymptotically stable and the worm dies out eventually, whereas, if R0 is greater than 1, one unique endemic equilibrium exists and it is globally asymptotically stable, thus the worm persists in the network. Finally, numerical simulations are given to illustrate the theoretical results. (general)
On the feasibility of a homogenised multi-group Monte Carlo method in reactor analysis
International Nuclear Information System (INIS)
The use of homogenised multi-group cross sections to speed up Monte Carlo calculation has been studied to some extent, but the method is not widely implemented in modern calculation codes. This paper presents a calculation scheme in which homogenised material parameters are generated using the PSG continuous-energy Monte Carlo reactor physics code and used by MORA, a new full-core Monte Carlo code entirely based on homogenisation. The theory of homogenisation and its implementation in the Monte Carlo method are briefly introduced. The PSG-MORA calculation scheme is put to practice in two fundamentally different test cases: a small sodium-cooled fast reactor (JOYO) and a large PWR core. It is shown that the homogenisation results in a dramatic increase in efficiency. The results are in a reasonably good agreement with reference PSG and MCNP5 calculations, although fission source convergence becomes a problem in the PWR test case. (authors)
The French 'CEA 86' multigroup cross-section library and its integral qualification
International Nuclear Information System (INIS)
This paper describe the up-dated 99 groups library of the APOLLO French neutron computer code, the denominated 'CEA 86' library. The multigroup cross-section sets are based on the more recent nuclear data evaluations. The THEMIS code was generally used for the JEF-1 processing. In order to account for recent differential measurements and to improve the consistency between calculation and integral experiments, we produced our own CEA evaluations for the actinide nuclides: 235U, 238U, 239Pu, 240Pu, 241Am. This new APOLLO library was checked against critical experiments and PWR measurements: computed Conversion Factor, Reactivity Coefficients, Multiplication Factor, and Pu build-up are now in good agreement with LWR experimental results. PWR Pu recycling calculations, as does as HCLWR design studies, are also significantly improved. (author)
Approximate analytical solution of two-dimensional multigroup P-3 equations
International Nuclear Information System (INIS)
Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (orig./RW)
A Method to Solve Multigroup P3 Equations in Cylindrical Geometry
International Nuclear Information System (INIS)
To determine the space-energy distribution of thermal neutrons in a reactor cell a combination of the spherical harmonics method and multigroup procedure has been chosen. In P-3 approximation and cylindrical geometry such a scheme implies the solution of an inhomogeneous system of six ordinary first order differential equations. The general solution of the corresponding homogeneous system is known in analytical form. The present work shows how the free term of the system can be approximated in order to find a particular solution, and thus the general solution, of the inhomogeneous system. The procedure has been applied to calculate thermal spectra in a number of different reactor cells. Some results are presented and discussed. (author)
Approximate analytical solution of two-dimensional multigroup P-3 equations
International Nuclear Information System (INIS)
Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (author)
Jones, Kelvyn; Johnston, Ron; Manley, David; Owen, Dewi; Charlton, Chris
2015-12-01
We develop and apply a multilevel modeling approach that is simultaneously capable of assessing multigroup and multiscale segregation in the presence of substantial stochastic variation that accompanies ethnicity rates based on small absolute counts. Bayesian MCMC estimation of a log-normal Poisson model allows the calculation of the variance estimates of the degree of segregation in a single overall model, and credible intervals are obtained to provide a measure of uncertainty around those estimates. The procedure partitions the variance at different levels and implicitly models the dependency (or autocorrelation) at each spatial scale below the topmost one. Substantively, we apply the model to 2011 census data for London, one of the world's most ethnically diverse cities. We find that the degree of segregation depends both on scale and group. PMID:26487190
The solution of the multigroup neutron transport equation using spherical harmonics
International Nuclear Information System (INIS)
A solution of the multi-group neutron transport equation in up to three space dimensions is presented. The flux is expanded in a series of unnormalised spherical harmonics. Using the various recurrence formulae a linked set of first order differential equations is obtained for the moments psisup(g)sub(lm)(r), γsup(g)sub(lm)(r). Terms with odd l are eliminated resulting in a second order system which is solved by two methods. The first is a finite difference formulation using an iterative procedure, secondly, in XYZ and XY geometry a finite element solution is given. Results for a test problem using both methods are exhibited and compared. (orig./RW)
International Nuclear Information System (INIS)
The importance of accounting for resonance self-screening effects in multigroup cross sections when calculating fast reactors and neutron shields is considered. Formulae for averaging cross sections over resonance features with the account of anisotropy for scattering with large energy losses are derived. The model calculations of neutron fluxes have been performed for a U-H mixture (rhosub(H)/rhosub(U)=0.1), a U-Fe-H mixture and for the latter with rhosub(5)/rhosub(Fe)=0.01-0.5. It is concluded that in hydrogen-containing reactors the effect may be significant if the core contains iron in large quantities. The cross section averaging is considered for 3 systems: the KBR-2 critical assembly, spherical model of a large breeder, critical sphere of UO2 with 30% enrichment. The scattering anisotropy changes the multiplication factors of the first two systems by about 0.3%
Analyzing Average and Conditional Effects with Multigroup Multilevel Structural EquationModels
Directory of Open Access Journals (Sweden)
Axel Mayer
2014-04-01
Full Text Available Conventionally, multilevel analysis of covariance (ML-ANCOVA has been therecommended approach for analyzing treatment effects in quasi-experimental multilevel designswith treatment application at the cluster-level. In this paper, we introduce the generalizedML-ANCOVA with linear effect functions that identifies average and conditional treatment effectsin the presence of treatment-covariate interactions. We show how the generalized ML-ANCOVAmodel can be estimated with multigroup multilevel structural equation models that offerconsiderable advances compared to traditional ML-ANCOVA. The proposed model takes intoaccount measurement error in the covariates, sampling error in contextual covariates,treatment-covariate interactions, and stochastic predictors. We illustrate the implementation ofML-ANCOVA with an example from educational effectiveness research where we estimateaverage and conditional effects of early transition to secondary schooling on readingcomprehension.
Release of the mtmg01ex NDI Neutron Multigroup Data Library
International Nuclear Information System (INIS)
We have released the multi-temperature neutron multigroup transport library mtmg01ex, consisting of 181 isotope tables from mtmg01 and 18 element tables calculated from the isotope tables, all at 15 temperatures. These data, based primarily on the evaluations that produced the lanl2006 library, include gamma production and americium branching data. They were subjected to our standard production library testing. Because there are still known problems with and unanswered questions about multi-temperature data, including data size and load time issues, we do not recommend this data for general use; however, its quality is good enough for production release, and we request user help in addressing the remaining problems.
ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma0 are not guaranteed because of the approximations used in the unresolved resonance region
Recent validation experience with multigroup cross-section libraries and scale
International Nuclear Information System (INIS)
This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment
International Nuclear Information System (INIS)
The MGPRAKTINETs computer code for the BESM-6 computer intended for calculation of zone average trmal neutron group fluxes and functionals is described. The neutron spatial-energy distribution in a multizone cyllindrically-symmetric reactor cell is calculated by the operator splitting method. For the solution of the spatial part of the problem the method of surface pseudosources (Gsub(N)-approximation) in approximation of plane derivatives from the energy neutron current is employed. The energy part of the problem is solved in a multigroup approximation. Computer code efficiency has been demonstrated by calculation of two-zone cells with internal and external sources of the cell with on additional absorber and RBMK cell with reduction of the latter to cylindrical geometry. It is shown that the approximation of plane derivatives of neutron energy current allows calculating reactor cell characteristics with a sufficient for design calculations accuracy
Zhang, Xi; Showman, Adam P.
2015-11-01
Most of the current atmospheric chemistry models for planets (e.g., Krasnopolsky & Parshev 1981; Yung & Demore 1982; Yung, Allen & Pinto 1984; Lavvas et al. 2008; Zhang et al. 2012) and exoplanets (e.g., Line, Liang & Yung 2010; Moses et al. 2011; Hu & Seager 2014) adopt a one-dimensional (1D) chemical-diffusion approach in the vertical coordinate. Although only a crude approximation, these 1D models have succeeded in explaining the global-averaged vertical profiles of many chemical species in observations. One of the important assumptions of these models is that all chemical species are transported via the same eddy diffusion profile--that is, the assumption is made that the eddy diffusivity is a fundamental property of the dynamics alone, and does not depend on the chemistry. Here we show that, as also noticed in the Earth community (e.g., Holton 1986), this “homogenous eddy diffusion” assumption generally breaks down. We first show analytically why the 1D eddy diffusivity must generally depend both on the horizontal eddy mixing and the chemical lifetime of the species. This implies that the long-lived species and short-lived chemical species will generally exhibit different eddy diffusion profiles, even in a given atmosphere with identical dynamics. Next, we present tracer-transport simulations in a 2D chemical-diffusion-advection model (Shia et al. 1989; Zhang, Shia & Yung 2013) and a 3D general circulation model (MITgcm, e.g., Liu & Showman 2013), for both rapid-rotating planets and tidally-locked exoplanets, to further explore the effect of chemical timescales on the eddy diffusivity. From the 2D and 3D simulation outputs, we derive effective 1D eddy diffusivity profiles for chemical tracers exhibiting a range of chemical timescales. We show that the derived eddy diffusivity can depend strongly on the horizontal eddy mixing and chemistry, although the dependences are more complex than the analytic model predicts. Overall, these results suggest that
Multigroup Albedo Method applied to coupled neutron-gamma radiations shielding
International Nuclear Information System (INIS)
Shielding calculations for neutron-gamma radiation are usually done by using the full Theory of Transport or the Monte Carlo Techniques. After some works based on the Albedo Method, the shielding calculations for neutron-gamma radiation have a reliable tool with great didactical value which shows its clarity and simplicity for the resolution of cases that involve neutrons and photon shielding in nonmultiplying media. The excellent results of these works have motivated the elaboration and the development of this study that will be presented in this dissertation. The balance of a neutronic current entering a shield of two layers considering the coupling neutron-gamma will be determined by the Albedo Method. The shield will be composed of a layer of iron and another one of manganese with 10 cm of thickness each. The arrays of the materials coefficients will be obtained from the ANISN code. ANISN is a one dimensional deterministic code that is based on transport equation. The final results obtained by the Albedo Method will be compared with the ANISN results for an order of angular quadrature S2. The angular quadrature S2 admits that the radiation has two routes in the same direction what better describes the Albedo Method behavior. The results obtained by using the Albedo Method show an excellent agreement with the values predicted by the adopted deterministic code ANISN. Due to the excellent results, the multigroup Albedo Method should be applied to the shielding calculations with multiple layers. In conclusion the multigroup Albedo Method has the great ability in solving shielding problems concerning to the Nuclear Engineering. (author)
ZZ AMZ, 70-Group 40 Isotope Multigroup Library for Fast Reactor Calculation
International Nuclear Information System (INIS)
1 - Description of program or function: format: EXPANDA; number of groups: 70-group library of multigroup constants; nuclides: H-1, Be-9, B-10, B-11, C-12, O-16, N-23, Mg, Al-27, Si, Ti, V, Cr, Mn-55, Fe, Ni, Cu, Ga, Zr, Nb-93, Mo, In-115, Sn, Pb, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, and lumped fission products of U-233, U-235, Pu-239. origin: ENDF/B-IV and ENDF/B-V; weighting spectrum: Fission products inventories for BBR reactor at 360 and 600 days of irradiation were calculated and used as weighting function. AMZ is a 70-group library of multigroup constants for the fast reactor nuclear design code EXPANDA. Data is stored for three temperatures (300 K, 900 K, 2100 K) and for seven background cross sections. The following isotopes are available: H1, Be9, B10, B11, C12, O16, N23, Mg, Al27, Si, Ti, V, Cr, Mn55, Fe, Ni, Cu, Ga, Zr, Nb93, Mo, In115, Sn, Pb, Th232, Pa233, U233, U234, U235, U236, U238, Pu238, Pu239, Pu240, Pu241, Pu242, Am241, and lumped fission products of U233, U235, Pu239. 2 - Method of solution: Nuclear cross sections, transfer matrices, and self-shielding factors were generated from ENDF/B-IV data using the codes NJOY (PSR-0171) and RGENDF
International Nuclear Information System (INIS)
The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. However, as presently formulated, it is both restricted to orthogonal geometries and susceptible to producing ray effects. In this work, a finite element formulation, utilizing a canonical form of the transport equation, is developed to obtain both integral and pointwise solutions to neutron transport problems. To facilitate its application to nonorthogonal planar geometries, the formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included in the formulation by employing discrete ordinates like approximations. In addition, multigroup source outer iteration techniques are employed to perform group dependent calculations. The ability of the formulation to substantially reduce ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal type lattice. A small high leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation
XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections
International Nuclear Information System (INIS)
1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections
Energy Technology Data Exchange (ETDEWEB)
Popovic, Marta; Zaja, Roko [Laboratory for Molecular Ecotoxicology, Division for Marine and Environmental Research, Rudjer Boskovic Institute, Bijenicka 54, 10 000 Zagreb (Croatia); Fent, Karl [University of Applied Sciences Northwestern Switzerland, School of Life Sciences, Gründenstrasse 40, CH-4132 Muttenz (Switzerland); Swiss Federal Institute of Technology (ETH Zürich), Department of Environmental System Sciences, Institute of Biogeochemistry and Pollution Dynamics, CH-8092 Zürich (Switzerland); Smital, Tvrtko, E-mail: smital@irb.hr [Laboratory for Molecular Ecotoxicology, Division for Marine and Environmental Research, Rudjer Boskovic Institute, Bijenicka 54, 10 000 Zagreb (Croatia)
2014-10-01
Polyspecific transporters from the organic anion transporting polypeptide (OATP/Oatp) superfamily mediate the uptake of a wide range of compounds. In zebrafish, Oatp1d1 transports conjugated steroid hormones and cortisol. It is predominantly expressed in the liver, brain and testes. In this study we have characterized the transport of xenobiotics by the zebrafish Oatp1d1 transporter. We developed a novel assay for assessing Oatp1d1 interactors using the fluorescent probe Lucifer yellow and transient transfection in HEK293 cells. Our data showed that numerous environmental contaminants interact with zebrafish Oatp1d1. Oatp1d1 mediated the transport of diclofenac with very high affinity, followed by high affinity towards perfluorooctanesulfonic acid (PFOS), nonylphenol, gemfibrozil and 17α-ethinylestradiol; moderate affinity towards carbaryl, diazinon and caffeine; and low affinity towards metolachlor. Importantly, many environmental chemicals acted as strong inhibitors of Oatp1d1. A strong inhibition of Oatp1d1 transport activity was found by perfluorooctanoic acid (PFOA), chlorpyrifos-methyl, estrone (E1) and 17β-estradiol (E2), followed by moderate to low inhibition by diethyl phthalate, bisphenol A, 7-acetyl-1,1,3,4,4,6-hexamethyl-1,2,3,4 tetrahydronapthalene and clofibrate. In this study we identified Oatp1d1 as a first Solute Carrier (SLC) transporter involved in the transport of a wide range of xenobiotics in fish. Considering that Oatps in zebrafish have not been characterized before, our work on zebrafish Oatp1d1 offers important new insights on the understanding of uptake processes of environmental contaminants, and contributes to the better characterization of zebrafish as a model species. - Highlights: • We optimized a novel assay for determination of Oatp1d1 interactors • Oatp1d1 is the first SLC characterized fish xenobiotic transporter • PFOS, nonylphenol, diclofenac, EE2, caffeine are high affinity Oatp1d1substrates • PFOA, chlorpyrifos
International Nuclear Information System (INIS)
Polyspecific transporters from the organic anion transporting polypeptide (OATP/Oatp) superfamily mediate the uptake of a wide range of compounds. In zebrafish, Oatp1d1 transports conjugated steroid hormones and cortisol. It is predominantly expressed in the liver, brain and testes. In this study we have characterized the transport of xenobiotics by the zebrafish Oatp1d1 transporter. We developed a novel assay for assessing Oatp1d1 interactors using the fluorescent probe Lucifer yellow and transient transfection in HEK293 cells. Our data showed that numerous environmental contaminants interact with zebrafish Oatp1d1. Oatp1d1 mediated the transport of diclofenac with very high affinity, followed by high affinity towards perfluorooctanesulfonic acid (PFOS), nonylphenol, gemfibrozil and 17α-ethinylestradiol; moderate affinity towards carbaryl, diazinon and caffeine; and low affinity towards metolachlor. Importantly, many environmental chemicals acted as strong inhibitors of Oatp1d1. A strong inhibition of Oatp1d1 transport activity was found by perfluorooctanoic acid (PFOA), chlorpyrifos-methyl, estrone (E1) and 17β-estradiol (E2), followed by moderate to low inhibition by diethyl phthalate, bisphenol A, 7-acetyl-1,1,3,4,4,6-hexamethyl-1,2,3,4 tetrahydronapthalene and clofibrate. In this study we identified Oatp1d1 as a first Solute Carrier (SLC) transporter involved in the transport of a wide range of xenobiotics in fish. Considering that Oatps in zebrafish have not been characterized before, our work on zebrafish Oatp1d1 offers important new insights on the understanding of uptake processes of environmental contaminants, and contributes to the better characterization of zebrafish as a model species. - Highlights: • We optimized a novel assay for determination of Oatp1d1 interactors • Oatp1d1 is the first SLC characterized fish xenobiotic transporter • PFOS, nonylphenol, diclofenac, EE2, caffeine are high affinity Oatp1d1substrates • PFOA, chlorpyrifos
NMR 1D-imaging of water infiltration into meso-porous matrices
International Nuclear Information System (INIS)
It is shown that coupling nuclear magnetic resonance (NMR) 1D-imaging with the measure of NMR relaxation times and self-diffusion coefficients can be a very powerful approach to investigate fluid infiltration into porous media. Such an experimental design was used to study the very slow seeping of pure water into hydrophobic materials. We consider here three model samples of nuclear waste conditioning matrices which consist in a dispersion of NaNO3 (highly soluble) and/or BaSO4 (poorly soluble) salt grains embedded in a bitumen matrix. Beyond studying the moisture progression according to the sample depth, we analyze the water NMR relaxation times and self-diffusion coefficients along its 1D-concentration profile to obtain spatially resolved information on the solution properties and on the porous structure at different scales. It is also shown that, when the relaxation or self-diffusion properties are multimodal, the 1D-profile of each water population is recovered. Three main levels of information were disclosed along the depth-profiles. They concern (i) the water uptake kinetics, (ii) the salinity and the molecular dynamics of the infiltrated solutions and (iii) the microstructure of the water-filled porosities: open networks coexisting with closed pores. All these findings were fully validated and enriched by NMR cryo-poro-metry experiments and by performing environmental scanning electronic microscopy observations. Surprisingly, results clearly show that insoluble salts enhance the water progression and thereby increase the capability of the material to uptake water. (authors)
Lozano Montero, Juan Andrés; Aragonés Beltrán, José María; García Herranz, Nuria
2009-01-01
More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be i...
The Gain Properties of 1-D Active Photonic Crystal
Institute of Scientific and Technical Information of China (English)
无
2003-01-01
The terminology 'ID frequency'(w ID) is proposed after analyzing the 1D active photonic crystal based on the transfer matrix method. The relationship between wID and the structure parameters of the photonic crystal is investigated.
Popovic, Marta; Zaja, Roko; Fent, Karl; Smital, Tvrtko
2014-10-01
Polyspecific transporters from the organic anion transporting polypeptide (OATP/Oatp) superfamily mediate the uptake of a wide range of compounds. In zebrafish, Oatp1d1 transports conjugated steroid hormones and cortisol. It is predominantly expressed in the liver, brain and testes. In this study we have characterized the transport of xenobiotics by the zebrafish Oatp1d1 transporter. We developed a novel assay for assessing Oatp1d1 interactors using the fluorescent probe Lucifer yellow and transient transfection in HEK293 cells. Our data showed that numerous environmental contaminants interact with zebrafish Oatp1d1. Oatp1d1 mediated the transport of diclofenac with very high affinity, followed by high affinity towards perfluorooctanesulfonic acid (PFOS), nonylphenol, gemfibrozil and 17α-ethinylestradiol; moderate affinity towards carbaryl, diazinon and caffeine; and low affinity towards metolachlor. Importantly, many environmental chemicals acted as strong inhibitors of Oatp1d1. A strong inhibition of Oatp1d1 transport activity was found by perfluorooctanoic acid (PFOA), chlorpyrifos-methyl, estrone (E1) and 17β-estradiol (E2), followed by moderate to low inhibition by diethyl phthalate, bisphenol A, 7-acetyl-1,1,3,4,4,6-hexamethyl-1,2,3,4 tetrahydronapthalene and clofibrate. In this study we identified Oatp1d1 as a first Solute Carrier (SLC) transporter involved in the transport of a wide range of xenobiotics in fish. Considering that Oatps in zebrafish have not been characterized before, our work on zebrafish Oatp1d1 offers important new insights on the understanding of uptake processes of environmental contaminants, and contributes to the better characterization of zebrafish as a model species. PMID:25088042
Numerical modeling of 1-D transient poroelastic waves in the low-frequency range
Chiavassa, Guillaume; Piraux, Joël
2007-01-01
Propagation of transient mechanical waves in porous media is numerically investigated in 1D. The framework is the linear Biot's model with frequency-independant coefficients. The coexistence of a propagating fast wave and a diffusive slow wave makes numerical modeling tricky. A method combining three numerical tools is proposed: a fourth-order ADER scheme with time-splitting to deal with the time-marching, a space-time mesh refinement to account for the small-scale evolution of the slow wave, and an interface method to incorporate the jump conditions at interfaces. Comparisons with analytical solutions confirm the validity of this approach.
TBC1D24 genotype–phenotype correlation
Balestrini, Simona; Milh, Mathieu; Castiglioni, Claudia; Lüthy, Kevin; Finelli, Mattea J.; Verstreken, Patrik; Cardon, Aaron; Stražišar, Barbara Gnidovec; Holder, J. Lloyd; Lesca, Gaetan; Mancardi, Maria M.; Poulat, Anne L.; Repetto, Gabriela M.; Banka, Siddharth; Bilo, Leonilda; Birkeland, Laura E.; Bosch, Friedrich; Brockmann, Knut; Cross, J. Helen; Doummar, Diane; Félix, Temis M.; Giuliano, Fabienne; Hori, Mutsuki; Hüning, Irina; Kayserili, Hulia; Kini, Usha; Lees, Melissa M.; Meenakshi, Girish; Mewasingh, Leena; Pagnamenta, Alistair T.; Peluso, Silvio; Mey, Antje; Rice, Gregory M.; Rosenfeld, Jill A.; Taylor, Jenny C.; Troester, Matthew M.; Stanley, Christine M.; Ville, Dorothee; Walkiewicz, Magdalena; Falace, Antonio; Fassio, Anna; Lemke, Johannes R.; Biskup, Saskia; Tardif, Jessica; Ajeawung, Norbert F.; Tolun, Aslihan; Corbett, Mark; Gecz, Jozef; Afawi, Zaid; Howell, Katherine B.; Oliver, Karen L.; Berkovic, Samuel F.; Scheffer, Ingrid E.; de Falco, Fabrizio A.; Oliver, Peter L.; Striano, Pasquale; Zara, Federico
2016-01-01
Objective: To evaluate the phenotypic spectrum associated with mutations in TBC1D24. Methods: We acquired new clinical, EEG, and neuroimaging data of 11 previously unreported and 37 published patients. TBC1D24 mutations, identified through various sequencing methods, can be found online (http://lovd.nl/TBC1D24). Results: Forty-eight patients were included (28 men, 20 women, average age 21 years) from 30 independent families. Eighteen patients (38%) had myoclonic epilepsies. The other patients carried diagnoses of focal (25%), multifocal (2%), generalized (4%), and unclassified epilepsy (6%), and early-onset epileptic encephalopathy (25%). Most patients had drug-resistant epilepsy. We detail EEG, neuroimaging, developmental, and cognitive features, treatment responsiveness, and physical examination. In silico evaluation revealed 7 different highly conserved motifs, with the most common pathogenic mutation located in the first. Neuronal outgrowth assays showed that some TBC1D24 mutations, associated with the most severe TBC1D24-associated disorders, are not necessarily the most disruptive to this gene function. Conclusions: TBC1D24-related epilepsy syndromes show marked phenotypic pleiotropy, with multisystem involvement and severity spectrum ranging from isolated deafness (not studied here), benign myoclonic epilepsy restricted to childhood with complete seizure control and normal intellect, to early-onset epileptic encephalopathy with severe developmental delay and early death. There is no distinct correlation with mutation type or location yet, but patterns are emerging. Given the phenotypic breadth observed, TBC1D24 mutation screening is indicated in a wide variety of epilepsies. A TBC1D24 consortium was formed to develop further research on this gene and its associated phenotypes. PMID:27281533
1D photonic crystal sensor integrated in a microfluidic system
DEFF Research Database (Denmark)
Nunes, Pedro; Mortensen, Asger; Kutter, Jörg Peter; Mogensen, Klaus Bo
2009-01-01
A refractive index sensor was designed as a 1D resonator incorporated in a microfluidic channel, where aqueous solutions were injected. A sensitivity of 480 nm/RIU and a minimum difference of Deltan = 0.002 were determined.......A refractive index sensor was designed as a 1D resonator incorporated in a microfluidic channel, where aqueous solutions were injected. A sensitivity of 480 nm/RIU and a minimum difference of Deltan = 0.002 were determined....
Supported plasma-made 1D heterostructures: perspectives and applications
Borras, Ana; Macias-Montero, Manuel; Romero-Gomez, Pablo; Gonzalez-Elipe, Agustin R
2011-01-01
Abstract Plasma related methods have been widely used in the fabrication of carbon nanotubes and nanofibres and semiconducting inorganic nanowires. A natural progression of the research in the field of 1D nanostructures is the synthesis of multicomponent nanowires and nanofibres. In this article we review the state of the art of the fabrication by plasma methods of 1D heterostructures including applications and perspectives. Furthermore, recent developments on the use of metal seeds (Ag, A...
International Nuclear Information System (INIS)
This paper presents the quantification of resonance interference effect for multi-group effective cross-section in lattice physics calculation. In the resonance self-shielding method based on the equivalence theory, the resonance interference effect among multiple nuclides cannot be treated directly to the multi-group effective cross-section. The continuous energy or the ultra-fine-group treatment can directly consider the effect, but the application to the fuel assembly geometry is not realistic with practical computation time. In the present study, the resonance interference effect to the multi-group effective cross-section is simply quantified by the resonance interference factor (RIF) in order to confirm the benefit for considering the effect. The RIF is generated for the typical pin-cell geometry of water moderated system. The multi-group effective cross-sections with and without RIFs are compared with the continuous energy Monte-Carlo result. As a result, the significant impact for considering the resonance interference effect is confirmed to the limited nuclide, reaction type and energy group. Fortunately, these have small effect on k-infinity because the resonance interference effect is mainly induced by the wide resonances of 238U to the other minor nuclides (e.g., 235U, 239Pu) in the limited resonance energy ranges. The results also show that the effect is small to the absorption cross-section of 238U, which is the dominant resonance nuclide in the fuel. The quantification results in the present study indicate a useful material to investigate the more advanced resonance treatment for the next generation lattice physics code. (author)
International Nuclear Information System (INIS)
A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author)
Light transport behaviours in quasi-1D disordered waveguides composed of random photonic lattices
Xu, Yuchen; Lin, Yujun; Zhu, Heyuan
2015-01-01
We present a numerical study on the light transport properties which are modulated by the disorder strength in quasi-one-dimensional disordered waveguide which consists of periodically arranged scatterers with random dielectric constant. The transport mean free path is found to stay inversely proportional to the square of the relative fluctuation of the dielectric constant as in the 1D and 2D cases but with . The transport properties of light through a sample with a fixed size can be modulated from ballistic to localized regime as well, and a generalized scaling function is defined to determine the light transport status in such a sample. The calculation of the diffusion coefficient and the energy density profile of the most transmitted eigenchannel clearly exhibits the transition of transport behaviour from diffusion to localization.
L(d1, d2,..., dt)-Number λ(Cn; d1, d2,...,dt) of Cycles
Institute of Scientific and Technical Information of China (English)
GAO Zhen Bin; ZHANG Xiao Dong
2009-01-01
An L(d1,d2,...,dt)-labeling of a graph G is a function f from its vertex set V(G) to the set {0, 1,..., k} for some positive integer k such that {f(x) - f(y)| ≥ di, if the distance between vertices x and y in G is equal to i for i = 1,2,...,t. The L(d1,d2,...,dt)-number λ(G;d1,d2,... ,dt) of G is the smallest integer number k such that G has an L(d1,d2,... ,dt)labeling with max{f(x)|x ∈ V(G)} = k. In this paper, we obtain the exact values for λ(Cn; 2, 2,1) and λ(Cn; 3, 2, 1), and present lower and upper bounds for λ(Cn; 2,..., 2,1,..., 1)
SNAP-3D: a three-dimensional neutron diffusion code
International Nuclear Information System (INIS)
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)
Resonant indirect exchange in 1D semiconductor nanostructures
International Nuclear Information System (INIS)
We consider resonant indirect exchange interaction between magnetic centers in 1D nanostructures. The magnetic centers are assumed to be coupled to the 1D conducting channel by the quantum tunneling which can be of resonant character. The indirect exchange between the centers is mediated by the free carriers of the channel. The two cases of quadratic and linear energy dispersion of the 1D free carriers are considered. The former case is attributed to conventional semiconductor (InGaAs based to be concrete) nanowires or nanowhiskers, while the latter case is associated with carbon nanotubes with magnetic adatoms. We demonstrate that whenever the energy of a bound state at the magnetic center lies within the continuum energy spectra of the delocalized carriers in the channel the indirect exchange is strongly enhanced due to effective tunnel hybridization of the bound states with the continuum. - Highlights: • A resonant indirect exchange interaction between magnetic centers mediated by a 1D conducting channel is considered. • It is shown that the indirect exchange is strongly enhanced due to resonant tunnel coupling of a magnetic bound state with the delocalized states. • The two cases of quadratic and linear energy dispersion of the 1D free carriers are considered. • Pecularities of the indirect exchange mediated by a carbon nanotube has been investigated
Resonant indirect exchange in 1D semiconductor nanostructures
Energy Technology Data Exchange (ETDEWEB)
Rozhansky, I.V., E-mail: rozhansky@gmail.com [Ioffe Institute, Russian Academy of Sciences, St.Petersburg 194021 (Russian Federation); Lappeenranta University of Technology, P.O. Box 20, FI-53851 Lappeenranta (Finland); St. Petersburg State Polytechnic University, St. Petersburg 195251 (Russian Federation); Krainov, I.V.; Averkiev, N.S. [Ioffe Institute, Russian Academy of Sciences, St.Petersburg 194021 (Russian Federation); Lähderanta, E. [Lappeenranta University of Technology, P.O. Box 20, FI-53851 Lappeenranta (Finland)
2015-06-01
We consider resonant indirect exchange interaction between magnetic centers in 1D nanostructures. The magnetic centers are assumed to be coupled to the 1D conducting channel by the quantum tunneling which can be of resonant character. The indirect exchange between the centers is mediated by the free carriers of the channel. The two cases of quadratic and linear energy dispersion of the 1D free carriers are considered. The former case is attributed to conventional semiconductor (InGaAs based to be concrete) nanowires or nanowhiskers, while the latter case is associated with carbon nanotubes with magnetic adatoms. We demonstrate that whenever the energy of a bound state at the magnetic center lies within the continuum energy spectra of the delocalized carriers in the channel the indirect exchange is strongly enhanced due to effective tunnel hybridization of the bound states with the continuum. - Highlights: • A resonant indirect exchange interaction between magnetic centers mediated by a 1D conducting channel is considered. • It is shown that the indirect exchange is strongly enhanced due to resonant tunnel coupling of a magnetic bound state with the delocalized states. • The two cases of quadratic and linear energy dispersion of the 1D free carriers are considered. • Pecularities of the indirect exchange mediated by a carbon nanotube has been investigated.
DYN1D-MSR dynamics code for molten salt reactors
International Nuclear Information System (INIS)
This paper reports about the DYN1D-MSR code development and dynamics studies of the molten salt reactors (MSR) - one of the 'Generation IV International Forum' concepts. In this forum the graphite-moderated channel type MSR based on the previous Oak Ridge National Laboratory research is considered. The liquid molten salt serves as a fuel and coolant, simultaneously and causes two physical peculiarities: the fission energy is released predominantly directly into the coolant and the delayed neutrons precursors are drifted by the fuel flow. The drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit and it can lead to a reactivity loss or gain in the case of fuel flow acceleration or deceleration, respectively. Therefore, specific 3D tool based on in house code DYN3D was developed in FZR. The code DYN3D-MSR is based on the solution of two-group neutron diffusion equation by the help of a nodal expansion method and it includes models of delayed neutrons drift and specific MSR heat release distribution. In this paper the development and verification of 1D version DYN1D-MSR of the code is described. The code has been validated with the experimental data gained from the molten salt reactor experiment performed in the Oak Ridge and after the validation it was applied to several typical transients (overcooling of fuel at the core inlet, reactivity insertion, and the fuel pump trip)
On the Extension of the Analytic Nodal Diffusion Solver ANDES to Sodium Fast Reactors
Ochoa Valero, Raquel; Herrero Carrascosa, José Javier; García Herranz, Nuria
2011-01-01
Within the framework of the Collaborative Project for a European Sodium Fast Reactor, the reactor physics group at UPM is working on the extension of its in-house multi-scale advanced deterministic code COBAYA3 to Sodium Fast Reactors (SFR). COBAYA3 is a 3D multigroup neutron kinetics diffusion code that can be used either as a pin-by-pin code or as a stand-alone nodal code by using the analytic nodal diffusion solver ANDES. It is coupled with thermalhydraulics codes such as COB...
2015-01-01
Diffusion Fundamentals is a peer-reviewed interdisciplinary open-access online journal published as a part of the website Diffusion-Fundamentals.org. It publishes original research articles in the field of diffusion and transport. Main research areas include theory, experiments applications, methods and diffusion-like phenomena. The readers of Diffusion Fundamentals are academic or industrial scientists in all research disciplines. The journal aims at providing a broad forum for their c...
1-D Air-snowpack modeling of atmospheric nitrous acid at South Pole during ANTCI 2003
Directory of Open Access Journals (Sweden)
W. Liao
2008-12-01
Full Text Available A 1-D air-snowpack model of HONO has been developed and constrained by observed chemistry and meteorology data. The 1-D model includes molecular diffusion and mechanical dispersion, windpumping in snow, gas phase to quasi-liquid layer phase HONO transfer and quasi-liquid layer nitrate and interstitial air HONO photolysis. Photolysis of nitrate is important as a dominant HONO source inside the snowpack, however, the observed HONO emission from the snowpack was triggered mainly by the equilibrium between quasi liquid layer nitrite and firn air HONO deep down the snow surface (i.e. 30 cm below snow surface. The high concentration of HONO in the firn air is subsequently transported above the snowpack by diffusion and windpumping. The model uncertainties come mainly from lack of measurements and the interpretation of the QLL properties based on the bulk snow measurements. One critical factor is the ionic strength of QLL nitrite, which is estimated here by the bulk snow pH, nitrite concentration, and QLL to bulk snow volume ratio.
Iris Feature Extraction Method Based on 1D Gabor Filter
Institute of Scientific and Technical Information of China (English)
XU Guang-zhu; MA Yi-de; ZHANG Zai-feng
2008-01-01
The normalized iris image was divided into eight sub-bands, and every column of each sub-band was averaged by rows to generate eight 1D iris signals. Then the even symmetry item of 1D Gabor filter was used to describe local characteristic blocks in 1D iris signals, and the results were quantified by their polarities to generate iris codes. In order to estimate the performance of the presented method, an iris recognition platform was produced and the Hamming distance between two iris codes was computed to measure the dissimilarity of them. The experimental results in CASIA v1 0 and Bath iris image databases show that the proposed iris feature extraction algorithm has a promising potential in iris recognition.
Development of a new two-dimensional Cartesian geometry nodal multigroup discrete-ordinates method
Energy Technology Data Exchange (ETDEWEB)
Pevey, R.E.
1982-07-01
The purpose of this work is the development and testing of a new family of methods for calculating the spatial dependence of the neutron density in nuclear systems described in two-dimensional Cartesian geometry. The energy and angular dependence of the neutron density is approximated using the multigroup and discrete ordinates techniques, respectively. The resulting FORTRAN computer code is designed to handle an arbitrary number of spatial, energy, and angle subdivisions. Any degree of scattering anisotropy can be handled by the code for either external source or fission systems. The basic approach is to (1) approximate the spatial variation of the neutron source across each spatial subdivision as an expansion in terms of a user-supplied set of exponential basis functions; (2) solve analytically for the resulting neutron density inside each region; and (3) approximate this density in the basis function space in order to calculate the next iteration flux-dependent source terms. In the general case the calculation is iterative due to neutron sources which depend on the neutron density itself, such as scattering interactions.
MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1
International Nuclear Information System (INIS)
A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35Cl and 233U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.
MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1
Energy Technology Data Exchange (ETDEWEB)
Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald Kent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gardiner, Steven J. [Univ. of California, Davis, CA (United States); Gray, Mark Girard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lee, Mary Beth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-12-17
A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of ^{35}Cl and ^{233}U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.
Definition and analysis of heavy water reactor benchmarks for testing new multigroup libraries
International Nuclear Information System (INIS)
A set of heavy water reactor benchmarks has been selected for testing new WIMS-D libraries. The libraries were constricted using data from ENDF/B-VI, Release 7, JENDL-3.2 and JEF-2.2 evaluated nuclear data files. The benchmarks cover a wide variety of reactor types and conditions, from fresh fuel to high burnup, and for natural and enriched uranium and Th-U fuels. The main parameters compared are the effective multiplication factor and other integral parameters, and isotopic composition of actinides on burnup cases. Besides, further investigations related with energy spectra used for preparation of WIMS-D libraries when applied on HWTR reactor calculations are included. Mostly of the benchmarks show a good agreement between experimental measurements and calculated values for all libraries. One exception is Th232 benchmark, were it is found that a library with JEND-3.2 Th232 data produces better results than ENDF/B-VI, R.7 and JEF-2.2 Th232 data. Results are slightly improved when HWTR spectra are used for weighting function to prepare the multi-group cross sections. This work is part of the International Atomic Energy Agency's Coordinated Research Project on 'Final Stage of WIMS-D Library Update Project'. (author)
A novel hybrid weighting scheme for multi-group cross section collapsing
International Nuclear Information System (INIS)
Multi-group cross section library generation plays an important role in deterministic transport simulations. In this paper, a new fine-group to broad-group cross section collapsing method is introduced. Rather than a traditional flux weighting, the new method uses a hybrid weighing scheme to collapse the scattering cross section matrix. Based upon a matrix analysis approach, we generalize different weighting schemes and derive the new hybrid weighting scheme, which mathematically shows that it is rational for the scattering cross section to be weighted by the (1) forward fluxes of the incoming/in-bound neutron groups and (2) the adjoint functions of the outgoing/out-bound neutron energy groups. This approach also makes physical sense, since it conserves the “importance flow” of particles through scattering while collapsing cross sections. To conserve the reaction rates at the same time, we re-normalize the hybrid weighted scattering cross section to the original library total scattering reaction rate. We demonstrate that the hybrid weighting scheme is more accurate, especially for the detector response simulation problem in a Dual-Range Coincidence Counter (DRCC) 3-D SN transport model. (author)
Krylov sub-space methods for K-eigenvalue problem in 3-D multigroup neutron transport
International Nuclear Information System (INIS)
The K-eigenvalue problem in nuclear reactor physics is often formulated in the framework of Neutron Transport Theory. The fundamental mode solution of this problem is usually obtained by the Power Iteration method. The present report is concerned with the use of a Krylov Sub-Space method. called ORTHOMIN, to obtain a more efficient solution of the K-eigenvalue problem. A matrix-free approach is proposed which can be easily implemented by using a transport code which can perform fixed source calculations. The Power Iteration and ORTHOMIN schemes are compared for two realistic 3-D multi-group cases: an LWR benchmark and the AHWR Critical Facility. The within-group iterations over self-scattering source are required in the solution of K-eigenvalue problem. They are also accelerated using another Krylov method called Conjugate Gradient method. In this work, the discretisation of Transport Equation is based on fmite-differencing and Sn-method and isotropic scattering is considered. (author)
Directory of Open Access Journals (Sweden)
Sanjukta Pookulangara
2010-12-01
Full Text Available This study examined channel-migration behavior using a decomposed Theory of Planned Behavior with crossover effects in brick-and-mortar stores and the Internet. An online survey was administered at four research sites (N = 547 and factor analysis and structural equation modeling, with multigroup analysis, were utilized for data analysis. Hedonic beliefs did not influence either of the channels, whereas, utilitarian beliefs were significant predictors in both brick-and-mortar stores and the Internet. Additionally, normative beliefs did not influence subjective norms in either of the channels, while self-efficacy influenced perceived behavioral control (PBC in both the channels. Attitude and subjective norms influenced channel-migration intentions for both channels; whereas, PBC was a significant predictor of channel-migration intentions on the Internet only. The crossover effects of normative beliefs and subjective norms on attitude was significant for the Internet. The crossover effects for subjective norms and PBC on attitude was significant for brick-and-mortar stores. Attitude toward channel migration from the Internet to brick-and-mortar stores yielded a negative influence.
Stability analysis of multi-group deterministic and stochastic epidemic models with vaccination rate
International Nuclear Information System (INIS)
We discuss in this paper a deterministic multi-group MSIR epidemic model with a vaccination rate, the basic reproduction number ℛ0, a key parameter in epidemiology, is a threshold which determines the persistence or extinction of the disease. By using Lyapunov function techniques, we show if ℛ0 is greater than 1 and the deterministic model obeys some conditions, then the disease will prevail, the infective persists and the endemic state is asymptotically stable in a feasible region. If ℛ0 is less than or equal to 1, then the infective disappear so the disease dies out. In addition, stochastic noises around the endemic equilibrium will be added to the deterministic MSIR model in order that the deterministic model is extended to a system of stochastic ordinary differential equations. In the stochastic version, we carry out a detailed analysis on the asymptotic behavior of the stochastic model. In addition, regarding the value of ℛ0, when the stochastic system obeys some conditions and ℛ0 is greater than 1, we deduce the stochastic system is stochastically asymptotically stable. Finally, the deterministic and stochastic model dynamics are illustrated through computer simulations. (general)
Development of a new two-dimensional Cartesian geometry nodal multigroup discrete-ordinates method
International Nuclear Information System (INIS)
The purpose of this work is the development and testing of a new family of methods for calculating the spatial dependence of the neutron density in nuclear systems described in two-dimensional Cartesian geometry. The energy and angular dependence of the neutron density is approximated using the multigroup and discrete ordinates techniques, respectively. The resulting FORTRAN computer code is designed to handle an arbitrary number of spatial, energy, and angle subdivisions. Any degree of scattering anisotropy can be handled by the code for either external source or fission systems. The basic approach is to (1) approximate the spatial variation of the neutron source across each spatial subdivision as an expansion in terms of a user-supplied set of exponential basis functions; (2) solve analytically for the resulting neutron density inside each region; and (3) approximate this density in the basis function space in order to calculate the next iteration flux-dependent source terms. In the general case the calculation is iterative due to neutron sources which depend on the neutron density itself, such as scattering interactions
JSD1000: multi-group cross section sets for shielding materials
International Nuclear Information System (INIS)
A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)
Development of a 3D multigroup program for Dancoff factor calculation in pebble bed reactors
International Nuclear Information System (INIS)
Highlights: • Development of a 3D Monte Carlo based code for pebble bed reactors. • Dancoff sensitivity to clad, moderator and fuel cross sections is considered. • Sensitivity of Dancoff to number of energy groups is considered. • Sensitivity of Dancoff to number of fuel and their arrangement is considered. • Excellent agreements vs. MCNP code. - Abstract: The evaluation of multigroup constants in reactor calculations depends on several parameters. One of these parameters is the Dancoff factor which is used for calculating the resonance integral and flux depression in the resonance region in heterogeneous systems. In the current paper, a computer program (MCDAN-3D) is developed for calculating three dimensional black and gray Dancoff coefficients, based on Monte Carlo, escape probability and neutron free flight methods. The developed program is capable to calculate the Dancoff factor for an arbitrary arrangement of fuel and moderator pebbles. Moreover this program can simulate fuels with homogeneous and heterogeneous compositions. It might generate the position of Triso particles in fuel pebbles randomly as well. It could calculate the black and gray Dancoff coefficients since fuel region might have different cross sections. Finally, the effects of clad and moderator are considered and the sensitivity of Dancoff factor with fuels arrangement variation, number of TRISO particles and neutron energy has been studied
Assessment and improvement of the 2D/1D method stability in DeCART
International Nuclear Information System (INIS)
As part of ongoing work with Consortium for Advanced Simulation of Light Water Reactors (CASL), the 2D/1D code, DeCART, has demonstrated some of the advantages of the 2D/1D method with respect to realistic, full-core analysis, particularly over explicit 3D transport methods, which generally have higher memory and computation requirements. The 2D/1D method performs 2D-radial transport sweeps coupled with ID-axial diffusion calculations to provide a full 3D simulation. DeCART employs the 2D method of characteristics for the radial sweeps and ID one-node nodal diffusion for the axial sweeps, coupling the two methods with transverse leakages to ensure a more consistent representation of the transport equation. It has been observed that refinement of the axial plane thickness leads to instabilities in the calculation scheme. This work assesses the sources of these instabilities and the approaches to improve them, especially with respect to negative scattering cross sections and the tightness of the 2D-radial/ID-axial coupling schemes. Fourier analyses show that the existing iteration scheme is not unconditionally stable, suggesting a tighter coupling scheme is required. For this reason 3D-CMFD has been implemented, among other developments, to ensure more stable calculation. A matrix of test cases has been used to assess the convergence, with the primary parameter being the axial plane thickness, which has been refined down to 1 cm. These cases demonstrate the issues observed and how the modification improve the stability. However, it is apparent that more work is necessary to ensure unconditional stability. (authors)
Assessment and improvement of the 2D/1D method stability in DeCART
Energy Technology Data Exchange (ETDEWEB)
Stimpson, S.; Young, M.; Collins, B.; Kelley, B.; Downar, T. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States)
2013-07-01
As part of ongoing work with Consortium for Advanced Simulation of Light Water Reactors (CASL), the 2D/1D code, DeCART, has demonstrated some of the advantages of the 2D/1D method with respect to realistic, full-core analysis, particularly over explicit 3D transport methods, which generally have higher memory and computation requirements. The 2D/1D method performs 2D-radial transport sweeps coupled with ID-axial diffusion calculations to provide a full 3D simulation. DeCART employs the 2D method of characteristics for the radial sweeps and ID one-node nodal diffusion for the axial sweeps, coupling the two methods with transverse leakages to ensure a more consistent representation of the transport equation. It has been observed that refinement of the axial plane thickness leads to instabilities in the calculation scheme. This work assesses the sources of these instabilities and the approaches to improve them, especially with respect to negative scattering cross sections and the tightness of the 2D-radial/ID-axial coupling schemes. Fourier analyses show that the existing iteration scheme is not unconditionally stable, suggesting a tighter coupling scheme is required. For this reason 3D-CMFD has been implemented, among other developments, to ensure more stable calculation. A matrix of test cases has been used to assess the convergence, with the primary parameter being the axial plane thickness, which has been refined down to 1 cm. These cases demonstrate the issues observed and how the modification improve the stability. However, it is apparent that more work is necessary to ensure unconditional stability. (authors)
Nonreciprocity of edge modes in 1D magnonic crystal
Energy Technology Data Exchange (ETDEWEB)
Lisenkov, I., E-mail: ivan.lisenkov@phystech.edu [Kotelnikov Institute of Radio-engineering and Electronics of RAS, 11-7 Mokhovaya st., Moscow 125009 (Russian Federation); Department of Physics, Oakland University, 2200 N. Squirrel Rd., Rochester, MI 48309 (United States); Moscow Institute of Physics and Technology, 9 Instituskij per., Dolgoprudny, 141700, Moscow Region (Russian Federation); Kalyabin, D., E-mail: dmitry.kalyabin@phystech.edu [Kotelnikov Institute of Radio-engineering and Electronics of RAS, 11-7 Mokhovaya st., Moscow 125009 (Russian Federation); Moscow Institute of Physics and Technology, 9 Instituskij per., Dolgoprudny, 141700, Moscow Region (Russian Federation); Osokin, S. [Kotelnikov Institute of Radio-engineering and Electronics of RAS, 11-7 Mokhovaya st., Moscow 125009 (Russian Federation); Moscow Institute of Physics and Technology, 9 Instituskij per., Dolgoprudny, 141700, Moscow Region (Russian Federation); Klos, J.W.; Krawczyk, M. [Adam Mickiewicz University in Poznan, Umultowska 85, Poznan 61-614 (Poland); Nikitov, S., E-mail: nikitov@cplire.ru [Kotelnikov Institute of Radio-engineering and Electronics of RAS, 11-7 Mokhovaya st., Moscow 125009 (Russian Federation); Moscow Institute of Physics and Technology, 9 Instituskij per., Dolgoprudny, 141700, Moscow Region (Russian Federation); Saratov State University, 112 Bol' shaya Kazach' ya, Saratov 410012 (Russian Federation)
2015-03-15
Spin waves propagation in 1D magnonic crystals is investigated theoretically. Mathematical model based on plane wave expansion method is applied to different types of magnonic crystals, namely bi-component magnonic crystal with symmetric/asymmetric boundaries and ferromagnetic film with periodically corrugated top surface. It is shown that edge modes in magnonic crystals may exhibit nonreciprocal behaviour at much lower frequencies than in homogeneous films. - Highlights: • Magnetostatic surface spin waves in 1D magnonic crystals were studied theoretically. • Mathematical model is based on plane wave method. • Mathematical model was applied to different types of magnonic crystals. • Stop band formation and nonreciprocity were obtained.
Nonreciprocity of edge modes in 1D magnonic crystal
International Nuclear Information System (INIS)
Spin waves propagation in 1D magnonic crystals is investigated theoretically. Mathematical model based on plane wave expansion method is applied to different types of magnonic crystals, namely bi-component magnonic crystal with symmetric/asymmetric boundaries and ferromagnetic film with periodically corrugated top surface. It is shown that edge modes in magnonic crystals may exhibit nonreciprocal behaviour at much lower frequencies than in homogeneous films. - Highlights: • Magnetostatic surface spin waves in 1D magnonic crystals were studied theoretically. • Mathematical model is based on plane wave method. • Mathematical model was applied to different types of magnonic crystals. • Stop band formation and nonreciprocity were obtained
1D antiferromagnetism in spin‐alternating bimetallic chains
Coronado Miralles, Eugenio; Sapiña Navarro, Fernando; Drillon, M.; De Jongh, L.J.
1990-01-01
The magnetic and thermal properties of the ordered bimetallic chain CoNi(EDTA)⋅6H2O in the very low‐temperature range are reported. The magnetic behavior does not exhibit the characteristic features of 1D ferrimagnets, but a continuous decrease of χmT towards zero at absolute zero. This 1D antiferromagnetic behavior results from an accidental compensation between the moments located at the two sublattices. This behavior, as well as the specific‐heat results, are modeled on the basis of an Isi...
Quantum electrodynamics with 1D arti cial atoms
DEFF Research Database (Denmark)
Javadi, Alisa
A 1D atom, a single quantum emitter coupled to a single optical mode, exhibits rich quantum electrodynamic (QED) e_ects and is thought to be the key ingredient for many applications in quantuminformation processing. Single quantum dots (QD) in photonic-crystal waveguides (PCW) constitute a robust...... photons as expected from the theory. The value of g(2)(0) is around 1.08. The results con_rm the observation of an on-chip giant optical nonlinearity and the 1D atom behavior. Another direction in this thesis has been to investigate the e_ect of Anderson localization on the electrodynamics of QDs in PCWs...
Axial transport solvers for the 2D/1D scheme in MPACT
International Nuclear Information System (INIS)
The MPACT code being developed collaboratively at the University of Michigan (UM) and Oak Ridge National Laboratory (ORNL) provides users with a variety of deterministic methods for solving the 2D and 3D Boltzmann transport equation. One of these methods, the 2D/1D technique, decomposes 3D problems into a 1D axial stack of 2D radial planes. In this scheme, the 2D planes are typically solved using a method such as the Method of Characteristics (MOC) to preserve the geometric heterogeneity in the radial direction. These planes are incorporated into a 1D axial solver, which can use a variety of methods. This work demonstrates the use of the traditional nodal methods for solving the 1D axial problem (finite difference, NEM, SANM, SP3), but also introduces a discrete ordinates (Sn) solver which uses up to cubic Legendre expansion spatially and can also incorporate higher order angular distributions of the radial transverse leakage. Several test cases are presented to demonstrate the accuracy of the solvers for various axial sizes. The first three are the 3D-C5G7 extension benchmark cases. The fourth case is a single quarter assembly benchmark problem with explicit nozzle, plenum, and core plate modelling known as AMA Problem 3. The final case is a quarter core benchmark problem that is an extension of the quarter assembly problem known as AMA Problem 5. In general, the diffusion-based axial solvers perform very well, though higher-order solvers provide some benefit in more difficult problems, particularly rodded cases. (author)
International Nuclear Information System (INIS)
Full text: The principal nuclear design tools available to the shielding designer include diffusion approximation, transport theory, and Monte Carlo techniques. Full transport theory or Monte Carlo methods are routinely used for shielding analyses, where penetration investigations are more sensitive to directional aspects. However, the aim of this paper is to illustrate the coupled neutron-gamma Albedo method particularly as applied to problems of shielding analysis. The multigroup Albedo method is applied to coupled neutron-gamma radiations considering 'n' neutron energy groups and 'g' gamma energy groups to estimate the probabilities of transmission through, absorption in, and reflection from shieldings composed by multiple material layers, 'm' slabs, in which no fission occurs. In this study, these energy groups were selected in order to minimize upscattering effects of the radiation from lower energy groups to higher energy groups. However, neutrons of all energies are assumed to generate gammas of all energies. The reflection coefficient or Albedo is defined as the current of the reflected radiation divided by the incident radiation current. The absorption coefficient is defined as the rate at which radiation is lost by absorption per second divided by the amount of incident radiation per second. The transmission coefficient is defined as the current of the transmitted radiation divided by the incident radiation current. The interaction probabilities can be arranged in matrix form where the rows indicate the energy group of the incident radiation and the columns indicate the energy group of the radiation after interaction. Thus, each material has 3 sets of distinct matrices, for the interactions neutron-neutron (N-N), neutron-gamma (N-G) and gamma-gamma (G-G). Each set is composed by 3 matrices, giving a total of 9 matrices per material. The first matrix set is for scattering/downscattering of neutrons (N-N); the next set is for scattering/downscattering of
Nonlinear ac conductivity of interacting 1d electron systems
Rosenow, Bernd; Nattermann, Thomas
2004-01-01
We consider low energy charge transport in one-dimensional (1d) electron systems with short range interactions under the influence of a random potential. Combining RG and instanton methods, we calculate the nonlinear ac conductivity and discuss the crossover between the nonanalytic field dependence of the electric current at zero frequency and the linear ac conductivity at small electric fields and finite frequency.
A 1D wavelet filtering for ultrasound images despeckling
Dahdouh, Sonia; Dubois, Mathieu; Frenoux, Emmanuelle; Osorio, Angel
2010-03-01
Ultrasound images appearance is characterized by speckle, shadows, signal dropout and low contrast which make them really difficult to process and leads to a very poor signal to noise ratio. Therefore, for main imaging applications, a denoising step is necessary to apply successfully medical imaging algorithms on such images. However, due to speckle statistics, denoising and enhancing edges on these images without inducing additional blurring is a real challenging problem on which usual filters often fail. To deal with such problems, a large number of papers are working on B-mode images considering that the noise is purely multiplicative. Making such an assertion could be misleading, because of internal pre-processing such as log compression which are done in the ultrasound device. To address those questions, we designed a novel filtering method based on 1D Radiofrequency signal. Indeed, since B-mode images are initially composed of 1D signals and since the log compression made by ultrasound devices modifies noise statistics, we decided to filter directly the 1D Radiofrequency signal envelope before log compression and image reconstitution, in order to conserve as much information as possible. A bi-orthogonal wavelet transform is applied to the log transform of each signal and an adaptive 1D split and merge like algorithm is used to denoise wavelet coefficients. Experiments were carried out on synthetic data sets simulated with Field II simulator and results show that our filter outperforms classical speckle filtering methods like Lee, non-linear means or SRAD filters.
Quantitative 1D saturation profiles on chalk by NMR
DEFF Research Database (Denmark)
Olsen, Dan; Topp, Simon; Stensgaard, Anders;
1996-01-01
Quantitative one-dimensional saturation profiles showing the distribution of water and oil in chalk core samples are calculated from NMR measurements utilizing a 1D CSI spectroscopy pulse sequence. Saturation profiles may be acquired under conditions of fluid flow through the sample. Results reveal...
Large Time existence For 1D Green-Naghdi equations
Israwi, Samer
2009-01-01
We consider here the $1D $ Green-Naghdi equations that are commonly used in coastal oceanography to describe the propagation of large amplitude surface waves. We show that the solution of the Green-Naghdi equations can be constructed by a standard Picard iterative scheme so that there is no loss of regularity of the solution with respect to the initial condition.
Lifescience Database Archive (English)
Full Text Available 1D6R 大豆 Soybean Glycine max (L.) Merrill Bowman-Birk Type Proteinase Inhibitor Precursor Glyci ... Warkentin, G.Wenzl, P.Flecker Crystal Structure Of Cancer ... Chemopreventive Bowman-Birk Inhibitor In Ternary C ...
Simulation of Organic Solar Cells Using AMPS-1D Program
Directory of Open Access Journals (Sweden)
Samah G. Babiker
2012-03-01
Full Text Available The analysis of microelectronic and photonic structure in one dimension program [AMPS-1D] program has been successfully used to study inorganic solar cells. In this work the program has been used to optimize the performance of the organic solar cells. The cells considered consist of poly(2-methoxy-5-(3,7- dimethyloctyloxy-1,4-phenylenevinylene [MDMO-PPV
NEW FEATURES OF HYDRUS-1D, VERSION 3.0
This paper briefly summarizes new features in version 3.0 of HYDRUS-1D, released in May 2005, as compared to version 2.1. The new features are a) new approaches to simulate preferential and nonequilibrium water flow and solute transport, b) a new hysteresis module that avoids the effects of pumpin...
Optical properties of LEDs with patterned 1D photonic crystal
Hronec, P.; Kuzma, A.; Å kriniarová, J.; Kováč, J.; Benčurová, A.; Haščík, Å.; Nemec, P.
2015-08-01
In this paper we focus on the application of the one-dimensional photonic crystal (1D PhC) structures on the top of Al0.295Ga0.705As/GaAs multi-quantum well light emitting diode (MQW LED). 1D PhC structures with periods of 600 nm, 700 nm, 800 nm, and 900 nm were fabricated by the E-Beam Direct Write (EBDW) Lithography. Effect of 1D PhC period on the light extraction enhancement was studied. 1D PhC LED radiation profiles were obtained from Near Surface Light Emission Images (NSLEI). Measurements showed the strongest light extraction enhancement using 800 nm period of PhC. Investigation of PhC LED radiation profiles showed strong light decoupling when light reaches PhC structure. Achieved LEE was from 22.6% for 600 nm PhC LED to 47.0% for 800 nm PhC LED. LED with PhC structure at its surface was simulated by FDTD simulation method under excitation of appropriate launch field.
Energy Technology Data Exchange (ETDEWEB)
Van Geemert, Rene [AREVA, AREVA NP, Erlangen (Germany)
2008-07-01
For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIA{sup R} next generation core simulation software package. ARCADIA{sup R} will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP{sub 3} nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIA{sup R}. For enabling a high computational performance, the SP{sub N} calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP{sub 3} is about 1.4 times as expensive computationally as SP{sub 1} (diffusion). The developed SP{sub 3} solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIA{sup R}. With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP{sub 3} instead of SP{sub 1} has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SP{sub N} results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO{sub 2}/MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP{sub 3} computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)
Bessel Series in the Space H1(D)%H1(D)空间的Bessel级数
Institute of Scientific and Technical Information of China (English)
木乐华
2001-01-01
An identity concerning the partial sums of Bessel series and power series for H1(D) functions is given.Based on it,many of precise extimates about the deviation of the partial sums of Bessel series can be obtained.%本文给出关于H1(D)空间中函数的Bessel级数的部分和用幂级数的部分和表示的一个恒等式.基于它，可以得到Bessel级数部分和偏差的诸多精确估计.
Diffusion of D-alpha-tocopherol (1); carbon dioxide (2)
Winkelmann, J.
This document is part of Subvolume A `Gases in Gases, Liquids and their Mixtures' of Volume 15 `Diffusion in Gases, Liquids and Electrolytes' of Landolt-Börnstein Group IV `Physical Chemistry'. It is part of the chapter of the chapter `Diffusion in Pure Gases' and contains data on diffusion of (1) D-alpha-tocopherol; (2) carbon dioxide
WESSELINGH, JA
1993-01-01
In this paper I first briefly look at diffusion in several controlled release devices. I have found that these exploit complicated diffusional mechanisms. These cannot be understood using the conventional description of diffusion. So, a general theory of multicomponent diffusion which can describe t
Fukuyama, Hidenao
Recent advances of magnetic resonance imaging have been described, especially stressed on the diffusion sequences. We have recently applied the diffusion sequence to functional brain imaging, and found the appropriate results. In addition to the neurosciences fields, diffusion weighted images have improved the accuracies of clinical diagnosis depending upon magnetic resonance images in stroke as well as inflammations.
MORSE-EMP, Monte-Carlo Neutron and Gamma Multigroup Transport with Array Geometry, for PC
International Nuclear Information System (INIS)
A - Description of program or function: MORSE-CGA was developed to add the capability of modeling rectangular lattices for nuclear reactor cores or for multi-partitioned structures. It thus enhances the capability of the MORSE code system. The MORSE code is a multipurpose neutron and gamma-ray transport Monte Carlo code. It has been designed as a tool for solving most shielding problems. Through the use of multigroup cross sections, the solution of neutron, gamma-ray, or coupled neutron-gamma-ray problems may be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry may be used with an albedo option available at any material surface. Isotropic or anisotropic scattering up to a P16 expansion of the angular distribution is allowed. B - Method of solution: Monte Carlo methods are used to solve the forward and the adjoint transport equations. Quantities of interest are then obtained by summing the contributions over all collisions, and frequently over most of phase space. Standard multigroup cross sections, such as those used in discrete ordinates codes, may be used as input; either CCC-254/ANISN, CCC-42/DTF-IV, or CCC-89/DOT cross section formats are acceptable. Anisotropic scattering is treated for each group-to-group transfer by utilizing a generalized Gaussian quadrature technique. The Morse code is organised into functional modules with simplified interfaces such that new modules may be incorporated with reasonable ease. The modules are (1) random walk, (2) cross section, (3) geometry, (4) analysis, and (5) diagnostic. The MARS module allows the efficient modeling of complex lattice geometries. Computer memory requirements are minimized because fewer body specifications are needed and nesting and repetition of arrays is allowed. While the basic MORSE code assumes the analysis module is user-written, a general analysis package, SAMBO is included. SAMBO handles some
Symmetry breaking in the opinion dynamics of a multi-group project organization
International Nuclear Information System (INIS)
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO
PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices
International Nuclear Information System (INIS)
PUFF-III is an extension of the previous PUFF-II code that was developed in the 1970s and early 1980s. The PUFF codes process the Evaluated Nuclear Data File (ENDF) covariance data and generate multigroup covariance matrices on a user-specified energy grid structure. Unlike its predecessor, PUFF-III can process the new ENDF/B-VI data formats. In particular, PUFF-III has the capability to process the spontaneous fission covariances for fission neutron multiplicity. With regard to the covariance data in File 33 of the ENDF system, PUFF-III has the capability to process short-range variance formats, as well as the lumped reaction covariance data formats that were introduced in ENDF/B-V. In addition to the new ENDF formats, a new directory feature is now available that allows the user to obtain a detailed directory of the uncertainty information in the data files without visually inspecting the ENDF data. Following the correlation matrix calculation, PUFF-III also evaluates the eigenvalues of each correlation matrix and tests each matrix for positive definiteness. Additional new features are discussed in the manual. PUFF-III has been developed for implementation in the AMPX code system, and several modifications were incorporated to improve memory allocation tasks and input/output operations. Consequently, the resulting code has a structure that is similar to other modules in the AMPX code system. With the release of PUFF-III, a new and improved covariance processing code is available to process ENDF covariance formats through Version VI
Symmetry breaking in the opinion dynamics of a multi-group project organization
Zhu, Zhen-Tao; Zhou, Jing; Li, Ping; Chen, Xing-Guang
2012-10-01
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO.
Symmetry breaking in the opinion dynamics of a multi-group project organization
Institute of Scientific and Technical Information of China (English)
Zhu Zhen-Tao; Zhou Jing; Li Ping; Chen Xing-Guang
2012-01-01
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces:(i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness.Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes,i.e.,a deadlock regime,a convergence regime,and a bifurcation regime in opinion dynamics.The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to.In the case of a three-group project with a symmetric social network,both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord,instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p.125 Fig.5),project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations,which urges that apart from divergence in participants' interests,nonlinear interaction can also make conflict inevitable in the PO.The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO.It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO.
PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices
Energy Technology Data Exchange (ETDEWEB)
Dunn, M.E.
2000-06-01
PUFF-III is an extension of the previous PUFF-II code that was developed in the 1970s and early 1980s. The PUFF codes process the Evaluated Nuclear Data File (ENDF) covariance data and generate multigroup covariance matrices on a user-specified energy grid structure. Unlike its predecessor, PUFF-III can process the new ENDF/B-VI data formats. In particular, PUFF-III has the capability to process the spontaneous fission covariances for fission neutron multiplicity. With regard to the covariance data in File 33 of the ENDF system, PUFF-III has the capability to process short-range variance formats, as well as the lumped reaction covariance data formats that were introduced in ENDF/B-V. In addition to the new ENDF formats, a new directory feature is now available that allows the user to obtain a detailed directory of the uncertainty information in the data files without visually inspecting the ENDF data. Following the correlation matrix calculation, PUFF-III also evaluates the eigenvalues of each correlation matrix and tests each matrix for positive definiteness. Additional new features are discussed in the manual. PUFF-III has been developed for implementation in the AMPX code system, and several modifications were incorporated to improve memory allocation tasks and input/output operations. Consequently, the resulting code has a structure that is similar to other modules in the AMPX code system. With the release of PUFF-III, a new and improved covariance processing code is available to process ENDF covariance formats through Version VI.
Optimization of multi-group cross sections for fast reactor analysis
International Nuclear Information System (INIS)
The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO2-UO2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)
PACER-IBM: A two dimensional Monte Carlo multigroup program for the IBM personal computer
International Nuclear Information System (INIS)
PACER-IBM is a Monte Carlo computer program written in BASIC at Bettis for the IBM personal computer. This program is capable of solving simple two dimensional neutron transport problems in X - Y geometry. The space - energy neutron flux distribution over the energy range of 10 MeV to 0 eV is calculated using fixed source starts within the source regions of the solution geometry. PACER-IBM accesses multigroup cross sections which have been prepared using the CDC-7600 program RCPL1, a program to prepare neutron and photon cross section libraries for RCP01, and down loaded from the CDC-7600. Neutron behavior in the PC program is simulated by a random walk, a process that is identical to that used in the CDC-7600 Monte Carlo program PACER. A neutron's location in phase space, i.e.., its position, direction, and energy is selected randomly and the neutron is tracked through the solution geometry (in free flight) until a collision occurs. In the collision analysis new neutron direction and energy are selected randomly based upon probabilities determined from basic neutron cross section data. The tracking process and the collision analysis is continued until a termination event, such as absorption, leakage, or slowing down below a specified energy, occurs. The set of calculations for one neutron from source to termination is called a neutron history. A large number of histories is processed to estimate the space-energy neutron flux. Comparisons of results with CDC-7600 PACER solutions are favorable for several two dimensional test problems. 2 refs
Developing 1D nanostructure arrays for future nanophotonics
Directory of Open Access Journals (Sweden)
Cooke DG
2006-01-01
Full Text Available AbstractThere is intense and growing interest in one-dimensional (1-D nanostructures from the perspective of their synthesis and unique properties, especially with respect to their excellent optical response and an ability to form heterostructures. This review discusses alternative approaches to preparation and organization of such structures, and their potential properties. In particular, molecular-scale printing is highlighted as a method for creating organized pre-cursor structure for locating nanowires, as well as vapor–liquid–solid (VLS templated growth using nano-channel alumina (NCA, and deposition of 1-D structures with glancing angle deposition (GLAD. As regards novel optical properties, we discuss as an example, finite size photonic crystal cavity structures formed from such nanostructure arrays possessing highQand small mode volume, and being ideal for developing future nanolasers.
On the 1D Coulomb Klein-Gordon equation
International Nuclear Information System (INIS)
For a single particle of mass m experiencing the potential -α/vertical bar x vertical bar, the 1D Klein-Gordon equation is mathematically underdefined even when α 2 the ground-state energy E decreases through zero, and soon after that mR reaches a finite critical value below which E becomes complex, signalling a breakdown of the single-particle theory. At this critical point of the curve E(mR) the Klein-Gordon norm changes sign: the curve has a lower branch describing a bound antiparticle state, with positive energy -E, which exists for mR between the critical and some higher value where E reaches -m. Though apparently unanticipated in this context, similar scenarios are in fact familiar for strong short-range potentials (1D or 3D), and also for strong 3D Coulomb potentials with α of order unity
Fuzzball solutions and D1-D5 microstates
Skenderis, K; Skenderis, Kostas; Taylor, Marika
2006-01-01
We revisit the relation between fuzzball solutions and D1-D5 microstates. A consequence of the fact that the RR ground states (in the usual basis) are eigenstates of the R-charge is that only neutral operators can have non-vanishing expectation values on these states. We compute the holographic 1-point functions of the fuzzball solutions and find that charged chiral primaries have non-zero expectation values, except when the curve characterizing the solution is circular. The non-zero vevs reflect the fact that a generic curve breaks R-symmetry completely. This implies that fuzzball solutions (excepting circular ones) can only correspond to superpositions of RR states. We construct new solutions by appropriately superimposing fuzzball solutions that have vanishing vevs for all charged chiral primary operators and can therefore correspond to D1-D5 microstates.
D1-D5-P microstates at the cap
Giusto, Stefano; Mathur, Samir D; Turton, David
2012-01-01
The geometries describing D1-D5-P bound states in string theory have three regions: flat asymptotics, an anti-de Sitter throat, and a 'cap' region at the bottom of the throat. We identify the CFT description of a known class of supersymmetric D1-D5-P microstate geometries which describe degrees of freedom in the cap region. The class includes both regular solutions and solutions with conical defects, and generalizes configurations with known CFT descriptions: a parameter related to spectral flow in the CFT is generalized from integer to fractional values. We provide strong evidence for this identification by comparing the massless scalar excitation spectrum between gravity and CFT and finding exact agreement.
Waves in a 1D electrorheological dusty plasma lattice
Rosenberg, M.
2015-08-01
The behavior of waves in a one-dimensional (1D) dusty plasma lattice where the dust interacts via Yukawa and electric dipole interactions is discussed theoretically. This study is motivated by recent reports on electrorheological dusty plasmas (e.g. Ivlev et al. 2008 Phys. Rev. Lett. 100, 095003) where the dipole interaction arises due to an external uniaxial AC electric field that distorts the Debye sphere surrounding each grain. Application to possible dusty plasma experimental parameters is discussed.
Dentin dysplasia type 1d: A rare case
Sujit Ranjan Sahoo; Sonia Aggarwal
2014-01-01
Dentin dysplasia is a rare hereditary disturbance of dentin formation characterized by a defective dentin development with clinically normal-appearing crowns, severe hypermobility of teeth and spontaneous dental abscesses or cysts. Radiographic analysis shows obliteration of all pulp chambers by pulp stones, short, blunted and malformed or absent roots, peri-apical radiolucencies of noncarious teeth. We present a case of dentin dysplasia type 1d in a 19-year-old boy along with the clinical, r...
Blind Detection of Severely Blurred 1D Barcode
Dridi, Noura; Delignon, Yves; Sawaya, Wadih; Septier, François
2010-01-01
In this paper, we present a joint blind channel estimation and symbol detection for decoding a blurred and noisy 1D barcode captured image. From an information transmission point of view, we show that the channel impulse response, the noise power and the symbols can be efficiently estimated by taking into account the signal structure such as the cyclostationary property of the hidden Markov process to estimate. Based on the Expectation-Maximisation method, we show that the new algorithm offer...
A study of slow light in 1D photonic crystals
Yudistira, D.; Hoekstra, H.J.W.M.; Hammer, M; Marpaung, D.A.I.
2005-01-01
Slow light (SL) states corresponding to wavelength regions near the bandgap edge of grating structure are known to show strong field enhancement. Such states may be excited efficiently by well-optimised adiabatic transitions in such structures, e.g., by slowly turning on the modulation depth. To study adiabatic excitations, a detailed research in 1D is performed to obtain insight into the relation between the device parameters and properties like enhancement and modal reflection. The results ...
Theory of slow light excitation in 1D photonic crystals
Yudistira, D.; Marpaung, D.A.I.; Handoyo, H.P.; Hoekstra, H.J.W.M.; Hammer, M; Tjia, M.O.; Iskandar, A.A.
2004-01-01
Slow light (SL) states corresponding to wavelength regions near the bandgap edge of grated structures are known to show strong eld enhancement. Such states may be excited efciently by well-optimised adiabatic transitions in grating structures, e.g., by slowly turning on the modulation depth. To study adiabatic excitations, a detailed investigation in 1D is performed to obtain insight into the relation between the device parameters and properties like eld enhancement and modal reection. The re...
Directory of Open Access Journals (Sweden)
Lena Nusser
2015-05-01
Full Text Available This article focuses on measurement invariance of the assessment of educationally relevant constructs via written questionnaires for students at special schools and at low track schools attending 5th grade. To examine optimal conditions of administration for students with special educational needs in the area of learning an experimental design was implemented. If accommodated questionnaires, different school enrollments as well as competence differences allow equivalent assessment of reading motivation and academic self-concepts will be investigated with multi-group comparison of confirmatory factor analysis. The results indicate that comparisons between groups of students at special schools and low track schools are meaningful for certain constructs.
FEM-RZ, 2-D Multigroup Neutron Transport in R-Z Geometry, Eigenvalue and Fixed Source Problems
International Nuclear Information System (INIS)
1 - Nature of the physical problem solved: FEM-RZ is a computer program for solving multi-group neutron transport problems in two-dimensional cylindrical (r,z) geometry. It can solve not only eigenvalue problems but also other problems, such as fixed source problems. 2 - Method of solution: The method of higher order finite elements is used for the spatial variables. It is based on the discontinuous method with Galerkin-type scheme. The discrete ordinate Sn method is used for the angular variables. 3 - Restrictions on the complexity of the problem: No restrictions except for computer size
International Nuclear Information System (INIS)
It is well known that the temperature and background dependent neutron cross-sections are conventionally represented, in a problem-independent multigroup cross-section set, by specifying, for each group and reaction, the unshielded cross-section along with a set of self-shielding factors for various background cross-sections and temperatures. Usually the unshielded group cross-section is assumed to be independent of temperature. The observation presented in this paper, with examples, shows that the unshielded cross-section could significantly depend on temperature, depending on the group boundaries. (author)
International Nuclear Information System (INIS)
MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (keff, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors)
Development of 1D Liner Compression Code for IDL
Shimazu, Akihisa; Slough, John; Pancotti, Anthony
2015-11-01
A 1D liner compression code is developed to model liner implosion dynamics in the Inductively Driven Liner Experiment (IDL) where FRC plasmoid is compressed via inductively-driven metal liners. The driver circuit, magnetic field, joule heating, and liner dynamics calculations are performed at each time step in sequence to couple these effects in the code. To obtain more realistic magnetic field results for a given drive coil geometry, 2D and 3D effects are incorporated into the 1D field calculation through use of correction factor table lookup approach. Commercial low-frequency electromagnetic fields solver, ANSYS Maxwell 3D, is used to solve the magnetic field profile for static liner condition at various liner radius in order to derive correction factors for the 1D field calculation in the code. The liner dynamics results from the code is verified to be in good agreement with the results from commercial explicit dynamics solver, ANSYS Explicit Dynamics, and previous liner experiment. The developed code is used to optimize the capacitor bank and driver coil design for better energy transfer and coupling. FRC gain calculations are also performed using the liner compression data from the code for the conceptual design of the reactor sized system for fusion energy gains.
MARG1D: One dimensional outer region matching data code
International Nuclear Information System (INIS)
A code MARG1D has been developed which computes outer region matching data of the one dimensional Newcomb equation. Matching data play an important role in the resistive (and non ideal) Magneto-hydrodynamic (MHD) stability analysis in a tokamak plasma. The MARG1D code computes matching data by using the boundary value method or by the eigenvalue method. Variational principles are derived for the problems to be solved and a finite element method is applied. Except for the case of marginal stability, the eigenvalue method is equivalent to the boundary value method. However, the eigenvalue method has the several advantages: it is a new method of ideal MHD stability analysis for which the marginally stable state can be identified, and it guarantees numerical stability in computing matching data close to marginal stability. We perform detailed numerical experiments for a model equation with analytical solutions and for the Newcomb equation in the m=1 mode theory. Numerical experiments show that MARG1D code gives the matching data with numerical stability and high accuracy. (author)
Supported plasma-made 1D heterostructures: perspectives and applications
Borras, Ana; Macias-Montero, Manuel; Romero-Gomez, Pablo; Gonzalez-Elipe, Agustin R.
2011-05-01
Plasma-related methods have been widely used in the fabrication of carbon nanotubes and nanofibres (NFs) and semiconducting inorganic nanowires (NWs). A natural progression of the research in the field of 1D nanostructures is the synthesis of multicomponent NWs and NFs. In this paper we review the state of the art of the fabrication by plasma methods of 1D heterostructures including applications and perspectives. Furthermore, recent developments on the use of metal seeds (Ag, Au, Pt) to obtain metal@oxide nanostructures are also extensively described. Results are shown for various metal substrates, either metal foils or supported nanoparticles/thin films of the metal where the effects of the size, surface coverage, percolation degree and thickness of the metal seeds have been systematically evaluated. The possibilities of the process are illustrated by the preparation of nanostructured films and supported NFs of different metal@oxides (Ag, Au and SiO2, TiO2, ZnO). Particularly, in the case of silver, the application of an oxygen plasma treatment prior to the deposition of the oxide was critical for efficiently controlling the growth of the 1D heterostructures. A phenomenological model is proposed to account for the thin-film nanostructuring and fibre formation by considering basic phenomena such as stress relaxation, inhomogeneities in the plasma sheath electrical field and the local disturbance of the oxide growth.
Supported plasma-made 1D heterostructures: perspectives and applications
International Nuclear Information System (INIS)
Plasma-related methods have been widely used in the fabrication of carbon nanotubes and nanofibres (NFs) and semiconducting inorganic nanowires (NWs). A natural progression of the research in the field of 1D nanostructures is the synthesis of multicomponent NWs and NFs. In this paper we review the state of the art of the fabrication by plasma methods of 1D heterostructures including applications and perspectives. Furthermore, recent developments on the use of metal seeds (Ag, Au, Pt) to obtain metal-oxide nanostructures are also extensively described. Results are shown for various metal substrates, either metal foils or supported nanoparticles/thin films of the metal where the effects of the size, surface coverage, percolation degree and thickness of the metal seeds have been systematically evaluated. The possibilities of the process are illustrated by the preparation of nanostructured films and supported NFs of different metal-oxides (Ag, Au and SiO2, TiO2, ZnO). Particularly, in the case of silver, the application of an oxygen plasma treatment prior to the deposition of the oxide was critical for efficiently controlling the growth of the 1D heterostructures. A phenomenological model is proposed to account for the thin-film nanostructuring and fibre formation by considering basic phenomena such as stress relaxation, inhomogeneities in the plasma sheath electrical field and the local disturbance of the oxide growth.
Domain walls and instantons in N=1, d=4 supergravity
Huebscher, M; Ortin, T
2009-01-01
We study the supersymmetric sources of (multi-) domain-wall and (multi-) instanton solutions of generic N=1, d=4 supergravities, that is: the worldvolume effective actions for said supersymmetric topological defects. The domain-wall solutions naturally couple to the two 3-forms recently found as part of the N=1, d=4 tensor hierarchy (i.e. they have two charges in general) and their tension is the absolute value of the superpotential section L. The introduction of sources (we study sources with finite and vanishing thickness) is equivalent to the introduction of local coupling constants and results in dramatic changes of the solutions. Our results call for a democratic reformulation of N=1,d=4 supergravity in which coupling constants are, off-shell, scalar fields. The effective actions for the instantons are always proportional to the coordinate orthogonal to the twist-free embedding of the null-geodesic (in the Wick-rotated scalar manifold) describing the instanton. We show their supersymmetry and find the as...
Examining Prebiotic Chemistry Using O(^1D) Insertion Reactions
Hays, Brian M.; Laas, Jacob C.; Weaver, Susanna L. Widicus
2013-06-01
Aminomethanol, methanediol, and methoxymethanol are all prebiotic molecules expected to form via photo-driven grain surface chemistry in the interstellar medium (ISM). These molecules are expected to be precursors for larger, biologically-relevant molecules in the ISM such as sugars and amino acids. These three molecules have not yet been detected in the ISM because of the lack of available rotational spectra. A high resolution (sub)millimeter spectrometer coupled to a molecular source is being used to study these molecules using O(^1D) insertion reactions. The O(^1D) chemistry is initiated using an excimer laser, and the products of the insertion reactions are adiabatically cooled using a supersonic expansion. Experimental parameters are being optimized by examination of methanol formed from O(^1D) insertion into methane. Theoretical studies of the structure and reaction energies for aminomethanol, methanediol, and methoxymethanol have been conducted to guide the laboratory studies once the methanol experiment has been optimized. The results of the calculations and initial experimental results will be presented.
International Nuclear Information System (INIS)
This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics
Energy Technology Data Exchange (ETDEWEB)
White, J.E.; Wright, R.Q.; Roussin, R.W.; Ingersoll, D.T.
1992-11-01
This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics.
Lozano Montero, Juan Andrés; Jiménez Escalante, Javier; García Herranz, Nuria; Aragonés Beltrán, José María
2010-01-01
In this paper the extension of the multigroup nodal diffusion code ANDES, based on the Analytic Coarse Mesh Finite Difference (ACMFD) method, from Cartesian to hexagonal geometry is presented, as well as its coupling with the thermal–hydraulic (TH) code COBRA-IIIc for hexagonal core analysis. In extending the ACMFD method to hexagonal assemblies, triangular-Z nodes are used. In the radial plane, a direct transverse integration procedure is applied along the three directions that are orthog...
International Nuclear Information System (INIS)
MOL1D is a FORTRAN subroutine package for the method of lines solution for systems of initial-boundary-value partial differential equations in one space dimension. Using the package, a programer with limited experience in numerical analysis can accurately solve linear and nonlinear hyperbolic equations with or without discontinuities, linear and nonlinear parabolic equations (including those arising in reaction-diffusion equations), and elliptic boundary-value problems when posed as the stable time-independent solution of a parabolic equation. Systems are handled as easily as single equations, and a wide variety of boundary conditions can be accommodated, including most that arise in applications. The major advantage of the package is that initial-value problems can be solved accurately with a minimum of programing effort and with moderate computer cost. 4 figures, 1 table
Initial Stage of the Microwave Ionization Wave Within a 1D Model
Semenov, V. E.; Rakova, E. I.; Glyavin, M. Yu.; Nusinovich, G. S.
2016-06-01
The dynamics of the microwave breakdown in a gas is simulated numerically within a simple 1D model which takes into account such processes as the impact ionization of gas molecules, the attachment of electrons to neutral molecules, and plasma diffusion. Calculations are carried out for different spatial distributions of seed electrons with account for reflection of the incident electromagnetic wave from the plasma. The results reveal considerable dependence of the ionization wave evolution on the relation between the field frequency and gas pressure, as well as on the existence of extended rarefied halo of seed electrons. At relatively low gas pressures (or high field frequencies), the breakdown process is accompanied by the stationary ionization wave moving towards the incident electromagnetic wave. In the case of a high gas pressure (or a relatively low field frequency), the peculiarities of the breakdown are associated with the formation of repetitive jumps of the ionization front.
Gildenburg, V B
2016-01-01
The initial stage of the small-scale ionization-induced instability developing inside the fused silica volume exposed to the femtosecond laser pulse is studied as a possible initial cause of the self-organized nanograting formation. We have calculated the spatial spectra of the instability with the electron-hole diffusion taken into account for the first time and have found that it results in the formation of some hybrid (diffusion-wave) 1D structure with the spatial period determined as geometrical mean of the laser wavelength and characteristic diffusion length of the process considered. Near the threshold of the instability this period occurs to be approximately equal to the laser half-wavelength in the silica, close to the one experimentally observed.
1D Measurement of Sodium Ion Flow in Hydrogel After a Bath Concentration Jump.
Roos, R W; Pel, L; Huinink, H P; Huyghe, J M
2015-07-01
NMR is used to measure sodium flow driven by a 1D concentration gradient inside poly-acrylamid (pAA) hydrogel. A sodium concentration jump from 0.5 M NaCl to 0 M NaCl is applied at the bottom of a cylindrical pAA sample. The sodium level and hydrogen level are measured as a function of time and position inside the sample for 5 days. Then a reversed step is applied, and ion flow is measured for another 5 days. During the measurement, the cylindrical sample is radially confined and allowed to swell in the axial direction. At the same time, sodium and moisture in the sample are measured on a 1D spatial grid in the axial direction. A quadriphasic mixture model (Huyghe and Janssen in Int J Eng Sci 35:793, 1997) is used to simulate the results and estimate the diffusion coefficient of sodium and chloride. The best fit results were obtained for D[Formula: see text] cm(2)/s and D[Formula: see text] cm(2)/s, at 25 degrees centigrade. Different time constants were observed for swelling and deswelling. PMID:25786888
International Nuclear Information System (INIS)
We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)
International Nuclear Information System (INIS)
As a result of the IAEA Co-ordinated Research Programme entitled 'Final Stage of the WIMS Library Update Project', new and updated WIMS-D libraries based upon ENDF/B-VI.5, JENDL-3.2 and JEF-2.2 have become available. A project to prepare an exhaustive handbook of WIMS-D cross sections from old and new libraries has been taken up by the authors. As part of this project, we have developed a computer program XnWlup with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualization of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. The current features of the software, on-line help manual and future plans for further development are described in this paper
Munayer, Salim J; Horenczyk, Gabriel
2014-10-01
Grounded in a contextual approach to acculturation of minorities, this study examines changes in acculturation orientations among Palestinian Christian Arab adolescents in Israel following the "lost decade of Arab-Jewish coexistence." Multi-group acculturation orientations among 237 respondents were assessed vis-à-vis two majorities--Muslim Arabs and Israeli Jews--and compared to 1998 data. Separation was the strongest endorsed orientation towards both majority groups. Comparisons with the 1998 data also show a weakening of the Integration attitude towards Israeli Jews, and also distancing from Muslim Arabs. For the examination of the "Westernisation" hypothesis, multi-dimensional scaling (MDS) analyses of perceptions of Self and group values clearly showed that, after 10 years, Palestinian Christian Arabs perceive Israeli Jewish culture as less close to Western culture, and that Self and the Christian Arab group have become much closer, suggesting an increasing identification of Palestinian Christian Arab adolescents with their ethnoreligious culture. We discuss the value of a multi-group, multi-method, and multi-wave approach to the examination of the role of the political context in acculturation processes. PMID:25178958
International Nuclear Information System (INIS)
Highlights: • Multi-group formulation for exact neutron elastic scattering kernel is developed. • Up-scattering effects are incorporated in the cross-section data for heavy nuclei. • Effects on Doppler Temperature Coefficient (DTC) are demonstrated using DRAGON. • Results show an increase in DTC values by almost 10% for UOX and MOX LWR fuels. - Abstract: A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects stemming from lattice atoms thermal motion and it accounts for them within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to −10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes research performed to date on this topic
Computational Study and Analysis of Structural Imperfections in 1D and 2D Photonic Crystals
Energy Technology Data Exchange (ETDEWEB)
K.R. Maskaly
2005-06-01
Dielectric reflectors that are periodic in one or two dimensions, also known as 1D and 2D photonic crystals, have been widely studied for many potential applications due to the presence of wavelength-tunable photonic bandgaps. However, the unique optical behavior of photonic crystals is based on theoretical models of perfect analogues. Little is known about the practical effects of dielectric imperfections on their technologically useful optical properties. In order to address this issue, a finite-difference time-domain (FDTD) code is employed to study the effect of three specific dielectric imperfections in 1D and 2D photonic crystals. The first imperfection investigated is dielectric interfacial roughness in quarter-wave tuned 1D photonic crystals at normal incidence. This study reveals that the reflectivity of some roughened photonic crystal configurations can change up to 50% at the center of the bandgap for RMS roughness values around 20% of the characteristic periodicity of the crystal. However, this reflectivity change can be mitigated by increasing the index contrast and/or the number of bilayers in the crystal. In order to explain these results, the homogenization approximation, which is usually applied to single rough surfaces, is applied to the quarter-wave stacks. The results of the homogenization approximation match the FDTD results extremely well, suggesting that the main role of the roughness features is to grade the refractive index profile of the interfaces in the photonic crystal rather than diffusely scatter the incoming light. This result also implies that the amount of incoherent reflection from the roughened quarterwave stacks is extremely small. This is confirmed through direct extraction of the amount of incoherent power from the FDTD calculations. Further FDTD studies are done on the entire normal incidence bandgap of roughened 1D photonic crystals. These results reveal a narrowing and red-shifting of the normal incidence bandgap with
TPHEX, MultiGroup Neutron Flux in Homogeneous Hexagonal LWR Cells
International Nuclear Information System (INIS)
1 - Description of program or function: This program is intended to calculate the multigroup neutron flux distribution in an assembly of homogenized hexagonal cells using a transmission probability (interface current) method. It is primarily intended for calculations on hexagonal LWR fuel assemblies, with each cell corresponding to a pin cell, but can be used for other purposes, although its accuracy in other applications must be established separately. The flux at each cell interface is divided azimuthally into 60-degree sectors, with two components (an incomplete P1 expansion) in each sector. The interface fluxes are connected by transmission of un-collided neutrons through the cell. AN isotropic source (from fission or scattering) within the cell with a parabolic spatial distribution also contributes. The boundary conditions may correspond to full reflection at the mid-planes of the peripheral cells or (approximately) to a diagonal albedo matrix. Periodic boundary conditions can easily be implemented. If the peripheral cells are not regular hexagons, an edge transport correction may be applied to decrease the error from treating them as regular. 2 - Method of solution: The flux in one group is solved in an inner iteration, which may be accelerated by successive over-relaxation and, optionally, renormalization. The fluxes in different groups, connected through scattering and fission, are solved by outer iteration. The coefficients needed by the program (transmission coefficients etc.) are interpolated from pre-calculated values stored in a file. 3 - Restrictions on the complexity of the problem: The optical thickness of the cells must be in the range from 0.1 to 5. These limits can be expanded if the coefficient file is recalculated, but the accuracy is best when the optical thickness is not too near the ends of this range. Variable dimensioning is used, so there are no fixed limits on the number of cells or groups. However, since 48 variables are needed to
MGAV10: the latest evolution in the multi-group analysis code - 16248
International Nuclear Information System (INIS)
At many points in the safe and transparent handling of plutonium materials the relative isotopic composition of the principle isotopes needs to be known. Sometimes this information may be of primary interest - such as in the verification of safeguard declarations or in the confirmation of the reactivity of mixed oxide fuel. At other times, e.g., for radioactive waste characterization, the isotopic composition may be needed to calculate specific thermal power or specific spontaneous fission rates for the item under study, which can subsequently be combined with calorimetric and correlated neutron counting measurements, respectively, in order to make quantitative assessments of the mass of Pu and associated nuclides that are present in an item. The Multi-Group Analysis code MGA is a highly regarded and widely used computer code for the analysis of high resolution gamma ray spectra in order to extract the relative isotopic composition of plutonium for a diversity of items with minimal prior information. It has been honed over many years to give reliable results for a broad range of measurement scenarios commonly encountered in the fuel cycle. The nuclear industry is not dormant however and the demands on such codes continue to shift as a combination of technology and necessity open up new application areas. For example, while MGA had its origins in the analysis of clean spectra on product material principally for nuclear safeguards applications taken with germanium detectors having good low-energy resolution, it is now widely applied to the characterization of drummed waste forms and the complex spectra from such items acquired with much larger volume and poorer resolution detectors often used in such applications for the dual use of quantitative assay of the many gamma-emitters. This new domain of operational experience resulted in the need to enhance MGA to deal with spectra of poor statistical quality and also to cope with some of the complications that arise in
Extended-Range Ultrarefractive 1D Photonic Crystal Prisms
Ting, David Z.
2007-01-01
A proposal has been made to exploit the special wavelength-dispersive characteristics of devices of the type described in One-Dimensional Photonic Crystal Superprisms (NPO-30232) NASA Tech Briefs, Vol. 29, No. 4 (April 2005), page 10a. A photonic crystal is an optical component that has a periodic structure comprising two dielectric materials with high dielectric contrast (e.g., a semiconductor and air), with geometrical feature sizes comparable to or smaller than light wavelengths of interest. Experimental superprisms have been realized as photonic crystals having three-dimensional (3D) structures comprising regions of amorphous Si alternating with regions of SiO2, fabricated in a complex process that included sputtering. A photonic crystal of the type to be exploited according to the present proposal is said to be one-dimensional (1D) because its contrasting dielectric materials would be stacked in parallel planar layers; in other words, there would be spatial periodicity in one dimension only. The processes of designing and fabricating 1D photonic crystal superprisms would be simpler and, hence, would cost less than do those for 3D photonic crystal superprisms. As in 3D structures, 1D photonic crystals may be used in applications such as wavelength-division multiplexing. In the extended-range configuration, it is also suitable for spectrometry applications. As an engineered structure or artificially engineered material, a photonic crystal can exhibit optical properties not commonly found in natural substances. Prior research had revealed several classes of photonic crystal structures for which the propagation of electromagnetic radiation is forbidden in certain frequency ranges, denoted photonic bandgaps. It had also been found that in narrow frequency bands just outside the photonic bandgaps, the angular wavelength dispersion of electromagnetic waves propagating in photonic crystal superprisms is much stronger than is the angular wavelength dispersion obtained
BGK electron solitary waves: 1D and 3D
Directory of Open Access Journals (Sweden)
L.-J. Chen
2002-01-01
Full Text Available This paper presents new results for 1D BGK electron solitary wave (phase-space electron hole solutions and, based on the new results, extends the solutions to include the 3D electrical interaction (E ~ 1/r 2 of charged particles. Our approach for extending to 3D is to solve the nonlinear 3D Poisson and 1D Vlasov equations based on a key feature of 1D electron hole (EH solutions; the positive core of an EH is screened by electrons trapped inside the potential energy trough. This feature has not been considered in previous studies. We illustrate this key feature using an analytical model and argue that the feature is independent of any specific model. We then construct azimuthally symmetric EH solutions under conditions where electrons are highly field-aligned and ions form a uniform background along the magnetic field. Our results indicate that, for a single humped electric potential, the parallel cut of the perpendicular component of the electric field (E⊥ is unipolar and that of the parallel component (E|| bipolar, reproducing the multi-dimensional features of the solitary waves observed by the FAST satellite. Our analytical solutions presented in this article capture the 3D electric interaction and the observed features of (E|| and E⊥. The solutions predict a dependence of the parallel width-amplitude relation on the perpendicular size of EHs. This dependence can be used in conjunction with experimental data to yield an estimate of the typical perpendicular size of observed EHs; this provides important information on the perpendicular span of the source region as well as on how much electrostatic energy is transported by the solitary waves.
Calculation of control assembly effectiveness in a LMFBR by a transport-diffusion equivalence method
International Nuclear Information System (INIS)
A method allowing the calculation of the control assembly effectiveness in a fast neutron reactor is proposed. For each type of heterogeneous assembly - control or follower - a polar parameter, taking into account the assembly absorption and the axial leakage of neutrons inside the assembly, is defined; in a similar way a bipolar parameter, taking into account the reaction of the assembly to a transverse flux gradient, is also defined. These two parameters, deduced from transport theory, are used to determine the absorption cross-section and the diffusion coefficient of an equivalent homogeneous control or follower assembly. These new parameters are to be introduced in an one-group diffusion code calculating the reactor as a whole with any number of control and follower assemblies. An approximate generalization to multigroup theory is proposed. Numerical comparisons show that this equivalent diffusion method gives results which are much closer to transport results than those obtained by the classical diffusion theory
Coherent thermal conductance of 1-D photonic crystals
Tschikin, Maria; Ben-Abdallah, Philippe; Biehs, Svend-Age
2012-10-01
We present an exact calculation of coherent thermal conductance in 1-D multilayer photonic crystals using the S-matrix method. In particular, we study the thermal conductance in a bilayer structure of Si/vacuum or Al2O3/vacuum slabs by means of the exact radiative heat flux expression. Based on the results obtained for the Al2O3/vacuum structure we show by comparison with previous works that the material losses and (localized) surface modes supported by the inner layers play a fundamental role and cannot be omitted in the definition of thermal conductance. Our results could have significant implications in the conception of efficient thermal barriers.
Spatial coherence of polaritons in a 1D channel
Energy Technology Data Exchange (ETDEWEB)
Savenko, I. G., E-mail: savenko.j@mail.ru [Russian Academy of Sciences, Academic University, Research and Education Center of Nanotechnologies (Russian Federation); Iorsh, I. V. [National Research University of Information Technologies, Mechanics and Optics (Russian Federation); Kaliteevski, M. A. [Russian Academy of Sciences, Academic University, Research and Education Center of Nanotechnologies (Russian Federation); Shelykh, I. A. [University of Iceland, Science Institute (Iceland)
2013-01-15
We analyze time evolution of spatial coherence of a polariton ensemble in a quantum wire (1D channel) under constant uniform resonant pumping. Using the theoretical approach based on the Lindblad equation for a one-particle density matrix, which takes into account the polariton-phonon and excitonexciton interactions, we study the behavior of the first-order coherence function g{sup 1} for various pump intensities and temperatures in the range of 1-20 K. Bistability and hysteresis in the dependence of the first-order coherence function on the pump intensity is demonstrated.
A Godunov method for Lagrangian hydrodynamics in 1D
Energy Technology Data Exchange (ETDEWEB)
Crowley, W.P.
1987-01-15
For transient problems involving strong shocks, the artificial viscosity method has been the standard in numerical hydrodynamics for many years. An alternative approach was suggested by Godunov and it is gaining acceptance. We consider a Godunov method for 1D Lagrangian calculations and show that in the case of a strong shock moving through a nonuniform mesh the Godunov solution is superior to the artificial viscosity solution. For uniform mesh shock problems in spherical geometry the two methods give comparable results. 4 refs., 9 figs.
1D-transport properties of single superconducting lead nanowires
DEFF Research Database (Denmark)
Michotte, S.; Mátéfi-Tempfli, Stefan; Piraux, L.
2003-01-01
nanowire is small enough to ensure a 1D superconducting regime in a wide temperature range below T. The non-zero resistance in the superconducting state and its variation caused by fluctuations of the superconducting order parameter were measured versus temperature, magnetic field, and applied DC current......We report on the transport properties of single superconducting lead nanowires grown by an electrodeposition technique, embedded in a nanoporous track-etched polymer membrane. The nanowires are granular, have uniform diameter of ̃40 nm and a very large aspect ratio (̃500). The diameter of the...
Breakdown of 1D water wires inside Charged Carbon Nanotubes
Pant, Shashank
2016-01-01
Using Molecular Dynamics approach we investigated the structure, dynamics of water confined inside pristine and charged 6,6 carbon nanotubes (CNTs). This study reports the breakdown of 1D water wires and the emergence of triangular faced water on incorporating charges in 6,6 CNTs. Incorporation of charges results in high potential barriers to the flipping of water molecules due to the formation of a large number of hydrogen bonds. The PMF analyses show the presence of ~2 kcal/mol barrier for the movement of water inside pristine CNT and almost negligible barrier in charged CNTs.
Restrained Dark $U(1)_d$ at Low Energies
Correia, F C
2016-01-01
We investigate a spontaneously broken $U(1)_d$ gauge symmetry with a muon-specific dark Higgs. Our first goal is to verify how the presence of a new dark Higgs, $\\phi$, and a dark gauge boson, $V$, can simultaneously face the anomalies from the muon magnetic moment and the proton charge radius. Secondly, by assuming that $V$ must decay to an electron-positron pair, we explore the corresponding parameter space determined with the low energy constraints coming from $ K \\to \\mu X$, electron $(g-2)_e$, $K \\to \\mu \
Phthalocyanine based 1D nanowires for device applications
Saini, Rajan; Mahajan, Aman; Bedi, R. K.
2012-06-01
1D nanowires (NWs) of Cu (II) 1,4,8,11,15,18,22,25-octabutoxy-29H,31H-Phthalocyanine (CuPc(OBu)8) molecule have been grown on different substrates by cost effective solution processing technique. The density of NWs is found to be strongly dependent on the concentration of solution. The possible formation mechanism of these structures is π-π interaction between phthalocyanine molecules. The improved conductivity of these NWs as compared to spin coated film indicates their potential for molecular device applications.
1-D ELECTRO-OPTIC BEAM STEERING DEVICE
Wang, Wei-Chih; Tsui, Chi Leung
2011-01-01
In this paper, we present the design and fabrication of a 1D beam steering device based on planar electro-optic thermal-plastic prisms and a collimator lens array. With the elimination of moving parts, the proposed device is able to overcome the mechanical limitations of present scanning devices, such as fatigue and low operating frequency, while maintaining a small system footprint (~0.5mm×0.5mm). From experimental data, our prototype device is able to achieve a maximum deflection angle of 5...
Coherent thermal conductance of 1-D photonic crystals
International Nuclear Information System (INIS)
We present an exact calculation of coherent thermal conductance in 1-D multilayer photonic crystals using the S-matrix method. In particular, we study the thermal conductance in a bilayer structure of Si/vacuum or Al2O3/vacuum slabs by means of the exact radiative heat flux expression. Based on the results obtained for the Al2O3/vacuum structure we show by comparison with previous works that the material losses and (localized) surface modes supported by the inner layers play a fundamental role and cannot be omitted in the definition of thermal conductance. Our results could have significant implications in the conception of efficient thermal barriers.
A 1D analysis of two high order MOC methods
International Nuclear Information System (INIS)
The work presented here provides two different methods for evaluating angular fluxes along long characteristics. One is based off a projection of the 1D transport equation onto a complete set of Legendre polynomials, while the other uses the 1D integral transport equation to evaluate the angular flux values at specific points along each track passing through a cell. The Moment Long Characteristic (M-LC) method is shown to provide 2(P+1) spatial convergence and significant gains in accuracy with the addition of only a few spatial degrees of freedom. The M-LC method, though, is shown to be ill-conditioned at very high order and for optically thin geometries. The Point Long Characteristic (P-LC) method, while less accurate, significantly improves stability to problems with optically thin cells. The P-LC method is also more flexible, allowing for extra angular flux evaluations along a given track to give a more accurate representation of the shape along each track. This is at the expense of increasing the degrees of freedom of the system, though, and requires an increase in memory storage. This work concludes that both may be used simultaneously within the same geometry to provide the best mix of accuracy and stability possible. (authors)
1-D DCT Using Latency Efficient Floating Point Algorithms
Directory of Open Access Journals (Sweden)
Viswanath Gowd A, Yedukondala Rao V, T. Shanmuganantham
2013-04-01
Full Text Available This paper presents the design of one-dimensional discrete cosine transform (DCT architecture for digital signal processing (DSP applications. DCT is a basic transformation for coding method which converts spatial domain to frequency domain of image. In 1-D DCT operation addition, subtraction, multiplication operations are required. These operations must be accurate, less latency. Floating point operations have dynamic range of representation, more accurate and perform millions of calculations per second. So the floating point operations are used for the above operations. In this floating point adder/subtractor is the most complex operation in a floating-point arithmetic and consists of many variable latency- and area dependent sub-operations. In floating-point addition implementations, latency is the primary performance bottleneck. So different types of floating point adder/subtractor algorithms such as LOD, LOP, Two-path are used to decrease the latency. The trade off is observed in 1-D DCT by changing different types of adders in place of summer. All architectures are designed and implemented using VHDL using Xillinx 13.1software.
Study of 1D Strange Charmed Meson Family Using HQET
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Pallavi Gupta
2016-01-01
Full Text Available Recently LHCb predicted spin 1 and spin 3 states Ds1⁎(2860 and Ds3⁎(2860 which are studied through their strong decays and are assigned to fit the 13D1 and 13D3 states in the charm spectroscopy. In this paper, using the heavy quark effective theory, we state that assigning Ds1⁎(2860 as the mixing of 13D1-23S1 states is rather a better justification to its observed experimental values than a pure state. We study its decay modes variation with hadronic coupling constant gxh and the mixing angle θ. We appoint spin 3 state Ds3⁎(2860 as the missing 1D 3-JP state and also study its decay channel behavior with coupling constant gyh. To appreciate the above results, we check the variation of decay modes for their spin partners states, that is, 1D2 and 1D2′, with their masses and strong coupling constant, that is, gxh and gyh. Our calculation using HQET approach gives mixing angle of the 13D1-23S1 state for Ds1⁎(2860 to lie in the range (-1.6 radians ≤θ≤-1.2 radians. Our calculation for coupling constant values gives gxh to lie within value range of 0.17–0.20 and gyh to be 0.40. We expect from experiments to observe this mixing angle to verify our results.
Modeling atrazine transport in soil columns with HYDRUS-1D
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John Leju CELESTINO LADU
2011-09-01
Full Text Available Both physical and chemical processes affect the fate and transport of herbicides. It is useful to simulate these processes with computer programs to predict solute movement. Simulations were run with HYDRUS-1D to identify the sorption and degradation parameters of atrazine through calibration from the breakthrough curves (BTCs. Data from undisturbed and disturbed soil column experiments were compared and analyzed using the dual-porosity model. The study results show that the values of dispersivity are slightly lower in disturbed columns, suggesting that the more heterogeneous the structure is, the higher the dispersivity. Sorption parameters also show slight variability, which is attributed to the differences in soil properties, experimental conditions and methods, or other ecological factors. For both of the columns, the degradation rates were similar. Potassium bromide was used as a conservative non-reactive tracer to characterize the water movement in columns. Atrazine BTCs exhibited significant tailing and asymmetry, indicating non-equilibrium sorption during solute transport. The dual-porosity model was verified to best fit the BTCs of the column experiments. Greater or lesser concentration of atrazine spreading to the bottom of the columns indicated risk of groundwater contamination. Overall, HYDRUS-1D successfully simulated the atrazine transport in soil columns.
The Cosmological Mass Function with 1D Gravity
Monaco, P; Monaco, Pierluigi; Murante, Giuseppe
1999-01-01
The cosmological mass function problem is analyzed in full detail in the case of 1D gravity, with analytical, semi-analytical and numerical techniques. The extended Press & Schechter theory is improved by detailing the relation between smoothing radius and mass of the objects. This is done by introducing in the formalism the concept of a growth curve for the objects. The predictions of the extended Press & Schechter theory are compared to large N-body simulations of flat expanding 1D universes with scale-free power spectra of primordial perturbations. The collapsed objects in the simulations are located with a clump-finding algorithm designed to find regions that have undergone orbit crossing or that are in the multi-stream regime (these are different as an effect of the finite size of the multi-stream regions). It is found that the semi-analytical mass function theory, which has no free parameters, is able to recover the properties of collapsed objects both statistically and object by object. In part...
Hamming Distance and Data Compression of 1-D CA
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Raied Salman
2013-05-01
Full Text Available In this paper an application of von Neumann correct ion technique to the output string of some chaotic rules of 1-D Cellular Automata that are uns uitable for cryptographic pseudo random number generation due to their non uniform distribu tion of the binary elements is presented. The one dimensional (1-D Cellular Automata (CA Ru le space will be classified by the time run of Hamming Distance (HD. This has the advantage of determining the rules that have short cycle lengths and therefore deemed to be unsuitable for cryptographic pseudo random number generation. The data collected from evolution of ch aotic rules that have long cycles are subjected to the original von Neumann density corre ction scheme as well as a new generalized scheme presented in this paper and tested for stati stical testing fitness using Diehard battery of tests. Results show that significant improvement in the statistical tests are obtained when the output of a balanced chaotic rule are mutually excl usive ORed with the output of unbalanced chaotic rule that have undergone von Neumann densit y correction.
Institute of Scientific and Technical Information of China (English)
CHEN Chao; YANG Yu-lin; LI Wei-sheng; LIU Yun-ling; YI Zhuo; GUO Yang-hong; PANG Wen-qin
2005-01-01
The transformation of titanium phosphate from 1-D chiral- chain(JTP-A) to 2-D layer(TP-J1) has been carefully investigated. Through a hydrolysis-condensation self-assembly pathway, the crystals of TP-J1 can be obtained from the JTP-A phase under hydrothermal conditions. An intermediate material with zigzag chain during the transformation was observed by XRD characterization. A hypothesis of the transformation mechanism is also described in this article. It is noteworthy that ethylenediamine plays an important role in the transformation.
International Nuclear Information System (INIS)
Highlights: • Fixed-source SN transport problems. • Energy multigroup model. • Anisotropic scattering. • Slab-geometry spectral nodal method. - Abstract: A generalization of the spectral Green’s function (SGF) method is developed for multigroup, fixed-source, slab-geometry discrete ordinates (SN) problems with anisotropic scattering. The offered SGF method with the one-node block inversion (NBI) iterative scheme converges numerical solutions that are completely free from spatial truncation errors for multigroup, slab-geometry SN problems with scattering anisotropy of order L, provided L < N. As a coarse-mesh numerical method, the SGF method generates numerical solutions that generally do not give detailed information on the problem solution profile, as the grid points can be located considerably away from each other. Therefore, we describe in this paper a technique for the spatial reconstruction of the coarse-mesh solution generated by the multigroup SGF method. Numerical results are given to illustrate the method’s accuracy
Sideridis, Georgios D.; Tsaousis, Ioannis; Al-harbi, Khaleel A.
2015-01-01
The purpose of the present study was to extend the model of measurement invariance by simultaneously estimating invariance across multiple populations in the dichotomous instrument case using multi-group confirmatory factor analytic and multiple indicator multiple causes (MIMIC) methodologies. Using the Arabic version of the General Aptitude Test…
Simplified 1D modelling of the HGA test
International Nuclear Information System (INIS)
Document available in extended abstract form only. The HGA test is located in the Mont Terri Rock Laboratory (Switzerland). It consists of a horizontal borehole of 1.00 m of diameter and 13.00 m of length excavated in the ultra-low permeable Opalinus clay. During the tunnel drilling, the Opalinus clay near the tunnel wall was damaged, giving rise to an EDZ (Excavation Damaged Zone) around the tunnel. A steel liner was placed along the 6.00 m close to the tunnel mouth in order to guarantee the stability. The last 4.00 m at the tunnel end were backfilled with gravel. Along the remaining 3.00 m, an inflatable rubber packer of 1.00 m in diameter, was installed and inflated, thereby compressing the EDZ that was created during the tunnel excavation. The test section was filled with de-aired water and care was taken in order to eliminate the air from this tunnel section. Subsequently, a series of water and gas injection tests were carried out with varying mega-packer pressure, whereby water or gas was injected into the test section and, due to the very low permeability of the intact Opalinus clay, forced to flow back along the EDZ. In order to model the water and gas flow through the EDZ, we have followed a two-track approach. On the one hand, a 2D axisymmetric numerical model using code-bright has been made. On the other hand, a 1D analytical-numerical model has been developed and implemented in an Excel spreadsheet, whereby the field equations defined on a 1D geometrical domain are numerically solved using the finite element method. The 1D model has been used in order to calibrate the 2D axisymmetric model. Both the Opalinus clay and the EDZ will be considered to be porous media, with an incompressible solid phase (clay), an incompressible liquid phase (water and air) and a gas phase (water and air). The properties of the liquid phase will be assumed to be independent of the concentration of dissolved air and the gas phase will be assumed to be a mixture of dry air and
International Nuclear Information System (INIS)
Single-zone modelling is used to assess different collections of impeller 1D loss models. Three collections of loss models have been identified in literature, and the background to each of these collections is discussed. Each collection is evaluated using three modern automotive turbocharger style centrifugal compressors; comparisons of performance for each of the collections are made. An empirical data set taken from standard hot gas stand tests for each turbocharger is used as a baseline for comparison. Compressor range is predicted in this study; impeller diffusion ratio is shown to be a useful method of predicting compressor surge in 1D, and choke is predicted using basic compressible flow theory. The compressor designer can use this as a guide to identify the most compatible collection of losses for turbocharger compressor design applications. The analysis indicates the most appropriate collection for the design of automotive turbocharger centrifugal compressors
Harley, P.; Spence, S.; Early, J.; Filsinger, D.; Dietrich, M.
2013-12-01
Single-zone modelling is used to assess different collections of impeller 1D loss models. Three collections of loss models have been identified in literature, and the background to each of these collections is discussed. Each collection is evaluated using three modern automotive turbocharger style centrifugal compressors; comparisons of performance for each of the collections are made. An empirical data set taken from standard hot gas stand tests for each turbocharger is used as a baseline for comparison. Compressor range is predicted in this study; impeller diffusion ratio is shown to be a useful method of predicting compressor surge in 1D, and choke is predicted using basic compressible flow theory. The compressor designer can use this as a guide to identify the most compatible collection of losses for turbocharger compressor design applications. The analysis indicates the most appropriate collection for the design of automotive turbocharger centrifugal compressors.
A solution of the neutron diffusion equation for a hemisphere containing a uniform source
International Nuclear Information System (INIS)
An analytic solution of the diffusion equation for a hemisphere of fissile or non-fissile material is presented which contains a spatially uniform neutron source. Numerical results are given for the flux distribution for one-speed fast neutrons in 235U and also for a non-fissile element of similar scattering properties. We use these results to check the accuracy of the finite element code EVENT. The procedure is also developed for multigroup calculations. In an Appendix we outline the procedure required when the hemisphere contains a source and is also irradiated by an external current of neutrons
Fleury, Leesa M.; Moore, Guy D.
2016-05-01
If the axion exists and if the initial axion field value is uncorrelated at causally disconnected points, then it should be possible to predict the efficiency of cosmological axion production, relating the axionic dark matter density to the axion mass. The main obstacle to making this prediction is correctly treating the axion string cores. We develop a new algorithm for treating the axionic string cores correctly in 2+1 dimensions. When the axionic string cores are given their full physical string tension, axion production is about twice as efficient as in previous simulations. We argue that the string network in 2+1 dimensions should behave very differently than in 3+1 dimensions, so this result cannot be simply carried over to the physical case. We outline how to extend our method to 3+1D axion string dynamics.
Fleury, Leesa M
2016-01-01
If the axion exists and if the initial axion field value is uncorrelated at causally disconnected points, then it should be possible to predict the efficiency of cosmological axion production, relating the axionic dark matter density to the axion mass. The main obstacle to making this prediction is correctly treating the axion string cores. We develop a new algorithm for treating the axionic string cores correctly in 2+1 dimensions. When the axionic string cores are given their full physical string tension, axion production is about twice as efficient as in previous simulations. We argue that the string network in 2+1 dimensions should behave very differently than in 3+1 dimensions, so this result cannot be simply carried over to the physical case. We outline how to extend our method to 3+1D axion string dynamics.
Slug modeling with 1D two-fluid model
International Nuclear Information System (INIS)
Simulations of condensation-induced water hammer with one-dimensional two-fluid model requires explicit modeling of slug formation, slug propagation, and in some cases slug decay. Stratified flow correlations that are more or less well known in 1D two-fluid models, are crucial for accurate description of the initial phase of the slug formation and slug propagation. Slug formation means transition to other flow regime that requires different set of correlations. To use such two-fluid model for condensation induced water hammer simulations, a single slug must be explicitly recognized and captured. In the present work two cases of condensation-induced water hammer simulations performed with WAHA code, are described and discussed: injection of cold liquid into horizontal pipe filled with steam and injection of hot steam into horizontal pipe partially filled with cold liquid. (author)
1D PIC simulation of relativistic Buneman instability
International Nuclear Information System (INIS)
Buneman instability in the relativistic regime has been studied using a 1D electrostatic particle-in-cell code. In the non-relativistic case, Hirose et al. (Plasma Phys. 20, 481(1978)) has shown that breakdown of linear growth (saturation) occurs when |E|2/16πW0 ∼ ζomax, where W0 is the initial beam kinetic energy density and ζomax is maximum growth rate of the instability. In the weakly relativistic case, it has been confirmed using PIC simulation that scaling of saturation of Buneman instability follows a similar behavior as the non-relativistic case, whereas in the strongly relativistic case our simulation results show significant deviation from Hirose's results. In the strongly relativistic case, growth rate reduces due to relativistic corrections; so saturation occurs at a lower value compared to the non-relativistic/weakly relativistic case. (author)
Assessment of the 2D/1D implementation in MPACT
International Nuclear Information System (INIS)
The 2D/1D method is used in the MPACT code to obtain 3D solutions of the Boltzmann transport equation for practical reactor geometries. The OECD C5G7 transport benchmark problem is used first to assess the accuracy of the method with a fixed set of cross-sections. The VERA Core Physics Progression Problems are then used to compare the accuracy of the transport solver using a 56-group library based on ENDFB-VII.0. Single assembly PWR designs are simulated, and the eigenvalue and pin powers are compared to continuous-energy Monte Carlo results. A 3x3 assembly cluster with a control rod inserted into the center assembly is then compared to Monte Carlo to assess the ability of MPACT to predict a control rod worth curve. Finally, MPACT is used to simulate the initial critical states of a full 3D initial core of a PWR at zero power conditions. (author)
1-D Molecular Chains of Thiophene on Ge(100)
Jeon, Seok Min; Jung, Soon Jung; Kim, Hyeong-Do; Lim, Do Kyung; Lee, Hangil; Kim, Sehun
2007-01-01
The adsorption geometry of thiophene on Ge(100) have been studied by high-resolution core-level photoemission spectroscopy (HRPES) using synchrotron radiation and scanning tunneling microscopy (STM). From the analysis of the Ge 3d, S 2p, and C 1s core-level photoemission spectra, we found three different adsorption geometries, which were assigned to a dative bonding feature, a [4+2] cycloaddition reaction product, and a desulfurization reaction product. Furthermore, we investigated that the ratio of the components induced by three adsorption geometries changed depending on the molecular coverage and the annealing temperature. At low coverage, the kinetically favorable dative bonding features favorably form 1-D molecular chains. Increasing the molecular coverage, the energetically more stable [4+2] cycloaddition reaction products are additionally created.
Microlens Masses from 1-D Parallaxes and Heliocentric Proper Motions
Gould, Andrew
2014-01-01
One-dimensional (1-D) microlens parallaxes can be combined with heliocentric lens-source relative proper motion measurements to derive the lens mass and distance, as suggested by Ghosh et al. (2004). Here I present the first mathematical anlysis of this procedure, which I show can be represented as a quadratic equation. Hence, it is formally subject to a two-fold degeneracy. I show that this degeneracy can be broken in many cases using the relatively crude 2-D parallax information that is often available for microlensing events. I also develop an explicit formula for the region of parameter space where it is more difficult to break this degeneracy. Although no mass/distance measurements have yet been made using this technique, it is likely to become quite common over the next decade.
A 1-D morphodynamic model of postglacial valley incision
Tunnicliffe, Jon F.; Church, Michael
2015-11-01
Chilliwack River is typical of many Cordilleran valley river systems that have undergone dramatic Holocene degradation of valley fills that built up over the course of Pleistocene glaciation. Downstream controls on base level, mainly blockage of valleys by glaciers, led to aggradation of significant glaciofluvial and glaciolacustrine valley fills and fan deposits, subsequently incised by fluvial action. Models of such large-scale, long-term degradation present a number of important challenges since the evolution of model parameters, such as the rate of bedload transport and grain size characteristics, are governed by the nature of the deposit. Sediment sampling in the Chilliwack Valley reveals a complex sequence of very coarse to fine textural modes. We present a 1-D numerical morphodynamic model for the river-floodplain system tailored to conditions in the valley. The model is adapted to dynamically adjust channel width to optimize sediment transporting capacity and to integrate relict valley fill material as the channel incises through valley deposits. Sensitivity to model parameters is studied using four principal criteria: profile concavity, rate of downstream grain size fining, bed surface sand content, and the timescale to equilibrium. Model results indicate that rates of abrasion and coarsening of the grain size distributions exert the strongest controls on all of the interrelated model performance criteria. While there are a number of difficulties in satisfying all model criteria simultaneously, results indicate that 1-D models of valley bottom sedimentary systems can provide a suitable framework for integrating results from sediment budget studies and chronologies of sediment evacuation established from dating.
Directory of Open Access Journals (Sweden)
Tongfeng Zhang
2016-01-01
Full Text Available A one-dimensional (1D hybrid chaotic system is constructed by three different 1D chaotic maps in parallel-then-cascade fashion. The proposed chaotic map has larger key space and exhibits better uniform distribution property in some parametric range compared with existing 1D chaotic map. Meanwhile, with the combination of compressive sensing (CS and Fibonacci-Lucas transform (FLT, a novel image compression and encryption scheme is proposed with the advantages of the 1D hybrid chaotic map. The whole encryption procedure includes compression by compressed sensing (CS, scrambling with FLT, and diffusion after linear scaling. Bernoulli measurement matrix in CS is generated by the proposed 1D hybrid chaotic map due to its excellent uniform distribution. To enhance the security and complexity, transform kernel of FLT varies in each permutation round according to the generated chaotic sequences. Further, the key streams used in the diffusion process depend on the chaotic map as well as plain image, which could resist chosen plaintext attack (CPA. Experimental results and security analyses demonstrate the validity of our scheme in terms of high security and robustness against noise attack and cropping attack.
Díez, C. J.; Cabellos, O.; Martínez, J. S.
2015-01-01
Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.
International Nuclear Information System (INIS)
A computational method for calculating multigroup self-shielded cross sections in heterogeneous media containing arbitrary mixtures of resonant isotopes is presented. The method accounts for resonance interference between immixed resonant nuclei as well as for spatial resonance interference between resonant isotopes in different geometrical locations. A general correction is used to generate an intermediary reaction-rate library for resonant isotopic mixtures from a single-isotope, standard preprocessed library. Reaction rates for the heterogeneous fine-structure equation are computed from the intermediary library by invoking an equivalence theorem either on a group basis or using Bell's factors defined on macrogroups. Results are presented for a homogeneous mixture of Uranium oxide as well as for a recycled-fuel PWR cell. A study of the radial dependence of self-shielding for a recycled mixture of Uranium-Plutonium oxide in a PWR cell and in a submoderated cell is also included
International Nuclear Information System (INIS)
In the present paper a generalization is performed of a procedure to solve multigroup spherical harmonics equations, which has originally been proposed and developed for one-dimensional systems in cylindrical or spherical geometry, and later extended for a special case of a two-dimensional system in r-z geometry. The expressions are derived for the axial and the radial dependence of the group values of the neutron flux moments, in the P-3 approximation of the spherical harmonics method, in a cylindrically symmetrical system with an arbitrary number of material regions in both r- and z-directions. In the special case of an axially homogeneous system, these expressions reduce to the relations derived previously. (author)
Tominaga, Nozomu; Blinnikov, Sergei I
2015-01-01
We develop a time-dependent multi-group multidimensional relativistic radiative transfer code, which is required to numerically investigate radiation from relativistic fluids involved in, e.g., gamma-ray bursts and active galactic nuclei. The code is based on the spherical harmonic discrete ordinate method (SHDOM) that evaluates a source function including anisotropic scattering in spherical harmonics and implicitly solves the static radiative transfer equation with a ray tracing in discrete ordinates. We implement treatments of time dependence, multi-frequency bins, Lorentz transformation, and elastic Thomson and inelastic Compton scattering to the publicly available SHDOM code. Our code adopts a mixed frame approach; the source function is evaluated in the comoving frame whereas the radiative transfer equation is solved in the laboratory frame. This implementation is validated with various test problems and comparisons with results of a relativistic Monte Carlo code. These validations confirm that the code ...
International Nuclear Information System (INIS)
A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs
International Nuclear Information System (INIS)
Subgroup 7 of the NEA NSC WP-IEC simulated interest in processing data from various international and regional evaluated cross-section libraries into the 174n, 42g VITAMIN-J multigroup energy for the purpose of intercomparison. Cooperation and participation came from numerous installations around the world. Most processing was done with the NJOY system, but some independent contributions were provided. At the WP-IEC meeting in June 1993, many contributions to the effort were described and the exercise proved to be useful from several aspects. It was decided to expand the role of the temporary subgroup into a long term subgroup to look at both format and processing problems. A summary of the progress of Subgroup 7 is provided and the objective and scope of the new entity, Subgroup B, is reported
International Nuclear Information System (INIS)
Highlights: • Code works based on Monte Carlo and escape probability methods. • Sensitivity of Dancoff factor to number of energy groups and type and arrangement of neighbor’s fuels is considered. • Sensitivity of Dancoff factor to control rod’s height is considered. • Dancoff factor high efficiency is achieved versus method sampling neutron flight direction from the fuel surface. • Sensitivity of K to Dancoff factor is considered. - Abstract: Evaluation of multigroup constants in reactor calculations depends on several parameters, the Dancoff factor amid them is used for calculation of the resonance integral as well as flux depression in the resonance region in the heterogeneous systems. This paper focuses on the computer program (MCDAN-3D) developed for calculation of the multigroup black and gray Dancoff factor in three dimensional geometry based on Monte Carlo and escape probability methods. The developed program is capable to calculate the Dancoff factor for an arbitrary arrangement of fuel rods with different cylindrical fuel dimensions and control rods with various lengths inserted in the reactor core. The initiative calculates the black and gray Dancoff factor versus generated neutron flux in cosine and constant shapes in axial fuel direction. The effects of clad and moderator are followed by studying of Dancoff factor’s sensitivity with variation of fuel arrangements and neutron’s energy group for CANDU37 and VVER1000 fuel assemblies. MCDAN-3D outcomes poses excellent agreement with the MCNPX code. The calculated Dancoff factors are then used for cell criticality calculations by the WIMS code
3D/1D Analysis of ICRF Antennas
Maggiora, Riccardo; Lancellotti, Vito; Vecchi, Giuseppe
2003-10-01
An innovative tool has been realized for the 3D/1D simulation of Ion Cyclotron Radio Frequency (ICRF), i.e. accounting for antennas in a realistic 3D geometry and with an accurate 1D plasma model. The approach to the problem is based on an integral-equation formulation for the self-consistent evaluation of the current distribution on the conductors. The environment has been subdivided in two coupled region: the plasma region and the vacuum region. The two problems are linked by means of a magnetic current (electric field) distribution on the aperture between the two regions. In the vacuum region all the calculations are executed in the spatial domain while in the plasma region an extraction in the spectral domain of some integrals is employed that permits to significantly reduce the integration support and to obtain a high numerical efficiency leading to the practical possibility of using a large number of sub-domain (rectangular or triangular) basis functions on each solid conductor of the system. The plasma enters the formalism of the plasma region via a surface impedance matrix; for this reason any plasma model can be used; at present the FELICE code has been adopted, that affords density and temperature profiles, and FLR effects. The source term directly models the TEM mode of the coax feeding the antenna and the current in the coax is determined self-consistently, giving the input impedance/admittance of the antenna itself. Calculation of field distributions (both magnetic and electric), useful for sheath considerations, is included. This tool has been implemented in a suite, called TOPICA, that is modular and applicable to ICRF antenna structures of arbitrary shape. This new simulation tool can assist during the detailed design phase and for this reason can be considered a "Virtual Prototyping Laboratory" (VPL). The TOPICA suite has been tested against assessed codes and against measurements and data of mock-ups and existing antennas. The VPL is being used in
Energy Technology Data Exchange (ETDEWEB)
Collins, John, E-mail: jcollins@wheatonma.edu [Department of Physics and Astronomy,Wheaton College, Norton, MA 02766 (United States); Geen, Megan [Department of Physics and Astronomy,Wheaton College, Norton, MA 02766 (United States); Bettinelli, M. [Dipartmento di Biotechnologie, Universita Delgi Studi di Verona, Verona (Italy); Di Bartolo, B. [Department of Physics, Boston College, Chestnut Hill, MA 02467 (United States)
2012-10-15
We report on the role of cross-relaxation in the decay of the {sup 1}D{sub 2} level of trivalent Pr in YPO{sub 4} in crystals with Pr concentrations of 0.1%, 1%, 2%, and 5%. We have found that the {sup 1}D{sub 2} level decay is purely radiative in the low-doped system. As the Pr concentration is increased, the {sup 1}D{sub 2} luminescence is quenched due to a cross-relaxation energy transfer between two Pr ions. The temporal behavior of the {sup 1}D{sub 2} luminescence following pulsed excitation has been monitored in each sample at temperatures between 30 K and 300 K, and all decay curves were fit to the Yokota-Tanimoto model. The decay times decrease as temperature increases, due to an increase in both the radiative rate and the energy transfer rate with temperature. There is little evidence of diffusion at any temperature, even in the more concentrated samples. We have also fit the decay curves using the LumiTrans computer simulation. A comparison of the fits to the decay curves of the two methods is presented. - Highlights: Black-Right-Pointing-Pointer We present data on the decay of the {sup 1}D{sub 2} level of Pr in YPO{sub 4} from 30-300 K. Black-Right-Pointing-Pointer We determine the {sup 1}D{sub 2} cross-relaxation rate throughout that temperature range. Black-Right-Pointing-Pointer Fits to the data indicate diffusion among the Pr ions is negligible. Black-Right-Pointing-Pointer Radiative efficiencies of the {sup 1}D{sub 2} level are determined.
Graphs on uniform points in [0,1]d
Appel, Martin J. B.; Russo, Ralph P.; Yang, King J.
1995-06-01
Statistical problems in pattern or structure recognition for a random multidimensional point set may be addressed by variations on the random graph model of Erdos and Renyui. The imposition of graph structure with a variable edge criterion on a large random point set allows a search for signature quantities or behavior under the given distributional hypothesis. The work is motivated by the question of how to make statistical inferences from sensed mine field data. This article describes recent results obtained in the following special cases. On independent random points U1,...,Un distributed uniformly on [0,1]d, a random graph Gn(x) is constructed in which two distinct such points are joined by an edge if the l(infinity )-distance between them is at most some prescribed value 0 graph are described. Almost-sure asymptotic rates of convergence/divergence are obtained for various quantities, including the maximum and minimum vertex degree of the random graph, its clique number, chromatic number, and independence number, as the number n of points becomes large and the edge distance x is allowed to vary with n. The connectivity distance cn, the smallest x such that Gn(x) is connected, and the largest nearest neighbor link dn, the smallest x such that Gn(x) has no vertices of degree zero, are asymptotic in ratio, as n becomes large, for d >= 2.
1D dynamic beam modulation: methods to counteract inertia effects
International Nuclear Information System (INIS)
Dynamic modulation can be affected by inaccuracies when the required acceleration is larger than the highest allowed by the mechanical characteristics of the whole apparatus. In this study, inertia effects have been investigated with regard to the single absorber 1D modulation, analysing primarily how the acceleration performed by the modulating system affects the realization of 'single absorber' fluence profiles and the type of correction which could be devised. The observed percentage deviations from desired modulation at the lowest fluence coordinate of single minimum fluence profiles, when no correction is applied, were almost negligible for 'easy' modulations of the incident fluence (i.e. slow gradients); deviations became increasingly relevant as the moving absorber executed steeper gradients (a 17.6% higher dose being delivered in the minimum position when a 0.2 modulation is required). By applying the proposed corrections, the single absorber performances were improved to a satisfactory level, with a maximum deviation from desired modulation in the minima within 1.6%. (author)
Nonclassical Particle Transport in 1-D Random Periodic Media
Vasques, Richard; Slaybaugh, Rachel N
2016-01-01
We investigate the accuracy of the recently proposed nonclassical transport equation. This equation contains an extra independent variable compared to the classical transport equation (the path-length $s$), and models particle transport taking place in homogenized random media in which a particle's distance-to-collision is not exponentially distributed. To solve the nonclassical equation one needs to know the $s$-dependent ensemble-averaged total cross section, $\\Sigma_t(\\mu,s)$, or its corresponding path-length distribution function, $p(\\mu,s)$. We consider a 1-D spatially periodic system consisting of alternating solid and void layers, randomly placed in the $x$-axis. We obtain an analytical expression for $p(\\mu,s)$ and use this result to compute the corresponding $\\Sigma_t(\\mu,s)$. Then, we proceed to numerically solve the nonclassical equation for different test problems in rod geometry; that is, particles can move only in the directions $\\mu=\\pm 1$. To assess the accuracy of these solutions, we produce ...
The molecular spin filter constructed from 1D organic chain
International Nuclear Information System (INIS)
We proposed a molecular spin filter, which is constructed from the 1D metallic organic chain (Fen+1(C6H4)n). The spin-polarized transport properties of the molecular spin filter are explored by combining density functional theory with nonequilibrium Green's function formalism. Theoretical results reveal that Fen+1(C6H4)n molecular chain exhibits robust spin filtering effect, and only the spin-down electrons can transmit through the molecular chain. At the given bias voltage window [−1 eV,1 eV], the calculated spin filter efficiency is close to 100% in the case of n≥3. We find that the effect of spin polarization origin from both Fen+1 and (C6H4)n. In addition, negative difference resistance behavior appears in Fen+1(C6H4)n molecular chain. The results can help us understand the spin transport properties of organic molecular chain. - Highlights: • Theoretical results reveal that Fen+1(C6H4)n molecular chain exhibits robust spin filtering effect. • The effect of spin polarization origin from both of Fen+1 and (C6H4)n. • Negative difference resistance behavior appears in Fen+1(C6H4)n molecular chain
Havlickova, E; Subba, F; Coster, D; Wischmeier, M; Fishpool, G
2013-01-01
A 1D code modelling SOL transport parallel to the magnetic field (SOLF1D) is benchmarked with 2D simulations of MAST-U SOL performed via the SOLPS code for two different collisionalities. Based on this comparison, SOLF1D is then used to model the effects of divertor leg stretching in 1D, in support of the planned Super-X divertor on MAST. The aim is to separate magnetic flux expansion from volumetric power losses due to recycling neutrals by stretching the divertor leg either vertically or radially.
Havlickova, E.; Fundamenski, W.; Subba, F.; Coster, D; Wischmeier, M; Fishpool, G.
2013-01-01
A 1D code modelling SOL transport parallel to the magnetic field (SOLF1D) is benchmarked with 2D simulations of MAST-U SOL performed via the SOLPS code for two different collisionalities. Based on this comparison, SOLF1D is then used to model the effects of divertor leg stretching in 1D, in support of the planned Super-X divertor on MAST. The aim is to separate magnetic flux expansion from volumetric power losses due to recycling neutrals by stretching the divertor leg either vertically or ra...
Indian Academy of Sciences (India)
PRITAM PATIL; GANESH GAIKWAD; D R PATIL; JITENDRA NAIK
2016-06-01
1-D ZnO nanorods and PPy/1-D ZnO nanocomposites were prepared by the surfactant-assisted precipitation and in situ polymerization method, respectively. The synthesized nanorods and nanocomposites were characterized by UV–Vis spectrophotometer, Fourier transform-infrared spectroscopy (FTIR), X-ray diffraction (XRD) and field emission scanning electron microscope (FE-SEM), which gave the evidence of 1-D ZnO nanorods, polymerization of pyrrole monomer and strong interaction between PPy and 1-D ZnO nanorods, respectively. Photocatalytic activity of 1-D ZnO nanorods was conducted by $3^3$ level full-factorial design to evaluate the effect of three independent process variables viz., dye concentration (crystal violet), catalyst concentration (1-D ZnO nanorods) and the reaction time on the preferred response: photodegradation efficiency (%). The PPy/1-D ZnO nanocompositeswere used for the sensing of NH$_3$, LPG, CO$_2$ and H$_2$S gases, respectively, at room temperature. It was observed that PPy/1-D ZnO nanocomposites with different 1-D ZnO nanorod weight ratios (15 and 25%) had better selectivity and sensitivity towards NH3 at room temperature.
Comparison of multigroup and few-group calculations of fast power reactor parameters
International Nuclear Information System (INIS)
The basic parameters of a fast breeder reactor in two-dimensional cylindrical geometry and in multi- and few-group diffusion approximation were calculated and compared. Two different types of reactor were considered, viz., homogeneous and heterogeneous. The results can serve as a quantitative aid for the choice of the proper number of groups for the calculations of various reactor parameters with required accuracy. (author)
Energy Technology Data Exchange (ETDEWEB)
Wilson, W.B.; England, T.R.; LaBauve, R.J.
1978-02-01
The ENDF/B-IV fission-product data file includes data describing 824 nuclides. Cross sections, given for 181 of these nuclides, were processed into 154 neutron energy groups. The production of the data file is described. The TOAFEW code, useful in collapsing the multigroup values to few-group cross sections, is presented with instructions and examples of its use. The file of multigroup cross sections is available on request. 3 figures, 11 tables.
Energy Technology Data Exchange (ETDEWEB)
Kostorz, G. [Eidgenoessische Technische Hochschule, Angewandte Physik, Zurich (Switzerland)
1996-12-31
While Bragg scattering is characteristic for the average structure of crystals, static local deviations from the average lattice lead to diffuse elastic scattering around and between Bragg peaks. This scattering thus contains information on the occupation of lattice sites by different atomic species and on static local displacements, even in a macroscopically homogeneous crystalline sample. The various diffuse scattering effects, including those around the incident beam (small-angle scattering), are introduced and illustrated by typical results obtained for some Ni alloys. (author) 7 figs., 41 refs.
International Nuclear Information System (INIS)
While Bragg scattering is characteristic for the average structure of crystals, static local deviations from the average lattice lead to diffuse elastic scattering around and between Bragg peaks. This scattering thus contains information on the occupation of lattice sites by different atomic species and on static local displacements, even in a macroscopically homogeneous crystalline sample. The various diffuse scattering effects, including those around the incident beam (small-angle scattering), are introduced and illustrated by typical results obtained for some Ni alloys. (author) 7 figs., 41 refs
Evidence against dopamine D1/D2 receptor heteromers
Frederick, Aliya L.; Yano, Hideaki; Trifilieff, Pierre; Vishwasrao, Harshad D.; Biezonski, Dominik; Mészáros, József; Sibley, David R.; Kellendonk, Christoph; Sonntag, Kai C.; Graham, Devon L.; Colbran, Roger J.; Stanwood, Gregg D.; Javitch, Jonathan A.
2014-01-01
Hetero-oligomers of G-protein-coupled receptors have become the subject of intense investigation because their purported potential to manifest signaling and pharmacological properties that differ from the component receptors makes them highly attractive for the development of more selective pharmacological treatments. In particular, dopamine D1 and D2 receptors have been proposed to form hetero-oligomers that couple to Gαq proteins, and SKF83959 has been proposed to act as a biased agonist that selectively engages these receptor complexes to activate Gαq and thus phospholipase C. D1/D2 heteromers have been proposed as relevant to the pathophysiology and treatment of depression and schizophrenia. We used in vitro bioluminescence resonance energy transfer (BRET), ex vivo analyses of receptor localization and proximity in brain slices, and behavioral assays in mice to characterize signaling from these putative dimers/oligomers. We were unable to detect Gαq or Gα11 protein coupling to homomers or heteromers of D1 or D2 receptors using a variety of biosensors. SKF83959-induced locomotor and grooming behaviors were eliminated in D1 receptor knockout mice, verifying a key role for D1-like receptor activation. In contrast, SKF83959-induced motor responses were intact in D2 receptor and Gαq knockout mice, as well as in knock-in mice expressing a mutant Ala286-CaMKIIα, that cannot autophosphorylate to become active. Moreover, we found that in the shell of the nucleus accumbens, even in neurons in which D1 and D2 receptor promoters are both active, the receptor proteins are segregated and do not form complexes. These data are not compatible with SKF83959 signaling through Gαq or through a D1–D2 heteromer and challenge the existence of such a signaling complex in the adult animals that we used for our studies. PMID:25560761
A new general 1-D vadose zone flow solution method
Ogden, Fred L.; Lai, Wencong; Steinke, Robert C.; Zhu, Jianting; Talbot, Cary A.; Wilson, John L.
2015-06-01
We have developed an alternative to the one-dimensional partial differential equation (PDE) attributed to Richards (1931) that describes unsaturated porous media flow in homogeneous soil layers. Our solution is a set of three ordinary differential equations (ODEs) derived from unsaturated flux and mass conservation principles. We used a hodograph transformation, the Method of Lines, and a finite water-content discretization to produce ODEs that accurately simulate infiltration, falling slugs, and groundwater table dynamic effects on vadose zone fluxes. This formulation, which we refer to as "finite water-content", simulates sharp fronts and is guaranteed to conserve mass using a finite-volume solution. Our ODE solution method is explicitly integrable, does not require iterations and therefore has no convergence limits and is computationally efficient. The method accepts boundary fluxes including arbitrary precipitation, bare soil evaporation, and evapotranspiration. The method can simulate heterogeneous soils using layers. Results are presented in terms of fluxes and water content profiles. Comparing our method against analytical solutions, laboratory data, and the Hydrus-1D solver, we find that predictive performance of our finite water-content ODE method is comparable to or in some cases exceeds that of the solution of Richards' equation, with or without a shallow water table. The presented ODE method is transformative in that it offers accuracy comparable to the Richards (1931) PDE numerical solution, without the numerical complexity, in a form that is robust, continuous, and suitable for use in large watershed and land-atmosphere simulation models, including regional-scale models of coupled climate and hydrology.
Turing pattern formation in the Brusselator system with nonlinear diffusion
Gambino, G.; Lombardo, M. C.; Sammartino, M.; Sciacca, V.
2013-01-01
In this work we investigate the effect of density dependent nonlinear diffusion on pattern formation in the Brusselator system. Through linear stability analysis of the basic solution we determine the Turing and the oscillatory instability boundaries. A comparison with the classical linear diffusion shows how nonlinear diffusion favors the occurrence of Turing pattern formation. We study the process of pattern formation both in 1D and 2D spatial domains. Through a weakly nonlinear multiple sc...
Dou, Remy; Hogan, DaNel; Kossover, Mark; Spuck, Timothy; Young, Sarah
2013-01-01
Diffusion has often been taught in science courses as one of the primary ways by which molecules travel, particularly within organisms. For years, classroom teachers have used the same common demonstrations to illustrate this concept (e.g., placing drops of food coloring in a beaker of water). Most of the time, the main contributor to the motion…
PPM1D exerts its oncogenic properties in human pancreatic cancer through multiple mechanisms.
Wu, Bo; Guo, Bo-Min; Kang, Jie; Deng, Xian-Zhao; Fan, You-Ben; Zhang, Xiao-Ping; Ai, Kai-Xing
2016-03-01
Protein phosphatase, Mg(2+)/Mn(2+) dependent, 1D (PPM1D) is emerging as an oncogene by virtue of its negative control on several tumor suppressor pathways. However, the clinical significance of PPM1D in pancreatic cancer (PC) has not been defined. In this study, we determined PPM1D expression in human PC tissues and cell lines and their irrespective noncancerous controls. We subsequently investigated the functional role of PPM1D in the migration, invasion, and apoptosis of MIA PaCa-2 and PANC-1 PC cells in vitro and explored the signaling pathways involved. Furthermore, we examined the role of PPM1D in PC tumorigenesis in vivo. Our results showed that PPM1D is overexpressed in human PC tissues and cell lines and significantly correlated with tumor growth and metastasis. PPM1D promotes PC cell migration and invasion via potentiation of the Wnt/β-catenin pathway through downregulation of apoptosis-stimulating of p53 protein 2 (ASPP2). In contrast to PPM1D, our results showed that ASPP2 is downregulated in PC tissues. Additionally, PPM1D suppresses PC cell apoptosis via inhibition of the p38 MAPK/p53 pathway through both dephosphorylation of p38 MAPK and downregulation of ASPP2. Furthermore, PPM1D promotes PC tumor growth in vivo. Our results demonstrated that PPM1D is an oncogene in PC. PMID:26714478
DEFF Research Database (Denmark)
Treebak, Jonas Thue; Pehmøller, Christian; Kristensen, Jonas Møller;
2014-01-01
We investigated the phosphorylation signatures of two Rab GTPase activating proteins TBC1D1 and TBC1D4 in human skeletal muscle in response to physical exercise and physiological insulin levels induced by a carbohydrate rich meal using a paired experimental design. Eight healthy male volunteers...... TBC1D4 in response to physiological stimuli in human skeletal muscle and support the idea that Akt and AMPK are upstream kinases. TBC1D1 phosphorylation signatures were comparable between in vitro contracted mouse skeletal muscle and exercised human muscle, and we show that AMPK was regulating...... phosphorylation of these sites in mouse muscle. Contraction and exercise elicited a different phosphorylation pattern of TBC1D4 in mouse compared with human muscle, and although different circumstances in our experimental setup may contribute to this difference, the observation exemplifies that transferring...
Testing the Early Mars H2-CO2 Greenhouse Hypothesis with a 1-D Photochemical Model
Batalha, Natasha; Ramirez, Ramses; Kasting, James
2015-01-01
A recent study by Ramirez et al. (2014) demonstrated that an atmosphere with 1.3-4 bar of CO2 and H2O, in addition to 5-20% H2, could have raised the mean annual and global surface temperature of early Mars above the freezing point of water. Such warm temperatures appear necessary to generate the rainfall (or snowfall) amounts required to carve the ancient martian valleys. Here, we use our best estimates for early martian outgassing rates, along with a 1-D photochemical model, to assess the conversion efficiency of CO, CH4, and H2S to CO2, SO2, and H2. Our outgassing estimates assume that Mars was actively recycling volatiles between its crust and interior, as Earth does today. H2 production from serpentinization and deposition of banded iron-formations is also considered. Under these assumptions, maintaining an H2 concentration of ~1-2% by volume is achievable, but reaching 5% H2 requires additional H2 sources or a slowing of the hydrogen escape rate below the diffusion limit. If the early martian atmosphere...
From nonfinite to finite 1D arrays of origami tiles.
Wu, Tsai Chin; Rahman, Masudur; Norton, Michael L
2014-06-17
average solution structures for blocks is more readily achieved using computer models than using direct imaging methods. The development of scalable 1D-origami arrays composed of uniquely addressable components is a logical, if not necessary, step in the evolution of higher order fully addressable structures. Our research into the fabrication of arrays has led us to generate a listing of several important areas of future endeavor. Of high importance is the re-enforcement of the mechanical properties of the building blocks and the organization of multiple arrays on a surface of technological importance. While addressing this short list of barriers to progress will prove challenging, coherent development along each of these lines of inquiry will accelerate the appearance of commercial scale molecular manufacturing. PMID:24803094
Vaz, Sharmila; Falkmer, Marita; Parsons, Richard; Passmore, Anne Elizabeth; Parkin, Timothy; Falkmer, Torbjörn
2014-01-01
The relationship between school belongingness and mental health functioning before and after the primary-secondary school transition has not been previously investigated in students with and without disabilities. This study used a prospective longitudinal design to test the bi-directional relationships between these constructs, by surveying 266 students with and without disabilities and their parents, 6-months before and after the transition to secondary school. Cross-lagged multi-group analy...
Ingrid Moons; Patrick De Pelsmacker
2015-01-01
An Extended Decomposed Theory of Planned Behaviour (DTPB) is developed that integrates emotions towards car driving and electric cars as well as car driving habits of the DTPB, and is empirically validated in a Belgian sample ( n = 1023). Multi-group comparisons explore how the determinants of usage intention are different between groups of consumers differing in environmentally-friendly behaviour, environmental concern, innovativeness and personal values. Besides attitudes, media, perceived ...
Energy Technology Data Exchange (ETDEWEB)
Kim, Jong Woon; Kim, Sang Ji; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of {sup 239}Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis.
International Nuclear Information System (INIS)
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of 239Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis
Energy Technology Data Exchange (ETDEWEB)
Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)
2008-05-16
Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies
International Nuclear Information System (INIS)
A highly accurate S4 eigenfunction-based nodal method has been developed to solve multi-group discrete ordinate neutral particle transport problems with a linearly anisotropic scattering in slab geometry. The new method solves the even-parity form of discrete ordinates transport equation with an arbitrary SN order angular quadrature using two sub-cell balance equations and the S4 eigenfunctions of within-group transport equation. The four eigenfunctions from S4 approximation have been chosen as basis functions for the spatial expansion of the angular flux in each mesh. The constant and cubic polynomial approximations are adopted for the scattering source terms from other energy groups and fission source. A nodal method using the conventional polynomial expansion and the sub-cell balances was also developed to be used for demonstrating the high accuracy of the new methods. Using the new methods, a multi-group eigenvalue problem has been solved as well as fixed source problems. The numerical test results of one-group problem show that the new method has third-order accuracy as mesh size is finely refined and it has much higher accuracies for large meshes than the diamond differencing method and the nodal method using sub-cell balances and polynomial expansion of angular flux. For multi-group problems including eigenvalue problem, it was demonstrated that the new method using the cubic polynomial approximation of the sources could produce very accurate solutions even with large mesh sizes. (author)
International Nuclear Information System (INIS)
Thermal reactor design calculations are being performed in India using the WIMS/D-4 multi group cross section library, obtained in late 60's, reflecting the status of the basic nuclear data and processing technology then available. Significant improvements in basic evaluated data files such as ENDF/B-IV to VI and JEF data files etc. have been made in the past four decades and the multigroup libraries have been updated world over using improved and comprehensive nuclear data processing code systems. A few of such updated multigroup cross sections in WIMS/D-4 format are available from KAERI and NEA data bank sources. This paper presents the analysis of a set of enriched UO2 and U-metal uniform critical lattice experiments. These include TRX(4), BAPL (3) and B and W (17) lattice, 64 enriched UO2 lattices complied in NEACRP-U-190 report, 56 enriched UO2 lattices and 61 U-metal lattices which were used for validating the WIMKAL-1988 library. Calculated reaction rate values from the participants of WIMS library update project (WLUP) are available for TRX, BAPL lattices. Integral data measured in the lattices of TRX, BAPL, B and W and NEACRP compilations are available in the open literature. Different calculational methods like J± and Pij, and resonance interpolation schemes were examined in the theoretical analysis. Possible shortcomings of the WIMS-D/4 multigroup cross section library currently being used are also identified. (author)
TBC1D1 Regulates Insulin- and Contraction-Induced Glucose Transport in Mouse Skeletal Muscle
Toyoda, Taro; Yu, Haiyan; Fujii, Nobuharu; Hirshman, Michael F.; An, Ding Jeff; Goodyear, Laurie Joy; Taylor, Eric B.
2010-01-01
OBJECTIVE: TBC1D1 is a member of the TBC1 Rab-GTPase family of proteins and is highly expressed in skeletal muscle. Insulin and contraction increase TBC1D1 phosphorylation on phospho-Akt substrate motifs (PASs), but the function of TBC1D1 in muscle is not known. Genetic linkage analyses show a TBC1D1 R125W missense variant confers risk for severe obesity in humans. The objective of this study was to determine whether TBC1D1 regulates glucose transport in skeletal muscle. RESEARCH DESIGN AND M...
International Nuclear Information System (INIS)
Two computer codes that are available at IPEN for analyses of static neutron diffusion problems are studied and applied. The CITATION code is animed at analyses of criticality, fuel burnup, flux and power distributions etc, in one, two, and three spatial dimensions in multigroup. The EXTERMINATOR code can be used for the same purposes as for CITATION with a limitation to one or two spatial dimensions. Basic theories and numerical techniques used in the codes are studied and summarized. Benchmark problems have been solved using the codes. Comparisons of the results show that both codes can be used with confidence in the analyses of nuclear reactor problems. (author)
Benchmark Tests of the Multigroup Cross Section Libraries for Fast Reactors
International Nuclear Information System (INIS)
In Korea, a design study for a fast breeder reactor named KALIMER (Korea Advanced LIquid MEtal Reactor) has been carried out. The simulations of the KALIMER core have been performed with the JEF-2.2- based 80-group neutron library KAFAX-F22 or the ENDF/B-VI.6-based 150-group neutron library KAFAXE66. Recently, newly evaluated nuclear data files such as ENDF/B-VII (beta 0 and 1), JEFF-3.1, and JENDL-3.3 have been released. And thus there is a need to update the libraries for the KALIMER by using the new data files. In this study, the fast cross section sets with 150 groups were prepared based on ENDF/B-VII beta 0, JEFF-3.1, and JENDL-3.3. The validations of the libraries have been carried out for 14 Cross Section Evaluation Working Group (CSEWG) fast benchmark problems through the 1-D and 2-D DANTSYS calculations. The effective multiplication factors (keff's) and central spectral indices have been compared with the experimental values and the results by the MCNPX calculations
Daskalov, George M; Baker, R S; Rogers, D W O; Williamson, J F
2002-02-01
Our purpose in this work is to demonstrate that the efficiency of dose-rate computations in 125I brachytherapy, using multigroup discrete ordinates radiation transport simulations, can be significantly enhanced using broad energy group cross sections without a loss of accuracy. To this end, the DANTSYS multigroup discrete ordinates neutral particle transport code was used to estimate the absorbed dose-rate distributions around an 125I-model 6702 seed in two-dimensional (2-D) cylindrical R-Z geometry for four different problems spanning the geometries found in clinical practice. First, simulations with a high resolution 210 energy groups library were used to analyze the photon flux spectral distribution throughout this set of problems. These distributions were used to design an energy group structure consisting of three broad groups along with suitable weighting functions from which the three-group cross sections were derived. The accuracy of 2-D DANTSYS dose-rate calculations was benchmarked against parallel Monte Carlo simulations. Ray effects were remedied by using the DANTSYS internal first collision source algorithm. It is demonstrated that the 125I primary photon spectrum leads to inappropriate weighting functions. An accuracy of +/-5% is achieved in the four problem geometries considered using geometry-independent three-group libraries derived from either material-specific weighting functions or a single material-independent weighting function. Agreement between Monte Carlo and the three-group DANTSYS calculations, within three standard Monte Carlo deviations, is observed everywhere except for a limited region along the Z axis of rotational symmetry, where ray effects are difficult to mitigate. The three-group DANTSYS calculations are 10-13 times faster than ones with a 210-group cross section library for 125I dosimetry problems. Compared to 2-D EGS4 Monte Carlo calculations, the 3-group DANTSYS simulations are a 100-fold more efficient. Provided that these
International Nuclear Information System (INIS)
Our purpose in this work is to demonstrate that the efficiency of dose-rate computations in 125I brachytherapy, using multigroup discrete ordinates radiation transport simulations, can be significantly enhanced using broad energy group cross sections without a loss of accuracy. To this end, the DANTSYS multigroup discrete ordinates neutral particle transport code was used to estimate the absorbed dose-rate distributions around an 125I-model 6702 seed in two-dimensional (2-D) cylindrical R-Z geometry for four different problems spanning the geometries found in clinical practice. First, simulations with a high resolution 210 energy groups library were used to analyze the photon flux spectral distribution throughout this set of problems. These distributions were used to design an energy group structure consisting of three broad groups along with suitable weighting functions from which the three-group cross sections were derived. The accuracy of 2-D DANTSYS dose-rate calculations was benchmarked against parallel Monte Carlo simulations. Ray effects were remedied by using the DANTSYS internal first collision source algorithm. It is demonstrated that the 125I primary photon spectrum leads to inappropriate weighting functions. An accuracy of ±5% is achieved in the four problem geometries considered using geometry-independent three-group libraries derived from either material-specific weighting functions or a single material-independent weighting function. Agreement between Monte Carlo and the three-group DANTSYS calculations, within three standard Monte Carlo deviations, is observed everywhere except for a limited region along the Z axis of rotational symmetry, where ray effects are difficult to mitigate. The three-group DANTSYS calculations are 10-13 times faster than ones with a 210-group cross section library for 125I dosimetry problems. Compared to 2-D EGS4 Monte Carlo calculations, the 3-group DANTSYS simulations are a 100-fold more efficient. Provided that these
Anderson, Robert C.
1976-06-22
1. A method for joining beryllium to beryllium by diffusion bonding, comprising the steps of coating at least one surface portion of at least two beryllium pieces with nickel, positioning a coated surface portion in a contiguous relationship with an other surface portion, subjecting the contiguously disposed surface portions to an environment having an atmosphere at a pressure lower than ambient pressure, applying a force upon the beryllium pieces for causing the contiguous surface portions to abut against each other, heating the contiguous surface portions to a maximum temperature less than the melting temperature of the beryllium, substantially uniformly decreasing the applied force while increasing the temperature after attaining a temperature substantially above room temperature, and maintaining a portion of the applied force at a temperature corresponding to about maximum temperature for a duration sufficient to effect the diffusion bond between the contiguous surface portions.
Grid Cell Responses in 1D Environments Assessed as Slices through a 2D Lattice.
Yoon, KiJung; Lewallen, Sam; Kinkhabwala, Amina A; Tank, David W; Fiete, Ila R
2016-03-01
Grid cells, defined by their striking periodic spatial responses in open 2D arenas, appear to respond differently on 1D tracks: the multiple response fields are not periodically arranged, peak amplitudes vary across fields, and the mean spacing between fields is larger than in 2D environments. We ask whether such 1D responses are consistent with the system's 2D dynamics. Combining analytical and numerical methods, we show that the 1D responses of grid cells with stable 1D fields are consistent with a linear slice through a 2D triangular lattice. Further, the 1D responses of comodular cells are well described by parallel slices, and the offsets in the starting points of the 1D slices can predict the measured 2D relative spatial phase between the cells. From these results, we conclude that the 2D dynamics of these cells is preserved in 1D, suggesting a common computation during both types of navigation behavior. PMID:26898777
International Nuclear Information System (INIS)
ONETRAN solves the one-dimensional multigroup transport equation in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous 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. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. Negative fluxes are eliminated by a local set-to-zero and correct algorithm. Standard inner (within-group) iteration cycles are accelerated by system rebalance, coarse-mesh rebalance, or Chebyshev acceleration. Outer iteration cycles are accelerated by coarse-mesh rebalance. Provision is made for creation of standard interface output files for S/sub N/ constants, inhomogeneous sources, angle-integrated fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and total or angular fluxes may be read. All binary operations are localized in subroutines REED and RITE. Flexible edit options, including restart capability are provided. ONETRAN is designed for use with CDC-7600 and IBM 360. (12 tables, 10 figures) (U.S.)
VELM61 and VELM22: Multigroup cross-section libraries for sodium-cooled reactor shield analysis
International Nuclear Information System (INIS)
Two coupled neutron and photon multigroup cross-section libraries, derived from ENDF/B-V nuclear data, are described. The energy group structures, 61n/23γ and 22n/10γ, are subsets of the Vitamin-E 174n/38γ group structure, and are tailored to the iron and sodium resonances, windows, and capture gamma-ray spectra. Each of the two libraries are available in two formats, the AMPX master format and the ANISN format. Cross sections for all materials in the Vitamin-E library were collapsed using a standard energy weighting function, and in addition, several cross-section sets for each of the major constituents of commercial grade sodium, stainless steel (types 304 and 316), and carbon steel were derived using several problem-dependent weighting functions for averaging the fine groups. Effects of various group structures and weighting functions on the accuracy of the broad group libraries are studied by ANISN analysis of a typical sodium-iron shield configuration
Roberts, Luke F; Haas, Roland; O'Connor, Evan P; Diener, Peter; Schnetter, Erik
2016-01-01
We report on a set of long-term general-relativistic three-dimensional (3D) multi-group (energy-dependent) neutrino-radiation hydrodynamics simulations of core-collapse supernovae. We employ a full 3D two-moment scheme with the local M1 closure, three neutrino species, and 12 energy groups per species. With this, we follow the post-core-bounce evolution of the core of a nonrotating $27$-$M_\\odot$ progenitor in full unconstrained 3D and in octant symmetry for $\\gtrsim$$ 380\\,\\mathrm{ms}$. We find the development of an asymmetric runaway explosion in our unconstrained simulation. We test the resolution dependence of our results and, in agreement with previous work, find that low resolution artificially aids explosion and leads to an earlier runaway expansion of the shock. At low resolution, the octant and full 3D dynamics are qualitatively very similar, but at high resolution, only the full 3D simulation exhibits the onset of explosion.
International Nuclear Information System (INIS)
A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)
Endogenous N-terminal Domain Cleavage Modulates α1D-Adrenergic Receptor Pharmacodynamics.
Kountz, Timothy S; Lee, Kyung-Soon; Aggarwal-Howarth, Stacey; Curran, Elizabeth; Park, Ji-Min; Harris, Dorathy-Ann; Stewart, Aaron; Hendrickson, Joseph; Camp, Nathan D; Wolf-Yadlin, Alejandro; Wang, Edith H; Scott, John D; Hague, Chris
2016-08-26
The α1D-adrenergic receptor (ADRA1D) is a key regulator of cardiovascular, prostate, and central nervous system functions. This clinically relevant G protein-coupled receptor has proven difficult to study, as it must form an obligate modular homodimer containing the PDZ proteins scribble and syntrophin or become retained in the endoplasmic reticulum as non-functional protein. We previously determined that targeted removal of the N-terminal (NT) 79 amino acids facilitates ADRA1D plasma membrane expression and agonist-stimulated functional responses. However, whether such an event occurs in physiological contexts was unknown. Herein, we report the ADRA1D is subjected to innate NT processing in cultured human cells. SNAP near-infrared imaging and tandem-affinity purification revealed the ADRA1D is expressed as both full-length and NT truncated forms in multiple human cell lines. Serial truncation mapping identified the cleavage site as Leu(90)/Val(91) in the 95-amino acid ADRA1D NT domain, suggesting human cells express a Δ1-91 ADRA1D species. Tandem-affinity purification MS/MS and co-immunoprecipitation analysis indicate NT processing of ADRA1D is not required to form scribble-syntrophin macromolecular complexes. Yet, label-free dynamic mass redistribution signaling assays demonstrate that Δ1-91 ADRA1D agonist responses were greater than WT ADRA1D. Mutagenesis of the cleavage site nullified the processing event, resulting in ADRA1D agonist responses less than the WT receptor. Thus, we propose that processing of the ADRA1D NT domain is a physiological mechanism employed by cells to generate a functional ADRA1D isoform with optimal pharmacodynamic properties. PMID:27382054
Levenson, L.
1963-09-01
A high-vacuum diffusion pump is described, featuring a novel housing geometry for enhancing pumping speed. An upright, cylindrical lower housing portion is surmounted by a concentric, upright, cylindrical upper housing portion of substantially larger diameter; an uppermost nozzle, disposed concentrically within the upper portion, is adapted to eject downwardly a conical sheet of liquid outwardly to impinge upon the uppermost extremity of the interior wall of the lower portion. Preferably this nozzle is mounted upon a pedestal rising coaxially from within the lower portion and projecting up into said upper portion. (AEC)
DEFF Research Database (Denmark)
Trinkaus, H.; Singh, Bachu Narain; Golubov, S.I.
formation) can be rationalised in terms of intra-cascade clustering of vacancies and self-interstitial atoms (SIAs), differences in the thermal stability and mobility of the resulting clusters and one-dimensional (1D) glide diffusion of SIA clusters (“production bias model”). The 1D diffusion of SIA...... clusters is generally disturbed by changes between equivalent 1D diffusion paths and by transversal diffusion by self-climb, resulting in diffusion reaction kinetics between the 1D and 3D limiting cases. In this paper, a general treatment of such kinetics operating in systems containing random...... distributions of sinks is presented. The complicated partial sink strengths of different components of the system for the annihilation of SIA clusters are expressed by those for the simple 1D and 3D limiting cases. The effects of direction changes and transversal diffusion are first considered separately and...
关于图的L(d1,d2)-标号问题%The L(d1, d2)-Labeling Problem on Graphs
Institute of Scientific and Technical Information of China (English)
邵振东; 刘家壮
2006-01-01
The L(2, 1)-labeling is formulated from the frequency assignment problem. We study the L(d1, d2)- labeling which is a generalization of the L(2, 1)-labeling. Vertex 2-coloring, 2-chromatic number and other related concepts are firstly defined, and the upper bound for 2-chromatic number is given; a very general relationship between λd1 ,d2 (G) and minimum degree δ(G) and maximum degree △(G) is then derived; finally, the upper bounds of L(d1, d2)-labelings of general and planar graphs are given.%图的L(2,1)-标号问题是由频率分配问题归结而来,本文研究作为L(2,1)-标号问题的推广的L(d1,d2)-标号问题.首先定义了顶点2-着色,2-色数及其它有关概念,给出了2-色数的上界.然后得出了λd1,d2(G)与δ(G)和△(G)的一般关系.最后得出了一般图与平面图的λd1,d2(G)的上界.
Exercise increases TBC1D1 phosphorylation in human skeletal muscle
Jessen, Niels; An, Ding; Lihn, Aina S.; Nygren, Jonas; Hirshman, Michael F.; Thorell, Anders; Goodyear, Laurie J.
2011-01-01
Exercise and weight loss are cornerstones in the treatment and prevention of type 2 diabetes, and both interventions function to increase insulin sensitivity and glucose uptake into skeletal muscle. Studies in rodents demonstrate that the underlying mechanism for glucose uptake in muscle involves site-specific phosphorylation of the Rab-GTPase-activating proteins AS160 (TBC1D4) and TBC1D1. Multiple kinases, including Akt and AMPK, phosphorylate TBC1D1 and AS160 on distinct residues, regulatin...
Benchmarks for multicomponent diffusion and electrochemical migration
DEFF Research Database (Denmark)
Rasouli, Pejman; Steefel, Carl I.; Mayer, K. Ulrich; Rolle, Massimo
2015-01-01
been published to date. This contribution provides a set of three benchmark problems that demonstrate the effect of electric coupling during multicomponent diffusion and electrochemical migration and at the same time facilitate the intercomparison of solutions from existing reactive transport codes...... considered in solute transport problems, electromigration can strongly affect mass transport processes. The number of reactive transport models that consider electromigration has been growing in recent years, but a direct model intercomparison that specifically focuses on the role of electromigration has not....... The first benchmark focuses on the 1D transient diffusion of HNO3 (pH = 4) in a NaCl solution into a fixed concentration reservoir, also containing NaCl—but with lower HNO3 concentrations (pH = 6). The second benchmark describes the 1D steady-state migration of the sodium isotope 22Na triggered by...
Column Testing and 1D Reactive Transport Modeling to Evaluate Uranium Plume Persistence Processes
Johnson, R. H.; Morrison, S.; Morris, S.; Tigar, A.; Dam, W. L.; Dayvault, J.
2015-12-01
At many U.S. Department of Energy Office of Legacy Management sites, 100 year natural flushing was selected as a remedial option for groundwater uranium plumes. However, current data indicate that natural flushing is not occurring as quickly as expected and solid-phase and aqueous uranium concentrations are persistent. At the Grand Junction, Colorado office site, column testing was completed on core collected below an area where uranium mill tailings have been removed. The total uranium concentration in this core was 13.2 mg/kg and the column was flushed with laboratory-created water with no uranium and chemistry similar to the nearby Gunnison River. The core was flushed for a total of 91 pore volumes producing a maximum effluent uranium concentration of 6,110 μg/L at 2.1 pore volumes and a minimum uranium concentration of 36.2 μg/L at the final pore volume. These results indicate complex geochemical reactions at small pore volumes and a long tailing affect at greater pore volumes. Stop flow data indicate the occurrence of non-equilibrium processes that create uranium concentration rebound. These data confirm the potential for plume persistence, which is occurring at the field scale. 1D reactive transport modeling was completed using PHREEQC (geochemical model) and calibrated to the column test data manually and using PEST (inverse modeling calibration routine). Processes of sorption, dual porosity with diffusion, mineral dissolution, dispersion, and cation exchange were evaluated separately and in combination. The calibration results indicate that sorption and dual porosity are major processes in explaining the column test data. These processes are also supported by fission track photographs that show solid-phase uranium residing in less mobile pore spaces. These procedures provide valuable information on plume persistence and secondary source processes that may be used to better inform and evaluate remedial strategies, including natural flushing.
Diffusion coefficient in photon diffusion theory
Graaff, R; Ten Bosch, JJ
2000-01-01
The choice of the diffusion coefficient to be used in photon diffusion theory has been a subject of discussion in recent publications on tissue optics. We compared several diffusion coefficients with the apparent diffusion coefficient from the more fundamental transport theory, D-app. Application to
On physical diffusion and stochastic diffusion
Narasimhan, T.N.
2010-01-01
Although the same mathematical expression is used to describe physical diffusion and stochastic diffusion, there are intrinsic similarities and differences in their nature. A comparative study shows that characteristic terms of physical and stochastic diffusion cannot be placed exactly in one-to-one correspondence. Therefore, judgment needs to be exercised in transferring ideas between physical and stochastic diffusion.
1D engine simulation of a turbocharged SI engine with CFD computation on components
Renberg, Ulrica
2008-01-01
Techniques that can increase the SI- engine efficiency while keeping the emissions very low is to reduce the engine displacement volume combined with a charging system. Advanced systems are needed for an effective boosting of the engine and today 1D engine simulation tools are often used for their optimization. This thesis concerns 1D engine simulation of a turbocharged SI engine and the introduction of CFD computations on components as a way to assess inaccuracies in the 1D model. 1D engine ...
Solution of diffusive wave equation using fem for flood forecasting
International Nuclear Information System (INIS)
Flood forecasting can be predicted numerically by the solution of 1D (One-Dimensional) unsteady diffusive wave equation. This research paper presents the development of a fem (Finite Element Model) for flood routing using diffusive wave with and without lateral inflow/outflow. FEM is based on two-step semi implicit Taylor Galerkin technique. The accuracy of the model has been verified by comparing computed results with available solution of the diffusive wave problems available in open literature. Numerical results demonstrate that the technique is an efficient and accurate tool to simulate diffusive wave equation for outflow hydrograph for temporal variation of lateral inflow and outflow. (author)
Faltermann, Susanne; Prétôt, René; Pernthaler, Jakob; Fent, Karl
2016-02-01
Microcystin-LR (MC-LR) and nodularin are hepatotoxins produced by several cyanobacterial species. Their toxicity is based on active cellular uptake and subsequent inhibition of protein phosphatases PP1/2A, leading to hyperphosphorylation and cell death. To date, uptake of MC-LR and nodularin in fish is poorly understood. Here, we investigated the role of the organic anion transporting polypeptide Oatp1d1 in zebrafish (drOatp1d1, Slco1d1) in cellular uptake in zebrafish. We stably transfected CHO and HEK293 cell lines expressing drOatp1d1. In both transfectants, uptake of MC-LR and nodularin was demonstrated by competitive inhibition of uptake with fluorescent substrate lucifer yellow. Direct uptake of MC-LR was demonstrated by immunostaining, and indirectly by the high cytotoxicity in stable transfectants. By means of a synthesized fluorescent labeled MC-LR derivative, direct uptake was further confirmed in HEK293 cells expressing drOatp1d1. Additionally, uptake and toxicity was investigated in the permanent zebrafish liver cell line ZFL. These cells had only a low relative abundance of drOatp1d1, drOatp2b1 and drOatp1f transcripts, which correlated with the lack of MC-LR induced cytotoxicity and transcriptional changes of genes indicative of endoplasmic reticulum stress, a known effect of this toxin. Our study demonstrates that drOatp1d1 functions as an uptake transporter for both MC-LR and nodularin in zebrafish. PMID:26769064
Bingi, Jayachandra; Murukeshan, Vadakke Matham
2016-02-01
Laser speckles and speckle patterns, which are formed by the random interference of scattered waves from optically rough surfaces, have found tremendous applications in a wide range of metrological and biomedical fields. Here, we demonstrate a novel edge diffraction phenomenon of individual speckle for the fabrication of 1D and 2D micron and sub-micron size random gratings. These random gratings exhibit broadband response with interesting diffusive diffraction patterns. As an immediate application for solar energy harvesting, significant reduction in transmission and enhanced absorption in thin “Si-random grating-Si” sandwich structure is demonstrated. This work has multifaceted significance where we exploited the individual speckle diffraction properties for the first time. Besides the solar harvesting applications, random gratings are suitable structures for fabrication of theoretically proposed random quantum well IR detectors and hence expected that this work will augur well for such studies in the near future.
Propagation in quantum walks and relativistic diffusions
Debbasch, Fabrice; Di Molfetta, Giuseppe; Espaze, David; Foulonneau, Vincent
2013-01-01
Propagation in quantum walks is revisited by showing that very general 1D discrete-time quantum walks with time- and space-dependent coefficients can be described, at the continuous limit, by Dirac fermions coupled to electromagnetic fields. Short-time propagation is also established for relativistic diffusions by presenting new numerical simulations of the Relativistic Ornstein-Uhlenbeck Process. A geometrical generalization of Fick's law is also obtained for this process. The results sugges...
Is Arnold diffusion relevant to global diffusion?
Honjo, Seiichiro; Kaneko, Kunihiko
2003-01-01
Global diffusion of Hamiltonian dynamical systems is investigated by using a coupled standard maps. Arnold web is visualized in the frequency space, using local rotation numbers, while Arnold diffusion and resonance overlaps are distinguished by the residence time distributions at resonance layers. Global diffusion in the phase space is shown to be accelerated by diffusion across overlapped resonances generated by the coupling term, rather than Arnold diffusion along the lower-order resonance...
ZZ VITENEA-J, AMPX 175-N,42-gamma multigroup X-sect. library for nuclear fusion applications
International Nuclear Information System (INIS)
1 - Description of program or function: VITENEA-J is a coupled multigroup cross section library in the standard VITAMIN-J energy group structure (175 n + 42 gamma) in AMPX format produced by ENEA-Bologna for nuclear fusion applications.It is based on nuclear data from the general purpose Fusion Evaluated Nuclear Data Library (FENDL/E-2.0) and ENDF-B/VI. This library has been widely used by ENEA group in neutron/gamma transport calculations for ITER safety assessment via the SCALENEA-1 multipurpose Sn calculation sequence. List of the included materials: 1-H-1, 1-H-1(water), 1-H-2, 1-H-3, 2-He-3, 2-He-4, 3-Li-6, 3-Li-7, 4-Be-9, 4-Be-9(metal) , 4-Be-9(oxide), 5-B-10, 5-B-11, C -nat, C -nat(graphite), 7-N-14, 7-N-15, 8-O-16, 9-F-19, 11-Na-23, Mg-nat, 13-Al-27, 14-Si-28, 14-Si-29, 14-Si-30, 15-P-31, S -nat, Cl-nat, K-nat, Ca-nat, Ti-nat, 23-V-51, 24-Cr-50, 24-Cr-52, 24-Cr-53, 24-Cr-54, 25-Mn-55, 26-Fe-54, 26-Fe-56, 26-Fe-57, 26-Fe-58, 27-Co-59, 28-Ni-58, 28-Ni-60, 28-Ni-61, 28-Ni-62, 28-Ni-64, 29-Cu-63, 29-Cu-65, Ga-nat, Zr-nat, 41-Nb-93, Mo-nat, Cd-nat, Sn-nat, Hf-nat, 72-Hf-176, 72-Hf-177 , 72-Hf-178, 72-Hf-179, 72-Hf-180, 73-Ta-181, W-nat, 74-W-182, 74-W-183 , 74-W-184, 74-W-186 , 79-Au-197 , 82-Pb-206, 82-Pb-207, 82-Pb-208, 83-Bi-209, 47-Ag-107 , 47-Ag-109 , ANSI-Flux. 2 - Methods: The data have been prepared by processing basic nuclear data into a fine group problem independent format through an automatic calculation procedure using the modules of NJOY-94.105 and AMPX-77 Processing Systems. The translation of the multigroup data from the GENDF format produced by NJOY to the AMPX Master Library format was performed by means of the AMPX-77 SMILER module. 3 - Restrictions on the complexity of the problem: The library contains 75 nuclides. Seventy nuclides were processed at four temperatures (300 K, 600 K, 900 K, 1500 K) at ten values for the background cross section s0. Thermal scattering cross sections were processed at the three temperatures (296 K, 600 K, 1200 K
MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX
International Nuclear Information System (INIS)
1 - Description of problem or function: MINX calculates fine-group averaged infinitely diluted cross sections and self-shielding factors from ENDF/B-IV data. Its primary purpose is to generate a pseudo-composition-independent multigroup library which is input to the SPHINX space-energy collapse program (2) (PSR-0129) through standard CCCC-III (8) interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (5) (NESC0388) and ENDRUN (9) and the high-order group-to-group transfer matrices of SUPERTOG (10) (PSR-0013) and ETOG (11). Fine group energy boundaries, Legendre expansion order, gross spectral shape component (in the Bondarenko flux model), temperatures and dilutions can all be used specifically. 2 - Method of solution: Infinitely dilute, un-broadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the
Energy Technology Data Exchange (ETDEWEB)
Ozaki, N.; Lappalainen, J.; Linnoila, M. [National Institute on Alcohol Abuse and Alcoholism, Rockville, MD (United States)] [and others
1995-04-24
Serotonin (5-HT){sub ID} receptors are 5-HT release-regulating autoreceptors in the human brain. Abnormalities in brain 5-HT function have been hypothesized in the pathophysiology of various psychiatric disorders, including obsessive-compulsive disorder, autism, mood disorders, eating disorders, impulsive violent behavior, and alcoholism. Thus, mutations occurring in 5-HT autoreceptors may cause or increase the vulnerability to any of these conditions. 5-HT{sub 1D{alpha}} and 5-HT{sub 1D{Beta}} subtypes have been previously localized to chromosomes 1p36.3-p34.3 and 6q13, respectively, using rodent-human hybrids and in situ localization. In this communication, we report the detection of a 5-HT{sub 1D{alpha}} receptor gene polymorphism by single strand conformation polymorphism (SSCP) analysis of the coding sequence. The polymorphism was used for fine scale linkage mapping of 5-HT{sub 1D{alpha}} on chromosome 1. This polymorphism should also be useful for linkage studies in populations and in families. Our analysis also demonstrates that functionally significant coding sequence variants of the 5-HT{sub 1D{alpha}} are probably not abundant either among alcoholics or in the general population. 14 refs., 1 fig., 1 tab.
Energy Technology Data Exchange (ETDEWEB)
Hill, T.R.; Reed, W.H.
1976-01-01
TIMEX solves the time-dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous 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. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time steps can be taken. Because no iteration is performed the method is exceptionally fast in terms of computing time per time step. Two acceleration methods, exponential extrapolation and rebalance, are utilized to improve the accuracy of the time differencing scheme. Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. The running time for TIMEX is highly problem-dependent, but varies almost linearly with the total number of unknowns and time steps. Provision is made for creation of standard interface output files for angular fluxes and angle-integrated fluxes. Five interface units (use of interface units is optional), five output units, and two system input/output units are required. A large bulk memory is desirable, but may be replaced by disk, drum, or tape storage. 13 tables, 9 figures. (auth)
International Nuclear Information System (INIS)
TIMEX solves the time-dependent, one-dimensional multigroup transport equation with delayed neutrons in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous 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. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. The time variable is differenced by an explicit technique that is unconditionally stable so that arbitrarily large time steps can be taken. Because no iteration is performed the method is exceptionally fast in terms of computing time per time step. Two acceleration methods, exponential extrapolation and rebalance, are utilized to improve the accuracy of the time differencing scheme. Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXCOR can be accommodated. The running time for TIMEX is highly problem-dependent, but varies almost linearly with the total number of unknowns and time steps. Provision is made for creation of standard interface output files for angular fluxes and angle-integrated fluxes. Five interface units (use of interface units is optional), five output units, and two system input/output units are required. A large bulk memory is desirable, but may be replaced by disk, drum, or tape storage. 13 tables, 9 figures
International Nuclear Information System (INIS)
Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group (1-g) cross sections must be provided in advance. This paper focuses on generating accurate 1-g cross section values that are necessary for evaluation of nuclide densities as a function of burnup. The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires extensive computational efforts. The method presented here is based on the multi-group (MG) approach, in which pre-generated MG sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate 1-g cross sections requires their tabulation against the background cross-section (σ0) to account for the self-shielding effect. However, in previous studies, the model that was used to calculate σ0 was simplified by fixing Bell and Dancoff factors. This work demonstrates that 1-g values calculated under the previous simplified model may not agree with the tallied values. Therefore, the original background cross section model was extended by implicitly accounting for the Dancoff and bell factors. The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented into BGCore code system. The 1-g cross section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement (<0.05%) in the 1-g cross values was observed. The method dose not carry any additional computational burden and it is universally applicable to the analysis of thermal as well as fast reactor systems. (author)
SIXTUS-3, 3-D Nodal Neutron Diffusion Criticality in Hexagonal Geometry
International Nuclear Information System (INIS)
1 - Description of program or function: Three-dimensional, nodal, neutron diffusion theory criticality code in hexagonal geometry. 2 - Method of solution: Intra-nodal elementary solutions with an exact multigroup Eigenvalue spectrum are spanned on two irreducible symmetry bases for the C6 and C2 groups of rotations for lateral and axial components, respectively. They represent a general homogeneous solution which is augmented with the special heterogeneous one determined by the transversal leakage terms, and from it the multigroup nodal response matrix for the partial current symmetry components on the node interfaces is computed. The response of a node to the incoming current is propagated in the system in a node sweeping process and after a prescribed number of sweeps the multiplication factor and the response matrix is recalculated. The iterations are terminated when the criteria of convergence for the multiplication factor, fission source and flux are met. An acceleration algorithm based on a special version of the Lyusternyk-Wagner extrapolation scheme is employed. The program contains a number of EISPACK routines. 3 - Restrictions on the complexity of the problem: Number of materials should not exceed 255; however, this restriction can be easily removed
International Nuclear Information System (INIS)
Highlights: • We compare the equal-probable cosine (EPC) method in NJOY and Gauss quadrature. • The EPC method has large discrepancy with analytic solutions. • Gauss quadrature does not suffer from the same problems. • We give quantitative comparisons between the methods for Legendre moments. - Abstract: As high-fidelity simulations become routine and computational modelers begin to ask questions about uncertainty in calculations, the understanding of uncertainties in nuclear data, including multigroup cross-sections to high scattering orders, is also becoming important. In this paper we look at how a widely-used data processing code, NJOY, processes thermal cross-section data. Using an alternative integration scheme, Gauss quadrature, for angular integration to generate thermal scattering cross sections from the thermal scattering laws for graphite and ZrHx, we observed discrepancies between these results and NJOY’s results using equal-probable cosines on high order Legendre moments. In order to find a reliable comparison between the methods in NJOY and alternatives, we derived novel analytical expressions for high order Legendre moments of the free gas scattering model. Such expressions are analytically tractable but complicated. Using these expressions we construct a semi-analytic benchmark for multigroup Legendre moments. By comparing the results among the benchmark, Gauss quadrature and NJOY, we found that Gauss quadrature can preserve comparable accuracy as NJOY for lower order Legendre moments for free gas scattering and outperforms NJOY in generating high order moments. Our findings indicate that for high-scattering moments, multigroup data could be an source of uncertainty in thermal reactor calculations
International Nuclear Information System (INIS)
A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (SN) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The SN discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the SN transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup SN eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)
An efficient diffusion approach for chaos-based image encryption
International Nuclear Information System (INIS)
One of the existing chaos-based image cryptosystems is composed of alternative substitution and diffusion stages. A multi-dimensional chaotic map is usually employed in the substitution stage for image pixel permutation while a one-dimensional (1D) chaotic map is used for diffusion purpose. As the latter usually involves real number arithmetic operations, the overall encryption speed is limited by the diffusion stage. In this paper, we propose a more efficient diffusion mechanism using simple table lookup and swapping techniques as a light-weight replacement of the 1D chaotic map iteration. Simulation results show that at a similar security level, the proposed cryptosystem needs about one-third the encryption time of a similar cryptosystem. The effective acceleration of chaos-based image cryptosystems is thus achieved.
Diffusive Dynamics of Contact Formation in Disordered Polypeptides.
Zerze, Gül H; Mittal, Jeetain; Best, Robert B
2016-02-12
Experiments measuring contact formation between probes in disordered chains provide information on the fundamental time scales relevant to protein folding. However, their interpretation usually relies on one-dimensional (1D) diffusion models, as do many experiments probing a single distance. Here, we use all-atom molecular simulations to capture both the time scales of contact formation, as well as the scaling with peptide length for tryptophan triplet quenching experiments, revealing the sensitivity of the experimental quenching times to the configurational space explored by the chain. We find a remarkable consistency between the results of the full calculation and from Szabo-Schulten-Schulten theory applied to a 1D diffusion model, supporting the validity of such models. The significant reduction in diffusion coefficient at the small probe separations which most influence quenching rate, suggests that contact formation and Förster resonance energy transfer correlation experiments provide complementary information on diffusivity. PMID:26919016
Diffusive Dynamics of Contact Formation in Disordered Polypeptides
Zerze, Gül H.; Mittal, Jeetain; Best, Robert B.
2016-02-01
Experiments measuring contact formation between probes in disordered chains provide information on the fundamental time scales relevant to protein folding. However, their interpretation usually relies on one-dimensional (1D) diffusion models, as do many experiments probing a single distance. Here, we use all-atom molecular simulations to capture both the time scales of contact formation, as well as the scaling with peptide length for tryptophan triplet quenching experiments, revealing the sensitivity of the experimental quenching times to the configurational space explored by the chain. We find a remarkable consistency between the results of the full calculation and from Szabo-Schulten-Schulten theory applied to a 1D diffusion model, supporting the validity of such models. The significant reduction in diffusion coefficient at the small probe separations which most influence quenching rate, suggests that contact formation and Förster resonance energy transfer correlation experiments provide complementary information on diffusivity.
International Nuclear Information System (INIS)
Selected neutron reaction nuclear data evaluations and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab
Identification of RAPD Marker for Chromosome 1D of Common Wheat
Directory of Open Access Journals (Sweden)
Imtiaz Ahmad Khan
2010-04-01
Full Text Available Development of genetically compensating nullisomic-tetrasomic and ditelosomic lines of commonwheat (Triticum aestivum L. have been widely used to construct high density genetic maps of homoeologouswheat chromosomes. During present research, easier, cheaper and quicker procedure of Polymerase ChainReaction (PCR was used to map Randomly Amplified Polymorphic DNA primers on chromosome 1D ofcommon wheat. Genomic DNA was isolated from two genetic stocks of wheat cultivar Chinese Spring viz;NT-1D1B and NT-2A2B. PCR were conducted using RAPD primers GLC-07 and GLC-11. RAPD primerGLC-11 amplified a polymorphic allele of approximately 500 bp, which was present in NT-2A2B (used aspositive control but was absent in NT-1D1B indicating that the locus is present on chromosome 1D of commonwheat. Hence this marker (GLC-11 can reliably be used to keep track of chromosome 1D of hexaploid wheat.
International Nuclear Information System (INIS)
A fully relativistic Dirac-Fock method with Breit and QED corrections has been employed to study energy levels and oscillator strengths for the ns(n-1)d 1D-ns21S transitions of the alkaline earth atoms. In calculation, the authors consider significant Breit and QED corrections, the results are in good agreements with recent experimental data and other theoretical values. The results show that it is feasible to obtain the highly Rybderg states of the alkaline earth atoms, especially the autoionization states, by use of quadrupole transitions as an intermediate resonance
Kobayashi, Hirokazu
2015-12-01
One-dimensional (1D) molecular chains of 4-substituted-2,2,6,6-tetramethyl-1-piperidinyloxyl (4-X-TEMPO) radicals were constructed in the crystalline 1D nanochannels of 2,4,6-tris(4-chlorophenoxy)-1,3,5-triazine (CLPOT) used as a template. The ESR spectra of CLPOT inclusion compounds (ICs) using 4-X-TEMPO were examined on the basis of spectral simulation using EasySpin program package for simulating and fitting ESR spectra. The ESR spectra of [(CLPOT)2-(TEMPO)1.0] IC were isotropic in the total range of temperatures. The peak-to-peak line width (ΔBpp) became monotonically narrower from 2.8 to 1.3 mT with increase in temperature in the range of 4.2-298 K. The effect of the rotational diffusion motion of TEMPO radicals in the CLPOT nanochannels for the inter-spin interaction of the [(CLPOT)2-(TEMPO)1.0] IC was found to be smaller than the case of [(TPP)2-(TEMPO)1.0] IC (TPP = tris(o-phenylenedioxy)cyclotriphosphazene) reported in our previous study. The ΔBpp of the [(CLPOT)2-(TEMPO)1.0] IC in the whole range of temperatures was much narrower than the estimation to be based on the Van Vleck's formula for the second moment of the rigid lattice model where the electron spin can be considered as fixed; 11 mT of Gaussian line-width component. This suggests the possibility of exchange narrowing in the 1D organic-radical chains of the [(CLPOT)2-(TEMPO)1.0] IC. On the other hand, the ESR spectra of [(CLPOT)2-(MeO-TEMPO)0.41] IC (MeO-TEMPO = 4-methoxy-TEMPO) were reproduced by a superposition of major broad isotropic adsorption line and minor temperature-dependent modulated triplet component. This suggests that the IC has the part of 1D organic-radical chains and MeO-TEMPO molecules isolated in the CLPOT nanochannels.
Energy Technology Data Exchange (ETDEWEB)
Kobayashi, Hirokazu [Department of Chemistry, College of Humanities and Sciences, Nihon University 3-25-40, Sakura-jo-sui, Setagaya-ku, Tokyo, 156-8550 (Japan)
2015-12-31
One-dimensional (1D) molecular chains of 4-substituted-2,2,6,6-tetramethyl-1-piperidinyloxyl (4-X-TEMPO) radicals were constructed in the crystalline 1D nanochannels of 2,4,6-tris(4-chlorophenoxy)-1,3,5-triazine (CLPOT) used as a template. The ESR spectra of CLPOT inclusion compounds (ICs) using 4-X-TEMPO were examined on the basis of spectral simulation using EasySpin program package for simulating and fitting ESR spectra. The ESR spectra of [(CLPOT){sub 2}-(TEMPO){sub 1.0}] IC were isotropic in the total range of temperatures. The peak-to-peak line width (ΔB{sub pp}) became monotonically narrower from 2.8 to 1.3 mT with increase in temperature in the range of 4.2–298 K. The effect of the rotational diffusion motion of TEMPO radicals in the CLPOT nanochannels for the inter-spin interaction of the [(CLPOT){sub 2}-(TEMPO){sub 1.0}] IC was found to be smaller than the case of [(TPP){sub 2}−(TEMPO){sub 1.0}] IC (TPP = tris(o-phenylenedioxy)cyclotriphosphazene) reported in our previous study. The ΔB{sub pp} of the [(CLPOT){sub 2}-(TEMPO){sub 1.0}] IC in the whole range of temperatures was much narrower than the estimation to be based on the Van Vleck’s formula for the second moment of the rigid lattice model where the electron spin can be considered as fixed; 11 mT of Gaussian line-width component. This suggests the possibility of exchange narrowing in the 1D organic-radical chains of the [(CLPOT){sub 2}-(TEMPO){sub 1.0}] IC. On the other hand, the ESR spectra of [(CLPOT){sub 2}-(MeO-TEMPO){sub 0.41}] IC (MeO-TEMPO = 4-methoxy-TEMPO) were reproduced by a superposition of major broad isotropic adsorption line and minor temperature-dependent modulated triplet component. This suggests that the IC has the part of 1D organic-radical chains and MeO-TEMPO molecules isolated in the CLPOT nanochannels.
International Nuclear Information System (INIS)
One-dimensional (1D) molecular chains of 4-substituted-2,2,6,6-tetramethyl-1-piperidinyloxyl (4-X-TEMPO) radicals were constructed in the crystalline 1D nanochannels of 2,4,6-tris(4-chlorophenoxy)-1,3,5-triazine (CLPOT) used as a template. The ESR spectra of CLPOT inclusion compounds (ICs) using 4-X-TEMPO were examined on the basis of spectral simulation using EasySpin program package for simulating and fitting ESR spectra. The ESR spectra of [(CLPOT)2-(TEMPO)1.0] IC were isotropic in the total range of temperatures. The peak-to-peak line width (ΔBpp) became monotonically narrower from 2.8 to 1.3 mT with increase in temperature in the range of 4.2–298 K. The effect of the rotational diffusion motion of TEMPO radicals in the CLPOT nanochannels for the inter-spin interaction of the [(CLPOT)2-(TEMPO)1.0] IC was found to be smaller than the case of [(TPP)2−(TEMPO)1.0] IC (TPP = tris(o-phenylenedioxy)cyclotriphosphazene) reported in our previous study. The ΔBpp of the [(CLPOT)2-(TEMPO)1.0] IC in the whole range of temperatures was much narrower than the estimation to be based on the Van Vleck’s formula for the second moment of the rigid lattice model where the electron spin can be considered as fixed; 11 mT of Gaussian line-width component. This suggests the possibility of exchange narrowing in the 1D organic-radical chains of the [(CLPOT)2-(TEMPO)1.0] IC. On the other hand, the ESR spectra of [(CLPOT)2-(MeO-TEMPO)0.41] IC (MeO-TEMPO = 4-methoxy-TEMPO) were reproduced by a superposition of major broad isotropic adsorption line and minor temperature-dependent modulated triplet component. This suggests that the IC has the part of 1D organic-radical chains and MeO-TEMPO molecules isolated in the CLPOT nanochannels
International Nuclear Information System (INIS)
We present a numerical study about the application of two versions of a second-degree iterative method for the solution of the sparse linear systems arising in the discretization of the 3D multi-group time-dependent Neutron Diffusion Equation. In addition, we propose some modifications to them, as well as a study of well-known preconditioning techniques in order to improve their convergence and accuracy when they are applied to a sequence of solutions in time of a real nuclear core transient. This is important for studies of stability and security of nuclear reactors. (authors)
Use of optimized 1D TOCSY NMR for improved quantitation and metabolomic analysis of biofluids
Energy Technology Data Exchange (ETDEWEB)
Sandusky, Peter [Eckerd College, Department of Chemistry (United States); Appiah-Amponsah, Emmanuel; Raftery, Daniel, E-mail: raftery@purdue.edu [Purdue University, Department of Chemistry (United States)
2011-04-15
One dimensional selective TOCSY experiments have been shown to be advantageous in providing improved data inputs for principle component analysis (PCA) (Sandusky and Raftery 2005a, b). Better subpopulation cluster resolution in the observed scores plots results from the ability to isolate metabolite signals of interest via the TOCSY based filtering approach. This report reexamines the quantitative aspects of this approach, first by optimizing the 1D TOCSY experiment as it relates to the measurement of biofluid constituent concentrations, and second by comparing the integration of 1D TOCSY read peaks to the bucket integration of 1D proton NMR spectra in terms of precision and accuracy. This comparison indicates that, because of the extensive peak overlap that occurs in the 1D proton NMR spectra of biofluid samples, bucket integrals are often far less accurate as measures of individual constituent concentrations than 1D TOCSY read peaks. Even spectral fitting approaches have proven difficult in the analysis of significantly overlapped spectral regions. Measurements of endogenous taurine made over a sample population of human urine demonstrates that, due to background signals from other constituents, bucket integrals of 1D proton spectra routinely overestimate the taurine concentrations and distort its variation over the sample population. As a result, PCA calculations performed using data matrices incorporating 1D TOCSY determined taurine concentrations produce better scores plot subpopulation cluster resolution.
The diffusion equation and the steady state. Chapter 2
International Nuclear Information System (INIS)
We shall now study the equations that govern the neutron field in a reactor. These equations are based on the concept of local neutron balance, which takes into account the reaction rates in an element of volume and the net leakage rates out of the volume. The reaction rates are written in terms of the local cross sections, assumed known from a preprocessed database (e.g., ENDF/B-VI). The starting equation is the Maxwell-Boltzmann transport equation, in its integro-differential form. The various approximations required to go from the transport equation to the neutron diffusion equation will be presented first, because all finite-reactor calculations are based on the diffusion approximation. We shall then discuss the multi-group formalism of the diffusion equations and study the mathematical properties of this equation in steady state. This preliminary step will allow us to derive in a more accurate way, in the next chapter, the reactor point-kinetics equations. In the diffusion approximation, neutrons diffuse from regions of high concentration to regions of low concentration, just as heat diffuses from regions of high temperature to those of low temperature, or, rather, as gas molecules diffuse to reduce spatial variations in concentration. While it is sufficiently accurate to treat the transport of gas molecules as a diffusion process, this approach is too limiting for neutron transport. In contrast to a gas, where collisions are very frequent, the cross sections for the interaction of neutrons with nuclei are relatively small, as we saw in chapter 1 (of the order of barns, i.e., 10-24cm2) . This implies that neutrons traverse appreciable distances (of the order of a centimetre) between collisions. This relatively long neutron mean free path, together with the heterogeneity of the physical medium, requires that a more complete treatment be carried out, taking account of variations in the angular distribution of neutron speed in the vicinity of highly absorbing
Parallel flow diffusion battery
Yeh, Hsu-Chi; Cheng, Yung-Sung
1984-08-07
A parallel flow diffusion battery for determining the mass distribution of an aerosol has a plurality of diffusion cells mounted in parallel to an aerosol stream, each diffusion cell including a stack of mesh wire screens of different density.
NIST Diffusion Data Center (Web, free access) The NIST Diffusion Data Center is a collection of over 14,100 international papers, theses, and government reports on diffusion published before 1980.
Charge Transport in 1-D Nanostructured CdS Dye Sensitized Solar Cell
International Nuclear Information System (INIS)
Charge transport in eosin yellow sensitized CdS 1-D nanostructures is studied. Direct conduction pathway for electron transport in nanowires enhances Voc in CdS nanowires compared to nanorods and nanoparticles. J-V characterization of nanowires results in improved efficiency of 0.184% due to fewer interparticle connections. Increase in Jsc is observed by coating CdS 1-D nanostructures on TiO2 substrate which reduces rate of recombination and photocorrosive nature of CdS photoanodes. Enhancement in efficiency up to 0.501% is achieved for CdS 1-D nanostructures DSSCs on TiO2 substrate.
Non-uniform black strings and the critical dimension in the $1/D$ expansion
Suzuki, Ryotaku; Tanabe, Kentaro
2015-01-01
Non-uniform black strings (NUBS) are studied by the large $D$ effective theory approach. By solving the near-horizon geometry in the $1/D$ expansion, we obtain the effective equation for the deformed horizon up to the next-to-next-to-leading order (NNLO) in $1/D$. We also solve the far-zone geometry by the Newtonian approximation. Matching the near and far zones, the thermodynamic variables are computed in the $1/D$ expansion. As the result, the large $D$ analysis gives a critical dimension $...
Development of a 1D neutron transport code employing the method of characteristics
International Nuclear Information System (INIS)
To investigate the 2D/1D fusion core analysis method, a 1D neutron transport problem solver, PEACH-ID, is developed. It is a code of method of characteristics (MOC), both the usual fiat-source step characteristics (SC) scheme and linear source (LS) approximation scheme are adopted for tracking calculation along the neutron flying trajectory. Exponential function interpolation table and fission source extrapolation are adopted as two major methods to accelerate the computational process. Numerical results demonstrate that PEACH-1D is accurate and efficient, and the proposed LS scheme is able to handle quite larger mesh division and deserves much more application in the MOC codes. (authors)
CD1d and invariant NKT cells at the human maternal–fetal interface
Boyson, Jonathan E.; Rybalov, Basya; Koopman, Louise A.; Exley, Mark; Balk, Steven P.; Racke, Frederick K.; Schatz, Frederick; Masch, Rachel; Wilson, S. Brian; Strominger, Jack L.
2002-01-01
Invariant CD1d-restricted natural killer T (iNKT) cells comprise a small, but significant, immunoregulatory T cell subset. Here, the presence of these cells and their CD1d ligand at the human maternal–fetal interface was investigated. Immunohistochemical staining of human decidua revealed the expression of CD1d on both villous and extravillous trophoblasts, the fetal cells that invade the maternal decidua. Decidual iNKT cells comprised 0.48% of the decidual CD3+ T cell population, a frequency...
International Nuclear Information System (INIS)
Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations
Estimation of multi-group cross section covariances for 235,238U, 239Pu, 241Am, 56Fe, 23Na and 27Al
International Nuclear Information System (INIS)
This paper presents the methodology used to estimate multi-group covariances for some major isotopes used in reactor physics. The starting point of this evaluation is the modelling of the neutron induced reactions based on nuclear reaction models with parameters. These latest are the vectors of uncertainties as they are absorbing uncertainties and correlation arising from the confrontation of nuclear reaction model to microscopic experiment. These uncertainties are then propagated towards multi-group cross sections. As major breakthroughs were then asked by nuclear reactor physicists to assess proper uncertainties to be used in applications, a solution is proposed by the use of integral experiment information at two different stages in the covariance estimation. In this paper, we will explain briefly the treatment of all type of uncertainties, including experimental ones (statistical and systematic) as well as those coming from validation of nuclear data on dedicated integral experiment (nuclear data oriented). We will illustrate the use of this methodology with various isotopes such as 235,238U, 239Pu, 241Am, 56Fe, 23Na and 27Al. (authors)
A diffusion-diffusion model for percutaneous drug absorption.
Kubota, K; Ishizaki, T
1986-08-01
Several theories describing percutaneous drug absorption have been proposed, incorporating the mathematical solutions of differential equations describing percutaneous drug absorption processes where the vehicle and skin are regarded as simple diffusion membranes. By a solution derived from Laplace transforms, the mean residence time MRT and the variance of the residence time VRT in the vehicle are expressed as simple elementary functions of the following five pharmacokinetic parameters characterizing the percutaneous drug absorption: kd, which is defined as the normalized diffusion coefficient of the skin, kc, which is defined as the normalized skin-capillary boundary clearance, the apparent length of diffusion of the skin 1d, the effective length of the vehicle lv, and the diffusion coefficient of the vehicle Dv. All five parameters can be obtained by the methods proposed here. Results of numerical computation indicate that: concentration-distance curves in the vehicle and skin approximate two curves which are simply expressed using trigonometric functions when sufficient time elapses after an ointment application; the most suitable condition for the assumption that the concentration of a drug in the uppermost epidermis can be considered unchanged is the case where the partition coefficient between vehicle and skin is small, and the constancy of drug concentration is even more valid when the effective length of the vehicle is large; and the amount of a drug in the vehicle or skin and the flow rate of the drug from vehicle into skin or from skin into blood becomes linear on a semilogarithmic scale, and the slopes of those lines are small when Dv is small, when the partition coefficient between vehicle and skin is small, when lv is large, or when kc is small. A simple simulation method is also proposed using a biexponential for the concentration-time curve for the skin near the skin-capillary boundary, that is, the flow rate-time curve for drug passing from skin
Energy Technology Data Exchange (ETDEWEB)
Kong, Chen; Lange, Jeffrey J.; Samovski, Dmitri [Department of Cell Biology and Physiology, Washington University School of Medicine, St. Louis, MO 63110 (United States); Su, Xiong [Department of Internal Medicine, Center for Human Nutrition Washington University School of Medicine, St. Louis, MO 63110 (United States); Liu, Jialiu [Department of Cell Biology and Physiology, Washington University School of Medicine, St. Louis, MO 63110 (United States); Sundaresan, Sinju [Department of Internal Medicine, Center for Human Nutrition Washington University School of Medicine, St. Louis, MO 63110 (United States); Stahl, Philip D., E-mail: pstahl@wustl.edu [Department of Cell Biology and Physiology, Washington University School of Medicine, St. Louis, MO 63110 (United States)
2013-05-03
Highlights: •Hominoid-specific oncogene TBC1D3 is targeted to plasma membrane by palmitoylation. •TBC1D3 is palmitoylated on two cysteine residues: 318 and 325. •TBC1D3 palmitoylation governs growth factors-induced TBC1D3 degradation. •Post-translational modifications may regulate oncogenic properties of TBC1D3. -- Abstract: Expression of the hominoid-specific oncoprotein TBC1D3 promotes enhanced cell growth and proliferation by increased activation of signal transduction through several growth factors. Recently we documented the role of CUL7 E3 ligase in growth factors-induced ubiquitination and degradation of TBC1D3. Here we expanded our study to discover additional molecular mechanisms that control TBC1D3 protein turnover. We report that TBC1D3 is palmitoylated on two cysteine residues: 318 and 325. The expression of double palmitoylation mutant TBC1D3:C318/325S resulted in protein mislocalization and enhanced growth factors-induced TBC1D3 degradation. Moreover, ubiquitination of TBC1D3 via CUL7 E3 ligase complex was increased by mutating the palmitoylation sites, suggesting that depalmitoylation of TBC1D3 makes the protein more available for ubiquitination and degradation. The results reported here provide novel insights into the molecular mechanisms that govern TBC1D3 protein degradation. Dysregulation of these mechanisms in vivo could potentially result in aberrant TBC1D3 expression and promote oncogenesis.
Non-uniform black strings and the critical dimension in the $1/D$ expansion
Suzuki, Ryotaku
2015-01-01
Non-uniform black strings (NUBS) are studied by the large $D$ effective theory approach. By solving the near-horizon geometry in the $1/D$ expansion, we obtain the effective equation for the deformed horizon up to the next-to-next-to-leading order (NNLO) in $1/D$. We also solve the far-zone geometry by the Newtonian approximation. Matching the near and far zones, the thermodynamic variables are computed in the $1/D$ expansion. As the result, the large $D$ analysis gives a critical dimension $D_*\\simeq13.5$ at which the translation-symmetry-breaking phase transition changes between first and second order. This value of $D_*$ agrees perfectly, within the precision of the $1/D$ expansion, with the result previously obtained by E. Sorkin through the numerical resolution. We also compare our NNLO results for the thermodynamics of NUBS to earlier numerical calculations, and find good agreement within the expected precision.
Non-uniform black strings and the critical dimension in the 1/D expansion
Suzuki, Ryotaku; Tanabe, Kentaro
2015-10-01
Non-uniform black strings (NUBS) are studied by the large D effective theory approach. By solving the near-horizon geometry in the 1 /D expansion, we obtain the effective equation for the deformed horizon up to the next-to-next-to-leading order (NNLO) in 1 /D. We also solve the far-zone geometry by the Newtonian approximation. Matching the near and far zones, the thermodynamic variables are computed in the 1 /D expansion. As the result, the large D analysis gives a critical dimension D * ≃ 13 .5 at which the translation-symmetry-breaking phase transition changes between first and second order. This value of D * agrees perfectly, within the precision of the 1 /D expansion, with the result previously obtained by E. Sorkin through the numerical resolution. We also compare our NNLO results for the thermodynamics of NUBS to earlier numerical calculations, and find good agreement within the expected precision.
Grinberg, L; Cheever, E; Anor, T; Madsen, J R; Karniadakis, G E
2011-01-01
We compare results from numerical simulations of pulsatile blood flow in two patient-specific intracranial arterial networks using one-dimensional (1D) and three-dimensional (3D) models. Specifically, we focus on the pressure and flowrate distribution at different segments of the network computed by the two models. Results obtained with 1D and 3D models with rigid walls show good agreement in massflow distribution at tens of arterial junctions and also in pressure drop along the arteries. The 3D simulations with the rigid walls predict higher amplitude of the flowrate and pressure temporal oscillations than the 1D simulations with compliant walls at various segments even for small time-variations in the arterial cross-sectional areas. Sensitivity of the flow and pressure with respect to variation in the elasticity parameters is investigated with the 1D model. PMID:20661645
International Nuclear Information System (INIS)
The neutron diffusion programs D3D and D3E solve the multigroup diffusion equations by the technique, outer iteration for the distribution of the source and inner iterations for the distribution of the group fluxes. The inner iterations consist of two nested block overrelaxations: with the planes as the blocks within the energy groups (called group iterations) and with tupels of lines as blocks within the planes (called plane iterations). The convergence behaviour of the outer iteration depends on the number of the group iterations, the behaviour of the group iterations on the number of the plane iterations. These numbers of the inner iterations are determined by upper bounds for the relative flux changes and, therefore, vary from outer iteration to outer iteration causing non-monotonic convergence behaviour and irregular jumps of numerical values of the outer iteration. This is demostrated by means of examples taken from the literature. In addition, ways to overcome such difficulties are indicated. (orig.)
1D model for the dynamics and expansion of elongated Bose-Einstein condensates
Massignan, Pietro; Modugno, Michele
2002-01-01
We present a 1D effective model for the evolution of a cigar-shaped Bose-Einstein condensate in time dependent potentials whose radial component is harmonic. We apply this model to investigate the dynamics and expansion of condensates in 1D optical lattices, by comparing our predictions with recent experimental data and theoretical results. We also discuss negative-mass effects which could be probed during the expansion of a condensate moving in an optical lattice.
Kayserili Karabey, Hülya; Schmidts, Miriam; Hou, Yuqing; Cortes, Claudio R.; Mans, Dorus A.; Huber, Celine; Boldt, Karsten; Patel, Mitali; van Reeuwijk, Jeroen; Plaza, Jean-Marc; van Beersum, Sylvia E. C.; Yap, Zhi Min; Letteboer, Stef J. F.; Taylor, S. Paige; Herridge, Warren; Johnson, Colin A.; Scambler, Peter J.; Ueffing, Marius; Krakow, Deborah; King, Stephen M.; Beales, Philip L.; Al-Gazali, Lihadh; Wicking, Carol; Cormier-Daire, Valerie; Roepman, Ronald; Mitchison, Hannah M.; Witman, George B.
2015-01-01
The analysis of individuals with ciliary chondrodysplasias can shed light on sensitive mechanisms controlling ciliogenesis and cell signalling that are essential to embryonic development and survival. Here we identify TCTEX1D2 mutations causing Jeune asphyxiating thoracic dystrophy with partially penetrant inheritance. Loss of TCTEX1D2 impairs retrograde intraflagellar transport (IFT) in humans and the protist Chlamydomonas, accompanied by destabilization of the retrograde IFT dynein motor. W...
Optimization of a cyclic peptide inhibitor of Ser/Thr phosphatase PPM1D (Wip1).
Hayashi, Ryo; Tanoue, Kan; Durell, Stewart R; Chatterjee, Deb K; Jenkins, Lisa M Miller; Appella, Daniel H; Appella, Ettore
2011-05-31
PPM1D (PP2Cδ or Wip1) was identified as a wild-type p53-induced Ser/Thr phosphatase that accumulates after DNA damage and classified into the PP2C family. It dephosphorylates and inactivates several proteins critical for cellular stress responses, including p38 MAPK, p53, and ATM. Furthermore, PPM1D is amplified and/or overexpressed in a number of human cancers. Thus, inhibition of its activity could constitute an important new strategy for therapeutic intervention to halt the progression of several different cancers. Previously, we reported the development of a cyclic thioether peptide with low micromolar inhibitory activity toward PPM1D. Here, we describe important improvements in the inhibitory activity of this class of cyclic peptides and also present a binding model based upon the results. We found that specific interaction of an aromatic ring at the X1 position and negative charge at the X5 and X6 positions significantly increased the inhibitory activity of the cyclic peptide, with the optimized molecule having a K(i) of 110 nM. To the best of our knowledge, this represents the highest inhibitory activity reported for an inhibitor of PPM1D. We further developed an inhibitor selective for PPM1D over PPM1A with a K(i) of 2.9 μM. Optimization of the cyclic peptide and mutagenesis experiments suggest that a highly basic loop unique to PPM1D is related to substrate specificity. We propose a new model for the catalytic site of PPM1D and inhibition by the cyclic peptides that will be useful both for the subsequent design of PPM1D inhibitors and for identification of new substrates. PMID:21528848
Momentum Conservation Implies Anomalous Energy Transport in 1D Classical Lattices
International Nuclear Information System (INIS)
Under quite general conditions, we prove that for classical many-body lattice Hamiltonians in one dimension (1D) total momentum conservation implies anomalous conductivity in the sense of the divergence of the Kubo expression for the coefficient of thermal conductivity, κ . Our results provide rigorous confirmation and explanation of many of the existing ''surprising'' numerical studies of anomalous conductivity in 1D classical lattices, including the celebrated Fermi-Pasta-Ulam problem. (c) 2000 The American Physical Society
Momentum conservation implies anomalous energy transport in 1d classical lattices
Prosen, T; Prosen, Tomaz; Campbell, David K.
2000-01-01
Under quite general conditions, we prove that for classical many-body lattice Hamiltonians in one dimension (1D) total momentum conservation implies anomalous conductivity in the sense of the divergence of the Kubo expression for the coefficient of thermal conductivity, $\\kappa$. Our results provide rigorous confirmation and explanation of many of the existing ``surprising'' numerical studies of anomalous conductivity in 1D classical lattices, including the celebrated Fermi-Pasta-Ulam problem.
User's manual of the REFLA-1D/MODE4 reflood thermo-hydrodynamic analysis code
International Nuclear Information System (INIS)
REFLA-1D/MODE4 code has been developed by incorporating local power effect model and fuel temperature profile effect model into REFLA-1D/MODE3 code. This code can calculate the temperature transient of local rod by considering radial power profile effect in core and simulate the thermal characteristics of the nuclear fuel rod. This manual describes the outline of incorporated models, modification of the code with incorporating models and provides application information required to utilize the code. (author)
*609850 TBC1 DOMAIN FAMILY, MEMBER 1; TBC1D1 [OMIM
Lifescience Database Archive (English)
Full Text Available FIELD NO 609850 FIELD TI 609850 TBC1 DOMAIN FAMILY, MEMBER 1; TBC1D1 ;;KIAA1108 FIELD TX DESCRIP ... o obesity, see BMIQ7 (608410). ANIMAL MODEL In the lean ... Swiss Jim Lambert (SJL) strain of obesity-resistan ... .; Joost, H.-G.; Al-Hasani, H.: Tbc1d1 mutation in lean ... mouse strain confers lean ness and protects from di ...
Protective mucosal immunity mediated by epithelial CD1d and IL-10.
Olszak, Torsten; Neves, Joana F; Dowds, C Marie; Baker, Kristi; Glickman, Jonathan; Davidson, Nicholas O; Lin, Chyuan-Sheng; Jobin, Christian; Brand, Stephan; Sotlar, Karl; Wada, Koichiro; Katayama, Kazufumi; Nakajima, Atsushi; Mizuguchi, Hiroyuki; Kawasaki, Kunito; Nagata, Kazuhiro; Müller, Werner; Snapper, Scott B; Schreiber, Stefan; Kaser, Arthur; Zeissig, Sebastian; Blumberg, Richard S
2014-05-22
The mechanisms by which mucosal homeostasis is maintained are of central importance to inflammatory bowel disease. Critical to these processes is the intestinal epithelial cell (IEC), which regulates immune responses at the interface between the commensal microbiota and the host. CD1d presents self and microbial lipid antigens to natural killer T (NKT) cells, which are involved in the pathogenesis of colitis in animal models and human inflammatory bowel disease. As CD1d crosslinking on model IECs results in the production of the important regulatory cytokine interleukin (IL)-10 (ref. 9), decreased epithelial CD1d expression--as observed in inflammatory bowel disease--may contribute substantially to intestinal inflammation. Here we show in mice that whereas bone-marrow-derived CD1d signals contribute to NKT-cell-mediated intestinal inflammation, engagement of epithelial CD1d elicits protective effects through the activation of STAT3 and STAT3-dependent transcription of IL-10, heat shock protein 110 (HSP110; also known as HSP105), and CD1d itself. All of these epithelial elements are critically involved in controlling CD1d-mediated intestinal inflammation. This is demonstrated by severe NKT-cell-mediated colitis upon IEC-specific deletion of IL-10, CD1d, and its critical regulator microsomal triglyceride transfer protein (MTP), as well as deletion of HSP110 in the radioresistant compartment. Our studies thus uncover a novel pathway of IEC-dependent regulation of mucosal homeostasis and highlight a critical role of IL-10 in the intestinal epithelium, with broad implications for diseases such as inflammatory bowel disease. PMID:24717441
Protective mucosal immunity mediated by epithelial CD1d and IL-10
Olszak, Torsten; Neves, Joana F.; Dowds, C. Marie; Baker, Kristi; Glickman, Jonathan; Davidson, Nicholas O; Lin, Chyuan-Sheng; Jobin, Christian; Brand, Stephan; Sotlar, Karl; Wada, Koichiro; Katayama, Kazufumi; Nakajima, Atsushi; Mizuguchi, Hiroyuki; Kawasaki, Kunito
2014-01-01
The mechanisms by which mucosal homeostasis is maintained are of central importance to inflammatory bowel disease. Critical to these processes is the intestinal epithelial cell (IEC), which regulates immune responses at the interface between the commensal microbiota and the host1,2. CD1d presents self and microbial lipid antigens to natural killer T (NKT) cells, which are involved in the pathogenesis of colitis in animal models and human inflammatory bowel disease3–8. As CD1d crosslinking on ...
Simple model of the density of states in 1D photonic crystal
Rudziński, Adam; Tyszka-Zawadzka, Anna; Szczepański, Paweł
2010-01-01
In this paper, we present a simple, yet versatile, analytical model of one-dimensional photonic crystal (1D PC). In our theoretical model, we take into account direction of propagation and therefore do not neglect anisotropic nature of photonic crystals. We derive analytical expressions for mode spectrum and density of states in 1D photonic crystal. With those formulas, we obtain mode spectrum characteristics, which depict formation of photonic band gap and reveal properties of photonic cryst...
One-Dimensional (1D) ZnS Nanomaterials and Nanostructures
Institute of Scientific and Technical Information of China (English)
Xiaosheng FANG; Lide ZHANG
2006-01-01
One-dimensional (1D) nanomaterials and nanostructures have received much attention due to their potential interest for understanding fundamental physical concepts and for applications in constructing nanoscale electric and optoelectronic devices. Zinc sulfide (ZnS) is an important semiconductor compound of Ⅱ-Ⅵ group,and the synthesis of 1D ZnS nanomaterials and nanostructures has been of growing interest owing to their promising application in nanoscale optoelectronic devices. This paper reviews the recent progress on 1D ZnS nanomaterials and nanostructures, including nanowires, nanowire arrays, nanorods, nanobelts or nanoribbons,nanocables, and hierarchical nanostructures etc. This article begins with a survey of various methods that have been developed for generating 1D nanomaterials and nanostructures, and then mainly focuses on structures,synthesis, characterization, formation mechanisms and optical property tuning, and luminescence mechanisms of 1D ZnS nanomaterials and nanostructures. Finally, this review concludes with personal views towards future research on 1D ZnS nanomaterials and nanostructures.
Dynamics of reactions O((1)D)+C(6)H(6) and C(6)D(6).
Chen, Hui-Fen; Liang, Chi-Wei; Lin, Jim J; Lee, Yuan-Pern; Ogilvie, J F; Xu, Z F; Lin, M C
2008-11-01
The reaction between O((1)D) and C(6)H(6) (or C(6)D(6)) was investigated with crossed-molecular-beam reactive scattering and time-resolved Fourier-transform infrared spectroscopy. From the crossed-molecular-beam experiments, four product channels were identified. The major channel is the formation of three fragments CO+C(5)H(5)+H; the channels for formation of C(5)H(6)+CO and C(6)H(5)O+H from O((1)D)+C(6)H(6) and OD+C(6)D(5) from O((1)D)+C(6)D(6) are minor. The angular distributions for the formation of CO and H indicate a mechanism involving a long-lived collision complex. Rotationally resolved infrared emission spectra of CO (1ratio of [CO]/[OH]=2.1+/-0.4 for O((1)D)+C(6)H(6) and [CO]/[OD]>2.9 for O((1)D)+C(6)D(6) is consistent with the expectation for an abstraction reaction. The mechanism of the reaction may be understood from considering the energetics of the intermediate species and transition states calculated at the G2M(CC5) level of theory for the O((1)D)+C(6)H(6) reaction. The experimentally observed branching ratios and deuterium isotope effect are consistent with those predicted from calculations. PMID:19045343
Novel hollow mesoporous 1D TiO2 nanofibers as photovoltaic and photocatalytic materials
Zhang, Xiang; Thavasi, Velmurugan; Mhaisalkar, S. G.; Ramakrishna, Seeram
2012-02-01
Hollow mesoporous one dimensional (1D) TiO2 nanofibers are successfully prepared by co-axial electrospinning of a titanium tetraisopropoxide (TTIP) solution with two immiscible polymers; polyethylene oxide (PEO) and polyvinylpyrrolidone (PVP) using a core-shell spinneret, followed by annealing at 450 °C. The annealed mesoporous TiO2 nanofibers are found to having a hollow structure with an average diameter of 130 nm. Measurements using the Brunauer-Emmett-Teller (BET) method reveal that hollow mesoporous TiO2 nanofibers possess a high surface area of 118 m2 g-1 with two types of mesopores; 3.2 nm and 5.4 nm that resulted from gaseous removal of PEO and PVP respectively during annealing. With hollow mesoporous TiO2 nanofibers as the photoelectrode in dye sensitized solar cells (DSSC), the solar-to-current conversion efficiency (η) and short circuit current (Jsc) are measured as 5.6% and 10.38 mA cm-2 respectively, which are higher than those of DSSC made using regular TiO2 nanofibers under identical conditions (η = 4.2%, Jsc = 8.99 mA cm-2). The improvement in the conversion efficiency is mainly attributed to the higher surface area and mesoporous TiO2 nanostructure. It facilitates the adsorption of more dye molecules and also promotes the incident photon to electron conversion. Hollow mesoporous TiO2 nanofibers with close packing of grains and crystals intergrown with each other demonstrate faster electron diffusion, and longer electron recombination time than regular TiO2 nanofibers as well as P25 nanoparticles. The surface effect of hollow mesoporous TiO2 nanofibers as a photocatalyst for the degradation of rhodamine dye was also investigated. The kinetic study shows that the hollow mesoporous surface of the TiO2 nanofibers influenced its interactions with the dye, and resulted in an increased catalytic activity over P25 TiO2 nanocatalysts.Hollow mesoporous one dimensional (1D) TiO2 nanofibers are successfully prepared by co-axial electrospinning of a titanium
From Reaction-Diffusion Systems to Confined Brownian Motion
Martens, Steffen
2016-01-01
In this note, we demonstrated for the first time that one can derive an expression for the effective diffusion coefficient, equal to the Lifson-Jackson formula, using a subsequent homogenization of the 1D reaction-diffusion-advection equation. The latter has been derived by applying asymptotic perturbation analysis to the underlying 3D reaction-diffusion equation with spatially dependent no-flux boundary conditions and incorporates the effects of boundary interactions on the reactants via a boundary-induced advection term [S. Martens et al, Phys. Rev. E 91, 022902 (2015)].
Riedl, C; Akopov, Z; Amarian, M; Ammosov, V V; Andrus, A; Aschenauer, E C; Augustyniak, W; Avakian, R; Avetisian, A; Avetissian, E; Bailey, P; Baturin, V; Baumgarten, C; Beckmann, M; Belostotskii, S; Bernreuther, S; Bianchi, N; Blok, H P; Böttcher, Helmut B; Borisov, A; Bouwhuis, M; Brack, J; Brüll, A; Bryzgalov, V V; Capitani, G P; Chiang, H C; Ciullo, G; Contalbrigo, M; Dalpiaz, P F; De Leo, R; De Nardo, L; De Sanctis, E; Devitsin, E G; Di Nezza, P; Düren, M; Ehrenfried, M; Elalaoui-Moulay, A; Elbakian, G M; Ellinghaus, F; Elschenbroich, U; Ely, J; Fabbri, R; Fantoni, A; Feshchenko, A; Felawka, L; Fox, B; Franz, J; Frullani, S; Gärber, Y; Gapienko, G; Gapienko, V; Garibaldi, F; Garrow, K; Garutti, E; Gaskell, D; Gavrilov, G E; Karibian, V; Graw, G; Grebenyuk, O; Greeniaus, L G; Hafidi, K; Hartig, M; Hasch, D; Heesbeen, D; Henoch, M; Hertenberger, R; Hesselink, W H A; Hillenbrand, A; Hoek, M; Holler, Y; Hommez, B; Iarygin, G; Ivanilov, A; Izotov, A; Jackson, H E; Jgoun, A; Kaiser, R; Kinney, E; Kiselev, A; Königsmann, K C; Kopytin, M; Korotkov, V A; Kozlov, V; Krauss, B; Krivokhizhin, V G; Lagamba, L; Lapikas, L; Laziev, A; Lenisa, P; Liebing, P; Lindemann, T; Lipka, K; Lorenzon, W; Lü, J; Maiheu, B; Makins, N C R; Marianski, B; Marukyan, H O; Masoli, F; Mexner, V; Meyners, N; Miklukho, O; Miller, C A; Miyachi, Y; Muccifora, V; Nagaitsev, A; Nappi, E; Naryshkin, Yu; Nass, A; Negodaev, M A; Nowak, Wolf-Dieter; Oganessyan, K; Ohsuga, H; Orlandi, G; Pickert, N; Potashov, S Yu; Potterveld, D H; Raithel, M; Reggiani, D; Reimer, P E; Reischl, A; Reolon, A R; Rith, K; Airapetian, A; Rosner, G; Rostomyan, A; Rubacek, L; Ryckbosch, D; Salomatin, Yu I; Sanjiev, I; Savin, I; Scarlett, C; Schäfer, A; Schill, C; Schnell, G; Schüler, K P; Schwind, A; Seele, J; Seidl, R; Seitz, B; Shanidze, R G; Shearer, C; Shibata, T A; Shutov, V B; Simani, M C; Sinram, K; Stancari, M D; Statera, M; Steffens, E; Steijger, J J M; Stewart, J; Stösslein, U; Tait, P; Tanaka, H; Taroian, S P; Tchuiko, B; Terkulov, A R; Tkabladze, A V; Trzcinski, A; Tytgat, M; Vandenbroucke, A; Van der Nat, P B; van der Steenhoven, G; Vetterli, Martin C; Vikhrov, V; Vincter, M G; Visser, J; Vogel, C; Vogt, M; Volmer, J; Weiskopf, C; Wendland, J; Wilbert, J; Ybeles-Smit, G V; Yen, S; Zihlmann, B; Zohrabyan, H G; Zupranski, P; Riedl, Caroline
2005-01-01
Final HERMES results on the proton, deuteron and neutron structure function g1 are presented in the kinematic range 0.0021
SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations
Energy Technology Data Exchange (ETDEWEB)
Adams, C. H.
1976-07-01
This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center.
Three-dimensional nodal diffusion and transport theory methods for hexagonal z-geometry
International Nuclear Information System (INIS)
This paper describes nodal methods for the solution of multigroup neutron diffusion and transport theory equations in three-dimensional hexagonal z-geometry. The code HEXNOD allows the accurate and efficient calculation of 3D problems for fast reactors and high converter light water reactors. A unique capability of HEXNOD is the solution of global 3D neutron transport problems for fast reactors with very small computing times. The accuracy of the nodal transport approximation is demonstrated by comparison with 3D Monte Carlo calculations. Based on the numerical results it is concluded that the code HEXNOD is well suited for 3D routine analysis of fast reactors and, in particular, as the neutronics module of the generalized quasistatic kinetics program HEXNODYN which is currently being developed as part of the European accident code EAC-2
Bi, Zhen; BenTov, Yoni; Xu, Cenke
2016-01-01
Motivated by recent studies of symmetry protected topological (SPT) phases, we explore the possible gapless quantum disordered phases in the $(2+1)d$ nonlinear sigma model defined on the Grassmannian manifold $\\frac{U(N)}{U(n)\\times U(N - n)}$ with a Wess-Zumino-Witten (WZW) term at level $k$, which is the effective low energy field theory of the boundary of certain $(3+1)d$ SPT states. With $k = 0$, this model has a well-controlled large-$N$ limit, $i.e.$ its renormalization group equations can be computed exactly with large-$N$. However, with the WZW term, the large-$N$ and large-$k$ limit alone is not sufficient for a reliable study of the nature of the quantum disordered phase. We demonstrate that at least for $n = 1$, through a combined large-$N$, large-$k$ and $3-\\epsilon$ generalization, a stable fixed point in the quantum disordered phase can be reliably located, which corresponds to a $(2+1)d$ strongly interacting conformal field theory. Extension of our method to $n > 1$ will also be discussed.
The drift-diffusion limit of thermal neutrons. Theoretical and numerical results
International Nuclear Information System (INIS)
This study presents a new approach to modeling neutron thermalization which can greatly simplify nuclear reactor simulations. A drift-diffusion equation is derived through an asymptotic analysis of the 1-D energy-dependent transport equation. Two test problems were used to provide numerical comparisons of this drift-diffusion approximation to MCNP6 calculations. Both of these test problems include a large 1-D graphite moderator with a steady-state temperature gradient. The results demonstrated that the drift-diffusion is able to capture energy-dependent effects and can successfully be used for modeling neutron thermalization in a large moderating material. (author)
Benchmarks and models for 1-D radiation transport in stochastic participating media
Energy Technology Data Exchange (ETDEWEB)
Miller, D S
2000-08-21
Benchmark calculations for radiation transport coupled to a material temperature equation in a 1-D slab and 1-D spherical geometry binary random media are presented. The mixing statistics are taken to be homogeneous Markov statistics in the 1-D slab but only approximately Markov statistics in the 1-D sphere. The material chunk sizes are described by Poisson distribution functions. The material opacities are first taken to be constant and then allowed to vary as a strong function of material temperature. Benchmark values and variances for time evolution of the ensemble average of material temperature energy density and radiation transmission are computed via a Monte Carlo type method. These benchmarks are used as a basis for comparison with three other approximate methods of solution. One of these approximate methods is simple atomic mix. The second approximate model is an adaptation of what is commonly called the Levermore-Pomraning model and which is referred to here as the standard model. It is shown that recasting the temperature coupling as a type of effective scattering can be useful in formulating the third approximate model, an adaptation of a model due to Su and Pomraning which attempts to account for the effects of scattering in a stochastic context. This last adaptation shows consistent improvement over both the atomic mix and standard models when used in the 1-D slab geometry but shows limited improvement in the 1-D spherical geometry. Benchmark values are also computed for radiation transmission from the 1-D sphere without material heating present. This is to evaluate the performance of the standard model on this geometry--something which has never been done before. All of the various tests demonstrate the importance of stochastic structure on the solution. Also demonstrated are the range of usefulness and limitations of a simple atomic mix formulation.
VIM4.0, Stead-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections
International Nuclear Information System (INIS)
1 - Description of program or function: VIM solves the steady-state neutron or photon transport problem in any detailed three-dimensional geometry using either continuous energy-dependent ENDF nuclear data or multigroup cross sections. Neutron transport is carried out in a criticality mode, or in a fixed source mode (optionally incorporating subcritical multiplication). Photon transport is simulated in the fixed source mode. The geometry options are infinite medium, combinatorial geometry, and hexagonal or rectangular lattices of combinatorial geometry unit cells, and rectangular lattices of cells of assembled plates. Boundary conditions include vacuum, specular and white reflection, and periodic boundaries for reactor cell calculations. The VIM 4.0 distribution includes data from ENDF/B-IV, ENDF/B-V, ENDF/B-VI and JEF2.2. Binary sequential data libraries for use with the code system on IBM or Sun workstations are included. ASCII data libraries and a convenient means to convert them to binary on a target machine are included for users on other systems. In addition to be included in the RSICC distribution files, the VIM User Guide is available on the developer's web site http://www.ra.anl.gov/vimguide/. 2 - Methods:VIM uses standard Monte Carlo methods for particle tracking with several optional variance-reduction techniques. These include splitting/Russian roulette, non-terminating absorption with non-analog weight cutoff energy. The keff is determined by the optimum linear combinations of two of the three eigenvalue estimates - analog, collision, and track length. Resonance and smooth cross sections are specified pointwise with linear-linear interpolation, frequently with many thousands of energy points. Unresolved resonances are described by the probability table method, which allows the statistical nature of the evaluated resonance cross sections to be incorporated naturally into self-shielding. Neutron interactions are elastic, inelastic and thermal scattering
ZZ-IRAN-LIB, Multigroup Neutron Gamma Cross-Section Library for 33 Elements in ANISN Format
International Nuclear Information System (INIS)
Description of program or function: - Format: ANISN/PC; - Number of groups: IRAN1.LIB (22 neutrons 18 gammas); IRAN2.LIB (17 neutrons, 18 gammas); IRAN3.LIB (7 neutrons, 18 gammas); IRAN4.LIB (7 neutrons, 6 gammas); IRAN5.LIB (5 neutrons, 4 gammas); IRAN6.LIB (2 neutrons, 4 gammas). - Nuclides: H-1, H-2, Li-6, Li-7, Be-9, B-10, C-12, N-14, O-16, Na, Mg, Al-27, Si, K, V, Cr, Mn-55, Fe, Ni, Nb-93, Pb, U-235, U-238, Pu-239, Ba-134, Ba-135, Ba-136, Ba-137, Ba-140, Bi-209, Ca-nat, Zr-nat, Cd-nat. - Origin: VITAMIN-4C; ENDF/B-IV and V, and JENDL-3. Weighting spectrum: IRAN.LIB's data (microscopic cross sections) is suitable for neutron, gamma and coupled neutron- gamma transport calculation (shielding). It is intended for use by the multigroup discrete ordinates code ANISN/PC (CCC-0514) using anisotropic scattering by Legendre expansion up to order P-3. IRAN.LIB is a collection of libraries for elements (H-1; H-2; Li-6; Li-7; Be-9; B-10; C-12; N-14; O-16; Na; Mg; Al-27; Si; K; V; Cr; Mn-55; Fe; Ni; Nb-93; Pb; U-235; U-238; Pu-239; Ba-134; Ba-135; Ba-136; Ba-137; Ba-140; Bi-209; Ca-nat; Zr-nat; Cd-nat) in ISOTXS format with a different group structure for each library, that is, IRAN1.LIB (22 neutrons, 18 gammas); IRAN2.LIB (17 neutrons, 18 gammas); IRAN3.LIB (7 neutrons, 18 gammas); IRAN4.LIB (7 neutrons, 6 gammas); IRAN5.LIB (5 neutrons, 4 gammas); IRAN6.LIB (2 neutrons, 4 gammas). 2 - Method of solution: The basic data sources were VITAMIN-4C; ENDF/B-IV and V and JENDL-3. Most of the data were taken from VITAMIN-4C (H-1, H-2, Li-6, Li-7, Be-9, B-10, C-12, N-14, O-16, Na, Mg, Al-27, Si, K, V, Cr, Mn-55, Fe, Ni, Nb-93, Pb, U-235, U-238, Pu-239) and collapsing them using AMPX-II modules. The AJAX module extracts the neutron cross sections of desired elements from VITAMIN-4C. CHOX module combines master neutron, gamma production and gamma interaction libraries into a coupled neutron-gamma library. MALOCS module collapses the cross sections into given energy groups and
Computational simulations of vorticity enhanced diffusion
Vold, Erik L.
1999-11-01
Computer simulations are used to investigate a phenomenon of vorticity enhanced diffusion (VED), a net transport and mixing of a passive scalar across a prescribed vortex flow field driven by a background gradient in the scalar quantity. The central issue under study here is the increase in scalar flux down the gradient and across the vortex field. The numerical scheme uses cylindrical coordinates centered with the vortex flow which allows an exact advective solution and 1D or 2D diffusion using simple numerical methods. In the results, the ratio of transport across a localized vortex region in the presence of the vortex flow over that expected for diffusion alone is evaluated as a measure of VED. This ratio is seen to increase dramatically while the absolute flux across the vortex decreases slowly as the diffusion coefficient is decreased. Similar results are found and compared for varying diffusion coefficient, D, or vortex rotation time, τv, for a constant background gradient in the transported scalar vs an interface in the transported quantity, and for vortex flow fields constant in time vs flow which evolves in time from an initial state and with a Schmidt number of order unity. A simple analysis shows that for a small diffusion coefficient, the flux ratio measure of VED scales as the vortex radius over the thickness for mass diffusion in a viscous shear layer within the vortex characterized by (Dτv)1/2. The phenomenon is linear as investigated here and suggests that a significant enhancement of mixing in fluids may be a relatively simple linear process. Discussion touches on how this vorticity enhanced diffusion may be related to mixing in nonlinear turbulent flows.
Hybrid finite element and Brownian dynamics method for diffusion-controlled reactions
Bauler, Patricia; Huber, Gary A.; McCammon, J. Andrew
2012-01-01
Diffusion is often the rate determining step in many biological processes. Currently, the two main computational methods for studying diffusion are stochastic methods, such as Brownian dynamics, and continuum methods, such as the finite element method. This paper proposes a new hybrid diffusion method that couples the strengths of each of these two methods. The method is derived for a general multidimensional system, and is presented using a basic test case for 1D linear and radially symmetri...
A Mathematical Model of T1D Acceleration and Delay by Viral Infection.
Moore, James R; Adler, Fred
2016-03-01
Type 1 diabetes (T1D) is often triggered by a viral infection, but the T1D prevalence is rising among populations that have a lower exposure to viral infection. In an animal model of T1D, the NOD mouse, viral infection at different ages may either accelerate or delay disease depending on the age of infection and the type of virus. Viral infection may affect the progression of T1D via multiple mechanisms: triggering inflammation, bystander activation of self-reactive T-cells, inducing a competitive immune response, or inducing a regulatory immune response. In this paper, we create mathematical models of the interaction of viral infection with T1D progression, incorporating each of these four mechanisms. Our goal is to understand how each viral mechanism interacts with the age of infection. The model predicts that each viral mechanism has a unique pattern of interaction with disease progression. Viral inflammation always accelerates disease, but the effect decreases with age of infection. Bystander activation has little effect at younger ages and actually decreases incidence at later ages while accelerating disease in mice that do get the disease. A competitive immune response to infection can decrease incidence at young ages and increase it at older ages, with the effect decreasing over time. Finally, an induced Treg response decreases incidence at any age of infection, but the effect decreases with age. Some of these patterns resemble those seen experimentally. PMID:27030351
Epitaxial 1D electron transport layers for high-performance perovskite solar cells.
Han, Gill Sang; Chung, Hyun Suk; Kim, Dong Hoe; Kim, Byeong Jo; Lee, Jin-Wook; Park, Nam-Gyu; Cho, In Sun; Lee, Jung-Kun; Lee, Sangwook; Jung, Hyun Suk
2015-10-01
We demonstrate high-performance perovskite solar cells with excellent electron transport properties using a one-dimensional (1D) electron transport layer (ETL). The 1D array-based ETL is comprised of 1D SnO2 nanowires (NWs) array grown on a F:SnO2 transparent conducting oxide substrate and rutile TiO2 nanoshells epitaxially grown on the surface of the 1D SnO2 NWs. The optimized devices show more than 95% internal quantum yield at 750 nm, and a power conversion efficiency (PCE) of 14.2%. The high quantum yield is attributed to dramatically enhanced electron transport in the epitaxial TiO2 layer, compared to that in conventional nanoparticle-based mesoporous TiO2 (mp-TiO2) layers. In addition, the open space in the 1D array-based ETL increases the prevalence of uniform TiO2/perovskite junctions, leading to reproducible device performance with a high fill factor. This work offers a method to achieve reproducible, high-efficiency perovskite solar cells with high-speed electron transport. PMID:26324759
Deconvolution of Complex 1D NMR Spectra Using Objective Model Selection.
Directory of Open Access Journals (Sweden)
Travis S Hughes
Full Text Available Fluorine (19F NMR has emerged as a useful tool for characterization of slow dynamics in 19F-labeled proteins. One-dimensional (1D 19F NMR spectra of proteins can be broad, irregular and complex, due to exchange of probe nuclei between distinct electrostatic environments; and therefore cannot be deconvoluted and analyzed in an objective way using currently available software. We have developed a Python-based deconvolution program, decon1d, which uses Bayesian information criteria (BIC to objectively determine which model (number of peaks would most likely produce the experimentally obtained data. The method also allows for fitting of intermediate exchange spectra, which is not supported by current software in the absence of a specific kinetic model. In current methods, determination of the deconvolution model best supported by the data is done manually through comparison of residual error values, which can be time consuming and requires model selection by the user. In contrast, the BIC method used by decond1d provides a quantitative method for model comparison that penalizes for model complexity helping to prevent over-fitting of the data and allows identification of the most parsimonious model. The decon1d program is freely available as a downloadable Python script at the project website (https://github.com/hughests/decon1d/.
Possible Dimensional Crossover to 1D of ^3He Fluid in Nanochannels Observed in Susceptibilities
Matsushita, Taku; Kurebayashi, Katsuya; Shibatsuji, Ryosuke; Hieda, Mitsunori; Wada, Nobuo
2016-05-01
Dimensional crossover to the one-dimensional (1D) state from higher dimensions has been studied for dilute ^3He fluid adsorbed in 2.4 nm ^4He-preplated nanochannels, by susceptibility measurements down to 70 mK using 4.29 MHz nuclear magnetic resonance. In nanochannels, since energy states of ^3He motion perpendicular to the channel axis are discrete, a genuine 1D ^3He fluid is expected when the Fermi energy is less than the first excitation Δ _{01} for azimuthal motion. The susceptibilities χ above 0.3 K show the Curie-law susceptibilities independent of the ^3He density, which are characteristic of nondegenerate fluid in higher dimensions. With decreasing the temperature, a significant reduction of χ T was observed from about 0.3 K for all ^3He densities. It is considered to be due to the dimensional crossover below Δ _{01}˜ 0.5 K to the 1D ^3He state in the semi-degenerate regime above the Fermi temperature. In the 1D state at lower temperatures, T-independent χ were observed for ^3He of 0.019 layers below 0.1 K. It suggests that the 1D ^3He fluid enters the quantum degenerate regime.
Phosphorus Doping of Polycrystalline CdTe by Diffusion
Energy Technology Data Exchange (ETDEWEB)
Colegrove, Eric; Albin, David S.; Guthrey, Harvey; Harvey, Steve; Burst, James; Moutinho, Helio; Farrell, Stuart; Al-Jassim, Mowafak; Metzger, Wyatt K.
2015-06-14
Phosphorus diffusion in single crystal and polycrystalline CdTe material is explored using various methods. Dynamic secondary ion mass spectroscopy (SIMS) is used to determine 1D P diffusion profiles. A 2D diffusion model is used to determine the expected cross-sectional distribution of P in CdTe after diffusion anneals. Time of flight SIMS and cross-sectional cathodoluminescence corroborates expected P distributions. Devices fabricated with diffused P exhibit hole concentrations up to low 1015 cm-3, however a subsequent activation anneal enabled hole concentrations greater than 1016 cm-3. CdCl2 treatments and Cu based contacts were also explored in conjunction with the P doping process.
Observation of single-file diffusion in a MOF.
Jobic, H
2016-06-29
The translational and rotational dynamics of neopentane adsorbed in the one-dimensional channels of MIL-47(V) has been studied by quasi-elastic neutron scattering. The rotational motion of neopentane is well-described by the rotational diffusion model, with a correlation time of 41 ps at 300 K. The translational motion of the molecule has been fitted by several models: isotropic diffusion, normal 1D and single-file diffusion. It is found that the observed line shapes can only be reproduced by the single-file diffusion model. The single-file mobility factor, F, is (8 ± 1) × 10(-14) m(2) s(-1/2) at 300 K. This is the first observation of this unusual diffusion behaviour in a MOF. PMID:26932296
Geiser, Christian; Griffin, Daniel; Shiffman, Saul
2016-01-01
Sometimes, researchers are interested in whether an intervention, experimental manipulation, or other treatment causes changes in intra-individual state variability. The authors show how multigroup-multiphase latent state-trait (MG-MP-LST) models can be used to examine treatment effects with regard to both mean differences and differences in state variability. The approach is illustrated based on a randomized controlled trial in which N = 338 smokers were randomly assigned to nicotine replacement therapy (NRT) vs. placebo prior to quitting smoking. We found that post quitting, smokers in both the NRT and placebo group had significantly reduced intra-individual affect state variability with respect to the affect items calm and content relative to the pre-quitting phase. This reduction in state variability did not differ between the NRT and placebo groups, indicating that quitting smoking may lead to a stabilization of individuals' affect states regardless of whether or not individuals receive NRT. PMID:27499744