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
Multigroup neutron dose calculations for proton therapy
International Nuclear Information System (INIS)
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.
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)
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)
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
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
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)
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
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)
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
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)
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.
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
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.)
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)
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
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 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)
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)
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)
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%
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.
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
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
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)
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.
The multigroup neutronics model of NuStar's 3D core code EGRET
International Nuclear Information System (INIS)
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)
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
AMPX-77, Modular System for Coupled Neutron-Gamma Multigroup Cross-Sections from ENDF/B-5
International Nuclear Information System (INIS)
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
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
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.
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)
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.
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
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)
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
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)
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
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
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
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
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
Energy Technology Data Exchange (ETDEWEB)
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.
International Nuclear Information System (INIS)
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)
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)
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.
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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
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
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
Institute of Scientific and Technical Information of China (English)
2008-01-01
The translation industry in China has to address myriad problems to reap huge returns from building the Tower of Babel By day, Chen Jing is a customs dec-laration clerk at a Shanghai-based shipping company.
Procedure to Generate the MPACT Multigroup Library
Energy Technology Data Exchange (ETDEWEB)
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
International Nuclear Information System (INIS)
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.
International Nuclear Information System (INIS)
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
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)
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)
International Nuclear Information System (INIS)
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)
Energy Technology Data Exchange (ETDEWEB)
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)
International Nuclear Information System (INIS)
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.)
International Nuclear Information System (INIS)
1 - Description of program or function: specified on ORNL-RSIC-25, shielding benchmark problems. - BP-3 (Neutron cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: air; Origin: ENDF/B; Weighting spectrum: 1/E; - BP-6 (neutron and gamma-ray cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: Borated Polyethylene (C-12, H, and B-10); Origin: ENDF/B-II. The cross section data can be used to repeat the Shielding Benchmark Problems 3.0 and 6.0 for testing against the results published in ORNL-RSIC-25. 2 - Method of solution: ZZ-BP-3 neutron cross sections from the CCC-17/05R library were processed into 104 neutron groups using the PSR-9/CSP code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The resulting multigroup cross sections are P5 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE. ZZ-BP-6 neutron and gamma-ray cross sections for 12C, H, and 10B were from ENDF/B-II data. The neutron multigroup cross sections were generated into 104 neutron groups using the PSR-13/SUPERTOG code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The gamma-ray multigroup cross sections were generated using PSR-7/MUG. The neutron-gamma-ray coupling utilized yield data from the DLC-12/POPOP4 library (data sets 010101, 060101, 060301, and 05100201). The neutron-gamma-ray coupled multigroup cross-section set was generated using the SAMPLE COUPLING CODE (ASCC). The multigroup cross sections are in a 22-18 group structure with P3 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE
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
International Nuclear Information System (INIS)
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.
A variational nodal expansion method for the solution of multigroup neutron diffusion equations
International Nuclear Information System (INIS)
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
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)
Development of 3D multi-group neutron diffusion code for hexagonal geometry
International Nuclear Information System (INIS)
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)
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)
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
Nodal deterministic simulation for problems of neutron shielding in multigroup formulation
International Nuclear Information System (INIS)
In this paper, we propose the use of some computational tools, with the implementation of numerical methods SGF (Spectral Green's Function), making use of a deterministic model of transport of neutral particles in the study and analysis of a known and simplified problem of nuclear engineering, known in the literature as a problem of neutron shielding, considering the model with two energy groups. These simulations are performed in MatLab platform, version 7.0, and are presented and developed with the help of a Computer Simulator providing a friendly computer application for their utilities
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
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
Energy Technology Data Exchange (ETDEWEB)
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)
Energy Technology Data Exchange (ETDEWEB)
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.
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
FORM-OTA, Multigroup Constant for Epithermal Neutron Slowing-Down in Homogeneous Media
International Nuclear Information System (INIS)
1 - Description of problem or function: FORM-OTA performs a multi- group slowing down calculation for a fundamental mode of given buckling in a homogeneous medium to obtain space-independent energy spectra for the epithermal neutron flux and current. Using the calculated flux and current spectra the program produces group constants for desired few group schemes. 2 - Method of solution: FORM-OTA is a member of the MUFT family of programs. The one-dimensional transport equation for the flux in plane geometry is solved by removing the spatial dependence by a Fourier transformation and by treating the angular dependence in either B1- or P1-approximation. Elastic slowing-down by hydrogen can be solved in an exact manner using a differential equation formulation. Elastic slowing-down by all non-hydrogen elements is lumped together and treated in the Greuling-Goertzel approximation. For resonance absorption a rather simple formulation is used. A flux peaking in the fuel can be assessed, too. The energy group structure (54 groups in the range 10 MeV - .625 eV) is pre-programmed into the code. A facility is provided to modify library data and to introduce entirely new data at run-time. 3 - Restrictions on the complexity of the problem: Maximum 18 elements (isotopes) in a mixture. Of these 18 elements maximum 10 elements can have resonance data. Maximum 6 few group schemes. Maximum 25 group in any few group scheme. Maximum 25 groups in a heterogeneous two-region (fuel-moderator) calculation
International Nuclear Information System (INIS)
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.
Gloria Regina Alves de Carvalho Amaral; Marcus Alexandre Motta
2013-01-01
A densidade da leitura lida e traduzida nas palavras de um outro. A densidade que permite esboroar fronteiras: culturais, de gênero, de línguas. Referências que atravessam, narrativas deslocadas, discursos embaralhados. A Trilogia de Nova Iorque é a Babel de Paul Auster: a literatura, a crítica, a história, a arte. Apresentando, representando, reconhecendo a leitura como a grande possibilidade de abarcar em uma mesma torre as diferenças, os atrasos, as discrepâncias, mas também aos encontros ...
Directory of Open Access Journals (Sweden)
Dwi Setiawan
2008-01-01
Full Text Available This article discusses how the plot of Babel reflects the principles of the quantum-physic theory of complex system such as complexity, indeterminacy and non-linearity. In terms of complexity, the movie exposes more than two distinct subplots with their subcomponents. Yet, every subplot is only meaningful when it is mentally put in relation to the others. Due to its complexity, it is highly difficult for the audience and the characters of Babel to determine the meaning or the significance of a component in the story. Finally, Babel also displays the phenomena of non-linearity and chaos. Babel's non-linearity expresses both positive and negative feedback loops, with the first being dominant. This is largely responsible for the chaotic development of the plot.
Dwi Setiawan; Liliek Soelistyo
2008-01-01
This article discusses how the plot of Babel reflects the principles of the quantum-physic theory of complex system such as complexity, indeterminacy and non-linearity. In terms of complexity, the movie exposes more than two distinct subplots with their subcomponents. Yet, every subplot is only meaningful when it is mentally put in relation to the others. Due to its complexity, it is highly difficult for the audience and the characters of Babel to determine the meaning or the significance of ...
Energy Technology Data Exchange (ETDEWEB)
Burns, Kimberly A. [Georgia Inst. of Technology, Atlanta, GA (United States)
2009-08-01
The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples.
Cross section probability tables in multi-group transport calculations
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
The KAFAX-F22 was developed from JEF-2.2, which is a MATXS format, multigroup library of fast reactor. The KAFAX-F22 has 80 and 24 energy group structures for neutron and photon, respectively. It includes 89 nuclide data processed by NJOY94.38. The TRANSX/TWODANT system was used for benchmark calculations of fast reactor and one- and two-dimensional calculations of ONEDANT and TWODANT were carried out with 80 group, P3S16 and with 25 group, P3S8, respectively. The average values of multiplication factors are 0.99652 for MOX cores, 1.00538 for uranium cores and 1.00032 for total cores. Various central reaction rate ratios also give good agreements with the experimental values considering experimental uncertainties except for VERA-11A, VERA-1B, ZPR-6-7 and ZPR-6-6A cores of which experimental values seem to involve some problems. (author). 13 refs., 18 tabs., 2 figs
International Nuclear Information System (INIS)
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: 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, as well as specialized one-dimensional geometry descriptions, 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. MORSE-E1 - This is a new analysis package written by ESIS at Ispra. It can be used with the O5R geometry or with the combinatorial geometry as with any other geometry compatible with MORSE. It contains a flexible set of subprograms tailored to solve a variety of shielding problems. It provides uniform source distributions of several geometrical shapes, and calculates particle fluxes and reaction rates integrated over the volumes defined by the user. Currents of particles through surfaces may be calculated. MORSE-H has been developed from MORSE-CG (CCC-0203) and MORSE-E. The special features of this version are: 1) Track-length (volume integrated flux) or next event (point flux) estimates; 2) multiple source region specification; 3) flexible source direction options; 4) restartable in all classes of problems; 5) eigenvalue (keff) solution obtainable even if keff is significantly different from unity. 2 - 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 ANISN, DTF-4 or DOT cross
Directory of Open Access Journals (Sweden)
Gloria Regina Alves de Carvalho Amaral
2013-11-01
Full Text Available A densidade da leitura lida e traduzida nas palavras de um outro. A densidade que permite esboroar fronteiras: culturais, de gênero, de línguas. Referências que atravessam, narrativas deslocadas, discursos embaralhados. A Trilogia de Nova Iorque é a Babel de Paul Auster: a literatura, a crítica, a história, a arte. Apresentando, representando, reconhecendo a leitura como a grande possibilidade de abarcar em uma mesma torre as diferenças, os atrasos, as discrepâncias, mas também aos encontros e as relevâncias. Mais que ficção ou crítica, a trilogia, qual performance, parece teorizar a filosofia, ficcionalizar a teoria, filosofar com a literatura, historicizar a... Ou nada disso. Talvez, e apenas talvez, seja nada mais que um chamado para dançar com as palavras, seguir os fios dos fios que seguem seus traços...
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)
Syntax and Semantics of Babel-17
Obua, Steven
2010-01-01
We present Babel-17, the first programming language for purely functional structured programming (PFSP). Earlier work illustrated PFSP in the framework of a toy research language. Babel-17 takes this earlier work to a new level by showing how PFSP can be combined with pattern matching, object oriented programming, and features like concurrency, lazy evaluation and memoization.
Syntax and Semantics of Babel-17
Obua, Steven
2010-01-01
We present Babel-17, the first programming language for purely functionalstructured programming (PFSP). Earlier work illustrated PFSP in the frameworkof a toy research language. Babel-17 takes this earlier work to a new level byshowing how PFSP can be combined with pattern matching, object orientedprogramming, and features like concurrency, lazy evaluation, memoization andsupport for lenses.
Cairns, John
2005-01-01
The key to long-term economic and social stability is the sustainable use of the planet. The Global Tower of Babel is such a formidable obstacle in achieving sustainable use of the planet that the barriers must be diminished. However, if a global consensus can be reached and compatible eco-ethics and sustainability ethics can be established, humankind may have a chance at achieving sustainability despite its cultural diversity.
Le mythe de Babel The Myth of Babel El mito de Babel
Directory of Open Access Journals (Sweden)
James Dauphiné
1996-05-01
Full Text Available Marqué par le triple sceau de la théologie, de la littérature et de la critique, le mythe de Babel engendre une réflexion sur les fondements de la pensée occidentale. Le texte de la Genèse XI est une source considérable de création et de questionnement qui, de saint Augustin à Joyce ou Perec, demeure particulièrement féconde.As a myth which bears the triple hallmark of theology, literature and criticism, Babel is an opportunity to take into consideration the very foundations of western thought and civilization. The « tale of origins » in Gen. XI has been a source of outstanding creation and questioning which, from saint Augustine to Joyce or Perec, has remained extraordinary fruitful.Como mito que lleva el triple sello de la teología, la literatura y la crítica, Babel permite reflexionar sobre los datos fundamentales del pensamiento y de la literatura occidentales. A partir del « relato de los orígenes » de Génesis XI, han brotado una creación y una interrogación dignas de consideración y siempre, desde San Agustín a Joyce o Perec, ha sido una fuente de inspiración asombrosamente fecunda.
International Nuclear Information System (INIS)
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 Genesis of the AFMLTA and Babel and the Babel of Genesis
Vale, David
2006-01-01
In this article, the author describes the genesis of the Australian Federation of Modern Language Teachers Associations (AFMLTA) and "Babel." With regard to the origin of the title of the journal, its name refers only indirectly to the Tower of Babel in Genesis. It comes in fact from the affectionate nickname that had been given to the building at…
How to Implement a Protocol for Babel RMI
Energy Technology Data Exchange (ETDEWEB)
Kumfert, G; Leek, J
2006-03-30
RMI support in Babel has two main goals: transparency & flexibility. Transparency meaning that the new RMI features are entirely transparent to existing Babelized code; flexibility meaning the RMI capability should also be flexible enough to support a variety of RMI transport implementations. Babel RMI is a big success in both areas. Babel RMI is completely transparent to already Babelized implementation code, allowing painless upgrade, and only very minor setup changes are required in client code to take advantage of RMI. The Babel RMI transport mechanism is also extremely flexible. Any protocol that implements Babel's minimal, but complete, interface may be used as a Babel RMI protocol. The Babel RMI API allows users to select the best protocol and connection model for their application, whether that means a WebServices-like client-server model for use over a WAP, or a faster binary peer-to-peer protocol for use on different nodes in a leadership-class supercomputer. Users can even change protocols without recompiling their code. The goal of this paper is to give network researchers and protocol implementors the information they need to develop new protocols for Babel RMI. This paper will cover both the high-level interfaces in the Babel RMI API, and the low level details about how Babel RMI handles RMI objects.
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
Pierre Bouretz, 22 variations sur Babel
Schellino, Andrea
2016-01-01
Pierre Bouretz, direttore di studi all’École des hautes études en sciences sociales di Parigi ed esperto dei rapporti tra messianismo e filosofia contemporanea (Témoins du futur. Philosophie et messianisme, 2003; Les Lumières du messianisme, 2008), propone con questo volume un’ermeneutica del celebre episodio biblico della torre di Babele (Genesi 11, 1-9). Più che una sintesi completa della fortuna del mito di Babele, lo studio raccoglie ventidue variazioni – di altrettante lettere è composto...
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)
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
International Nuclear Information System (INIS)
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.
Open Babel: An open chemical toolbox
Directory of Open Access Journals (Sweden)
O'Boyle Noel M
2011-10-01
Full Text Available Abstract Background A frequent problem in computational modeling is the interconversion of chemical structures between different formats. While standard interchange formats exist (for example, Chemical Markup Language and de facto standards have arisen (for example, SMILES format, the need to interconvert formats is a continuing problem due to the multitude of different application areas for chemistry data, differences in the data stored by different formats (0D versus 3D, for example, and competition between software along with a lack of vendor-neutral formats. Results We discuss, for the first time, Open Babel, an open-source chemical toolbox that speaks the many languages of chemical data. Open Babel version 2.3 interconverts over 110 formats. The need to represent such a wide variety of chemical and molecular data requires a library that implements a wide range of cheminformatics algorithms, from partial charge assignment and aromaticity detection, to bond order perception and canonicalization. We detail the implementation of Open Babel, describe key advances in the 2.3 release, and outline a variety of uses both in terms of software products and scientific research, including applications far beyond simple format interconversion. Conclusions Open Babel presents a solution to the proliferation of multiple chemical file formats. In addition, it provides a variety of useful utilities from conformer searching and 2D depiction, to filtering, batch conversion, and substructure and similarity searching. For developers, it can be used as a programming library to handle chemical data in areas such as organic chemistry, drug design, materials science, and computational chemistry. It is freely available under an open-source license from http://openbabel.org.
Human Language Technology: The Babel Fish
Gambäck, Björn
1999-01-01
The essay describes some of the main problems which meet us when trying to process human language on a computer. The overall approach is to look at what we would need to do in order to be able to build a device with the same general functionality as Douglas Adams' Babel fish. That is, a device which can take utterances spoken in one language and instantly translate them into speech in some other language.
Multi-language Struct Support in Babel
Energy Technology Data Exchange (ETDEWEB)
Ebner, D; Prantl, A; Epperly, T W
2011-03-22
Babel is an open-source language interoperability framework tailored to the needs of high-performance scientific computing. As an integral element of the Common Component Architecture (CCA) it is used in a wide range of research projects. In this paper we describe how we extended Babel to support interoperable tuple data types (structs). Structs are a common idiom in scientific APIs; they are an efficient way to pass tuples of nonuniform data between functions, and are supported natively by most programming languages. Using our extended version of Babel, developers of scientific code can now pass structs as arguments between functions implemented in any of the supported languages. In C, C++ and Fortran 2003, structs can be passed without the overhead of data marshaling or copying, providing language interoperability at minimal cost. Other supported languages are Fortran 77, Fortran 90, Java and Python. We will show how we designed a struct implementation that is interoperable with all of the supported languages and present benchmark data compare the performance of all language bindings, highlighting the differences between languages that offer native struct support and an object-oriented interface with getter/setter methods.
International Nuclear Information System (INIS)
GANTRAS is a system of codes for neutron transport calculations in which the anisotropy of elastic and inelastic (including (n,n'x)-reactions) scattering is fully taken into account. This is achieved by employing a rigorous method, so-called I*-method, to represent the scattering term of the transport equation and with the use of double-differential cross-sections for the description of the emission of secondary neutrons. The I*-method was incorporated into the conventional transport code ONETRAN. The ONETRAN subroutines were modified for the new purpose. An implementation of the updated version ANTRA1 was accomplished for plane and spherical geometry. ANTRA1 was included in GANTRAS and linked to another modules which prepare angle-dependent transfer matrices. The GANTRAS code consists of three modules: 1. The CROMIX code which calculates the macroscopic transfer matrices for mixtures on the base of microscopic nuclide-dependent data. 2. The ATP code which generates discretized angular transfer probabilities (i.e. discretizes the I*-function). 3. The ANTRA1 code to perform SN transport calculations in one-dimensional plane and spherical geometries. This structure of GANTRAS allows to accommodate the system to various transport problems. (orig.)
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
Energy Technology Data Exchange (ETDEWEB)
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.
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.)
International Nuclear Information System (INIS)
The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as 6Li, 7Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa)
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
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.
Reflections on After Babel:Aspects of Language and Translation
Institute of Scientific and Technical Information of China (English)
陶子凤
2015-01-01
The publication of his book After Babel:Aspects of Language and Translation in 1975,which was a landmark in the field of translation and linguistics and the first systematical study in translation theory since the 1800s in western academic circles,brought George Steiner worldwide attention.This paper will mainly introduce the hermeneutic motion of After Babel:Aspects of Language and Translation and present application of Steiner’s hermeneutic motion in analyzing translator’s subjectivity.
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)
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
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.
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 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)
Babel 1.0 Release Criteria: A Working Document
Energy Technology Data Exchange (ETDEWEB)
Kumfert, G; Dahlgren, T; Epperly, T; Leek, J
2004-10-19
In keeping with the Open Source tradition, we want our Babel 1.0 release to indicate a certain level of capability, maturity, and stability. From our first release (version 0.5.0) in July of 2001 to our current (18th) release (version 0.9.6) we have continued to add capabilities in response to customer feedback, our observations in the field, and a consistent vision for interoperability. The key to our maturity is without a doubt the ever-increasing demands of our growing user base... both in terms of sheer size and sophistication with the underlying technology. Stability is a special challenge for any research project. With our 1.0 release, we will branch and maintain a stable Babel 1.0 code line for at least a full year. This means no new features and no backward incompatible changes, only bug fixes. All continuing R&D will be performed on a separate development tree. Currently, Babel has a quarterly release cycle with no guarantee for backward compatibility from one release to the next (though we certainly try to make migration as painless as possible). Now is the time where we can see a good point for a Babel 1.0 release. But, seeing that point is different from being there. This list enumerates and explains the outstanding technical issues to be resolved to minimize volatility and help ensure stability for the 1.0 line. The first draft of this document was circulated internally in June 2003. A revised draft was then presented at the July 2003 CCA meeting. A third revision was made into the current working document form & circulated for general comment on the babel-users mailing list and Babel's homepage. The working document was intended to be an open record tracking progress in subsequent Babel releases. A major revision of the document (including adding new items and promoting/demoting items) was done in October 2004, well after the 0.9.6 release.
A numerical model for multigroup radiation hydrodynamics
International Nuclear Information System (INIS)
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.
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)
Babel Fortran 2003 Binding for Structured Data Types
Energy Technology Data Exchange (ETDEWEB)
Muszala, S; Epperly, T; Wang, N
2008-05-02
Babel is a tool aimed at the high-performance computing community that addresses the need for mixing programming languages (Java, Python, C, C++, Fortran 90, FORTRAN 77) in order to leverage the specific benefits of those languages. Scientific codes often rely on structured data types (structs, derived data types) to encapsulate data, and Babel has been lacking in this type of support until recently. We present a new language binding that focuses on their interoperability of C/C++ with Fortran 2003. The new binding builds on the existing Fortran 90 infrastructure by using the iso-c-binding module defined in the Fortran 2003 standard as the basis for C/C++ interoperability. We present the technical approach for the new binding and discuss our initial experiences in applying the binding in FACETS (Framework Application for Core-Edge Transport Simulations) to integrate C++ with legacy Fortran codes.
La Tour de Babel ou la Part du Diable
Directory of Open Access Journals (Sweden)
James Dauphiné
2000-06-01
Full Text Available Denis de Rougemont dans son essai La Part du Diable s’est efforcé de prouver que la Tour de Babel est exemplaire de l’action du « diable dans nos dieux et dans nos maladies ». Plus proche des analyses contenues dans les Mythologies de Barthes que de celles rencontrées au fil des traités de démonologie, Denis de Rougemont dénonce le modernisme qui a, de fait, consacré Babel « grand mythe de notre temps » (p. 146. La thèse avancée a pour fondement « la babélisation des cadres matériels de notr...
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)
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.
EL ESCORIAL COMO ANTITESIS DE LA TORRE DE BABEL
Arciniega García, Luis
1992-01-01
LA ARQUITECTURA BIBLICA FUE UTILIZADA "A POSTERIORI" PARA LEGITIMAR LA FIGURA DEL REY FELIPE II Y SU ARQUITECTURA. LOS CRONISTAS, CIRCULOS INTELECTUALES Y ARTISTAS, POR UN LADO, VINCULARON AL REY HISPANO Y SU MONASTERIO CON SALOMON Y EL TEMPLO QUE MANDO CONSTRUIR; POR OTRO LADO, CONTRAPUSIERON AL PRIMERO Y SU ARQUITECTURA, UNA VEZ CONSTRUIDA LA MISMA, CON LA CONSTRUCCION DE LA TORRE DE BABEL (ABC/LAG).
Multigroup albedo method applied to gamma radiation shielding
International Nuclear Information System (INIS)
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)
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...
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)
Energy Technology Data Exchange (ETDEWEB)
Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi
2012-10-01
PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICS (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable 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 RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.
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)
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)
Parallel computation of multigroup reactivity coefficient using iterative method
International Nuclear Information System (INIS)
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
International Nuclear Information System (INIS)
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
Before Babel: Ancient Tales from Genesis Retold in Reconstructed Proto-Indo-European
Barrois, Bertrand
2015-01-01
Before Babel retells the legends of the Creation, the Garden of Eden, the Flood, the Tower of Babel, and Psalm 104 in reconstructed Proto-Indo-European, with abundant grammatical commentary. This “translation” provides a more satisfying exemplar of the proto-language than Schleicher’s little fable and demonstrates its narrative power. Appendices on the author’s orthographic and grammatical conventions and a mini-lexicon are included.
Satirising the Norwegian language conflict: Gabriel Scott's Babels taarn contextually reconsidered
Hale, Frederick
2013-01-01
Gabriel Scott’s comedy Babels taarn (Babel Tower), first performed at the National Theatre in Kristiania in 1911, satirises the language controversy that was raging in Norway at the time. The play is regarded as important in linguistic and literary terms, but has been largely forgotten. This article argues that Scott was disillusioned by the politicisation of the language controversy and regarded the advance of landsmål as an artificial and unwelcome phenomenon in the unfolding of Norwegian c...
Pybel: a Python wrapper for the OpenBabel cheminformatics toolkit
Morley Chris; O'Boyle Noel M; Hutchison Geoffrey R
2008-01-01
Abstract Background Scripting languages such as Python are ideally suited to common programming tasks in cheminformatics such as data analysis and parsing information from files. However, for reasons of efficiency, cheminformatics toolkits such as the OpenBabel toolkit are often implemented in compiled languages such as C++. We describe Pybel, a Python module that provides access to the OpenBabel toolkit. Results Pybel wraps the direct toolkit bindings to simplify common tasks such as reading...
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.
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…
International Nuclear Information System (INIS)
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.
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
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)
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.
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)
Multigroup Free-atom Doppler-broadening Approximation. Theory
Energy Technology Data Exchange (ETDEWEB)
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 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
A multigroup treatment of radiation transport
International Nuclear Information System (INIS)
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)
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
International Nuclear Information System (INIS)
The reaction rates in the multi-layers containing depleted uranium were measured by activation foils and micro-fission chambers. The analysis of the experiment was carried out by using the multi-group transport calculation code, DOT 3.5 and the continuous energy Monte Carlo code, MCNP. The multi-group calculation overpredicted the low energy reaction rates in the DU layers, while the continuous energy calculation agreed well. The multi-group and continuous energy calculation was compared for the one-dimensional transmission of iron spheres. The results revealed overprediction of the multi-group calculation near the fast neutron source. The averaging of the resonance shapes in generating the multi-group cross sections made minima of the resonance valleys higher than that of the pointwise cross section. This increased the scattering of the neutrons inside and caused the overprediction of the multi-group calculation
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)
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
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
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.
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
New Reflections on Mirror Neuron Research, the Tower of Babel, and Intercultural Education
Westbrook, Timothy Paul
2015-01-01
Studies of the human mirror neuron system demonstrate how mental mimicking of one's social environment affects learning. The mirror neuron system also has implications for intercultural encounters. This article explores the common ground between the mirror neuron system and theological principles from the Tower of Babel narrative and applies them…
Divided by a Common Language: The Babel Proclamation and Its Influence in Iowa History
Frese, Stephen J.
2005-01-01
The anti-German sentiment during World War I reached a point where "people speaking German on the street were attacked and rebuked." Iowa Governor William L. Harding legitimized such expressions of prejudice and war-time fanaticism when he issued "The Babel Proclamation" on May 23, 1918. Antagonism toward Germans and their language escalated…
Pybel: a Python wrapper for the OpenBabel cheminformatics toolkit
Directory of Open Access Journals (Sweden)
Morley Chris
2008-03-01
Full Text Available Abstract Background Scripting languages such as Python are ideally suited to common programming tasks in cheminformatics such as data analysis and parsing information from files. However, for reasons of efficiency, cheminformatics toolkits such as the OpenBabel toolkit are often implemented in compiled languages such as C++. We describe Pybel, a Python module that provides access to the OpenBabel toolkit. Results Pybel wraps the direct toolkit bindings to simplify common tasks such as reading and writing molecular files and calculating fingerprints. Extensive use is made of Python iterators to simplify loops such as that over all the molecules in a file. A Pybel Molecule can be easily interconverted to an OpenBabel OBMol to access those methods or attributes not wrapped by Pybel. Conclusion Pybel allows cheminformaticians to rapidly develop Python scripts that manipulate chemical information. It is open source, available cross-platform, and offers the power of the OpenBabel toolkit to Python programmers.
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
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)
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)
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)
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
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
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…
Babel. Revista de Libros: formular el propio presente entre los finales y el fin
Directory of Open Access Journals (Sweden)
Mariana Catalin
2013-08-01
Full Text Available El presente artículo realiza un abordaje de la revista Babel. Revista de libros, publicada en Buenos Aires entre abril de 1988 y marzo de 1991, a partir de un eje singular: la temporalidad que la revista construye y problematiza a partir de pensar su propio presente como un presente en crisis. Como modo de insertarse en el campo intelectual y literario argentino y como estrategia para poder articular las lecturas que le interesa volver centrales, Babel construye una temporalidad entre dos épocas, que supone discutir el fin de la modernidad. Para ver cómo esa temporalidad se construye en la revista, tomaremos dos caminos: por una parte, analizaremos el funcionamiento del discurso sobre lo posmoderno en los primeros dos números de la revista y, por otra parte, intentaremos un recorrido por una sección central de la misma: los “Dossier”.
Interpretations of the Tower of Babel narrative in the African context
Rathbone, M
2014-01-01
Biblical scholarship from the African context provides possible new and creative perspectives for the interpretation of the Tower of Babel narrative because of uniquely African questions that structure the interpretative process. These unique questions relate to the cultures of African people, the injustice of colonialism, apartheid and so forth. The problem is that some of these new perspectives are influenced by rationalism that may result in reductionist interpretations of the Tower of Bab...
The Yearning for Unity and the Eternal Return of the Tower of Babel
Carty, Anthony
2007-01-01
International lawyers frequently aspire to affirm the existence of international community and the presence of authority to speak on its behalf. However by forcing a hierarchical representation of legal values upon nations, which have not accepted them, international lawyers, and the politicians whom they advise, risk unleashing a whirlwind of violence. The myth or the Biblical story of the Tower of Babel, is a millenniums old warning of the presumption which can lie behind an apparently reas...
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)
Cross-language Babel structs—making scientific interfaces more efficient
International Nuclear Information System (INIS)
Babel is an open-source language interoperability framework tailored to the needs of high-performance scientific computing. As an integral element of the Common Component Architecture, it is employed in a wide range of scientific applications where it is used to connect components written in different programming languages. In this paper we describe how we extended Babel to support interoperable tuple data types (structs). Structs are a common idiom in (mono-lingual) scientific application programming interfaces (APIs); they are an efficient way to pass tuples of nonuniform data between functions, and are supported natively by most programming languages. Using our extended version of Babel, developers of scientific codes can now pass structs as arguments between functions implemented in any of the supported languages. In C, C++, Fortran 2003/2008 and Chapel, structs can be passed without the overhead of data marshaling or copying, providing language interoperability at minimal cost. Other supported languages are Fortran 77, Fortran 90/95, Java and Python. We will show how we designed a struct implementation that is interoperable with all of the supported languages and present benchmark data to compare the performance of all language bindings, highlighting the differences between languages that offer native struct support and an object-oriented interface with getter/setter methods. A case study shows how structs can help simplify the interfaces of scientific codes significantly. (paper)
Cross-language Babel structs—making scientific interfaces more efficient
Prantl, Adrian; Ebner, Dietmar; Epperly, Thomas G. W.
2013-01-01
Babel is an open-source language interoperability framework tailored to the needs of high-performance scientific computing. As an integral element of the Common Component Architecture, it is employed in a wide range of scientific applications where it is used to connect components written in different programming languages. In this paper we describe how we extended Babel to support interoperable tuple data types (structs). Structs are a common idiom in (mono-lingual) scientific application programming interfaces (APIs); they are an efficient way to pass tuples of nonuniform data between functions, and are supported natively by most programming languages. Using our extended version of Babel, developers of scientific codes can now pass structs as arguments between functions implemented in any of the supported languages. In C, C++, Fortran 2003/2008 and Chapel, structs can be passed without the overhead of data marshaling or copying, providing language interoperability at minimal cost. Other supported languages are Fortran 77, Fortran 90/95, Java and Python. We will show how we designed a struct implementation that is interoperable with all of the supported languages and present benchmark data to compare the performance of all language bindings, highlighting the differences between languages that offer native struct support and an object-oriented interface with getter/setter methods. A case study shows how structs can help simplify the interfaces of scientific codes significantly.
Application de la methode des sous-groupes au calcul Monte-Carlo multigroupe
Martin, Nicolas
This thesis is dedicated to the development of a Monte Carlo neutron transport solver based on the subgroup (or multiband) method. In this formalism, cross sections for resonant isotopes are represented in the form of probability tables on the whole energy spectrum. This study is intended in order to test and validate this approach in lattice physics and criticality-safety applications. The probability table method seems promising since it introduces an alternative computational way between the legacy continuous-energy representation and the multigroup method. In the first case, the amount of data invoked in continuous-energy Monte Carlo calculations can be very important and tend to slow down the overall computational time. In addition, this model preserves the quality of the physical laws present in the ENDF format. Due to its cheap computational cost, the multigroup Monte Carlo way is usually at the basis of production codes in criticality-safety studies. However, the use of a multigroup representation of the cross sections implies a preliminary calculation to take into account self-shielding effects for resonant isotopes. This is generally performed by deterministic lattice codes relying on the collision probability method. Using cross-section probability tables on the whole energy range permits to directly take into account self-shielding effects and can be employed in both lattice physics and criticality-safety calculations. Several aspects have been thoroughly studied: (1) The consistent computation of probability tables with a energy grid comprising only 295 or 361 groups. The CALENDF moment approach conducted to probability tables suitable for a Monte Carlo code. (2) The combination of the probability table sampling for the energy variable with the delta-tracking rejection technique for the space variable, and its impact on the overall efficiency of the proposed Monte Carlo algorithm. (3) The derivation of a model for taking into account anisotropic
System of adjoint P1 equations for neutron moderation
International Nuclear Information System (INIS)
In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, this procedure is questioned and the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. (author)
Energy Technology Data Exchange (ETDEWEB)
Martinez, Aquilino Senra; Silva, Fernando Carvalho da; Cardoso, Carlos Eduardo Santos [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear
2000-07-01
In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, this procedure is questioned and the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. (author)
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.)
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
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...
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...
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...
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)
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)
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
Coupled neutron and photon cross sections for transport calculations
International Nuclear Information System (INIS)
A compact set of multigroup cross sections and transfer tables for use in neutron and photon transport calculations was prepared from ENDF/B-IV using the NJOY processing system. The library includes prompt and steady-state coupled sets for neutrons and photons in FIDO format, prompt and steady-state fission spectra (chi vectors) for the fissionable isotopes, and a table of useful response functions including heating and gas production. These multigroup constants should be useful for a wide variety of problems where self-shielding is not important. 15 references
Directory of Open Access Journals (Sweden)
Liliana Laura Rega
2006-12-01
Full Text Available El Proyecto Alfa Biblioteca de Babel que reúne bibliotecas universitarias de América Latina y Europa comenzó en marzo de 2005 y prevé su conclusión en marzo de 2007. El presente trabajo intenta describir los objetivos y los resultados esperados, e informa las actividades realizadas por la Red Biblioteca de Babel. Finalmente se analizan las propuestas del proyecto en cuanto al rol de las bibliotecas universitarias, y su relación con la innovación en las prácticas pedagógicas.Biblioteca de Babel Alfa Project that assembles academic libraries from Latin America and Europe was approved on March, 2005 and its conclusion is foreseen on March, 2007. This article attempts to describe the aims and the expected results, and reports the activities of the Biblioteca de Babel Network. Finally it analizes the proposals of the project about the role of academic libraries and their relationship with innovations in pedagogical practices.
Foreign accents,the obstacle for building the modern Tower of Babel in workplace
Institute of Scientific and Technical Information of China (English)
张晓铃
2014-01-01
<正>There is a little story of the Tower of Babel from the Bible.At that point of time,the whole world had one common language.The people of the earth became skilled in construction and decided to build a city with a tower that would reach to heaven.God came to see their city and the tower they were building.He found their intention that the people build the tower as a stairway to heaven.As a result,God confused their language,causing them to speak different languages so
ÎN CĂUTAREA LIMBII CREAŢIEI SAU OBSESIA TURNULUI BABEL
Ana Daniela Gheorghe
2008-01-01
The work with the title „Looking for creation language or the obsession of the Babel Tower ” propossesto treat the idea of the perfect language as an act of communication because the language is an extremellynecessary analogic code for the act of communication.Taking into consideration this perspective, we considerinteresting the works of the two cultural personalities: Ioan Petru Culianu’s „The Creation Language ” andUmberto Eco’s „Looking for the Perfection of Language in European Culture ”...
Espace et langage: La Tour d’amour de Rachilde et la Tour de Babel
Directory of Open Access Journals (Sweden)
Pablo Justel
2016-04-01
Full Text Available In this article I analyze the relationships between space and the main characters in La Tour d’amour, by Rachilde. More specif-ically I focus on how space has already stunned one of the character’s speech and communicative abilities and it is now in the process of impairing speech in the other. By analyzing the novelist’s use of myths, refer-ences to the divinity and, especially, the many instances that offer evidence of wide-spread corruption in the characters’ lan-guage, I show how the lighthouse in which the characters dwell can be interpreted as a decadent Tower of Babel.
Babel: Cine y comunicación en un mundo globalizado
Pereira Domínguez, Carmen; Solé Blanch, Jordi; Valero Iglesias, Luis Fernando
2012-01-01
En este artículo se presenta una propuesta formativa utilizando el cine como material cultural y fuente de conocimiento. Una película como Babel permite trabajar la globalización y la educación de la ciudadanía, con planteamientos que exigen un nuevo humanismo, una nueva relación interpersonal, conscientes de los problemas de comunicación, prejuicios y choques culturales derivados del desarrollo tecnológico. La película cuestiona esta existencia en un mundo global interrelacionado, evocando e...
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
International Nuclear Information System (INIS)
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.)
Smith, Jerry
2015-01-01
This paper discusses the similarities between the Bible record of the Tower of Babel and the resulting confusion of languages and how it relates to modern times and the trend we see of English as an International Language (EIL). This paper then briefly examines the trend of being culturally sensitive in EIL by accepting cultural or "world…
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
Rega, Liliana Laura
2006-01-01
Biblioteca de Babel Alfa Project that assembles academic libraries from Latin America and Europe was approved on March, 2005 and its conclusion is foreseen on March, 2007. This article attempts to describe the aims and the expected results, and reports the activities of the Biblioteca de Babel Network. Finally it analizes the proposals of the project about the role of academic libraries and their relationship with innovations in pedagogical practices
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
Korja, Annakaisa
2016-04-01
The Precambrian Svecofennian orogen is characterized by LP- HT metamorphism and voluminous granitoid magmatism that usually develop in transitional to plateau stages of a collisional orogeny. Deep seismic reflection profiles BABEL and FIRE have been interpreted using PURC concepts: prowedge, retrowedg, uplifted plug, subduction conduit and elevated plateau. BABEL profiles image a transitional orogen with several nuclei displaying prowedge-uplifted plug-retrowedge architecture above paleo-subduction conduits. Prowedge and -continent are on the south-southwestern side and retrowedge and -continent on the north-northwestern side. This implies a long-lived southwesterly retreating convergent margin, where transitional accretionary orogens have developed. FIRE1-3 profiles images a hot orogen with a pronounced super-infra structure, typical of an elevated plateau stage, below the Central Finland Granitoid Complex. Large volumes of granitoid intrusions suggest large scale melting of the middle and/or lower crust. Reflection structures, analogue and numerical modeling suggest midcrustal flow. The plateau is flanked by prowedges that are characterized by HT-LP migmatite belts. The Svecofennian orogeny has progressed to an elevated plateau stage in the thickest core of the orogen, west of the arc-continent collision zone.
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)
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)
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
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
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)
An approach to neutronics analysis of candu reactors
International Nuclear Information System (INIS)
An attempt is made to tackle the problem of neutronics analysis of CANDU reactors. Until now CANDU reactors have been analysed by the methods developed at AECL and CGE using mainly receipe methods. Relying on multigroup transport codes GAM-GATHER in combination with diffusion code CITATION a package of codes is established to use it for survey as well as production purposes. (authors)
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
Adjoint P1 equations solution for neutron slowing down
International Nuclear Information System (INIS)
In some applications of perturbation theory, it is necessary know the adjoint neutron flux, which is obtained by the solution of adjoint neutron diffusion equation. However, the multigroup constants used for this are weighted in only the direct neutron flux, from the solution of direct P1 equations. In this work, the adjoint P1 equations are derived by the neutron transport equation, the reversion operators rules and analogies between direct and adjoint parameters. The direct and adjoint neutron fluxes resulting from the solution of P1 equations were used to three different weighting processes, to obtain the macrogroup macroscopic cross sections. It was found out noticeable differences among them. (author)
ATR neutron spectral characterization
International Nuclear Information System (INIS)
The Advanced Test Reactor (ATR) at INEL provides intense neutron fields for irradiation-effects testing of reactor material samples, for production of radionuclides used in industrial and medical applications, and for scientific research. Characterization of the neutron environments in the irradiation locations of the ATR has been done by means of neutronics calculations and by means of neutron dosimetry based on the use of neutron activation monitors that are placed in the various irradiation locations. The primary purpose of this report is to present the results of an extensive characterization of several ATR irradiation locations based on neutron dosimetry measurements and on least-squares-adjustment analyses that utilize both neutron dosimetry measurements and neutronics calculations. This report builds upon the previous publications, especially the reference 4 paper. Section 2 provides a brief description of the ATR and it tabulates neutron spectral information for typical irradiation locations, as derived from the more historical neutron dosimetry measurements. Relevant details that pertain to the multigroup neutron spectral characterization are covered in section 3. This discussion includes a presentation on the dosimeter irradiation and analyses and a development of the least-squares adjustment methodology, along with a summary of the results of these analyses. Spectrum-averaged cross sections for neutron monitoring and for displacement-damage prediction in Fe, Cr, and Ni are given in section 4. In addition, section4 includes estimates of damage generation rates for these materials in selected ATR irradiation locations. In section 5, the authors present a brief discussion of the most significant conclusions of this work and comment on its relevance to the present ATR core configuration. Finally, detailed numerical and graphical results for the spectrum-characterization analyses in each irradiation location are provided in the Appendix
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)
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)
Intragroup Socialization for Adult Korean Adoptees: A Multigroup Analysis
Directory of Open Access Journals (Sweden)
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
Silvana Borutti – Ute Heidmann, La Babele in cui viviamo. Traduzioni, Riscritture, Culture
Directory of Open Access Journals (Sweden)
Manfredi Bernardini
2013-12-01
Full Text Available Cosa implica l’atto di tradurre da una lingua all’altra? Come si pone la traduzione in rapporto al concetto di cultura? È possibile rintracciare un’etica della traduzione che valorizzi le differenze linguistiche, e quindi culturali, piuttosto che annullarle? Che mutazioni subisce l’identità nelle sue varie sfaccettature nel corso del processo della traduzione? Sono questi alcuni degli interrogativi cui cercano di dare risposta Silvana Borutti e Ute Heidmann in La Babele in cui viviamo. Traduzioni, Riscritture, Culture. La prima insegna Filosofia teoretica all’Università di Pavia, mentre Ute Heidmann è docente di Letterature comparate all’Università di Losanna. Prendendo le mosse da una doppia prospettiva fornita dalla filosofia del linguaggio e dalla comparatistica, le autrici offrono una lettura interdisciplinare del tema della traduzione.
Common language or Tower of Babel? On the evolutionary dynamics of signals and their meanings.
van Baalen, Minus; Jansen, Vincent A A
2003-01-01
We investigate how the evolution of communication strategies affects signal credibility when there is common interest as well as a conflict between communicating individuals. Taking alarm calls as an example, we show that if the temptation to cheat is low, a single signal is used in the population. If the temptation increases cheaters will erode the credibility of a signal, and an honest mutant using a different signal ('a private code') will be very successful until this, in turn, is cracked by cheaters. In such a system, signal use fluctuates in time and space and hence the meaning of a given signal is not constant. When the temptation to cheat is too large, no honest communication can maintain itself in a Tower of Babel of many signals. We discuss our analysis in the light of the Green Beard mechanism for the evolution of altruism. PMID:12590773
"A snake of black language": il processo come struttura narrativa in Babel Tower di A.S. Byatt
Beatrice Seligardi
2012-01-01
The article offers an analysis of the narrative dimension of the legal episodes in A.S. Byatt’s Babel Tower. The theoretical framework of the investigation is constituted by Nelson Goodman’s theory of worldmaking processes and, more specifically, its application in contemporary cultural narratology. The analysis focuses in particular on the function assumed by specific narrative techniques. The presence of metafictional devices on the one hand, and, on the other hand, of specific ...
La torre de Babel, Heródoto y los primeros viajeros europeos por tierras mesopotámicas
Montero Fenollós, Juan-Luis
2008-01-01
Until the beginning of archaeological research in Babylon in 1899 the city was only known in Europe through the information provided by the Old Testament, classical geographers and historians (specially Herodotus), and the stories of many adventurers. In fact many western travellers, who for different reasons visited the Near East, sought the most important Mesopotamian city and its legendary tower, the Tower of Babel, using only the information provided by the Bible and classical sources.
Luján La Torre Perregrini, Esperanza
2016-01-01
This studyfocuses on the theory of intertextuality and on the most important approachesof Julia Kristeva, Gerard Genette and Ronald Barthestothis theory. It also examines the intertextual relationships in twoworksof Jorge Luis Borges:Pierre Menard, author of the Quixoteand The library of Babel. This studyconcludes that intertextual relations and issues are very often used in the works of Jorge Luis Borges. Analysis of histwoworks has shown the most obvious indicators of intertextuality such a...
Directory of Open Access Journals (Sweden)
Emanulele Serrelli
2013-06-01
Full Text Available If, by “Babel”, we mean the set languages that have appeared in the world, we may want to research the ‘boundaries of Babel’ by asking whether the expansion of Babel is prevented (i.e., whether unobserved languages are impossible languages, and, if so, by which factors. The boundaries of Babel are being explored by partnerships of linguists and neuroscientists. Neo-chomskian approaches find evidence of neural networks dedicated to language processing, and study how these networks constrain the space of possible grammars, whereas lexico-grammar looks at neuroscientific evidence that syntax is not a separate function in the brain. Research questions also expand beyond a tight focus on the brain-language relationship. By “foundations of Babel” we refer to broader, ancient brain functions in which articulated language is embedded. Imitation can be one of those functions. “Physics of Babel” refers to many extra-brain factors that are lacking in non-human species, and that together make language possible. Research on the boundaries of Babel is a fascinating and open scenario, not only interdisciplinary, but also multi-directional, beyond the language function and beyond the exclusive role of the brain.
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)
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)
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)
Energy Technology Data Exchange (ETDEWEB)
Wilcox, T. P.
1973-09-20
The code ANISN-L solves the one-dimensional, multigroup, time-independent Boltzmann transport equation by the method of discrete ordinates. In problems involving a fissionable system, it can calculate the system multiplication or alpha. In such cases, it is also capable of determining isotopic concentrations, radii, zone widths, or buckling in order to achieve a given multiplication or alpha. The code may also calculate fluxes caused by a specified fixed source. Neutron, gamma, and coupled neutron--gamma problems may be solved in either the forward or adjoint (backward) modes. Cross sections describing upscatter, as well as the usual downscatter, may be employed. This report describes the use of ANISN-L; this is a revised version of ANISN which handles both large and small problems efficiently on CDC-7600 computers. (RWR)
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)
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
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.
Neutronics equations: Positiveness; compactness; spectral theory; time asymptotic behavior
International Nuclear Information System (INIS)
Neutronics equations are studied: the continuous model (with and without delayed neutrons) and the multigroup model. Asymptotic descriptions of these equations (t→+∞) are obtained, either by the Dunford method or by using semigroup perturbation techniques, after deriving the spectral theory for the equations. Compactness problems are reviewed, and a general theory of compact injection in neutronic functional space is derived. The effects of positiveness in neutronics are analyzed: the irreducibility of the transport semigroup, and the properties of the main eigenvalue (existence, nonexistence, frame, strict dominance, strict monotony in relation to all the parameters). A class of transport operators whose real spectrum can be completely described is shown
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)
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.)
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
Steiner, R
1994-12-01
The author uses private correspondence and documents referring to the foundation of the 'International Journal of Psycho-Analysis' and the 'Glossary' for translating Freud's work, to try to delineate the political and cultural strategy of Jones in founding and developing the 'International Journal of Psycho-Analysis'. Both strategies were based on the wish to have administrative and cultural control of psychoanalysis in the English-speaking countries. In the end Jones and his colleagues succeeded in making the language they created the official language of the IPA; through control of Freud's translations, through the 'Glossary' and particularly through its diffusion in the 'Journal'. The author briefly illustrates the various cultural sources of this attempt and tries to show the similarities between the project of Jones and the first generation of pioneers of psychoanalysis in Great Britain and the myth of the tower of Babel--one of its most important foundation stones being the 'International Journal'. Finally, the author stresses that those issues are still extremely alive in psychoanalysis today. But, confronted with the near-Babel of languages of contemporary psychoanalysis, can we still imply the existence of this universal common language and use it? Can the 'International Journal' still maintain its hegemony? Do we really understand each other even when we use the same technical terminology in English? Or shall we accept that today we should live without a tower of Babel in psychoanalysis? The author concludes that there is some hope, provided that we do not pursue meanings to the forbidden limit of the absolute. PMID:7713667
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)
Fernandes, Renata Sieiro; Park, Margareth Brandini
2010-01-01
O objetivo deste artigo é discutir as formas de construção de conhecimento da realidade sendo esta fragmentada, caótica, em des-ordem, a partir de imagens-metáforas da Torre de Babel e do caleidoscópio. Permeando isso, em diálogo com literatura e com filmes, apresentamos a possibilidade de surgimento de novo, do original, do inovador, ao lado da presença do tradicional, especialmente trazidos pelo potencial revolucionário dos jovens ou da juventude. O contato com o novo carrega em si o potenc...
Crossover accelerates evolution in GAs with a Babel-like fitness landscape: mathematical analyses.
Suzuki, H; Iwasa, Y
1999-01-01
The effectiveness of crossover in accelerating evolution in genetic algorithms (GAs) is studied with a haploid finite population of bit sequences. A Babel-like fitness landscape is assumed. There is a single bit sequence (schema) that is significantly more advantageous than all the others. We study the time until domination of the advantageous schema (Τ&subd;). Evolution proceeds with appearance, spread, and domination of the advantageous schema. The most important process determining Τ&subd; is the appearance (creation) of the advantageous schema. Crossover helps this creation process and enhances the rate of evolution. To study this effect, we first establish an analytical method to estimate Τ&subd; with or without crossover. Then, we conduct a numerical analysis using the frequency vector representation of the population with the recurrence relations formulated after GA operations. Finally, we carry out direct computer simulations with simple GAs operating on a population of binary strings directly prepared in the computer memory to examine the performance of the two analytical methods. It is shown that Τ&subd; is reduced greatly by crossover with a mildly high rate when the mutation rate is adjusted to a moderate value and that an advantageous schema has a fairly larger order (the number of bits). From these observations, we can determine implementation criteria for GAs, which are useful when we are applying GAs to engineering problems having a conspicuously discontinuous fitness landscape. PMID:10491466
ÎN CĂUTAREA LIMBII CREAŢIEI SAU OBSESIA TURNULUI BABEL
Directory of Open Access Journals (Sweden)
Ana Daniela Gheorghe
2008-05-01
Full Text Available The work with the title „Looking for creation language or the obsession of the Babel Tower ” propossesto treat the idea of the perfect language as an act of communication because the language is an extremellynecessary analogic code for the act of communication.Taking into consideration this perspective, we considerinteresting the works of the two cultural personalities: Ioan Petru Culianu’s „The Creation Language ” andUmberto Eco’s „Looking for the Perfection of Language in European Culture ”.The two authors discurs aboutthe concept of creation language even if the first one creates a story which plot is among and around theoriginal language and undelines the existence of creation language inside a misterious box and the second onediscusses the theme from the linguistic point of view and considers that the perfect language is the original one itis the one given to Adam by Good, the one which must be regained.More and more than this, from the European Integration perspective, we can aim to this ideal ilustratedby the „perfection of language ” sentence without affecting the national boundaries which every nationality has.The creation language could become a kind of connection even if only at the utopia level.
Picturing the world—cinematic globalization in the deserts of Babel
Directory of Open Access Journals (Sweden)
Mads Anders Baggesgaard
2013-11-01
Full Text Available Globalization remains a challenge for the art of cinema. No art form is more suited to the task of showing clashes between cultures and the internal conflicts of a society, but as films are both narratively and physically dependent on locations—even if these can be multiple and dispersed throughout the world—and because of the logistics and the finances required for the production of film, cinema has almost always been placed in a national or regional framework. Reflecting the totality and networked nature of the globalized world seems more readily attainable for more conceptual forms of art. This article discusses Alejandro Gonzales Iñárritu's 2006 film Babel, often cited as the “first film of globalization,” asking the question of whether this claim can be substantiated alone with reference to the networked narrative of the film and use of multiple locations, suggesting that the relationship between cinema and globalization should in fact be understood on the terms of the medium as a visual reflection of images of the globe. Drawing on theories on the visual nature of globalization by Arjun Appadurai, Martin Heidegger, and W. J. T. Mitchell, this article thus argues for a different conception of cinematic globalization rooted in the history of cinema rather than in theories of globalization.
Rem Koolhaas y la nueva Babel. De la torre metropolitana al monumento al vacío
Directory of Open Access Journals (Sweden)
José Antonio Tallón
2015-05-01
Full Text Available Un primer acercamiento a las reflexiones de Rem Koolhaas en torno a la tipología de torre introduce al rascacielos neoyorquino como la alegoría del “automonumento”: una construcción en esencia destinada a reafirmar su sola presencia y que se distingue del resto por medio de su estatura, que la monumentaliza. La torre de Babel, símbolo inquebrantable de la leyenda de la construcción en altura, escenifica una historia de construcción y destrucción que está vinculada ineludiblemente al pensamiento crítico de Rem Koolhaas en torno a la torre como tipología desacreditada. Un recorrido por las distintas “Babel” que Rem Koolhaas cataloga en el glosario de términos incluido en el texto SMLXL construye un discurso en torno a la destrucción de la torre bíblica y la construcción de la nueva Babel koolhaasiana que inicia su recorrido con el rascacielos para acabar reclamando un nuevo estado de monumentalidad: la ausencia en su estado más puro representado por el muro, el máximo ejemplo de ausencia como la forma más elevada de presencia monumental. Una mirada crítica que comienza con la torre metropolitana como la nueva Babel para finalizar con el muro como el monumento al vacío
International Nuclear Information System (INIS)
Sensitivity and uncertainty calculations methods of neutronics parameters in pressurized light water reactors have been developed. The sensitivity is composed of three terms; the first is the sensitivity of cell-averaged multi-group cross-sections relative to multi-group infinite dilution cross-sections, the second is the sensitivity of assembly averaged few-group macroscopic cross-sections relative to cell-averaged multi-group cross-sections, and the third is the sensitivity of neutronics parameters in PWR cores relative to few-group macroscopic cross-sections. Combining the three sensitivities, the sensitivity of neutronics parameters in PWR cores relative to multi-group infinite dilution cross-sections is obtained. The discussion of this method will be presented in two papers; the present paper is part I, where the theory and some numerical results for typical pin cells, fuel assemblies and a simple PWR core are shown. The present method gives us multi-group sensitivities for individual nuclides in each reaction type, and wide ranges of applications are possible to the fields such as cross-section adjustment and uncertainty reduction. (author)
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.
Rem Koolhaas y la nueva Babel. De la torre metropolitana al monumento al vacío
José Antonio Tallón
2015-01-01
Un primer acercamiento a las reflexiones de Rem Koolhaas en torno a la tipología de torre introduce al rascacielos neoyorquino como la alegoría del “automonumento”: una construcción en esencia destinada a reafirmar su sola presencia y que se distingue del resto por medio de su estatura, que la monumentaliza. La torre de Babel, símbolo inquebrantable de la leyenda de la construcción en altura, escenifica una historia de construcción y destrucción que está vinculada ineludiblemente al pensami...
The background cross section method for calculating the epithermal neutron spectra
International Nuclear Information System (INIS)
We have developed a new methodology to the multigroup constants calculations, for thermal and fast reactors. The method to obtain the constants is extremely fast and simple, and it avoid repeated computations of the detailed neutron spectrum for different cell configurations (composition, geometry and temperature). (author)
Consistency of differential and integral thermonuclear neutronics data
International Nuclear Information System (INIS)
To increase the accuracy of the neutronics analysis of nuclear reactors, physicists and engineers have employed a variety of techniques, including the adjustment of multigroup differential data to improve consistency with integral data. Of the various adjustment strategies, a generalized least-squares procedure which adjusts the combined differential and integral data can significantly improve the accuracy of neutronics calculations compared to calculations employing only differential data. This investigation analyzes 14 MeV neutron-driven integral experiments, using a more extensively developed methodology and a newly developed computer code, to extend the domain of adjustment from the energy range of fission reactors to the energy range of fusion reactors
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