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
International Nuclear Information System (INIS)
1 - Description of program or function: VITENEA-E is a coupled 174-neutron, 38-gamma-ray multigroup cross sections library in AMPX format suitable for multidimensional radiation transport calculations and dose evaluation. To produce that library, the file ENDF/B-VI has been chosen as primary source of basic nuclear data because it is suitable for processing using the NJOY-SMILER-SCALE sequence. That file is adequate for fusion calculations since it allows important features, like: - data for angle-energy correlation of high energy neutrons, charged particles and recoil nuclei; - separate isotopic evaluations for the main structural materials; - photon production data for the main structural materials. The neutron weighting function is taken from Vitamin-E. It has the form: - Maxwellian thermal spectrum (from 10-5 to 0.414 eV), - 1/E slowing down region (from 0.414 eV to 2.1225 MeV), - fission spectrum (from 2.1225 to 10 MeV), - 1/E (from 10 to 12.52 MeV), - DT fusion neutron spectrum (from 12.52 to 15.68 MeV) - 1/E (from 15.68 to 19.64 MeV). The photon weighting spectrum is assumed to be constant both for production from neutron reactions and for interactions among photons only. The scattering Legendre expansion order is assumed to be 8, both for the neutron and the photon interactions.The library is based upon the Bondarenko method for the resonance self-shielding and the temperature effects. The materials were processed at 293.6, 900, 2100 K. Ten values of the background cross sections are considered: 10+10, 1000, 300, 100, 30, 10, 3, 1, 0.1, 10-6. The basic nuclear data (ENDF-B/VI) related to the following materials have been processed and included into VITENEA-E: H (free gas), H (water), H-3, He-4, Li-6, Li-7, Be-9 (metallic), Be-9 (BeO), C (free gas), C (graphite), O, Al-27, Si,, Ti, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Mo, W-nat, W-182, W183, W184, W186, Hf-176, Hf-177, Hf-178
MGAV10: the latest evolution in the multi-group analysis code - 16248
International Nuclear Information System (INIS)
At many points in the safe and transparent handling of plutonium materials the relative isotopic composition of the principle isotopes needs to be known. Sometimes this information may be of primary interest - such as in the verification of safeguard declarations or in the confirmation of the reactivity of mixed oxide fuel. At other times, e.g., for radioactive waste characterization, the isotopic composition may be needed to calculate specific thermal power or specific spontaneous fission rates for the item under study, which can subsequently be combined with calorimetric and correlated neutron counting measurements, respectively, in order to make quantitative assessments of the mass of Pu and associated nuclides that are present in an item. The Multi-Group Analysis code MGA is a highly regarded and widely used computer code for the analysis of high resolution gamma ray spectra in order to extract the relative isotopic composition of plutonium for a diversity of items with minimal prior information. It has been honed over many years to give reliable results for a broad range of measurement scenarios commonly encountered in the fuel cycle. The nuclear industry is not dormant however and the demands on such codes continue to shift as a combination of technology and necessity open up new application areas. For example, while MGA had its origins in the analysis of clean spectra on product material principally for nuclear safeguards applications taken with germanium detectors having good low-energy resolution, it is now widely applied to the characterization of drummed waste forms and the complex spectra from such items acquired with much larger volume and poorer resolution detectors often used in such applications for the dual use of quantitative assay of the many gamma-emitters. This new domain of operational experience resulted in the need to enhance MGA to deal with spectra of poor statistical quality and also to cope with some of the complications that arise in
International Nuclear Information System (INIS)
The procedures for the numerical and experimental determination of the neutron flux in the zones with the strong neutron absorption and leakage are described in this paper. The proposed procedures have been applied for the determination of the neutron flux in the internal neutron converter used with the RB reactor core configuration number 114. This paper shows: a) that the full heterogeneous core of complicate geometry, as the RB reactor core configuration No. 114, can be reliably calculated using the equivalence approach implemented in the VEGA2DAN code as a complement to the KENO-V.a multigroup Monte Carlo code; and b) a possibility of the RB reactor core configuration No. 114 for the irradiation of large samples in the field of fast neutrons.
A multilevel method for coupling the neutron kinetics and heat transfer equations
International Nuclear Information System (INIS)
We present a computational method for adequate and efficient coupling the neutron transport equation with the precursor and heat transfer equations. It is based on the multilevel nonlinear quasidiffusion (QD) method for solving the multigroup transport equation. The system of equations includes the time-dependent high-order transport equation and time-dependent multigroup and effective one-group low-order QD equations. We also formulate a method applying the α-approximation for the time-dependent high-order transport equation. This approach enables one to avoid storing the angular flux from the previous time step. Numerical results for a model transient problem are presented. (authors)
International Nuclear Information System (INIS)
The spatial eigenfunction expansion method is used to solve the multigroup time-dependent diffusion equation when the absorption cross-section in the thermal group is a function of time. An expression for the multi region reactor transfer function is obtained. Some numerical results for two energy groups are also presented. (author)
SIXTUS-2. A two dimensional multigroup diffusion theory code in hexagonal geometry. Pt. 1
International Nuclear Information System (INIS)
A new algorithm for solving the 2-dimensional multigroup diffusion equations in hexagonal geometry is described. It is based on three novel ideas: analytic intranodal solutions, use of the group irreducible representations and an explicit scheme for solving the response matrix equations. The resulting computer code SIXTUS-2 has been found to be very accurate and effective. (Auth.)
Jones, K.; Johnston, R.; Manley, D.J.; Owen, D.; Charlton, C.
2015-01-01
We develop and apply a multilevel modeling approach that is simultaneously capable of assessing multigroup and multiscale segregation in the presence of substantial stochastic variation that accompanies ethnicity rates based on small absolute counts. Bayesian MCMC estimation of a log-normal Poisson
A discretization of the multigroup PN radiative transfer equation on general meshes
Hermeline, F.
2016-05-01
We propose and study a finite volume method of discrete duality type for discretizing the multigroup PN approximation of radiative transfer equation on general meshes. This method is second order-accurate on a very large variety of meshes, stable under a Courant-Friedrichs-Lewy condition and it preserves naturally the diffusion asymptotic limit.
International Nuclear Information System (INIS)
In this work we address the development and implementation of the analytic coarse-mesh finite-difference (ACMFD) method in a nodal neutron diffusion solver called ANDES. The first version of the solver is implemented in any number of neutron energy groups, and in 3D Cartesian geometries; thus it mainly addresses PWR and BWR core simulations. The details about the generalization to multigroups and 3D, as well as the implementation of the method are given. The transverse integration procedure is the scheme chosen to extend the ACMFD formulation to multidimensional problems. The role of the transverse leakage treatment in the accuracy of the nodal solutions is analyzed in detail: the involved assumptions, the limitations of the method in terms of nodal width, the alternative approaches to implement the transverse leakage terms in nodal methods - implicit or explicit -, and the error assessment due to transverse integration. A new approach for solving the control rod 'cusping' problem, based on the direct application of the ACMFD method, is also developed and implemented in ANDES. The solver architecture turns ANDES into an user-friendly, modular and easily linkable tool, as required to be integrated into common software platforms for multi-scale and multi-physics simulations. ANDES can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. The verification and performance of the solver are demonstrated using both proof-of-principle test cases and well-referenced international benchmarks
Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data
International Nuclear Information System (INIS)
Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author)
Benchmark calculations in multigroup and multidimensional time-dependent transport
International Nuclear Information System (INIS)
It is widely recognized that reliable benchmarks are essential in many technical fields in order to assess the response of any approximation to the physics of the problem to be treated and to verify the performance of the numerical methods used. The best possible benchmarks are analytical solutions to paradigmatic problems where no approximations are actually introduced and the only error encountered is connected to the limitations of computational algorithms. Another major advantage of analytical solutions is that they allow a deeper understanding of the physical features of the model, which is essential for the intelligent use of complicated codes. In neutron transport theory, the need for benchmarks is particularly great. In this paper, the authors propose to establish accurate numerical solutions to some problems concerning the migration of neutron pulses. Use will be made of the space asymptotic theory, coupled with a Laplace transformation inverted by a numerical technique directly evaluating the inversion integral
A New Code for Proto-Neutron Star Evolution
Roberts, Luke F.
2012-01-01
A new code for following the evolution and emissions of proto-neutron stars during the first minute of their lives is developed and tested. The code is one dimensional, fully implicit, and general relativistic. Multi-group, multi-flavor neutrino transport is incorporated that makes use of variable Eddington factors obtained from a formal solution of the static general relativistic Boltzmann equation with linearized scattering terms. The timescales of neutrino emission and spectral evolution o...
Interpretation of active neutron measurements by the heterogeneous theory
International Nuclear Information System (INIS)
In this paper are presented results from a study on the application of the heterogeneous method for the interpretation of active neutron measurements. The considered apparatus consists out of a cylindrical lead pile, which is provided with two axial channels: a central channel incorporates an antimony beryllium photoneutron source and an excentric channel serves for the insertion of the sample to be assayed for fissionable materials contents. The mathematical model of this apparatus is the heterogeneous group diffusion theory. Sample and source channel are described by multigroup monopolar and dipolar sources and sinks. Monopolar sources take account of neutron production within energy group and in-scatter from upper groups. Monopolar sinks represent neutron removal by absorption within energy group and outscatter to lower groups. Dipol sources describe radial streaming of neutrons across the sample channel. Multigroup diffusion theory is applied throughout the lead pile. The strengths of the monopolar and dipolar sources and sinks are determined by linear extrapolation distances of azimuthal mean and first harmonic flux values at the channels' surface. In an experiment we may measure the neutrons leaking out of the lead pile and linear extrapolation distances at the channels' surface. Such informations are utilized for interpretation in terms of fission neutron source strengh and mean neutron flux values in the sample. In this paper we summarized the theoretical work in course
International Nuclear Information System (INIS)
This paper is the first in a series of publications dedicated to the description of the new Russian multigroup data set BNAB-93. The first part of this series is devoted to the description of the neutron and photon data and their formats, and to their use in calculations. (author). 12 refs, 8 tabs
International evaluation cooperation Subgroup 7: Multigroup cross section processing
Energy Technology Data Exchange (ETDEWEB)
Roussin, R.W.; White, J.E. (Oak Ridge National Lab., TN (USA)); Sartori, E. (NEA Data Bank, 91 - Gif-sur-Yvette (France)); Panini, G. (ENEA, Bologna (Italy)); MacFarlane, R. (Los Alamos National Lab., NM (USA)); Muir, D. (International Atomic Energy Agency, Vienna (Austria). Nuclear Data Section); Mattes, M. (Stuttgart Univ. (Germany, F.R.). Inst. fuer Kernenergetik und Energiesysteme); Hasegawa, I
1991-01-01
The chairmen of the ENDF/B, JEF, EFF, and JENDL evaluated data files adopted a proposal to develop a fine-group processed cross section library based on the VITAMIN'' concept. The authors listed above, with support from others, are participating in this project. The end result will be a pseudo-problem-independent fine-group cross section library generated from the latest evaluated data in ENDF/B-VI, JEF-2, EFF-2, and JENDL-3. Initial applications of the library will be for shielding, fast reactor physics, and fusion neutronics. Progress made to date will be discussed. 8 refs.
On the neutron-noise transmission studies for non-multiplying media using transport theory
International Nuclear Information System (INIS)
This paper reports the results of our investigations on the neutron-noise transmission characteristics of non-multiplying media using transport theory. The study has been carried out systematically by first considering the infinite medium case for monoenergetic neutrons and then extending it to the finite media, multigroup and anisotropic scattering cases. The results are particularly related with the problems and prospects of the neutron-noise studies by excore detectors in fast reactors and would be particularly useful in developing the technology of malfunction detection by neutron-noise methods. (author)
Modeling of neutron elastic scattering energy deposition in proton recoil counters
International Nuclear Information System (INIS)
For the purpose of determining the neutron energy deposition in proton-recoil detectors, a model based on the multigroup transport theory is developed. The matrix of the averaged recoil nucleus energies represents the entire process of neutron kinetic energy transfer to the target nuclei. The averaged energy recoil nucleus receive is correspondent to the energy loss of a neutron that suffers collision within detector volume. The necessary algorithm for the matrix elements determination is developed. Computer code EESCAT is developed to calculate elastic scattering matrices and recoil nucleus energies received from elastically scattered neutrons. (author)
Neutronic method of soil moisture measurement
International Nuclear Information System (INIS)
The neutronic method is first outlined: analysis of neutron-nucleus interaction phenomena shows that the neutronic parameters of damp soils depend on the water itself, then on the dry density and the total chemical composition. A physical model representing the neutron moisture gage is worked out next, which leads to the establishment of a simple new mathematical expression applicable to two-dimensional geometry using the multigroup theory diffusion approximation. Following a brief description of the principle and technology of moisture gages two essential problems are dealt with: the calibration curve established by a quick and reliable method involving direct measurement of the thermal neutron constants of soils, and the measurement volume which must be approached by the use of a two-dimensional physical model to describe the geometrical conditions correctly. The problem of the vertical and horizontal resolution power of the neutron probe and the limiting case of surface measurements are discussed. Some possibilities offered by epithermal and fast neutron detection are suggested: epicadmic information represents a step forward as a complement to the thermal measurement since it supplies the principle of a moisture-gage independent of the dry density and allows the calibration curves of conventional instruments to be determined in situ. An experimental study of fast neutron space-energy distribution is described extended and specified by a physical model constructed on a Monte Carlo code; this gives the basis of a technique to measure water contents by fast neutron transmission
Directory of Open Access Journals (Sweden)
Fernandes, Renata Sieiro
2010-11-01
Full Text Available 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 potencial para se romper com o que está estabelecido, trazendo a possibilidade da transformação, através do uso da imaginação, da criatividade, da criação, da projeção e da realização – e os jovens podem ser os propiciadores ou os instauradores dessa outra ordem.The aim of this paper is to discuss ways of building knowledge of reality as something that is fragmented, chaotic, in dis-order, taking as its starting point the image-metaphors of the Tower of Babel and the kaleidoscope. While establishing dialogues with literature and films, a discussion of the possible emergence of the new, the original, the innovative along with the traditional is presented. Those elements are brought by the revolutionary potential of young people or youth. Contact with the new carries the potential to break with what has been established, bringing the possibility of transformation through the use of imagination, creativity, creation, projection and realization - and young people can be the enablers or the founders of this new order.
International Nuclear Information System (INIS)
A hybrid multigroup/continuous-energy Monte Carlo algorithm is developed for solving the Boltzmann-Fokker-Planck equation. This algorithm differs significantly from previous charged-particle Monte Carlo algorithms. Most importantly, it can be used to perform both forward and adjoint transport calculations, using the same basic multigroup cross-section data. The new algorithm is fully described, computationally tested, and compared with a standard condensed history algorithm for coupled electron-photon transport calculations
"A snake of black language": il processo come struttura narrativa in Babel Tower di A.S. Byatt
Directory of Open Access Journals (Sweden)
Beatrice Seligardi
2012-04-01
Full Text Available 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 diegetic strategies employed by the narrator convey the narrative dramatization of the conflicts between different, juxtaposed master narratives. The existence and the influence of cultural and gender paradigms are reflected in these discursive instances. The resulting law in literature perspective (the presence in a narrative form of scenes and plots dealing with the legal field could be interpreted as a highly self-reflexive tool, which testimonies the pervasive omnipresence of extra-literary narrative dynamics.
International Nuclear Information System (INIS)
Criticality studies in nuclear fuel cycle are based on Monte Carlo method. These codes use multigroup cross sections which can verify by experimental configurations or by use of reference codes such Tripoli 2. In this Tripoli 2 code nuclear data are errors attached and asked for experimental studies with critical experiences. This is one of the aim of this thesis. To calculate the keff of interacted fissile units we have used the multigroup Monte Carlo code Moret with convergence problems. A new estimator of reactions rates permit to better approximate the neutrons exchange between units and a new importance function has been tested. 2 annexes
Processing ENDF/B-V uncertainty data into multigroup covariance matrices
International Nuclear Information System (INIS)
The purpose of this work is to develop and demonstrate the capability of processing Evaluated Nuclear Data File, system B, version five (ENDF/B-V) uncertainty data into multigroup covariance matrices. These covariances may then be folded with sensitivity coefficients to obtain uncertainties in selected integral parameters such as K-effective and breeding ratio. The project consisted of separating the previous uncertainty processor (PUFF) from the basic nuclear data cross section processor (MINX), updating the uncertanty processor to theENDF/B-V format, programming the processor for new uncertainty data, and demonstrating the processor capabilities by producing a multigroup covariance library. These capabilities were verified in various ways including hand calculations and comparisons with other known results. A computer code named PUFF-II was written to perform the task described above
A conservative multi-group approach to the Boltzmann equations for reactive gas mixtures
Bisi, M.; Rossani, A.; Spiga, G.
2015-11-01
Starting from a simple kinetic model for a quaternary mixture of gases undergoing a bimolecular chemical reaction, multi-group integro-differential equations are derived for the particle distribution functions of all species. The procedure takes advantage of a suitable probabilistic formulation, based on the underlying collision frequencies and transition probabilities, of the relevant reactive kinetic equations of Boltzmann type. Owing to an appropriate choice of a sufficiently large number of weight functions, it is shown that the proposed multi-group equations are able to fulfil exactly, at any order of approximation, the correct conservation laws that must be inherited from the original kinetic equations, where speed was a continuous variable. Future developments are also discussed.
Correction of multigroup cross sections for resolved resonance interference in mixed absorbers
International Nuclear Information System (INIS)
The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed
Second order time evolution of the multigroup diffusion and P1 equations for radiation transport
International Nuclear Information System (INIS)
Highlights: → An existing multigroup transport algorithm is extended to be second-order in time. → A new algorithm is presented that does not require a grey acceleration solution. → The two algorithms are tested with 2D, multi-material problems. → The two algorithms have comparable computational requirements. - Abstract: An existing solution method for solving the multigroup radiation equations, linear multifrequency-grey acceleration, is here extended to be second order in time. This method works for simple diffusion and for flux-limited diffusion, with or without material conduction. A new method is developed that does not require the solution of an averaged grey transport equation. It is effective solving both the diffusion and P1 forms of the transport equation. Two dimensional, multi-material test problems are used to compare the solution methods.
TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies
International Nuclear Information System (INIS)
TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies
International Nuclear Information System (INIS)
1 - Description of program or function: DIF3D solves multigroup diffusion theory eigenvalue, adjoint, fixed source and criticality (concentration search) problems in 1-, 2- and 3-space dimensions for orthogonal (rectangular or cylindrical), triangular and hexagonal geometries. Anisotropic diffusion coefficients are permitted. Flux and power density maps by mesh cell and region-wise balance integrals are provided. Although primarily designed for fast reactor problems, up-scattering and internal black boundary conditions are also treated. The DIF3D8.0/VARIANT8.0 release differs from the previous DIF3D7.0 release in that it includes a significantly expanded set of solution techniques using variational nodal methods. DIF3D's nodal option solves the multigroup steady state neutron diffusion equation in two- and three-dimensional hexagonal and cartesian geometries and solves the transport equation in two-and three-dimensional cartesian geometries. Eigenvalue, adjoint, fixed source and criticality (concentration) search problems are permitted as are anisotropic diffusion coefficients. Flux and power density maps by mesh cell and region-wise balance integrals are provided. Although primarily designed for fast reactor problems, up-scattering and for finite difference option only internal black boundary conditions are also treated. VARIANT solves the multigroup steady-state neutron diffusion and transport equations in two- and three-dimensional Cartesian and hexagonal geometries using variational nodal methods. The transport approximations involve complete spherical harmonic expansions up to order P5. Eigenvalue, adjoint, fixed source, gamma heating, and criticality (concentration) search problems are permitted. Anisotropic scattering is treated, and although primarily designed for fast reactor problems, up-scattering options are also included. Related and Auxiliary Programs: DIF3D reads and writes the standard interface files specified by the Committee on Computer Code
MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX
International Nuclear Information System (INIS)
reaction cross section of interest and the gross spectral shape. The integration scheme actually employed in MINX is adaptive Simpson's procedure for which the initial estimate is based on the unionized grid described above. The computation of elastic and discrete group- to-group matrices is based upon a semi-analytic scheme which treats the rapidly fluctuating cross-section behaviour analytically. Where this laboratory-system-based scheme becomes difficult to implement (e.g., light nuclei, inelastic thresholds), an alternative numerical integration in the center-of mass system is employed. Multigroup transfer matrices for processes in which the outgoing neutron energy and angular distribution is uncoupled are computed by direct numerical integration. 3 - Restrictions on the complexity of the problem: The principal restriction is the computing time available for a given desired accuracy, number of groups, and Legendre order. The paging technique and variable dimensioning make efficient use of available core storage; very large problems have been run with MINX (e.g. a complete 171-group P3 neutron library at ORNL and an extensive 240-group P4 library at LASL)
International Nuclear Information System (INIS)
This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)
International Nuclear Information System (INIS)
For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIAR next generation core simulation software package. ARCADIAR will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP3 nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIAR. For enabling a high computational performance, the SPN calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP3 is about 1.4 times as expensive computationally as SP1 (diffusion). The developed SP3 solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIAR. With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP3 instead of SP1 has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SPN results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO2/MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP3 computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)
Development of Library Processing System for Neutron Transport Calculation
Energy Technology Data Exchange (ETDEWEB)
Song, J. S.; Park, S. Y.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)
2008-12-15
A system for library generation was developed for the lattice neutron transport program for pressurized water reactor core analysis. The system extracts multi energy group nuclear data for requested nuclides from ENDF/B whose data are based on continuous energy, generates hydrogen equivalent factor and resonance integral table as functions of temperature and background cross section for resonance nuclides, generates subgroup data for the lattice program to treat resonance exactly as possible, and generates multi-group neutron library file including nuclide depletion data for use of the lattice program.
Neutron Spectra Near to a Temperature Discontinuity in Graphite
International Nuclear Information System (INIS)
Thermal neutron spectra have been measured on either side of a temperature discontinuity in a large graphite stack. The stack was fed with thermal neutrons from the source reactor LIDO. Spectra were measured by the time-of-flight method using a chopper, and also by integral activation techniques. Simple diffusion theory calculations considering two Maxwellian groups give as good agreement with the measured 'neutron temperature' distributions as multigroup transport theory. Direction-averaged spectra were calculated using the transport theory code WDSN, with 39 thermal energy groups. Results using heavy gas (mass 30 or 35) scattering kernels for graphite or realistic kernels based on the measured scattering law give agreement with the measured spectra to better than 10%. The measured spectra of fluxes in the direction of the neutron current were found to be dependent on the orientation of the graphite blocks since the mean free path of neutrons in graphite depends on its extrusion direction. (author)
Basic neutronics. Neutrons migration
International Nuclear Information System (INIS)
This article presents the basic neutronics necessary for the understanding of the operation of the different types of nuclear reactors: 1 - introduction to neutronics: principle of fission chain reactions, fast neutron reactors and thermal neutron reactors, capture, neutron status, variations with the reactor lattices; 2 - Boltzmann equation: neutrons population, neutrons migration, characterization of neutrons population and reactions, integral form of the Boltzmann equation, integral-differential form, equivalence between the two forms; 3 - reactor kinetics: fast neutrons and delayed neutrons, kinetic equations in punctual model, Nordheim equation, reactivity jumps, reactivity ramp; 4 - diffusion equation: local neutron status, Fick's law, diffusion equation, initial, boundary and interface conditions, nuclei in infinite and homogenous medium, some examples of solutions, developments in Eigenmodes; 5 - one-group theory: equation of the 'one-group - diffusion' theory, critical condition of the naked and homogenous reactor, critical condition of a reactor with reflectors, generalizations; 6 - neutrons moderation: different moderation mechanisms, elastic shock laws, moderation equation, some examples of solutions; 7 - resonance absorption of neutrons: advantage of the discontinuous moderation character, advantage of an heterogenous disposition, classical formula of the anti-trap factor in homogenous and heterogenous situation; 8 - neutrons thermalization: notions of thermalization mechanisms, thermalization equation, Maxwell spectrum, real spectrum, classical formula of the thermal utilisation factor, classical formula of the reproduction factor, moderation optimum. (J.S.)
A solution of the neutron diffusion equation for a hemisphere containing a uniform source
International Nuclear Information System (INIS)
An analytic solution of the diffusion equation for a hemisphere of fissile or non-fissile material is presented which contains a spatially uniform neutron source. Numerical results are given for the flux distribution for one-speed fast neutrons in 235U and also for a non-fissile element of similar scattering properties. We use these results to check the accuracy of the finite element code EVENT. The procedure is also developed for multigroup calculations. In an Appendix we outline the procedure required when the hemisphere contains a source and is also irradiated by an external current of neutrons
Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor
International Nuclear Information System (INIS)
A detailed MCNP model of the Advanced Neutron Source Reactor has been developed. All reactor components inside the reflector tank were included, and all components were highly segmented. Neutron and photon multigroup flux spectra have been calculated for each segment in the model, and thermal-to-fast neutron flux ratios were determined for each component segment. Axial profiles of the spectra are provided for all components of the reactor. Individual segment statistical uncertainties were limited wherever possible, and the group fluxes for all important reflector components have a standard deviation below 10%
ZZ VITENEA-J, AMPX 175-N,42-gamma multigroup X-sect. library for nuclear fusion applications
International Nuclear Information System (INIS)
1 - Description of program or function: VITENEA-J is a coupled multigroup cross section library in the standard VITAMIN-J energy group structure (175 n + 42 gamma) in AMPX format produced by ENEA-Bologna for nuclear fusion applications.It is based on nuclear data from the general purpose Fusion Evaluated Nuclear Data Library (FENDL/E-2.0) and ENDF-B/VI. This library has been widely used by ENEA group in neutron/gamma transport calculations for ITER safety assessment via the SCALENEA-1 multipurpose Sn calculation sequence. List of the included materials: 1-H-1, 1-H-1(water), 1-H-2, 1-H-3, 2-He-3, 2-He-4, 3-Li-6, 3-Li-7, 4-Be-9, 4-Be-9(metal) , 4-Be-9(oxide), 5-B-10, 5-B-11, C -nat, C -nat(graphite), 7-N-14, 7-N-15, 8-O-16, 9-F-19, 11-Na-23, Mg-nat, 13-Al-27, 14-Si-28, 14-Si-29, 14-Si-30, 15-P-31, S -nat, Cl-nat, K-nat, Ca-nat, Ti-nat, 23-V-51, 24-Cr-50, 24-Cr-52, 24-Cr-53, 24-Cr-54, 25-Mn-55, 26-Fe-54, 26-Fe-56, 26-Fe-57, 26-Fe-58, 27-Co-59, 28-Ni-58, 28-Ni-60, 28-Ni-61, 28-Ni-62, 28-Ni-64, 29-Cu-63, 29-Cu-65, Ga-nat, Zr-nat, 41-Nb-93, Mo-nat, Cd-nat, Sn-nat, Hf-nat, 72-Hf-176, 72-Hf-177 , 72-Hf-178, 72-Hf-179, 72-Hf-180, 73-Ta-181, W-nat, 74-W-182, 74-W-183 , 74-W-184, 74-W-186 , 79-Au-197 , 82-Pb-206, 82-Pb-207, 82-Pb-208, 83-Bi-209, 47-Ag-107 , 47-Ag-109 , ANSI-Flux. 2 - Methods: The data have been prepared by processing basic nuclear data into a fine group problem independent format through an automatic calculation procedure using the modules of NJOY-94.105 and AMPX-77 Processing Systems. The translation of the multigroup data from the GENDF format produced by NJOY to the AMPX Master Library format was performed by means of the AMPX-77 SMILER module. 3 - Restrictions on the complexity of the problem: The library contains 75 nuclides. Seventy nuclides were processed at four temperatures (300 K, 600 K, 900 K, 1500 K) at ten values for the background cross section s0. Thermal scattering cross sections were processed at the three temperatures (296 K, 600 K, 1200 K
Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh
International Nuclear Information System (INIS)
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
Application of equivalence methods on Monte Carlo method based homogenization multi-group constants
International Nuclear Information System (INIS)
The multi-group constants generated via continuous energy Monte Carlo method do not satisfy the equivalence between reference calculation and diffusion calculation applied in reactor core analysis. To the satisfaction of the equivalence theory, general equivalence theory (GET) and super homogenization method (SPH) were applied to the Monte Carlo method based group constants, and a simplified reactor core and C5G7 benchmark were examined with the Monte Carlo constants. The results show that the calculating precision of group constants is improved, and GET and SPH are good candidates for the equivalence treatment of Monte Carlo homogenization. (authors)
Global analysis on a class of multi-group SEIR model with latency and relapse.
Wang, Jinliang; Shu, Hongying
2016-02-01
In this paper, we investigate the global dynamics of a multi-group SEIR epidemic model, allowing heterogeneity of the host population, delay in latency and delay due to relapse distribution for the human population. Our results indicate that when certain restrictions on nonlinear growth rate and incidence are fulfilled, the basic reproduction number R0 plays the key role of a global threshold parameter in the sense that the long-time behaviors of the model depend only on R0. The proofs of the main results utilize the persistence theory in dynamical systems, Lyapunov functionals guided by graph-theoretical approach. PMID:26776266
Simulate-HEX - The multi-group diffusion equation in hexagonal-z geometry
International Nuclear Information System (INIS)
The multigroup diffusion equation is solved for the hexagonal-z geometry by dividing each hexagon into 6 triangles. In each triangle, the Fourier solution of the wave equation is approximated by 8 plane waves to describe the intra-nodal flux accurately. In the end an efficient Finite Difference like equation is obtained. The coefficients of this equation depend on the flux solution itself and they are updated once per power/void iteration. A numerical example demonstrates the high accuracy of the method. (authors)
Multi-group pin power reconstruction method based on colorset form functions
International Nuclear Information System (INIS)
A multi-group pin power reconstruction method that fully exploits nodal information obtained from global coarse mesh solution has been developed. It expands the intra-nodal flux distributions into nonseparable semi-analytic basis functions, and a colorset based form function generating method is proposed, which can accurately model the spectral interaction occurring at assembly interface. To demonstrate its accuracy and applicability to realistic problems, the new method is tested against two benchmark problems, including a mixed-oxide fuel problem. The results show that the new methods is comparable in accuracy to fine-mesh methods. (authors)
Directory of Open Access Journals (Sweden)
Xiaoming Fan
2014-01-01
Full Text Available We discuss multigroup SIRS (susceptible, infectious, and recovered epidemic models with random perturbations. We carry out a detailed analysis on the asymptotic behavior of the stochastic model; when reproduction number ℛ0>1, we deduce the globally asymptotic stability of the endemic equilibrium by measuring the difference between the solution and the endemic equilibrium of the deterministic model in time average. Numerical methods are employed to illustrate the dynamic behavior of the model and simulate the system of equations developed. The effect of the rate of immunity loss on susceptible and recovered individuals is also analyzed in the deterministic model.
On the dynamics of a class of multi-group models for vector-borne diseases
Iggidr, Abderrahman; Sallet, Gauthier; Souza, Max O.
2016-01-01
The resurgence of vector-borne diseases is an increasing public health concern, and there is a need for a better understanding of their dynamics. For a number of diseases, e.g. dengue and chikungunya, this resurgence occurs mostly in urban environments, which are naturally very heterogeneous, particularly due to population circulation. In this scenario, there is an increasing interest in both multi-patch and multi-group models for such diseases. In this work, we study the dynamics of a vector...
Geospatial Data Fusion and Multigroup Decision Support for Surface Water Quality Management
Sun, A. Y.; Osidele, O.; Green, R. T.; Xie, H.
2010-12-01
Social networking and social media have gained significant popularity and brought fundamental changes to many facets of our everyday life. With the ever-increasing adoption of GPS-enabled gadgets and technology, location-based content is likely to play a central role in social networking sites. While location-based content is not new to the geoscience community, where geographic information systems (GIS) are extensively used, the delivery of useful geospatial data to targeted user groups for decision support is new. Decision makers and modelers ought to make more effective use of the new web-based tools to expand the scope of environmental awareness education, public outreach, and stakeholder interaction. Environmental decision processes are often rife with uncertainty and controversy, requiring integration of multiple sources of information and compromises between diverse interests. Fusing of multisource, multiscale environmental data for multigroup decision support is a challenging task. Toward this goal, a multigroup decision support platform should strive to achieve transparency, impartiality, and timely synthesis of information. The latter criterion often constitutes a major technical bottleneck to traditional GIS-based media, featuring large file or image sizes and requiring special processing before web deployment. Many tools and design patterns have appeared in recent years to ease the situation somewhat. In this project, we explore the use of Web 2.0 technologies for “pushing” location-based content to multigroups involved in surface water quality management and decision making. In particular, our granular bottom-up approach facilitates effective delivery of information to most relevant user groups. Our location-based content includes in-situ and remotely sensed data disseminated by NASA and other national and local agencies. Our project is demonstrated for managing the total maximum daily load (TMDL) program in the Arroyo Colorado coastal river basin
International Nuclear Information System (INIS)
The good features of the Analytic Function Expansion Nodal (AFEN) method are utilized to develop a practical scheme for the multigroup diffusion problems, in combination with the polynomial expansion nodal (PEN) method. The thermal group fluxes exhibiting strong gradients are solved by the AFEN method, while the fast group fluxes that are smoother than the thermal group fluxes are solved by the PEN method. The scheme is developed for cores of rectangular and hexagonal geometries. In particular, to model the fast group fluxes in the hexagonal geometry by the PEN method, a polynomial function set which shows good performance in accuracy and numerical stability is derived, in premiere. (author)
A 3D multigroup transport kinetics code in hexagonal geometry for fast reactor transient analysis
International Nuclear Information System (INIS)
A description of the 3D multigroup diffusion/transport kinetics code HEXNODYN is given and numerical results are reported. HEXNODYN couples time integration by the quasi-static method with space integration by HEXNOD's analytic (diffusion option) or discrete ordinates (transport option) nodal method. An equivalent hexagonal version of the KfK rod ejection problem has been set up to validate the diffusion option by comparison with available 2D diffusion codes. The transport option has been validated by comparison with the diffusion option. Numerical results indicate that the diffusion option may be considered as fully validated while the transport version is at least internally consistent
International Nuclear Information System (INIS)
In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core
Specifications for a neutron scattering code for an XY geometry: DAIXY
International Nuclear Information System (INIS)
This report describes the specifications of the DAIXY programme which resolves the difference equations system analogue to the homogeneous problem of multigroup neutron scattering, in a two-dimensional space with XY geometry. The maximum number of points treated is 10 050. The number of groups must be less than 4. The internal iterations are dealt with by the alternating directions method; the external iterations are accelerated by the extrapolation method due to Tchebychev. (authors)
Korja, A.; Lahtinen, R.; Heikkinen, P.; Kukkonen, I. T.; Fire Working Group
2003-04-01
The Karelian - Svecofennian boundary zone has long been recognized as an important suture zone. Three competing models have been proposed for this boundary: continent-arc/continent collision zone, back-arc/retro-arc basin related to NE-directed subduction occurring further SW and strike-slip model, when all the Svecofennian parts are considered exotic. The geometry and style of deformation at depth has not been known and thus, the models have been rather schematic. The new reflection data along FIRE1-profile (2001) and older reflection data (BABEL 2&3) suggest a crocodile structure composed of Karelian passive margin wedge that has caused the splitting of Svecofennian island arc material into crocodile jaws. The reflection and refraction data together with lithological and structural evidence indicate that westward subduction consumed the ocean floor between the continent and the arc. In the onset of the collision, the remnants of the ocean floor (ophiolites) together with sedimentary material were obducted on the advancing continental margin. During continued convergence the young and hot island arc part split after which it was both under and overthrust on the continent. The collision was locked when the thickened continental margin was amalgamated with hard core of the continental island arc.
Multi-group unified nodal method with two-group coarse-mesh finite difference formulation
International Nuclear Information System (INIS)
The one-node kernels of the unified nodal method (UNM) which were originally developed for two-group (2G) problems are extended to solve multi-group (MG) problems within the framework of the 2G coarse-mesh finite difference (CMFD) formulation. The analytic nodal method (ANM) kernel of UNM is reformulated for the MG application by adopting the Pade approximation to avoid the similarity transform required to diagonalize the G x G buckling matrix. In addition, a one-node semi-analytic nodal method (SANM) kernel which is considered adequate for multi-group calculations is also integrated into the UNM formulation by expressing it in the form consistent with the other UNM kernels. As an efficient global solution framework, the 2G CMFD formulation with dynamic group condensation and prolongation is established and the performance of the various MG kernels is examined using various static and transient benchmark problems. It turns out that the SANM kernel is the best one for MG problems not only because it retains accuracy comparable to MGANM with a shorter computing time but also because its accuracy or its convergence does not depend on the eigenvalue range of the buckling matrix of the system. The 2G CMFD formulation with MG one-node UNM kernels turns out to be very effective in that it conveniently accelerates the MG source iteration
The group-level consequences of sexual conflict in multigroup populations.
Directory of Open Access Journals (Sweden)
Omar Tonsi Eldakar
Full Text Available In typical sexual conflict scenarios, males best equipped to exploit females are favored locally over more prudent males, despite reducing female fitness. However, local advantage is not the only relevant form of selection. In multigroup populations, groups with less sexual conflict will contribute more offspring to the next generation than higher conflict groups, countering the local advantage of harmful males. Here, we varied male aggression within- and between-groups in a laboratory population of water striders and measured resulting differences in local population growth over a period of three weeks. The overall pool fitness (i.e., adults produced of less aggressive pools exceeded that of high aggression pools by a factor of three, with the high aggression pools essentially experiencing no population growth over the course of the study. When comparing the fitness of individuals across groups, aggression appeared to be under stabilizing selection in the multigroup population. The use of contextual analysis revealed that overall stabilizing selection was a product of selection favoring aggression within groups, but selected against it at the group-level. Therefore, this report provides further evidence to show that what evolves in the total population is not merely an extension of within-group dynamics.
Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen IV Reactors
Energy Technology Data Exchange (ETDEWEB)
Rahnema, Farzad; Haghighat, Alireza; Ougouag, Abderrafi
2013-11-29
The objective of this proposal is the development of a consistent multi-group theory that accurately accounts for the energy-angle coupling associated with collapsed-group cross sections. This will allow for coarse-group transport and diffusion theory calculations that exhibit continuous energy accuracy and implicitly treat cross- section resonances. This is of particular importance when considering the highly heterogeneous and optically thin reactor designs within the Next Generation Nuclear Plant (NGNP) framework. In such reactors, ignoring the influence of anisotropy in the angular flux on the collapsed cross section, especially at the interface between core and reflector near which control rods are located, results in inaccurate estimates of the rod worth, a serious safety concern. The scope of this project will include the development and verification of a new multi-group theory enabling high-fidelity transport and diffusion calculations in coarse groups, as well as a methodology for the implementation of this method in existing codes. This will allow for a higher accuracy solution of reactor problems while using fewer groups and will reduce the computational expense. The proposed research represents a fundamental advancement in the understanding and improvement of multi- group theory for reactor analysis.
Amiri, Imen
2011-01-01
The literary myth of the Grail, created by Chrétien de Troyes, has common roots with thecosmological myth of Babel which explains the plurality of languages. Can we consider themyth of the Grail as a myth of a language lost, sought and finally regained? The adventure of evangelism in Estoire del Saint Grail and the Holy Grail in the Queste, inspired by theCrusades and the Franciscan missions, converge in a dynamic unity symbolized by the Grail,which prevents language . They tend to gather mat...
Creación de empresa de servicios multilingües Babel Institute modalidad creación de empresa
2013-01-01
Babel Institute será una escuela de idioma que brindara a los colombianos el aprendizaje del idioma inglés con estándares de calidad y certificados por el marco común europeo, por lo cual presenta un de plan de negocio como una alternativa para las personas y empresas que deseen competir en un mercado global con buenas bases de comunicación y dominio de un idioma que es el inglés como una herramienta indispensable
International Nuclear Information System (INIS)
The Generation IV [1] International forum identified six advanced reactor concepts and related fuel cycles along with the R and D programs necessary to achieve the four key goals: (1) sustainability, (2) safety and reliability, (3) economics, (4) proliferation resistance and physical protection. Among these six promising reactor concepts, the lead-cooled fast reactor (LFR) has been selected for development by EURATOM, which in 2006 decided to finance the European Lead Cooled System (ELSY) project. The aim of the project is to demonstrate the possibility to design a safe and competitive lead-cooled fast power reactor using simple engineering solutions. This paper demonstrates the use of the code package SCALE5.1 and its NEWT/TRITON modules [3] for preliminary neutronic core analysis of a LFR within Generation IV Nuclear Energy systems program. More specifically, the analysis of the reference design of the ELSY-600 open square fuel assembly is presented. In particular, the use of ENDF/B-V and ENDF/B-VI.7 and multigroup energy structure was investigated. The homogenized cross sections calculated for the ELSY fuel assembly 2D model have been evaluated and compared to the results obtained with calculations performed with the deterministic code ERANOS/ECCO using JEFF2.2 cross section library. A good agreement has been observed in the energy range of interests, and generally for energy above 1 eV. (authors)
PALLAS-TS: a one-dimensional neutron transport code for analyzing fusion blanket neutronics
International Nuclear Information System (INIS)
The one-dimensional neutron transport code PALLAS-TS has been developed for solving the transport equation by direct numerical integration method. Group-transference kernels are accurately obtained from the double-differential cross section data using the energy and scattering angle correlation relation for elastic and inelastic (discrete levels) scattering. In addition, a usual multigroup model is adopted in calculation of spatial and angular flux distribution so as to make it possible to use iteration technique with neutron rebalancing in each group. This code uses a 120-group data library for 29 nuclides prepared temporarily by processing the ENDF/B-IV file, though the nuclear data file available now is incomplete for accounting fully the anisotropy of scattering. Results of test calculation for a 4-region system consisting of lithium and carbon were compared with the P5-S8 calculations by the ANISN code. The present code is the first trial of incorporating the multigroup to the direct integration method for solving the transport equation. It is observed that computing time by this code is shorter than that of the usual S sub(n) method by a factor of 2 or 3. (author)
International Nuclear Information System (INIS)
Liquid Salt Cooled Reactors (LSCRs) are high temperature reactors, cooled by liquid salt, with a TRISO-particle based fuel in a solid form (stationary fuel elements or circulating fuel pebbles); this paper is focusing on the former. In either case, due to the double heterogeneity, core physics analysis require different considerations with more complex approaches than LWRs core physics calculations. Additional challenges appear when using the multi-group approach. In this paper we examine the use of SCALE6.1.1. Double heterogeneity may be accounted for through the Dancoff factor, however, SCALE6.1.1 does not provide an automated method to calculate Dancoff Factors for fuel planks with TRISO fuel particles. Therefore, depletion with continuous energy Monte Carlo Transport (CE depletion) in SCALE6.2 beta was used to generate MC Dancoff factors for multi-group calculations. MCDancoff corrected multi-group depletion agrees with the results for CE depletion within ±100 pcm, and within ±2σ. Producing MCDancoff factors for multi-group (MG) depletion calculations is necessary to LSCR analysis because CE depletion runtime and memory requirements are prohibitive for routine use. MG depletion with MCDancoff provides significantly shorter runtime and lower memory requirements while providing results of acceptable accuracy. (author)
International Nuclear Information System (INIS)
More accurate modelling of physical phenomena involved in present and future nuclear reactors requires a multi-scale and multi-physics approach. This challenge can be accomplished by the coupling of best-estimate core-physics, thermal-hydraulics and multi-physics solvers. In order to make viable that coupling, the current trends in reactor simulations are along the development of a new generation of tools based on user-friendly, modular, easily linkable, faster and more accurate codes to be integrated in common platforms. These premises are in the origin of the NURESIM Integrated Project within the 6. European Framework Program, which is envisaged to provide the initial step towards a Common European Standard Software Platform for nuclear reactors simulations. In the frame of this project and to reach the above-mentioned goals, a 3-D multigroup nodal solver for neutron diffusion calculations called ANDES (Analytic Nodal Diffusion Equation Solver) has been developed and tested in-depth in this Thesis. ANDES solves the steady-state and time-dependent neutron diffusion equation in three-dimensions and any number of energy groups, utilizing the Analytic Coarse-Mesh Finite-Difference (ACMFD) scheme to yield the nodal coupling equations. It can be applied to both Cartesian and triangular-Z geometries, so that simulations of LWR as well as VVER, HTR and fast reactors can be performed. The solver has been implemented in a fully encapsulated way, enabling it as a module to be readily integrated in other codes and platforms. In fact, it can be used either as a stand-alone nodal code or as a solver to accelerate the convergence of whole core pin-by-pin code systems. Verification of performance has shown that ANDES is a code with high order definition for whole core realistic nodal simulations. In this paper, the methodology developed and involved in ANDES is presented. (authors)
MINARET: Towards a time-dependent neutron transport parallel solver
International Nuclear Information System (INIS)
We present the newly developed time-dependent 3D multigroup discrete ordinates neutron transport solver that has recently been implemented in the MINARET code. The solver is the support for a study about computing acceleration techniques that involve parallel architectures. In this work, we will focus on the parallelization of two of the variables involved in our equation: the angular directions and the time. This last variable has been parallelized by a (time) domain decomposition method called the para-real in time algorithm. (authors)
MAGGENTA: Multiassembly General Geometry Neutron Transport Theory Code
International Nuclear Information System (INIS)
MAGGENTA solves the multigroup steady-state neutron integral transport equation in arbitrary two-dimensional multi-assembly geometries that can be described by combinatorial geometry. Given transport corrected macroscopic cross sections, MAGGENTA solves an eigenvalue problem and calculates the volumetric flux and incoming/outgoing current distributions. MAGGENTA utilizes the p4 Parallel Programming System on a network of workstations or other supercomputers to solve large multi-assembly problems. The solver is optimized for vectro processing on vector machines. A graphical interface has been developed to simplify the assembly layout construction and processor assignments
Global dynamics of a novel multi-group model for computer worms
International Nuclear Information System (INIS)
In this paper, we study worm dynamics in computer networks composed of many autonomous systems. A novel multi-group SIQR (susceptible-infected-quarantined-removed) model is proposed for computer worms by explicitly considering anti-virus measures and the network infrastructure. Then, the basic reproduction number of worm R0 is derived and the global dynamics of the model are established. It is shown that if R0 is less than or equal to 1, the disease-free equilibrium is globally asymptotically stable and the worm dies out eventually, whereas, if R0 is greater than 1, one unique endemic equilibrium exists and it is globally asymptotically stable, thus the worm persists in the network. Finally, numerical simulations are given to illustrate the theoretical results. (general)
International Nuclear Information System (INIS)
A code called COMESH based on corner mesh finite difference scheme has been developed to solve multigroup diffusion theory equations. One can solve 1-D, 2-D or 3-D problems in Cartesian geometry and 1-D (r) or 2-D (r-z) problem in cylindrical geometry. On external boundary one can use either homogeneous Dirichlet (θ-specified) or Neumann (∇θ specified) type boundary conditions or a linear combination of the two. Internal boundaries for control absorber simulations are also tackled by COMESH. Many an acceleration schemes like successive line over-relaxation, two parameter Chebyschev acceleration for fission source, generalised coarse mesh rebalancing etc., render the code COMESH a very fast one for estimating eigenvalue and flux/power profiles in any type of reactor core configuration. 6 refs. (author)
On the feasibility of a homogenised multi-group Monte Carlo method in reactor analysis
International Nuclear Information System (INIS)
The use of homogenised multi-group cross sections to speed up Monte Carlo calculation has been studied to some extent, but the method is not widely implemented in modern calculation codes. This paper presents a calculation scheme in which homogenised material parameters are generated using the PSG continuous-energy Monte Carlo reactor physics code and used by MORA, a new full-core Monte Carlo code entirely based on homogenisation. The theory of homogenisation and its implementation in the Monte Carlo method are briefly introduced. The PSG-MORA calculation scheme is put to practice in two fundamentally different test cases: a small sodium-cooled fast reactor (JOYO) and a large PWR core. It is shown that the homogenisation results in a dramatic increase in efficiency. The results are in a reasonably good agreement with reference PSG and MCNP5 calculations, although fission source convergence becomes a problem in the PWR test case. (authors)
Radiation Transport for Explosive Outflows: A Multigroup Hybrid Monte Carlo Method
Wollaeger, Ryan T; Graziani, Carlo; Couch, Sean M; Jordan, George C; Lamb, Donald Q; Moses, Gregory A
2013-01-01
We explore the application of Implicit Monte Carlo (IMC) and Discrete Diffusion Monte Carlo (DDMC) to radiation transport in strong fluid outflows with structured opacity. The IMC method of Fleck & Cummings is a stochastic computational technique for nonlinear radiation transport. IMC is partially implicit in time and may suffer in efficiency when tracking Monte Carlo particles through optically thick materials. The DDMC method of Densmore accelerates an IMC computation where the domain is diffusive. Recently, Abdikamalov extended IMC and DDMC to multigroup, velocity-dependent neutrino transport with the intent of modeling neutrino dynamics in core-collapse supernovae. Densmore has also formulated a multifrequency extension to the originally grey DDMC method. In this article we rigorously formulate IMC and DDMC over a high-velocity Lagrangian grid for possible application to photon transport in the post-explosion phase of Type Ia supernovae. The method described is suitable for a large variety of non-mono...
Approximate analytical solution of two-dimensional multigroup P-3 equations
International Nuclear Information System (INIS)
Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (orig./RW)
Approximate analytical solution of two-dimensional multigroup P-3 equations
International Nuclear Information System (INIS)
Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (author)
Jones, Kelvyn; Johnston, Ron; Manley, David; Owen, Dewi; Charlton, Chris
2015-12-01
We develop and apply a multilevel modeling approach that is simultaneously capable of assessing multigroup and multiscale segregation in the presence of substantial stochastic variation that accompanies ethnicity rates based on small absolute counts. Bayesian MCMC estimation of a log-normal Poisson model allows the calculation of the variance estimates of the degree of segregation in a single overall model, and credible intervals are obtained to provide a measure of uncertainty around those estimates. The procedure partitions the variance at different levels and implicitly models the dependency (or autocorrelation) at each spatial scale below the topmost one. Substantively, we apply the model to 2011 census data for London, one of the world's most ethnically diverse cities. We find that the degree of segregation depends both on scale and group. PMID:26487190
Analyzing Average and Conditional Effects with Multigroup Multilevel Structural EquationModels
Directory of Open Access Journals (Sweden)
Axel Mayer
2014-04-01
Full Text Available Conventionally, multilevel analysis of covariance (ML-ANCOVA has been therecommended approach for analyzing treatment effects in quasi-experimental multilevel designswith treatment application at the cluster-level. In this paper, we introduce the generalizedML-ANCOVA with linear effect functions that identifies average and conditional treatment effectsin the presence of treatment-covariate interactions. We show how the generalized ML-ANCOVAmodel can be estimated with multigroup multilevel structural equation models that offerconsiderable advances compared to traditional ML-ANCOVA. The proposed model takes intoaccount measurement error in the covariates, sampling error in contextual covariates,treatment-covariate interactions, and stochastic predictors. We illustrate the implementation ofML-ANCOVA with an example from educational effectiveness research where we estimateaverage and conditional effects of early transition to secondary schooling on readingcomprehension.
ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma0 are not guaranteed because of the approximations used in the unresolved resonance region
Recent validation experience with multigroup cross-section libraries and scale
International Nuclear Information System (INIS)
This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment
A New Monte Carlo Neutron Transport Code at UNIST
International Nuclear Information System (INIS)
Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results
ZZ AMZ, 70-Group 40 Isotope Multigroup Library for Fast Reactor Calculation
International Nuclear Information System (INIS)
1 - Description of program or function: format: EXPANDA; number of groups: 70-group library of multigroup constants; nuclides: H-1, Be-9, B-10, B-11, C-12, O-16, N-23, Mg, Al-27, Si, Ti, V, Cr, Mn-55, Fe, Ni, Cu, Ga, Zr, Nb-93, Mo, In-115, Sn, Pb, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, and lumped fission products of U-233, U-235, Pu-239. origin: ENDF/B-IV and ENDF/B-V; weighting spectrum: Fission products inventories for BBR reactor at 360 and 600 days of irradiation were calculated and used as weighting function. AMZ is a 70-group library of multigroup constants for the fast reactor nuclear design code EXPANDA. Data is stored for three temperatures (300 K, 900 K, 2100 K) and for seven background cross sections. The following isotopes are available: H1, Be9, B10, B11, C12, O16, N23, Mg, Al27, Si, Ti, V, Cr, Mn55, Fe, Ni, Cu, Ga, Zr, Nb93, Mo, In115, Sn, Pb, Th232, Pa233, U233, U234, U235, U236, U238, Pu238, Pu239, Pu240, Pu241, Pu242, Am241, and lumped fission products of U233, U235, Pu239. 2 - Method of solution: Nuclear cross sections, transfer matrices, and self-shielding factors were generated from ENDF/B-IV data using the codes NJOY (PSR-0171) and RGENDF
Dorval, Eric
2016-01-01
Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient model...
Prismatic VHTR neutronic benchmark problems
Energy Technology Data Exchange (ETDEWEB)
Connolly, Kevin John, E-mail: connolly@gatech.edu [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States); Rahnema, Farzad, E-mail: farzad@gatech.edu [Nuclear and Radiological Engineering and Medical Physics Programs, George W. Woodruff School, Georgia Institute of Technology, Atlanta, GA (United States); Tsvetkov, Pavel V. [Department of Nuclear Engineering, Texas A& M University, College Station, TX (United States)
2015-04-15
Highlights: • High temperature gas-cooled reactor neutronics benchmark problems. • Description of a whole prismatic VHTR core in its full heterogeneity. • Modeled using continuous energy nuclear data at a representative hot operating temperature. • Benchmark results for core eigenvalue, block-averaged power, and some selected pin fission density results. - Abstract: This paper aims to fill an apparent scarcity of benchmarks based on high temperature gas-cooled reactors. Within is a description of a whole prismatic VHTR core in its full heterogeneity and modeling using continuous energy nuclear data at a representative hot operating temperature. Also included is a core which has been simplified for ease in modeling while attempting to preserve as faithfully as possible the neutron physics of the core. Fuel and absorber pins have been homogenized from the particle level, however, the blocks which construct the core remain strongly heterogeneous. A six group multigroup (discrete energy) cross section set has been developed via Monte Carlo using the original heterogeneous core as a basis. Several configurations of the core have been solved using these two cross section sets; eigenvalue results, block-averaged power results, and some selected pin fission density results are presented in this paper, along with the six-group cross section data, so that method developers may use these problems as a standard reference point.
Prismatic VHTR neutronic benchmark problems
International Nuclear Information System (INIS)
Highlights: • High temperature gas-cooled reactor neutronics benchmark problems. • Description of a whole prismatic VHTR core in its full heterogeneity. • Modeled using continuous energy nuclear data at a representative hot operating temperature. • Benchmark results for core eigenvalue, block-averaged power, and some selected pin fission density results. - Abstract: This paper aims to fill an apparent scarcity of benchmarks based on high temperature gas-cooled reactors. Within is a description of a whole prismatic VHTR core in its full heterogeneity and modeling using continuous energy nuclear data at a representative hot operating temperature. Also included is a core which has been simplified for ease in modeling while attempting to preserve as faithfully as possible the neutron physics of the core. Fuel and absorber pins have been homogenized from the particle level, however, the blocks which construct the core remain strongly heterogeneous. A six group multigroup (discrete energy) cross section set has been developed via Monte Carlo using the original heterogeneous core as a basis. Several configurations of the core have been solved using these two cross section sets; eigenvalue results, block-averaged power results, and some selected pin fission density results are presented in this paper, along with the six-group cross section data, so that method developers may use these problems as a standard reference point
International Nuclear Information System (INIS)
1 - Description of problem or function: KENO is a multigroup, Monte Carlo criticality code containing a special geometry package which allows easy description of systems composed of cylinders, spheres, and cuboids (rectangular parallelepipeds) arranged in any order with only one restriction. They cannot be rotated or translated. Each geometrical region must be described as completely enclosing all regions interior to it. For systems not describable using this special geometry package, the program can use the generalized geometry package (GEOM) developed for the O5R Monte Carlo code. It allows any system that can be described by a collection of planes and/or quadratic surfaces, arbitrarily oriented and intersecting in arbitrary fashion. The entire problem can be mocked up in generalized geometry, or one generalized geometry unit or box type can be used alone or in combination with standard KENO units or box types. Rectangular arrays of fissile units are allowed with or without external reflector regions. Output from KENO consists of keff for the system plus an estimate of its standard deviation and the leakage, absorption, and fissions for each energy group plus the totals for all groups. Flux as a function of energy group and region and fission densities as a function of region are optional output. KENO-4: Added features include a neutron balance edit, PICTURE routines to check the input geometry, and a random number sequencing subroutine written in FORTRAN-4. 2 - Method of solution: The scattering treatment used in KENO assumes that the differential neutron scattering cross section can be represented by a P1 Legendre polynomial. Absorption of neutrons in KENO is not allowed. Instead, at each collision point of a neutron tracking history the weight of the neutron is reduced by the absorption probability. When the neutron weight has been reduced below a specified point for the region in which the collision occurs, Russian roulette is played to determine if the
Reddy, A. R.; Rao, M. V. N.
2012-01-01
The field of neutron radiography with special reference to isotopic neutron radiography has been reviewed. Different components viz., sources, collimators, imaging systems are described. Various designs of neutron radiography facilities, their relative merits and demerits , the appropriateness of each design depending on the object to be radiographed, and economics of each technique are also dealt. The applications of neutron radiography are also briefly presented.
International Nuclear Information System (INIS)
Neutron research where reflection, refraction, and interference play an essential role is generally referred to as 'neutron optics'. The neutron wavelength, the scattering length density and the magnetic properties of the material determine the critical angle for total reflection. The theoretical background of neutron reflection, experimental methods and the interpretation of reflection data are presented. (K.A.)
International Nuclear Information System (INIS)
An extended version of Hassitt's one-dimension Multi-group diffusion programme has been prepared which allows for a maximum of twenty-two energy groups rather than sixteen. It also permits the use of drum storage by programme and data up to the full machine capacity of 1024 sectors, rather than 478 sectors. Some minor corrections have been made, A binary tape of an 830-sector version has been prepared (Winfrith P.5272). (author)
International Nuclear Information System (INIS)
This paper presents the quantification of resonance interference effect for multi-group effective cross-section in lattice physics calculation. In the resonance self-shielding method based on the equivalence theory, the resonance interference effect among multiple nuclides cannot be treated directly to the multi-group effective cross-section. The continuous energy or the ultra-fine-group treatment can directly consider the effect, but the application to the fuel assembly geometry is not realistic with practical computation time. In the present study, the resonance interference effect to the multi-group effective cross-section is simply quantified by the resonance interference factor (RIF) in order to confirm the benefit for considering the effect. The RIF is generated for the typical pin-cell geometry of water moderated system. The multi-group effective cross-sections with and without RIFs are compared with the continuous energy Monte-Carlo result. As a result, the significant impact for considering the resonance interference effect is confirmed to the limited nuclide, reaction type and energy group. Fortunately, these have small effect on k-infinity because the resonance interference effect is mainly induced by the wide resonances of 238U to the other minor nuclides (e.g., 235U, 239Pu) in the limited resonance energy ranges. The results also show that the effect is small to the absorption cross-section of 238U, which is the dominant resonance nuclide in the fuel. The quantification results in the present study indicate a useful material to investigate the more advanced resonance treatment for the next generation lattice physics code. (author)
International Nuclear Information System (INIS)
A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author)
Sensitivity analysis in the fast neutron field
International Nuclear Information System (INIS)
Report presents first part of activities which were completed within task 7: 'Sensitivity analysis in the fast neutron field'. It includes general mathematical formulation of linear and bilinear relevant functionals, as well as special forms of characteristic values. In addition, explicit form of transport equation is derived. It should be solved for the need of sensitivity analysis. Based on presented mathematical method and review of existing methods, a computation procedure is conceived. It is made of 3 parts: preparation of multigroup constants, solving the transport equations and calculation of functionals. ENDF/B-IV data, service code NJOY, RFPN code for solving transport equations and ANOS code for calculating the functionals and sensitivity analysis will be used. RFPN code ws adapted for this purpose and the ANOS code needs additional work in the forthcoming phase
International Nuclear Information System (INIS)
The code system which is a compilation of well known codes NJOY, AMPX, ANISN and DOT-3 adapted for NDP Fortran on IBM PC/AT is described. The 171-group library DLC-41/VITAMIN-C with Bondarenko Factors based on the ENDF/B-4 is introduced to the system as a main data library. Modification and development of this library are realized by NJOY code on the base of ENDF/B-6 files. A 171-group problem-oriented data library is usually used in 1 dimensional calculations by the code ANISN. This library is being generated by the AMPX modules (BONAMI and others). Data libraries with smaller number of groups are being used in 2 dimensional calculations by DOT-3. These libraries have been gotten from 171-group problem-oriented libraries which are averaged by corresponding spectra from 1 dimensional calculations. Using of the described system is being demonstrated with two examples. (authors). 6 refs., 2 tabs., 5 figs
International Nuclear Information System (INIS)
In providing THERM-126 with cross section matrices for deuterium bound in heavy water the IKE phonon spectrum was reevaluated. The changes are modifications in the acoustic part and in the frequency of the second oscillator. Contrary to the phonon spectrum model for D in D2O in ENDF/B-IV the broad band of hindered rotations is assumed to be temperature dependent taking into account the diffusive motion of the molecule. With the new model scattering law data S (α, β) are generated in the temperature range 293.6 K-673.6 K. The THERM-126 scattering cross section matrices are calculated up to P3. As a validity check a lot of differential and integral cross sections are compared to experiments and benchmarks are recalculated. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Van Geemert, Rene [AREVA, AREVA NP, Erlangen (Germany)
2008-07-01
For satisfaction of future global customer needs, dedicated efforts are being coordinated internationally and pursued continuously at AREVA NP. The currently ongoing CONVERGENCE project is committed to the development of the ARCADIA{sup R} next generation core simulation software package. ARCADIA{sup R} will be put to global use by all AREVA NP business regions, for the entire spectrum of core design processes, licensing computations and safety studies. As part of the currently ongoing trend towards more sophisticated neutronics methodologies, an SP{sub 3} nodal transport concept has been developed for ARTEMIS which is the steady-state and transient core simulation part of ARCADIA{sup R}. For enabling a high computational performance, the SP{sub N} calculations are accelerated by applying multi-level coarse mesh re-balancing. In the current implementation, SP{sub 3} is about 1.4 times as expensive computationally as SP{sub 1} (diffusion). The developed SP{sub 3} solution concept is foreseen as the future computational workhorse for many-group 3D pin-by-pin full core computations by ARCADIA{sup R}. With the entire numerical workload being highly parallelizable through domain decomposition techniques, associated CPU-time requirements that adhere to the efficiency needs in the nuclear industry can be expected to become feasible in the near future. The accuracy enhancement obtainable by using SP{sub 3} instead of SP{sub 1} has been verified by a detailed comparison of ARTEMIS 16-group pin-by-pin SP{sub N} results with KAERI's DeCart reference results for the 2D pin-by-pin Purdue UO{sub 2}/MOX benchmark. This article presents the accuracy enhancement verification and quantifies the achieved ARTEMIS-SP{sub 3} computational performance for a number of 2D and 3D multi-group and multi-box (up to pin-by-pin) core computations. (authors)
Riva, Silvia. – Rulli di tam-tam dalla torre di Babele. Storia della letteratura del Congo-Kinshasa
Morabito, Vittorio
2007-01-01
C’est le « chant lourd du tam-tam intérieur » du poète Mukala Kadima-Nzuji qu’évoquerait le titre « rulli di tamburi ». La deuxième référence, la « tour de Babel », n’est pas, nous voulons le croire, une assemblée où tout le monde parle sans s’entendre, où personne n’est d’accord, ni l’annonce de la fin d’un monde rêvé. Au contraire, ce serait l’espérance d’atteindre un monde désiré. L’auteur, Silvia Riva, jalonne son parcours à travers les étapes de la formation de la République démocratique...
Deterministic adjoint transport applications for He-3 neutron detector design
International Nuclear Information System (INIS)
This work focuses on the determination of predicted neutron detector response accomplished using neutron importance derived from an adjoint discrete ordinates (SN) transport calculation. A hypothetical detector apparatus, intended to detect fast neutrons, was modeled using He-3 tubes with graphite moderation using the PENTRANTM 3-D multi-group discrete ordinates parallel transport code system. The detector geometry was modeled using z-axis symmetry and discretized into 30,280 3-D Cartesian cells. The material spatial mesh was generated using the PENMSHTM code in the PENTRAN system. The 47-group BUGLE-96 neutron cross section library was used for construction of macroscopic neutron cross sections. Results from an S8 angular quadrature using P3 anisotropy are presented. An adjoint transport source was established in the model using group dependent He-3 response cross sections. Each He-3 tube contained an adjoint source aliased to group He-3 absorption cross sections to permit assessment of detector performance. The spectrally dependent detector response from neutron capture in He-3 tubes from an arbitrary source can, therefore, be readily determined. This response comes from the complete integral of the actual source strength weighted by the adjoint function at the source location for any source distribution scenario. For selected neutron energies, an equivalent forward MCNP Monte Carlo model was used to demonstrate good agreement with the detector response determined from the adjoint calculation. The graphite used in this design has a large impact on detector performance due to the increasing sensitivity inherent in He-3 gas as neutrons thermalize. Computational adjoint results presented here predict a fast neutron detector design that yields efficiencies between 30 and 50% for neutron energies below 3 keV, and up to 30% efficiencies for neutron energies between 3 keV and 1 MeV. Overall, the methodology applied here highlights the elegant nature of an adjoint
FTR Set 500: a multigroup cross-section set for FTR analysis
International Nuclear Information System (INIS)
FTR Set 500 is a 53-neutron-group, 20-photon-group, cross-section set based on ENDF/B-V cross sections and neutron spectra typical of the Fast Test Reactor (FTR). This report describes the specifications and processing of Set 500 and provides one-group values of this set for use in limited FTR analyses
Neutron Skins and Neutron Stars
Piekarewicz, J
2013-01-01
The neutron-skin thickness of heavy nuclei provides a fundamental link to the equation of state of neutron-rich matter, and hence to the properties of neutron stars. The Lead Radius Experiment ("PREX") at Jefferson Laboratory has recently provided the first model-independence evidence on the existence of a neutron-rich skin in 208Pb. In this contribution we examine how the increased accuracy in the determination of neutron skins expected from the commissioning of intense polarized electron be...
A multigroup analysis from a continuos energy spectrum approach by a MC method
International Nuclear Information System (INIS)
In this work, the Monte Carlo method is applied to the energy dependent three- dimensional neutron transport equation, in order to analyze the change in the spectrum energy depending on the Monte Carlo step. The present work is a first step into a new direction where spectral influence on criticality may be analyzed. The method is based on the monitoring of a large number of individual realizations of neutron histories (i.e. microscopic interaction sequence) where the average behavior of neutrons yields an approximate solution for the neutron transport equation. The Monte Carlo is implemented using continuous functions, with respect to energy, for the cross sections of materials, functions which are obtained by parametrizations of the cross sections. The type of interaction that the neutron will suffer and the characteristics of their displacement in the element are estimated randomly following the relevant probability distributions. (author)
FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: FEM-2D solves the two-dimensional diffusion equation in x-y geometry. This is done by the finite elements method. 2 - Method of solution: FEM-2D uses triangular elements with first and second order Lagrange approximations. The systems equations are formulated in multigroup form and solved by Cholesky procedure which operates only on nonzero elements. Various acceleration techniques are available for the outer iteration. Fluxes along various lines and rates in arbitrary zones may be output. 3 - Restrictions on the complexity of the problem: The code uses variable dimensioning. Thus, the problem size is restricted by the largest array which usually is the systems matrix. Fluxes of all groups are kept in memory. This might become another restrictive data set for a large number of groups. The validity of the results is restricted by the approximations used. FEM-2D requires a finite element net which allows the approximation of fluxes by at most parabolas. The node distribution should be more dense in areas of heavy flux changes (near absorbers or the reflector)
Reflector modelling of small high leakage cores making use of multi-group nodal equivalence theory
International Nuclear Information System (INIS)
This research focuses on modelling reflectors in typical material testing reactors (MTRs). Equivalence theory is used to homogenise and collapse detailed transport solutions to generate equivalent nodal parameters and albedo boundary conditions for reflectors, for subsequent use in full core nodal diffusion codes. This approach to reflector modelling has been shown to be accurate for two-group large commercial light water reactor (LWR) analysis, but has not been investigated for MTRs. MTRs are smaller, with much larger leakage, environment sensitivity and multi-group spectrum dependencies than LWRs. This study aims to determine if this approach to reflector modelling is an accurate and plausible homogenisation technique for the modelling of small MTR cores. The successful implementation will result in simplified core models, better accuracy and improved efficiency of computer simulations. Codes used in this study include SCALE 6.1, OSCAR-4 and EQUIVA (the last two codes are developed and used at Necsa). The results show a five times reduction in calculational time for the proposed reduced reactor model compared to the traditional explicit model. The calculated equivalent parameters however show some sensitivity to the environment used to generate them. Differences in the results compared to the current explicit model, require more careful investigation including comparisons with a reference result, before its implementation can be recommended. (authors)
Definition and analysis of heavy water reactor benchmarks for testing new multigroup libraries
International Nuclear Information System (INIS)
A set of heavy water reactor benchmarks has been selected for testing new WIMS-D libraries. The libraries were constricted using data from ENDF/B-VI, Release 7, JENDL-3.2 and JEF-2.2 evaluated nuclear data files. The benchmarks cover a wide variety of reactor types and conditions, from fresh fuel to high burnup, and for natural and enriched uranium and Th-U fuels. The main parameters compared are the effective multiplication factor and other integral parameters, and isotopic composition of actinides on burnup cases. Besides, further investigations related with energy spectra used for preparation of WIMS-D libraries when applied on HWTR reactor calculations are included. Mostly of the benchmarks show a good agreement between experimental measurements and calculated values for all libraries. One exception is Th232 benchmark, were it is found that a library with JEND-3.2 Th232 data produces better results than ENDF/B-VI, R.7 and JEF-2.2 Th232 data. Results are slightly improved when HWTR spectra are used for weighting function to prepare the multi-group cross sections. This work is part of the International Atomic Energy Agency's Coordinated Research Project on 'Final Stage of WIMS-D Library Update Project'. (author)
Multigroup radiation hydrodynamics with flux-limited diffusion and adaptive mesh refinement
González, Matthias; Commerçon, Benoît; Masson, Jacques
2015-01-01
Radiative transfer plays a key role in the star formation process. Due to a high computational cost, radiation-hydrodynamics simulations performed up to now have mainly been carried out in the grey approximation. In recent years, multi-frequency radiation-hydrodynamics models have started to emerge, in an attempt to better account for the large variations of opacities as a function of frequency. We wish to develop an efficient multigroup algorithm for the adaptive mesh refinement code RAMSES which is suited to heavy proto-stellar collapse calculations. Due to prohibitive timestep constraints of an explicit radiative transfer method, we constructed a time-implicit solver based on a stabilised bi-conjugate gradient algorithm, and implemented it in RAMSES under the flux-limited diffusion approximation. We present a series of tests which demonstrate the high performance of our scheme in dealing with frequency-dependent radiation-hydrodynamic flows. We also present a preliminary simulation of a three-dimensional p...
Directory of Open Access Journals (Sweden)
Sanjukta Pookulangara
2010-12-01
Full Text Available This study examined channel-migration behavior using a decomposed Theory of Planned Behavior with crossover effects in brick-and-mortar stores and the Internet. An online survey was administered at four research sites (N = 547 and factor analysis and structural equation modeling, with multigroup analysis, were utilized for data analysis. Hedonic beliefs did not influence either of the channels, whereas, utilitarian beliefs were significant predictors in both brick-and-mortar stores and the Internet. Additionally, normative beliefs did not influence subjective norms in either of the channels, while self-efficacy influenced perceived behavioral control (PBC in both the channels. Attitude and subjective norms influenced channel-migration intentions for both channels; whereas, PBC was a significant predictor of channel-migration intentions on the Internet only. The crossover effects of normative beliefs and subjective norms on attitude was significant for the Internet. The crossover effects for subjective norms and PBC on attitude was significant for brick-and-mortar stores. Attitude toward channel migration from the Internet to brick-and-mortar stores yielded a negative influence.
Stability analysis of multi-group deterministic and stochastic epidemic models with vaccination rate
International Nuclear Information System (INIS)
We discuss in this paper a deterministic multi-group MSIR epidemic model with a vaccination rate, the basic reproduction number ℛ0, a key parameter in epidemiology, is a threshold which determines the persistence or extinction of the disease. By using Lyapunov function techniques, we show if ℛ0 is greater than 1 and the deterministic model obeys some conditions, then the disease will prevail, the infective persists and the endemic state is asymptotically stable in a feasible region. If ℛ0 is less than or equal to 1, then the infective disappear so the disease dies out. In addition, stochastic noises around the endemic equilibrium will be added to the deterministic MSIR model in order that the deterministic model is extended to a system of stochastic ordinary differential equations. In the stochastic version, we carry out a detailed analysis on the asymptotic behavior of the stochastic model. In addition, regarding the value of ℛ0, when the stochastic system obeys some conditions and ℛ0 is greater than 1, we deduce the stochastic system is stochastically asymptotically stable. Finally, the deterministic and stochastic model dynamics are illustrated through computer simulations. (general)
International Nuclear Information System (INIS)
SCALE 6 computes problem-dependent multigroup (MG) cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic pointwise (PW) transport calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The PW calculation is performed by the CENTRM code using a 1-D cylindrical Wigner-Seitz model with the white boundary condition instead of the real rectangular cell shape to represent a lattice unit cell. The pointwise fluxes computed by CENTRM are not exact because a 1-D model is used for the transport calculation, which introduces discrepancies in the MG self-shielded cross sections, resulting in some deviation in the eigenvalue. In order to solve this problem, the method of characteristics (MOC) has been applied to enable the CENTRM PW transport calculation for a 2-D square pin cell. The computation results show that the new BONAMI/CENTRM-MOC procedure produces very precise self-shielded cross sections compared to MCNP reaction rates.
A New Code for Proto-Neutron Star Evolution
Roberts, Luke F
2012-01-01
A new code for following the evolution and emissions of proto-neutron stars during the first minute of their lives is developed and tested. The code is one dimensional, fully implicit, and general relativistic. Multi-group, multi-flavor neutrino transport is incorporated that makes use of variable Eddington factors obtained from a formal solution of the static general relativistic Boltzmann equation with linearized scattering terms. The timescales of neutrino emission and spectral evolution obtained using the new code are broadly consistent with previous results. Unlike other recent calculations, however, the new code predicts that the neutrino-driven wind will be characterized, at least for part of its existence, by a neutron excess. This change, potentially consequential for nucleosynthesis in the wind, is due to an improved treatment of the charged-current interactions of electron flavored neutrinos and anti-neutrinos with nucleons. A comparison is also made between the results obtained using either variab...
Energy Technology Data Exchange (ETDEWEB)
Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)
1995-08-01
The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)
International Nuclear Information System (INIS)
Neutron shielding characteristics of the Waste Isolation Pilot Plant facility cask have been quantified for a variety of combinations of neutron sources and waste matrices which would potentially be handled in waste containers. The neutron attenuation and neutron environment of the waste container and the facility cask have been analyzed to ensure that the design requirement of neutron dose rate will be met under the combinations of the source and waste matrix conditions. The analyses considered the ranges of neutron source spectrum and waste matrices which combine to produce the minimum neutron shielding worth of the facility cask. One-dimensional analyses were performed with discrete ordinate transport theory methods using multigroup neutron cross section data. The results discussed in this report demonstrate the effect of source spectrum and waste container matrix on predicted neutron dose rates adjacent to the unshielded waste container and the surface of the facility cask. An evaluation of the uncertainties in predicted neutron dose rates is provided which results in an assessment of the maximum measured neutron dose rate external to the facility cask. A description of the analytical models developed, the analysis methodology, the neutron source spectra, and the detailed results are described in this report. 10 refs., 50 figs., 39 tabs
Neutron importance in source-driven systems
International Nuclear Information System (INIS)
A study of integral indicators of the neutron source importance in source-driven systems is carried out and their dependence on the phase-space characteristics of the neutron source is investigated in the first part of the paper. The second part is devoted to the analysis of the solution of the source-driven adjoint model, introducing different detectors as far as the spatial location and the energy is concerned. Spatial, angular and spectral effects are evidenced, solving the transport equation for a 2-dimensional x-y configuration in the multigroup SN approximation. Various definitions of the adjoint problem may be used in the interpretation of local flux measurements in source-driven subcritical systems and in the weighting procedures for the development of computational methods for transient analyses. The definition of the appropriate problem-dependent detector is still an open question and an object of discussion within the accelerator-driven system community. Some results showing the effects of different choices of the adjoint source on the effective mean neutron lifetime are illustrated. (authors)
International Nuclear Information System (INIS)
1 - Description of program or function: MATJEF22.BOLIB /1/ is a multigroup coupled (199 neutron groups + 42 photon groups) pseudo-problem-independent cross section library in MATXS format for nuclear fission applications. MATJEF22.BOLIB is based on the JEF-2.2 /2/ European nuclear data file and it was processed, through the NJOY /3/ system, in the VITAMIN-B6 /4/ group structure using the same parameters and calculation procedures. Original nuclear data file: JEF-2.2. Data processing systems: NJOY-94.66. Format: MATXS. Number of groups: 199 neutron groups and 42 photon groups. Thermal neutron groups: 36 groups below 5.043 eV with up-scatter cross sections. Neutron energy range: 1.0 E-5 eV - 19.64 MeV. Photon energy range: 1.0 keV - 30.0 MeV. Temperatures [K]: 300, 600, 1000, and 2100 (same values as in VITAMIN-B6). Background cross sections (Sigma-zeros) [barns]: 1, 10, 50, 100, 300, 1.0 E+3, 1.0 E+4, 1.0 E+5, 1.0 E+6, 1.0 E+10 (infinite dilution) (same values as in VITAMIN-B6). Legendre order: P7 for materials with Z ≤ 29 (copper); P5 for the remainders. Number of materials: 138 (list included). Neutron weighting spectrum: Corresponding to the IWT=4 input option in the GROUPR module of NJOY: from 1.0 E-5 eV to 0.125 eV → Maxwellian thermal spectrum (kT = 0.025 eV); from 0.125 eV to 820.8 keV → '1/E' slowing-down spectrum ; from 820.8 keV to 19.64 MeV → fission spectrum (fission temperature = 1.273 MeV). Photon weighting spectrum: Corresponding to the IWT=3 input option in the GAMINR module of NJOY: '1/E' spectrum with a 'roll-off' at lower energies to represent photoelectric absorption and a similar 'drop-off' at higher energies corresponding to the Q-value for neutron capture. The prompt neutron fission spectra for U-235, U-238 and Pu-239 are included in tabulated form in the package together with the neutron and photon group energy boundaries, the neutron and photon group lethargy boundaries and the neutron and photon group lethargy widths. The neutron
Directory of Open Access Journals (Sweden)
A. R. Reddy
1982-07-01
Full Text Available The field of neutron radiography with special reference to isotopic neutron radiography has been reviewed. Different components viz., sources, collimators, imaging systems are described. Various designs of neutron radiography facilities, their relative merits and demerits , the appropriateness of each design depending on the object to be radiographed, and economics of each technique are also dealt. The applications of neutron radiography are also briefly presented.
International Nuclear Information System (INIS)
Highlights: • Calculation of effective delayed neutron fraction in circulating-fuel reactors. • Extension of the Monte Carlo SERPENT-2 code for delayed neutron precursor tracking. • Forward and adjoint multi-group diffusion eigenvalue problems in OpenFOAM. • Analytical approach for βeff calculation in simple geometries and flow conditions. • Good agreement among the three proposed approaches in the MSFR test-case. - Abstract: This paper deals with the calculation of the effective delayed neutron fraction (βeff) in circulating-fuel nuclear reactors. The Molten Salt Fast Reactor is adopted as test case for the comparison of the analytical, deterministic and Monte Carlo methods presented. The Monte Carlo code SERPENT-2 has been extended to allow for delayed neutron precursors drift, according to the fuel velocity field. The forward and adjoint eigenvalue multi-group diffusion problems are implemented and solved adopting the multi-physics tool-kit OpenFOAM, by taking into account the convective and turbulent diffusive terms in the precursors balance. These two approaches show good agreement in the whole range of the MSFR operating conditions. An analytical formula for the circulating-to-static conditions βeff correction factor is also derived under simple hypotheses, which explicitly takes into account the spatial dependence of the neutron importance. Its accuracy is assessed against Monte Carlo and deterministic results. The effects of in-core recirculation vortex and turbulent diffusion are finally analysed and discussed
Symmetry breaking in the opinion dynamics of a multi-group project organization
International Nuclear Information System (INIS)
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO
Symmetry breaking in the opinion dynamics of a multi-group project organization
Zhu, Zhen-Tao; Zhou, Jing; Li, Ping; Chen, Xing-Guang
2012-10-01
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces: (i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness. Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes, i.e., a deadlock regime, a convergence regime, and a bifurcation regime in opinion dynamics. The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to. In the case of a three-group project with a symmetric social network, both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord, instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p. 125 Fig. 5), project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations, which urges that apart from divergence in participants' interests, nonlinear interaction can also make conflict inevitable in the PO. The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO. It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO.
Symmetry breaking in the opinion dynamics of a multi-group project organization
Institute of Scientific and Technical Information of China (English)
Zhu Zhen-Tao; Zhou Jing; Li Ping; Chen Xing-Guang
2012-01-01
A bounded confidence model of opinion dynamics in multi-group projects is presented in which each group's opinion evolution is driven by two types of forces:(i) the group's cohesive force which tends to restore the opinion back towards the initial status because of its company culture; and (ii) nonlinear coupling forces with other groups which attempt to bring opinions closer due to collaboration willingness.Bifurcation analysis for the case of a two-group project shows a cusp catastrophe phenomenon and three distinctive evolutionary regimes,i.e.,a deadlock regime,a convergence regime,and a bifurcation regime in opinion dynamics.The critical value of initial discord between the two groups is derived to discriminate which regime the opinion evolution belongs to.In the case of a three-group project with a symmetric social network,both bifurcation analysis and simulation results demonstrate that if each pair has a high initial discord,instead of symmetrically converging to consensus with the increase of coupling scale as expected by Gabbay's result (Physica A 378 (2007) p.125 Fig.5),project organization (PO) may be split into two distinct clusters because of the symmetry breaking phenomenon caused by pitchfork bifurcations,which urges that apart from divergence in participants' interests,nonlinear interaction can also make conflict inevitable in the PO.The effects of two asymmetric level parameters are tested in order to explore the ways of inducing dominant opinion in the whole PO.It is found that the strong influence imposed by a leader group with firm faith on the flexible and open minded follower groups can promote the formation of a positive dominant opinion in the PO.
Radiation Transport for Explosive Outflows: A Multigroup Hybrid Monte Carlo Method
Wollaeger, Ryan T.; van Rossum, Daniel R.; Graziani, Carlo; Couch, Sean M.; Jordan, George C., IV; Lamb, Donald Q.; Moses, Gregory A.
2013-12-01
We explore Implicit Monte Carlo (IMC) and discrete diffusion Monte Carlo (DDMC) for radiation transport in high-velocity outflows with structured opacity. The IMC method is a stochastic computational technique for nonlinear radiation transport. IMC is partially implicit in time and may suffer in efficiency when tracking MC particles through optically thick materials. DDMC accelerates IMC in diffusive domains. Abdikamalov extended IMC and DDMC to multigroup, velocity-dependent transport with the intent of modeling neutrino dynamics in core-collapse supernovae. Densmore has also formulated a multifrequency extension to the originally gray DDMC method. We rigorously formulate IMC and DDMC over a high-velocity Lagrangian grid for possible application to photon transport in the post-explosion phase of Type Ia supernovae. This formulation includes an analysis that yields an additional factor in the standard IMC-to-DDMC spatial interface condition. To our knowledge the new boundary condition is distinct from others presented in prior DDMC literature. The method is suitable for a variety of opacity distributions and may be applied to semi-relativistic radiation transport in simple fluids and geometries. Additionally, we test the code, called SuperNu, using an analytic solution having static material, as well as with a manufactured solution for moving material with structured opacities. Finally, we demonstrate with a simple source and 10 group logarithmic wavelength grid that IMC-DDMC performs better than pure IMC in terms of accuracy and speed when there are large disparities between the magnitudes of opacities in adjacent groups. We also present and test our implementation of the new boundary condition.
Neutron Capture and Neutron Halos
A.Mengoni; Otsuka, T; Nakamura, T.(International Center for Elementary Particle Physics and Department of Physics, The University of Tokyo, Tokyo, Japan); Ishihara, M.
1996-01-01
The connection between the neutron halo observed in light neutron rich nuclei and the neutron radiative capture process is outlined. We show how nuclear structure information such as spectroscopic factors and external components of the radial wave function of loosely bound states can be derived from the neutron capture cross section. The link between the direct radiative capture and the Coulomb dissociation process is elucidated.
International Nuclear Information System (INIS)
The Analytic Coarse-Mesh Finite-Difference method is developed in detail for multi-group and multi-dimensional diffusion calculations, including the general and particular modal solutions in the complex space for any number of groups. For rectangular multidimensional geometries, the Chao's generalized relations with transverse integration provide a high-order approximation of the ACMFD method, where all energy groups are coupled by matrix-vector FD relations and the errors are limited to the ones incurred by the interpolation of the transverse interface currents, in a non-linear iterative scheme. The implementation of the method in a multigroup 3D rectangular geometry nodal solver called ANDES is discussed, pointing out the encapsulation achieved for integration of the solver as an optional module within larger code systems. The performance of the ANDES solver in 3D rectangular (X-Y-Z) geometry and multi-groups is verified by its application to several 2D-3D model and international benchmarks (NEA-OECD), with given diffusion cross section sets in few-groups (2 to 8). The extensive verification, always required for new methods and codes, shows a quite fast convergence of ANDES in both the eigenvalue and transverse leakage iteration loops and with the nodal coarse-mesh size, allowing to reach the conclusion that quite high accuracy is achieved with rather large nodes, one node or four nodes per PWR fuel assembly, as compared with reference solutions obtained with fine-mesh finite-difference diffusion calculations using mesh sizes 64 to 128 times smaller than the ANDES nodes. (authors)
Directory of Open Access Journals (Sweden)
Lena Nusser
2015-05-01
Full Text Available This article focuses on measurement invariance of the assessment of educationally relevant constructs via written questionnaires for students at special schools and at low track schools attending 5th grade. To examine optimal conditions of administration for students with special educational needs in the area of learning an experimental design was implemented. If accommodated questionnaires, different school enrollments as well as competence differences allow equivalent assessment of reading motivation and academic self-concepts will be investigated with multi-group comparison of confirmatory factor analysis. The results indicate that comparisons between groups of students at special schools and low track schools are meaningful for certain constructs.
An analytical approach to fast neutron spectra by the modified Wigner approximation
International Nuclear Information System (INIS)
For these several years there has been considerable interest in the application of continuous slowing down (CSD) theory to problems in Fast Reactor Analysis. In such applications it is very important how to redefine the moderating parameters and how to treat inelastic scatterings in a resolved region and in an unresolved region. Treating inelastic and elastic scattering separately Stacey expanded the total collision density in a two-term Taylor series and gave an accurate neutron spectrum for a representative fast reactor composition, while Dunn and Becker incorporated inelastic scatterings into their moderating parameters by using the multigroup inelastic scattering matrix. In this paper we extend the CSD theory to the space-dependent problem by assuming the factorized neutron flux so as to derive the modified diffusion equation. In order to treat analytically the neutron flux in a finite bulk medium it is desired that the overall moderating process is described by as few moderating parameters as possible which can be defined for any energy region and any composition of materials by the unified formalism. To satisfy this requirement we propose the modified Wigner approximation (MWA) which is the CSD theory of the Wigner-type and its moderating parameter xi(u)-circumflex is given iteratively by the simple definition. For rapid computations of our parameter xi(u) we use the separate-type synthetic kernels for elastic scattering and inelastic scatterings. For the space-dependent problem in a finite bulk medium an simple analytical formula is derived by solving the modified diffusion equation and is used to study the space-dependence of fast neutron fluxes and the leakage effects on fast neutron fluxes at various points. This analytical solution brings out the fine structure of the fast neutron spectrum in greater detail than comparable multigroup treatments and allows simple analyses of fast neutron time-of-flight spectra
International Nuclear Information System (INIS)
This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics
Energy Technology Data Exchange (ETDEWEB)
White, J.E.; Wright, R.Q.; Roussin, R.W.; Ingersoll, D.T.
1992-11-01
This report discusses specifications which have been developed for a new multigroup cross section library based on ENDF/B-VI data for light water reactor shielding and reactor pressure vessel dosimetry applications. The resulting broad-group library and an intermediate fine-group library are defined by the specifications provided in this report. Processing ENDF/B-VI into multigroup format for use in radiation transport codes will provide radiation shielding analysts with the most currently available nuclear data. it is expected that the general nature of the specifications will be useful in other applications such as reactor physics.
Analysis of neutron transmission and reflection experiments with JENDL-3
International Nuclear Information System (INIS)
The ANISN-JR and DIAC codes of one-dimensional discrete ordinates transport were used to analyze the results of two benchmark test experiments irradiated by D-T neutrons. One was to measure angular neutron flux spectra of transmission through important materials as fusion blankets, i.e., graphite, beryllium and lithium, of which thickness was 0.6 to 5 mean free path and the leaking angles were 12.2deg, 24.9deg, 41.8deg and 66.8deg for 14.8 MeV neutrons. The other was to measure neutron flux of reflection by shielding materials such as iron and aluminum. The measurement was performed at 120deg for 15 MeV neutrons generated by an accelerator. The multi-group cross section data of 110 groups up to 16 MeV were generated from JENDL-3 nuclear data by using RADHEAT-V4 code system. The results of both transmission and reflection calculations agreed well with the measurements as well as the calculated results by the Monte Carlo codes, i.e., MORSE-DD and MCNP, in the energy region up to 10 MeV. On the other side, there were some discrepancies between calculations and measurements above 10 MeV and leaking angle of 66.8deg. (author)
Neutron cross section covariances in the resolved resonance region.
Energy Technology Data Exchange (ETDEWEB)
Herman,M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.T.; Rochman, D.
2008-04-01
We present a detailed analysis of the impact of resonance parameter uncertainties on covariances for neutron capture and fission cross sections in the resolved resonance region. Our analysis uses the uncertainties available in the recently published Atlas of Neutron Resonances employing the Multi-Level Breit-Wigner formalism. We consider uncertainties on resonance energies along with those on neutron-, radiative-, and fission-widths and examine their impact on cross section uncertainties and correlations. We also study the effect of the resonance parameter correlations deduced from capture and fission kernels and illustrate our approach on several practical examples. We show that uncertainties of neutron-, radiative- and fission-widths are important, while the uncertainties of resonance energies can be effectively neglected. We conclude that the correlations between neutron and radiative (fission) widths should be taken into account. The multi-group cross section uncertainties can be properly generated from both the resonance parameter covariance format MF32 and the cross section covariance format MF33, though the use of MF32 is more straightforward and hence preferable.
Directory of Open Access Journals (Sweden)
Alex Willener
2008-01-01
Full Text Available Resumo: A comunidade de Baselstrasse/Bernstrasse em Lucerna, Suíça, é caracterizada por conter uma população multinacional formada por mais de 70 nações, todas mantendo seus específicos estilos de vida. O aumento da população economicamente pobre refletiu na reputação da comunidade de forma negativa. A dinâmica social dessa comunidade repercutiu, também, nos aspectos educacionais e sócio-culturais da região, levando professores e outros profissionais a enfrentarem grandes desafios. Uma equipe interdisciplinar formada por professores da Universidade de Ciências Sociais Aplicadas de Lucerna aceitou o desafio de trabalhar com a comunidade e desenvolveu um projeto denominado BaBel. O principal objetivo do Projeto BaBel era desenvolver um trabalho que trouxesse benefícios para os diferentes grupos que vivem na comunidade, para as instituições presentes no local, bem como criar espaços para os futuros stakeholders. Todos esses atores foram envolvidos no desenvolvimento do projeto. Diferentes metodologias foram utilizadas para fomentar a participação progressiva da população, tais como: ‘desenvolvimento de cenários’, grupos de intervenção, análise da comunidade realizada pelas crianças e pelos jovens. A implementação do projeto envolveu 16 diferentes campos de ação, os quais abarcaram temas como economia de energia e poluição sonora (projetos pilotos, projetos na área da infância (curricular e extracurricular, melhoria das áreas de proteção ambiental das margens do rio que corta a comunidade, melhoria dos espaços de lazer, saúde e prevenção, estabelecimento de pontos de encontro na área, melhoria nas estruturas comerciais e otimização do tráfego local.Abstract: The Baselstrasse/Bernstrasse neighbourhood is characterized by its multinational population hailing from 70 nations, all maintaining their specific lifestyles. The increase in the number of economically weak people to a level above average
An evaluation of multigroup flux predictions in the EBR-II core
International Nuclear Information System (INIS)
The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required
International Nuclear Information System (INIS)
A method of calculation is given which assists the analyses of chopper measurements of spectra from ZENITH and enables complex multigroup theoretical calculations of the spectra to be put into a form which may be compared with experiment. In addition the theory of the cut-off function has been extended to give analytical expressions which take into account the effects of sub-collimators, off centre slits and of a rotor made of a material partially transparent to neutrons. The theoretical cut-off function suggested shows good agreement with experiment. (author)
Cross-section evaluation utilizing integral reaction-rate measurements in fast neutron fields
International Nuclear Information System (INIS)
The role of integral reaction-rate data for cross-section evaluation is reviewed. The subset of integral data considered comprises integral reaction rates measured for dosimeter, fission-product, and actinide-type materials irradiated in reactor dosimetry fast neutron benchmark fields and in the EBR-II. Utilization of these integral data for integral testing, multigroup cross-section adjustment and pointwise cross section adjustment is treated in some detail. Examples are given that illustrate the importance of considering a priori uncertainty and correlation information for these analyses. 3 figures, 3 tables
International Nuclear Information System (INIS)
In this dissertation we use the Laplace transform to derive expressions for nonstandard albedo boundary conditions for one and two non-multiplying regions at the ends of one dimensional domains. In practice, the fuel regions of reactor cores are surrounded by reflector regions that reduce neutron leakage. In order to exclude the reflector regions from the calculations, we introduce a reflection coefficient or albedo. We use the present albedo boundary conditions to solve numerically slab-geometry monoenergetic and multigroup diffusion equations using the conventional finite difference method. Numerical results are generated for fixed source and eigenvalue diffusion problems in slab geometry(author)
A New Method for Predicting the Penetration and Slowing-Down of Neutrons in Reactor Shields
International Nuclear Information System (INIS)
A new approach is presented in the formulation of removal-diffusion theory. The 'removal cross-section' is redefined and the slowing-down between the multigroup diffusion equations is treated with a complete energy transfer matrix rather than in an age theory approximation. The method, based on the new approach contains an adjustable parameter. Examples of neutron spectra and thermal flux penetrations are given in a number of differing shield configurations and the results compare favorably with experiments and Moments Method calculations
International Nuclear Information System (INIS)
1 - Description: VITJEFF31.BOLIB /1/ /2/ /3/ is a multigroup coupled (199 neutron groups + 42 photon groups) pseudo-problem-independent cross section library in AMPX /4/ format for nuclear fission applications. VITJEFF31.BOLIB is based on the JEFF-3.1 /3/ /5/ European nuclear data file processed, through the NJOY-99.160 /6/ system with an ENEA-Bologna revised GROUPR module /7/ and the ENEA-Bologna 2007 Revision /8/ of the SCAMPI /9/ system, in the VITAMIN-B6 /10/ energy group structure using the same parameters and calculation procedures. Original nuclear data file: JEFF-3.1. Data processing systems: ENEA-Bologna Revised Versions of NJOY-99.160 and SCAMPI. Format: AMPX. Number of groups: 199 neutron groups and 42 photon groups; Thermal neutron groups: 36 groups below 5.043 eV with up-scattering cross sections; Neutron energy range: 1.0 E-5 eV - 19.64 MeV; Photon energy range: 1.0 keV - 30.0 MeV; Temperatures [K]: 300, 600, 1000, and 2100 (same values as in VITAMIN-B6); Background cross sections (Sigma-zeros) [barns]: 1, 10, 50, 100, 300, 1.0 E+3, 1.0 E+4, 1.0 E+5, 1.0 E+6, 1.0 E+10 (infinite dilution) (same values as in VITAMIN-B6); Legendre order: P7 for materials with Z ≤ 29 (copper); P5 for the remainders; Number of materials: 182. Neutron Weighting spectrum: Corresponding to the IWT=4 input option in the GROUPR module of NJOY: from 1.0 E-5 eV to 0.125 eV → Maxwellian thermal spectrum (kT = 0.025 eV); from 0.125 eV to 820.8 keV → '1/E' slowing-down spectrum; from 820.8 keV to 19.64 MeV → fission spectrum (fission temperature = 1.273 MeV). Photon Weighting spectrum: Corresponding to the IWT=3 input option in the GAMINR module of NJOY: '1/E' spectrum with a 'roll-off' at lower energies to represent photoelectric absorption and a similar 'drop-off' at higher energies corresponding to the Q-value for neutron capture. The total (prompt + delayed) neutron fission spectra for U-235, U-238 and Pu-239 are included in tabulated form in the package together with
International Nuclear Information System (INIS)
As neutron scattering experiments have grown more and more demanding with respect to resolution and quality, it became more and more necessary to include the neutron source itself in the design of an experimental setup. In this sense the generic representation of a neutron scattering arrangement includes the primary neutron source and the associated spectrum shifter (or moderator). In fact, the design of a modern neutron source will start from a set of users requirements and will proceed 'inwards' through a selection of the moderators (spectrum shifters) to the primary source best suited to meet these often conflicting needs. This paper aims at explaining the options source designers have to match the neutron source performance to the users' demands. (author)
Stephan, Andrew C.; Jardret; Vincent D.
2011-04-05
A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.
Consideration of neutron flux gradients for sophisticated evaluation of irradiation experiments
International Nuclear Information System (INIS)
A joint Russian/German irradiation experiment was performed at the pressurised water reactor WWER 2 of the Rheinsberg NPP (Germany). The experiment comprises about 800 Charpy V-notch, SENB and CT specimens made from 24 different heats of Russian type RPV base and weld metals. Comprehensive calculations of the neutron fluence were carried out. A multigroup Monte Carlo method allows the calculation of the neutron fluence of each specimen or of different points within a large specimen under consideration of the details of the geometric arrangement. As the calculations shown the neutron fluence considerably varies over the cross section of an irradiation rig. Therefore, influence of the flux gradients on testing of Charpy V-notch and CT-specimens is evaluated. Methods taking into account a fluence correction of the measured absorbed energies are presented and discussed. (author)
Spectral effects at core-reflector interface in fast neutron systems
International Nuclear Information System (INIS)
The existence of space and energy neutron flux distribution transients is a well known phenomena which affects the calculation of fast reactors where the core is surrounded by fertile blankets and/or reflectors. The present interest for fast neutron reactors as TRU burners of relatively low power and small core size (high leakage) underlines the need for an accurate treatment of neutron reflection. In fact, the treatment of reflector effects can introduce significant uncertainties on Keff and power distribution calculated values [1]. A recent work [2] pointed out that a detailed multigroup energy treatment to account for spectrum transient at interfaces dramatically improves the agreement with respect to a reference continuous energy Monte Carlo calculation. In the present work we report on a preliminary analysis of experiments relevant for our purpose (the MUSE experiments at MASURCA - Cadarache) and on a simplified calculation procedure which takes into account the fundamental spectrum effects, but which can be used in standard design calculations
International Nuclear Information System (INIS)
The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results
International Nuclear Information System (INIS)
We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)
Munayer, Salim J; Horenczyk, Gabriel
2014-10-01
Grounded in a contextual approach to acculturation of minorities, this study examines changes in acculturation orientations among Palestinian Christian Arab adolescents in Israel following the "lost decade of Arab-Jewish coexistence." Multi-group acculturation orientations among 237 respondents were assessed vis-à-vis two majorities--Muslim Arabs and Israeli Jews--and compared to 1998 data. Separation was the strongest endorsed orientation towards both majority groups. Comparisons with the 1998 data also show a weakening of the Integration attitude towards Israeli Jews, and also distancing from Muslim Arabs. For the examination of the "Westernisation" hypothesis, multi-dimensional scaling (MDS) analyses of perceptions of Self and group values clearly showed that, after 10 years, Palestinian Christian Arabs perceive Israeli Jewish culture as less close to Western culture, and that Self and the Christian Arab group have become much closer, suggesting an increasing identification of Palestinian Christian Arab adolescents with their ethnoreligious culture. We discuss the value of a multi-group, multi-method, and multi-wave approach to the examination of the role of the political context in acculturation processes. PMID:25178958
International Nuclear Information System (INIS)
According to Russian federal norms and the safety guide of the nuclear regulatory body of Russia, the maximum fast neutron fluence above 0.5 MeV at critical positions of the reactor pressure vessel (RPV) of VVER-type reactors is used for prediction of the RPV lifetime. For the computation of neutron fluences in the RPV near the reactor core midplane level, the three-dimensional (3-D) synthesis method based on two- and one-dimensional SN calculations may be acceptable but needs validation. The present validation analysis was carried out on the basis of neutron transport calculations for a VVER-1000 model by means of the well-known codes DORT (R, Θ- and R, Z geometry) and ANISN (R geometry) using the multigroup library BUGLE-96. The 3-D spatial neutron source distribution, including pin-to-pin power variations and the complex baffle construction, were modeled in detail
Neutron spectrum obtained with Monte Carlo and transport theory
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The development of the computer, resulting in increasing memory capacity and processing speed, has enabled the application of Monte Carlo method to estimate the fluxes in thousands of fine bin energy structure. Usually the MC calculation is made using continuous energy nuclear data and exact geometry. Self shielding and interference of nuclides resonances are properly considered. Therefore, the fluxes obtained by this method may be a good estimation of the neutron energy distribution (spectrum) for the problem. In an early work it was proposed to use these fluxes as weighting spectrum to generate multigroup cross section for fast reactor analysis using deterministic codes. This non-traditional use of MC calculation needs a validation to gain confidence in the results. The work presented here is the validation start step of this scheme. The spectra of the JOYO first core fuel assembly MK-I and the benchmark Godiva were calculated using the tally flux estimator of the MCNP code and compared with the reference. Also, the two problems were solved with the multigroup transport theory code XSDRN of the AMPX system using the 171 energy groups VITAMIN-C library. The spectra differences arising from the utilization of these codes, the influence of evaluated data file and the application to fast reactor calculation are discussed. (author)
International Nuclear Information System (INIS)
Additional calibrations of the University of California double-scatter neutron and additional analysis corrections lead to the slightly changed neutron fluxes reported here. The theoretical angular distributions of Merker (1975) are in general agreement with our experimental fluxes but do not give the peaks for vertical upward and downward moving neutrons. The theoretical neutron escape current J2/sub pi/ (Merker, 1972; Armstrong et al., 1973) is in agreement with the experimental values from 10 to 100 MeV. Our experimental fluxes agree with those of the Kanbach et al. (1974) in the overlap region from 70 to 100 MeV
International Nuclear Information System (INIS)
A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq 241 Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s-1 and 0,5 μSv s-1. A calibrated 50 nSv s-1 thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the 241 Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold 241 Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,α) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kVpp cm-1, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46± 0,09) 104 tracks cm-2 mSv-1 for thermal neutrons, (9±3) 102 tracks cm-2 mSV-1 for intermediate neutrons and (26±4) tracks cm-2 mSv-1 for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990's ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is sufficiently sensitive to thermal and intermediate neutrons but fast neutron monitoring ar radiological protection level
Riempp, Gerold; Kremer, Stefan
2001-01-01
Portale leben von Inhalten - also letztlich von Informationen. Um diese für den effizienten Gebrauch zu erschliessen, genügt es nicht, intern eine Suchmaschine einzusetzen. Glossare und Taxonomien sind erforderlich, damit die Nutzer finden, was sie wirklich suchen.
Balasubramanian, V; Simón, J; Balasubramanian, Vijay; Jejjala, Vishnu; Sim\\'on, Joan
2005-01-01
We show that heavy pure states of gravity can appear to be mixed states to almost all probes. Our arguments are made for $\\rm{AdS}_5$ Schwarzschild black holes using the field theory dual to string theory in such spacetimes. Our results follow from applying information theoretic notions to field theory operators capable of describing very heavy states in gravity. For certain supersymmetric states of the theory, our account is exact: the microstates are described in gravity by a spacetime ``foam'', the precise details of which are invisible to almost all probes.
Edson, Lee
1982-01-01
How children acquire language is a riddle for developmental linguists and the subject of debate among them. Some linguists argue that children acquire language through a universal process regardless of their native tongues. Evidence of the innateness of language capacity has also appeared in studies of deaf children. (Author/JN)
Barbier-Bouvet, Jean-François
2013-01-01
Submergée par les demandeurs des sa mise en service en février 1977, agrandie en 1980, la Médiathèque de langues de la B.P.I. accueille quotidiennement près de 500 personnes. Mais qui sont ces utilisateurs persévérants (parfois une heure d'attente pour accéder a une cabine de langues). D'où viennent-il ? Qu'attendent-ils de cet espace babélien ? Quelles langues - parmi les 80 qui leur sont proposées - viennent-ils étudier ? Quelles sont leurs motivations ? Comment travaillent-ils ? En collabo...
International Nuclear Information System (INIS)
This document summarises a neutron activation cross-section database in processed in two formats as generated by F.M. Mann within the project of the Fusion Evaluated Nuclear Data Library (FENDL): FENDL/PA in continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP; and FENDL/PA-175G, in ASCII 175 group multigroup format as used by the transmutation code REAC*2/3. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command. (author)
International Nuclear Information System (INIS)
Highlights: • New type of multi-level rebalancing approach for nodal transport. • Generally improved and more mesh-independent convergence behavior. • Importance for intended regime of 3D pin-by-pin core computations. - Abstract: A new multi-level surface rebalancing (MLSR) approach has been developed, aimed at enabling an improved non-linear acceleration of nodal flux iteration convergence in 3D steady-state and transient reactor simulation. This development is meant specifically for anticipating computational needs for solving envisaged multi-group diffusion-like SPN calculations with enhanced mesh resolution (i.e. 3D multi-box up to 3D pin-by-pin grid). For the latter grid refinement regime, the previously available multi-level coarse mesh rebalancing (MLCMR) strategy has been observed to become increasingly inefficient with increasing 3D mesh resolution. Furthermore, for very fine 3D grids that feature a very fine axial mesh as well, non-convergence phenomena have been observed to emerge. In the verifications pursued up to now, these problems have been resolved by the new approach. The novelty arises from taking the interface current balance equations defined over all Cartesian box edges, instead of the nodal volume-integrated process-rate balance equation, as an appropriate restriction basis for setting up multi-level acceleration of fine grid interface current iterations. The new restriction strategy calls for the use of a newly derived set of adjoint spectral equations that are needed for computing a limited set of spectral response vectors per node. This enables a straightforward determination of group-condensed interface current spectral coupling operators that are of crucial relevance in the new rebalancing setup. Another novelty in the approach is a new variational method for computing the neutronic eigenvalue. Within this context, the latter is treated as a control parameter for driving another, newly defined and numerically more fundamental
International Nuclear Information System (INIS)
The present work is a contribution to the neutronics calculational methods of fast neutron reactors. The first step is devoted to the analysis of the validity of the few-groups (of the order of 25) multigroup scheme, and of the transport-correction approximation for the treatment of the scattering anisotropy. This analysis includes both the reactor core, where the usual approximations are found to be satisfactory, and the reflector, where it turns out that the rapid variations of the neutron flux and of it's spectrum necessitate the improvement of the multigroup cross-sections' generation. Therefore, a zero-dimensional simple and accurate model for the average spectrum in the reflector is developed by the space-energy synthesis method. Finally using the Rayleigh-Ritz method, a model is developed in which the flux is spatially represented by an analytical function. This model is applied to the analysis of the sensitivity of reflector neutronics parameters to the variations of the cross sections
Micromegas neutron beam monitor neutronics.
Stephan, Andrew C; Miller, Laurence F
2005-01-01
The Micromegas is a type of ionising radiation detector that consists of a gas chamber sandwiched between two parallel plate electrodes, with the gas chamber divided by a Frisch grid into drift and amplification gaps. Investigators have applied it to a number of different applications, such as charged particle, X-ray and neutron detection. A Micromegas device has been tested as a neutron beam monitor at CERN and is expected to be used for that purpose at the Spallation Neutron Source (SNS) under construction in Oak Ridge, TN. For the Micromegas to function effectively as neutron beam monitor, it should cause minimal disruption to the neutron beam in question. Specifically, it should scatter as few neutrons as possible and avoid neutron absorption when it does not contribute to generating useful information concerning the neutron beam. Here, we present the results of Monte Carlo calculations of the effect of different types of wall materials and detector gases on neutron beams and suggest methods for minimising disruption to the beam. PMID:16381746
Neutron dosimetry; Dosimetria de neutrons
Energy Technology Data Exchange (ETDEWEB)
Fratin, Luciano
1993-12-31
A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is
International Nuclear Information System (INIS)
Highlights: • Fixed-source SN transport problems. • Energy multigroup model. • Anisotropic scattering. • Slab-geometry spectral nodal method. - Abstract: A generalization of the spectral Green’s function (SGF) method is developed for multigroup, fixed-source, slab-geometry discrete ordinates (SN) problems with anisotropic scattering. The offered SGF method with the one-node block inversion (NBI) iterative scheme converges numerical solutions that are completely free from spatial truncation errors for multigroup, slab-geometry SN problems with scattering anisotropy of order L, provided L < N. As a coarse-mesh numerical method, the SGF method generates numerical solutions that generally do not give detailed information on the problem solution profile, as the grid points can be located considerably away from each other. Therefore, we describe in this paper a technique for the spatial reconstruction of the coarse-mesh solution generated by the multigroup SGF method. Numerical results are given to illustrate the method’s accuracy
Sideridis, Georgios D.; Tsaousis, Ioannis; Al-harbi, Khaleel A.
2015-01-01
The purpose of the present study was to extend the model of measurement invariance by simultaneously estimating invariance across multiple populations in the dichotomous instrument case using multi-group confirmatory factor analytic and multiple indicator multiple causes (MIMIC) methodologies. Using the Arabic version of the General Aptitude Test…
International Nuclear Information System (INIS)
The neutron gun combines a new core ion source of the cold type based on X-ray ionization and new cold type of neutron source working with core ion generation. The neutrons are formed from the impact of core ions on the negatively charged anode. Based on a new conversion function, the function of the positive anode becomes analogous to the beta-unstable decomposition of a neutron. A core ion and neutron amplifier in the sense of amplifying the number is derived from the beta-unstable neutron decomposition, in order to raise the output of a gun in pulsed operation by using the number amplification in the intervals between the pulses. The method of construction is simple and cheap, the equipment has purely linear acceleration or operation with circular acceleration with linear pre-acceleration as an alternative. Purely linear operation should be sufficient for medical applications, e.g. for neutron photography to replay X-ray photography and particularly for neutron scalpels in the surgical treatment of tumours. (orig./HP)
International Nuclear Information System (INIS)
This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality
Leung, Ka-Ngo; Lou, Tak Pui; Reijonen, Jani
2008-03-11
A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium/tritium (or even just a tritium) plasma is generated in the chamber and D or D/T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.
Babel en España : A propósito de la recurrencia del relato bíblico: el caso de la General Estoria
Rodríguez Temperley, María Mercedes
2007-01-01
La Edad Media, período en el cual Europa se transforma en una sociedad de textos gracias a un proceso complejo en el que confluye la alternancia de dos tradiciones, oralidad y escritura, tiene en la Biblia el Libro por excelencia, texto que lee como escuela de aprendizaje y que utiliza para validar otros discursos. En el contexto del Antiguo Testamento, el relato de la construcción de la torre de Babel (Génesis 11, 1-9) reúne en su austera complejidad una serie de temas y motivos que tendr...
Amaia Arizaleta
2012-01-01
Se comentan aquí una serie” de discursos redactados entre 1200 y 1250, todos ellos relacionados con el relato de la torre de Babel. Dichos textos dan fe de la existencia de un pensamiento común relativo a las lenguas y los pueblos en clérigos contemporáneos y de semejante alta cultura libresca, castellanos de origen o bien afectos a la causa de Castilla: Diego García, Rodrigo Jiménez de Rada y el anónimo de cuyo cálamo surgió el (Libro de) Alexandre. Estos letrados, que participaron en perman...
International Nuclear Information System (INIS)
1 - Description or function: VITJEF22.BOLIB /1/ is a multigroup coupled (199 neutron groups + 42 photon groups) pseudo-problem-independent cross section library in AMPX /2/ format for nuclear fission applications. VITJEF22.BOLIB is based on the JEF-2.2 /3/ European nuclear data file and it was processed, through the NJOY /4/ and SCAMPI /5/ systems, in the VITAMIN-B6 /6/ group structure using the same parameters and calculation procedures. Original nuclear data file: JEF-2.2. Data processing systems: NJOY-94.66 and SCAMPI. Format: AMPX. Number of groups: 199 neutron groups and 42 photon groups. Thermal neutron groups: 36 groups below 5.043 eV with up-scattering cross sections. Neutron energy range: 1.0 E-5 eV - 19.64 MeV. Photon energy range: 1.0 keV - 30.0 MeV. Temperatures [K]: 300, 600, 1000, and 2100 (same values as in VITAMIN-B6). Background cross sections (SIGMA-Zeros) [barns]: 1, 10, 50, 100, 300, 1.0 E+3, 1.0 E+4, 1.0 E+5, 1.0 E+6, 1.0 E+10 (infinite dilution) (same values as in VITAMIN-B6). Legendre order: P7 for materials with Z ≤ 29 (copper); P5 for the remainders. Number of materials: 138. Materials included: one file per material (nat=natural): H-H2O, H-CH2, H2-D2O, H-3, He-3, He-4, Li-6, Li-7, Be-9, Be-TH, B-10, B-11, C-nat, C-GPH, N-14, N-15, O-16, O-17, F-19, Na-23, Mg-nat, Al-27, Si-nat, P-31, S-32, S-33, S-34, S-36, Cl-nat, K-nat, Ca-nat, Ti-nat, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-nat, Ga-nat, Y-89, Zr-nat, Nb-93, Mo-nat, Ag-107, Ag-109, Cd-nat, Cd-106, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Ba-138, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-nat, Bi-209, Th-230, Th-232, Pa-231, Pa-233, U-232, U-233, U-234, U-235, U-236
Energy Technology Data Exchange (ETDEWEB)
Shestakov, A I; Offner, S R
2007-03-02
We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory
Two-dimensional multigroup finite element calculation of fast reactor in diffusion approximation
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When a linear element of triangular shape is used the actual finite element calculation is relatively simple. Extensive programs for mesh generation were written for easy inputting the configuration of reactors. A number of other programs were written for plotting neutron flux fields in individual groups, the power distribution, spatial plotting of fields, etc. The operation of selected programs, data preparation and operating instructions are described and examples given of data and results. All programs are written in GIER ALGOL. The used method and the developed programs have demonstrated that they are a useful instrument for the calculation of criticality and the distribution of neutron flux and power of both fast and thermal reactors. (J.B.)
Simulation of Space Shuttle neutron measurements with FLUKA
International Nuclear Information System (INIS)
FLUKA is an integrated particle transport code that has enhanced multigroup low-energy neutron transport capability similar to the well-known MORSE transport code. Gammas are produced in groups but many important individual lines are specifically included, and subsequently transported by the main FLUKA routines which use a modified version of EGS4 for electromagnetic (EM) transport. Recoil protons are also transported by the primary FLUKA transport simulation. The neutron cross-section libraries employed within FLUKA were supplied by Giancarlo Panini (ENEA, Italy) based upon the most recent data from JEF-1, JEF-2.2, ENDF/B-VI, JENDL-3, etc. More than 60 different materials are included in the FLUKA databases with temperature ranges including down to cryogenic temperatures. This code has been used extensively to model the neutron environments near high-energy physics experiment shielding. A simulation of the Space Shuttle based upon a spherical aluminum equivalent shielding distribution has been performed with reasonable results. There are good prospects for extending this calculation to a more realistic 3-D geometrical representation of the Shuttle including an accurate representation of its composition, which is an essential ingredient for the improvement of the predictions. A proposed project to develop a combined analysis and simulation package based upon FLUKA and the analysis infrastructure provided by the ROOT software is under active consideration. The code to be developed for this project will be of direct application to the problem of simulating the neutron environment in space, including the albedo effects
A NEW CODE FOR PROTO-NEUTRON STAR EVOLUTION
Energy Technology Data Exchange (ETDEWEB)
Roberts, L. F., E-mail: lroberts@ucolick.org [Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States)
2012-08-20
A new code for following the evolution and emissions of proto-neutron stars during the first minute of their lives is developed and tested. The code is one dimensional, fully implicit, and general relativistic. Multi-group, multi-flavor neutrino transport is incorporated that makes use of variable Eddington factors obtained from a formal solution of the static general relativistic Boltzmann equation with linearized scattering terms. The timescales of neutrino emission and spectral evolution obtained using the new code are broadly consistent with previous results. Unlike other recent calculations, however, the new code predicts that the neutrino-driven wind will be characterized, at least for part of its existence, by a neutron excess. This change, potentially consequential for nucleosynthesis in the wind, is due to an improved treatment of the charged current interactions of electron-flavored neutrinos and anti-neutrinos with nucleons. A comparison is also made between the results obtained using either variable Eddington factors or simple equilibrium flux-limited diffusion. The latter approximation, which has been frequently used in previous studies of proto-neutron star cooling, accurately describes the total neutrino luminosities (to within 10%) for most of the evolution, until the proto-neutron star becomes optically thin.
A New Code for Proto-neutron Star Evolution
Roberts, L. F.
2012-08-01
A new code for following the evolution and emissions of proto-neutron stars during the first minute of their lives is developed and tested. The code is one dimensional, fully implicit, and general relativistic. Multi-group, multi-flavor neutrino transport is incorporated that makes use of variable Eddington factors obtained from a formal solution of the static general relativistic Boltzmann equation with linearized scattering terms. The timescales of neutrino emission and spectral evolution obtained using the new code are broadly consistent with previous results. Unlike other recent calculations, however, the new code predicts that the neutrino-driven wind will be characterized, at least for part of its existence, by a neutron excess. This change, potentially consequential for nucleosynthesis in the wind, is due to an improved treatment of the charged current interactions of electron-flavored neutrinos and anti-neutrinos with nucleons. A comparison is also made between the results obtained using either variable Eddington factors or simple equilibrium flux-limited diffusion. The latter approximation, which has been frequently used in previous studies of proto-neutron star cooling, accurately describes the total neutrino luminosities (to within 10%) for most of the evolution, until the proto-neutron star becomes optically thin.
International Nuclear Information System (INIS)
This introduction is addressed to an audience active in diverse forms of neutron source applications but not directly familiar with neutron radiography. Neutron radiography is, of course, similar to, and complementary to, radiography using x-rays. However, neutrons, being sensitive to the nuclear properties of materials, provide information fundamentally different from x-rays. For example, neutrons can penetrate many dense metals such as uranium, lead, bismuth or steel, and can reveal details of internal hydrogenous components: explosives, lubricants and gaskets. For nuclear fuel inspection neutron radiography offers the ability to penetrate dense uranium-238 and contrast the isotopes U-235 or Pu-239 and also offers the ability to discriminate against unwanted interference from gamma radiation. In addition to advantages in industrial applications, there are special situations in fields such as medical diagnostics, dentistry, agriculture and forensic science. Comprehensive accounts of applications in the field can be found in the proceedings of the world conferences on neutron radiography: USA (1981), FRANCE (1986). A third conference in this series is scheduled for May 1989 in Japan
Benchmarks on neutron leakage from iron and Beryllium slavs and spheres
International Nuclear Information System (INIS)
Five benchmarks, recommended by the IAEA for nuclear power engineering have been calculated for an assessment of the Iron and Beryllium neutron data from the recent FENDL-1 version. The FENDL/MG-1.0 multigroup data processed in the IAEA by NJOY code are in VITAMIN-J energy structure in MATXS format. These data have been converted to ANISN format by TRANSX code and collected to binary library by LIBFENDL code. The neutron transport calculations have been carried out by the codes ANISN, GRTUNCL and DORT. Two benchmarks corresponding to the 14 MeV neutron transmission through Iron sphere shell (Simakov S.P. at al, IPPE, Obninsk) and Iron slabs (Y. Oyama and H. Maekawa, FNS/JAERI) permit to test the FENDL-1 Iron data for fusion application. The benchmark on neutron leakage from 25 cm radius Iron sphere with 252Cf source allows to show the FENDL-1 Iron data applicability in LWRs tasks. The comparison of the calculated and measured results demonstrates discouraged inconsistency when material thickness exceeds 20 cm . Modelling of the 14 MeV neutrons' transmission through Beryllium slabs (H. Maekawa and Y. Oyama at FNS/JAERI), and through sphere shell (Simakov S.P. at al in IPPE, Obninsk) has been carried out to test the multiplication data for the Beryllium as a fusion blanket material . The calculated angular neutron leakage from the slabs and the scalar neutron leakage from the sphere are in relatively good consistency with the measured ones. (author)
International Nuclear Information System (INIS)
The digital processing of the neutron radiography images gives the possibility for data quantification. In this case an exact relation between the measured neutron attenuation and the real macroscopic attenuation coefficient for every point of the sample is required. The assumption that the attenuation of the neutron beam through the sample is exponential is valid only in an ideal case where a monochromatic beam, non scattering sample and non background contribution are assumed. In the real case these conditions are not fulfilled and in dependence on the sample material we have more or less deviation from the exponential attenuation law. Because of the high scattering cross-sections of hydrogen (σs=80.26 barn) for thermal neutrons, the problem with the scattered neutrons at quantitative radiography investigations of hydrogenous materials (as PE, Oil, H2O, etc) is not trivial. For these strong scattering materials the neutron beam attenuation is no longer exponential and a dependence of the macroscopic attenuation coefficient on the material thickness and on the distance between the sample and the detector appears. When quantitative radiography (2 D) or tomography investigations (3 D) are performed, some image correction procedures for a description of the scattering effect are required. This thesis presents a method that can be used to enhance the neutron radiography image for objects with high scattering materials like hydrogen, carbon and other light materials. This method uses the Monte Carlo code, MCNP5, to simulate the neutron radiography process and get the flux distribution for each pixel of the image and determine the scattered neutrons distribution that causes the image blur and then subtract it from the initial image to improve its quality.
Díez, C. J.; Cabellos, O.; Martínez, J. S.
2015-01-01
Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.
International Nuclear Information System (INIS)
A computational method for calculating multigroup self-shielded cross sections in heterogeneous media containing arbitrary mixtures of resonant isotopes is presented. The method accounts for resonance interference between immixed resonant nuclei as well as for spatial resonance interference between resonant isotopes in different geometrical locations. A general correction is used to generate an intermediary reaction-rate library for resonant isotopic mixtures from a single-isotope, standard preprocessed library. Reaction rates for the heterogeneous fine-structure equation are computed from the intermediary library by invoking an equivalence theorem either on a group basis or using Bell's factors defined on macrogroups. Results are presented for a homogeneous mixture of Uranium oxide as well as for a recycled-fuel PWR cell. A study of the radial dependence of self-shielding for a recycled mixture of Uranium-Plutonium oxide in a PWR cell and in a submoderated cell is also included
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A multigroup diffusion theory code, TRIHEX-3D, has been developed for hexagonal lattice core analyses. For 2-D problems one can use hexagonal or triangular centre-mesh finite difference (FD) schemes. The geometrical description of the problem is for hexagonal geometry only. Subdivision of each hexagon into uniform triangles is facilitated by a built-in auto-triangularisati on procedure. One can analyse any symmetric part of the core or the whole core as well. Reflective (30deg, 60deg, 90deg, 120deg and 180deg) and rotational (60deg, 120deg and 180deg) symmetry boundary conditions are allowed. For 3-D problems one can use a direct 3-D FDM or an axial flux synthesis method. TRIHEX-3D can be used for the core design problems of VVER type of hexagonal lattice cores. The code has been validated against a LMFBR SNR-300 benchmark problem. (author). 8 tabs., 9 figs., 9 refs., 5 appendixes
Tominaga, Nozomu; Blinnikov, Sergei I
2015-01-01
We develop a time-dependent multi-group multidimensional relativistic radiative transfer code, which is required to numerically investigate radiation from relativistic fluids involved in, e.g., gamma-ray bursts and active galactic nuclei. The code is based on the spherical harmonic discrete ordinate method (SHDOM) that evaluates a source function including anisotropic scattering in spherical harmonics and implicitly solves the static radiative transfer equation with a ray tracing in discrete ordinates. We implement treatments of time dependence, multi-frequency bins, Lorentz transformation, and elastic Thomson and inelastic Compton scattering to the publicly available SHDOM code. Our code adopts a mixed frame approach; the source function is evaluated in the comoving frame whereas the radiative transfer equation is solved in the laboratory frame. This implementation is validated with various test problems and comparisons with results of a relativistic Monte Carlo code. These validations confirm that the code ...
International Nuclear Information System (INIS)
Subgroup 7 of the NEA NSC WP-IEC simulated interest in processing data from various international and regional evaluated cross-section libraries into the 174n, 42g VITAMIN-J multigroup energy for the purpose of intercomparison. Cooperation and participation came from numerous installations around the world. Most processing was done with the NJOY system, but some independent contributions were provided. At the WP-IEC meeting in June 1993, many contributions to the effort were described and the exercise proved to be useful from several aspects. It was decided to expand the role of the temporary subgroup into a long term subgroup to look at both format and processing problems. A summary of the progress of Subgroup 7 is provided and the objective and scope of the new entity, Subgroup B, is reported
SNAP-3D: a three-dimensional neutron diffusion code
International Nuclear Information System (INIS)
A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)
Scattering anisotropy and neutron leakage in a reactor lattice
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Scattering anisotropy is often taken into account, in an isotropic formalism by a transport correction. This correction which, even in a homogeneous medium, is known to be false in multigroup theory, is always incorrect for the calculation of neutron leakages in a lattice. The method presented here allows to calculate the diffusion coefficients in a Wigner-Seitz cell, at the zeroth order in powers of the buckling, for a linearly anisotropic scattering law; it allows to test the degree of approximation of the transport correction in various types of lattices. This correction appears to be a good approximation as far as the radial diffusion coefficient is concerned, but it may strongly underestimate the axial coefficient in certain types of lattices. The method allows to study the problems of correlation between groups which appear in the calculation of diffusion coefficients
Preliminary neutronic design of TRIGA Mark II Reactor
International Nuclear Information System (INIS)
It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible
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In this paper a survey is given of recent developments in selected areas of neutron tomography, within the context of several applications Argonne is involved in, including high penetration of reactor-fuel bundles in thick containers (involving TREAT and NRAD facilities), dual-energy hydrogen imaging (performed at IPNS), dynamic coarse-resolution emission tomography of rector fuel under test (a proposed modification to the TREAT hodoscope), and an associated-particle system that uses neutron flight-time to electronically collimate transmitted neutrons and to tomographically image nuclides identified by reaction gamma-rays
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This report contains the text of 16 lectures given at the Summer School and the report on a panel discussion entitled ''the relative merits and complementarities of x-rays, synchrotron radiation, steady- and pulsed neutron sources''. figs., tabs., refs
Cardone, Fabio; Petrucci, Andrea
2008-01-01
We report the results of neutron measurements carried out during the application of ultrasounds to a solution containing only stable elements like Iron and Chlorine, without any other radioactive source of any kind. These measurements, carried out by CR39 detectors and a Boron TriFouride electronic detector, evidenced the emission of neutron pulses. These pulses stand well above the electronic noise and the background of the laboratory where the measurements were carried out.
Convergence of finite element calculations of shielding factors in multigroup reactor analysis
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Finite Element-Spherical Harmonics Transport calculations can lead to accurate values of group shielding factors in reactor analysis. A benchmark problem in resonance absorption in a lattice unit cell is considered to study the convergence of the results obtained by this approach. The treatment refers to a representative symmetry portion of the unit cell for which the boundary conditions can be handled precisely for any specified order of the angular expansion of the neutron flux. Complementary approaches using variational methods are available to bracket desired results of the analysis, such as shielding factors or effective resonance integrals, between upper and lower bounds. This convergence is discussed
International Nuclear Information System (INIS)
The neutron probe is a standard tool for measuring soil water content. This article provides an overview of the underlying theory, describes the methodology for its calibration and use, discusses example applications, and identifies the safety issues. Soil water makes land-based life possible by satisfying plant water requirements, serving as a medium for nutrient movement to plant roots and nutrient cycling, and controlling the fate and transport of contaminants in the soil environment. Therefore, a successful understanding of the dynamics of plant growth, nutrient cycling, and contaminant behavior in the soil requires knowledge of the soil water content as well as its spatial and temporal variability. After more than 50 years, neutron probes remain the most reliable tool available for field monitoring of soil water content. Neutron probes provide integrated measurements over relatively large volumes of soil and, with proper access, allow for repeated sampling of the subsurface at the same locations. The limitations of neutron probes include costly and time-consuming manual operation, lack of data automation, and costly regulatory requirements. As more non-radioactive systems for soil water monitoring are developed to provide automated profiling capabilities, neutron-probe usage will likely decrease. Until then, neutron probes will continue to be a standard for reliable measurements of field water contents in soils around the globe
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Over the past ten years research has been going on at the P.N. Lebedev Physics Institute on the non-stationary moderation of neutrons in heavy media, the development of a method of neutron spectrometry based on the slowing-down time and the use of this method in studying the energy dependence of the cross-sections of nuclear reactions produced by neutrons with energy up to 30 keV. The authors review this work and discuss the results achieved. After a brief discussion of the theory of the non-stationary moderation and thermalization of neutrons the authors set forth the results of experimental studies of neutron moderation in graphite, iron and lead, and of neutron thermalization in lead. Using a pulsed neutron source and resonance detectors the distribution of slowing-down times was measured up to a series of fixed values for final neutron energy. The results are compared with theory, which takes into account the thermal motion of the moderator atoms; in the case of lead this thermal motion leads to a measurable spread in the slowing-down times at energies below 10 eV. The relationship between the mean velocity of neutrons in lead and the slowing-down time is measured in the subcadmium energy range and a comparison made with multigroup theory. The procedure for determining the energy dependence of neutron reaction cross-sections by slowing-down time is described and the potentialities of this method of spectrometry discussed. There follows a brief discussion of the results obtained in two fields of spectrometric measurement. Firstly, precise measurement of the relative excitation functions of the following reactions: He3(n, p), Li6(n, α), B10(n, α) and N14(n, p) - the most interesting results being the discovery of a constant negative component of the reaction cross-section and indications of the existence of an excited He4 level. Secondly, measurement of the energy dependence of averaged radiative capture cross-sections. Measurements carried out on a large
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This report describes a major revision of the DIAMANT2 code, which evolved from the DIAMANT-code. DIAMANT2 solves the static multigroup neutron transport equation in planar geometry using the SN-method. Spatial discretization is accomplished by taking finite differences on a meshgrid composed of equilateral triangles. This new version of DIAMANT2 is written in FORTRAN 77 and designed to run on scalar and vector computers. This report contains a detailed documentation of the program and the input description as well as some experiences gained with the code on several vector computers. (orig.)
Integral Data Test of HENDL1.0/MG and VisualBUS with Neutronics Shielding Experiments (Ⅰ)
Institute of Scientific and Technical Information of China (English)
高纯静; 许德政; 李静惊; 吴宜灿; 邓铁如
2004-01-01
HENDL1.0/MG, a multi-group working library of the Hybrid Evaluated Nuclear Data Library, was home-developed by the FDS Team of ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences) on the basis of several national data libraries. To validate and qualify the process of producing HENDL1.0/MG, simulating calculations of a series of existent spherical shell benchmark experiments (Al, Mo, Co, Ti, Mn, W, Be and V) have been performed with HENDL1.0/MG and the multifunctional neutronics code system named VisualBUS home-developed also by FDS Team.
International Nuclear Information System (INIS)
A generalized least-squares algorithm which refiens a prior multi-group energy-differential neutron-reaction cross-section evaluation by addition of new experimental data is described. Complete covariance information for the prior evaluation and for the new experimentla information is required in this procedure. The result is a revised best-estimate multi-group cross-section evaluation with complete covariance information. The algorithm tests the consistency of the ew and apriori information, and it readily indicates whether the new data significantly improve the knowledge of the differential cross section. These new data need not be specific differential cross sections. Therefore, the experimenter is not limited to measurements which involve only conventional monoenergetic techniques. This opportunity suggests exploration of diverse new experimental methods, e.g., ones which can exploit the high yield and favorable neutron-energy ranges offered by certain unconventional neutron sources which have received little past attention. This method is demonstratedby the detailed analysis of several hypothetical numerical examples. The understanding of the method's potential and limitations which has emerged from the present investigation is discussed
FOREWORD: Neutron metrology Neutron metrology
Thomas, David J.; Nolte, Ralf; Gressier, Vincent
2011-12-01
The International Committee for Weights and Measures (CIPM) has consultative committees covering various areas of metrology. The Consultative Committee for Ionizing Radiation (CCRI) differs from the others in having three sections: Section (I) deals with radiation dosimetry, Section (II) with radionuclide metrology and Section (III) with neutron metrology. In 2003 a proposal was made to publish special issues of Metrologia covering the work of the three Sections. Section (II) was the first to complete their task, and their special issue was published in 2007, volume 44(4). This was followed in 2009 by the special issue on radiation dosimetry, volume 46(2). The present issue, volume 48(6), completes the trilogy and attempts to explain neutron metrology, the youngest of the three disciplines, the neutron only having been discovered in 1932, to a wider audience and to highlight the relevance and importance of this field. When originally approached with the idea of this special issue, Section (III) immediately saw the value of a publication specifically on neutron metrology. It is a topic area where papers tend to be scattered throughout the literature in journals covering, for example, nuclear instrumentation, radiation protection or radiation measurements in general. Review articles tend to be few. People new to the field often ask for an introduction to the various topics. There are some excellent older textbooks, but these are now becoming obsolete. More experienced workers in specific areas of neutron metrology can find it difficult to know the latest position in related areas. The papers in this issue attempt, without presenting a purely historical outline, to describe the field in a sufficiently logical way to provide the novice with a clear introduction, while being sufficiently up-to-date to provide the more experienced reader with the latest scientific developments in the different topic areas. Neutron radiation fields obviously occur throughout the nuclear
International Nuclear Information System (INIS)
A computer code APPLE-2 which plots the spatial distribution of energy spectra of multi-group neutron and/or gamma ray fluxes, and reaction rates has been developed. This code is an improved version of the previously developed APPLE code and has the following features: (1) It plots energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT and MORSE. (2) It calculates and plots the spatial distribution of neutron and gamma ray fluxes and various types of reaction rates such as nuclear heating rates, operational dose rates, displacement damage rates. (3) Input data specification is greatly simplified by the use of standard, response libraries and by close coupling with radiation transport calculation codes. (4) Plotting outputs are given in camera ready form. (author)
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A new proposed method for solving the space-energy dependent spherical harmonics equations represents a methodological contribution to neutron transport theory. The proposed method was applied for solving the problem of spec-energy transport of fast and resonance neutrons in multi-zone, cylindrical y symmetric infinite reactor cell and is related to previously developed procedure for treating the thermal energy region. The advantages of this method are as follows: a unique algorithm was obtained for detailed determination of spatial and energy distribution of neutrons (from thermal to fast) in the reactor cell; these detailed distributions enable more precise calculations of criticality conditions, obtaining adequate multigroup data and better interpretation of experimental data; computing time is rather short
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RCPL1 is a FORTRAN digital computer program designed and developed to prepare neutron and photon cross section libraries for the RCP01 Monte Carlo computer program for solving neutron and photon transport problems in three-dimensional geometry with detailed energy description. The neutron libraries prepared by RCPL1 contain detailed Doppler-broadened resonance cross sections from unresolved and either single-level or multilevel resonance parameters, for any number of nuclides, within an arbitrary energy structure, and the photon libraries contain tabulations of the interaction cross sections and gamma emission spectra. This report describes the various RCPL1 program options, calculational details, and input requirements. All data used for library construction are extracted from a multigroup cross section library system XAP, described in an appendix to the report, which contains Evaluated Nuclear Data File (ENDF) data. 5 figures, 6 tables
Benchmark Tests of the Multigroup Cross Section Libraries for Fast Reactors
International Nuclear Information System (INIS)
In Korea, a design study for a fast breeder reactor named KALIMER (Korea Advanced LIquid MEtal Reactor) has been carried out. The simulations of the KALIMER core have been performed with the JEF-2.2- based 80-group neutron library KAFAX-F22 or the ENDF/B-VI.6-based 150-group neutron library KAFAXE66. Recently, newly evaluated nuclear data files such as ENDF/B-VII (beta 0 and 1), JEFF-3.1, and JENDL-3.3 have been released. And thus there is a need to update the libraries for the KALIMER by using the new data files. In this study, the fast cross section sets with 150 groups were prepared based on ENDF/B-VII beta 0, JEFF-3.1, and JENDL-3.3. The validations of the libraries have been carried out for 14 Cross Section Evaluation Working Group (CSEWG) fast benchmark problems through the 1-D and 2-D DANTSYS calculations. The effective multiplication factors (keff's) and central spectral indices have been compared with the experimental values and the results by the MCNPX calculations
International Nuclear Information System (INIS)
This document summarizes a neutron activation cross-section database processed in two formats as generated by F.M. Mann within the project of the Fusion Evaluated Nuclear Data Library (FENDL): in continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP; and in 175 group multigroup format with VIT-E weighting spectrum, as used by the transmutation code REAC*2/3. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author). 2 refs, 1 tab
Using of discrete ordinate method in the spallation target neutronics and shielding calculations
International Nuclear Information System (INIS)
A discrete ordinate algorithm for coupled charged/neutral particle transport calculations in 2D pencil beam problems is developed. It is based on the use of the second order of accuracy adaptive WDD (AWDD) scheme for approximation both the continuous slowing down (CSD) and streaming terms of the charged particle transport equation in z geometry, and a suitable algorithm for treatment of the extended uncollided flux from an initially monodirectional beam of charged particles with given radial distribution. The developed algorithm is implemented in the 2D transport code KASKAD-S-1.5 and is applied to the high-energy coupled proton-pion-neutron-photon transport calculations. The multigroup cross-section library SADCO-2 for nucleon-meson cascade calculations coupled with standard neutron and gamma-ray cross-section libraries below 20 MeV is used. Some numerical examples are given.(author)
Neutron Scattering in Hydrogenous Moderators, Studied by Time Dependent Reaction Rate Method
International Nuclear Information System (INIS)
The moderation and absorption of a neutron burst in water, poisoned with the non-1/v absorbers cadmium and gadolinium, has been followed on the time scale by multigroup calculations, using scattering kernels for the proton gas and the Nelkin model. The time dependent reaction rate curves for each absorber display clear differences for the two models, and the separation between the curves does not depend much on the absorber concentration. An experimental method for the measurement of infinite medium reaction rate curves in a limited geometry has been investigated. This method makes the measurement of the time dependent reaction rate generally useful for thermalization studies in a small geometry of a liquid hydrogenous moderator, provided that the experiment is coupled to programs for the calculation of scattering kernels and time dependent neutron spectra. Good agreement has been found between the reaction rate curve, measured with cadmium in water, and a calculated curve, where the Haywood kernel has been used
Manuel, Oliver K
2011-01-01
Earth is connected gravitationally, magnetically and electrically to its heat source - a neutron star that is obscured from view by waste products in the photosphere. Neutron repulsion is like the hot filament in an incandescent light bulb. Excited neutrons are emitted from the solar core and decay into hydrogen that glows in the photosphere like a frosted light bulb. Neutron repulsion was recognized in nuclear rest mass data in 2000 as the overlooked source of energy, the keystone of an arch that locked together these puzzling space-age observations: 1.) Excess 136Xe accompanied primordial helium in the stellar debris that formed the solar system (Fig. 1); 2.) The Sun formed on the supernova core (Fig. 2); 3.) Waste products from the core pass through an iron-rich mantle, selectively carrying lighter elements and lighter isotopes of each element into the photosphere (Figs. 3-4); and 4.) Neutron repulsion powers the Sun and sustains life (Figs. 5-7). Together these findings offer a framework for understanding...
RADSAT Benchmarks for Prompt Gamma Neutron Activation Analysis Measurements
Energy Technology Data Exchange (ETDEWEB)
Burns, Kimberly A.; Gesh, Christopher J.
2011-07-01
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. High-resolution gamma-ray spectrometers are used in these applications to measure the spectrum of the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used simulation tool for this type of problem, but computational times can be prohibitively long. This work explores the use of multi-group deterministic methods for the simulation of coupled neutron-photon problems. The main purpose of this work is to benchmark several problems modeled with RADSAT and MCNP to experimental data. Additionally, the cross section libraries for RADSAT are updated to include ENDF/B-VII cross sections. Preliminary findings show promising results when compared to MCNP and experimental data, but also areas where additional inquiry and testing are needed. The potential benefits and shortcomings of the multi-group-based approach are discussed in terms of accuracy and computational efficiency.
International Nuclear Information System (INIS)
Neutron radiography (or radiology) is a diverse filed that uses neutrons of various energies, subthermal, thermal, epithermal or fast in either steady state or pulsed mode to examine objects for industrial, medical, or other purposes, both microscopic and macroscopic. The applications include engineering design, biological studies, nondestructive inspection and materials evaluation. In the past decade, over 100 different centers in some 30 countries have published reports of pioneering activities using reactors, accelerators and isotopic neutron sources. While film transparency and electronic video are most common imaging methods for static or in motion objects respectively, there are other important data gathering techniques, including track etch, digital gauging and computed tomography. A survey of the world-wide progress shows the field to be gaining steadily in its diversity, its sophistication and its importance. (author)
Vaz, Sharmila; Falkmer, Marita; Parsons, Richard; Passmore, Anne Elizabeth; Parkin, Timothy; Falkmer, Torbjörn
2014-01-01
The relationship between school belongingness and mental health functioning before and after the primary-secondary school transition has not been previously investigated in students with and without disabilities. This study used a prospective longitudinal design to test the bi-directional relationships between these constructs, by surveying 266 students with and without disabilities and their parents, 6-months before and after the transition to secondary school. Cross-lagged multi-group analy...
Ingrid Moons; Patrick De Pelsmacker
2015-01-01
An Extended Decomposed Theory of Planned Behaviour (DTPB) is developed that integrates emotions towards car driving and electric cars as well as car driving habits of the DTPB, and is empirically validated in a Belgian sample ( n = 1023). Multi-group comparisons explore how the determinants of usage intention are different between groups of consumers differing in environmentally-friendly behaviour, environmental concern, innovativeness and personal values. Besides attitudes, media, perceived ...
Energy Technology Data Exchange (ETDEWEB)
Kim, Jong Woon; Kim, Sang Ji; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of {sup 239}Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis.
International Nuclear Information System (INIS)
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of 239Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis
Energy Technology Data Exchange (ETDEWEB)
Heger, G. [Rheinisch-Westfaelische Technische Hochschule Aachen, Inst. fuer Kristallographie, Aachen (Germany)
1996-12-31
X-ray diffraction using conventional laboratory equipment and/or synchrotron installations is the most important method for structure analyses. The purpose of this paper is to discuss special cases, for which, in addition to this indispensable part, neutrons are required to solve structural problems. Even though the huge intensity of modern synchrotron sources allows in principle the study of magnetic X-ray scattering the investigation of magnetic structures is still one of the most important applications of neutron diffraction. (author) 15 figs., 1 tab., 10 refs.
International Nuclear Information System (INIS)
A highly accurate S4 eigenfunction-based nodal method has been developed to solve multi-group discrete ordinate neutral particle transport problems with a linearly anisotropic scattering in slab geometry. The new method solves the even-parity form of discrete ordinates transport equation with an arbitrary SN order angular quadrature using two sub-cell balance equations and the S4 eigenfunctions of within-group transport equation. The four eigenfunctions from S4 approximation have been chosen as basis functions for the spatial expansion of the angular flux in each mesh. The constant and cubic polynomial approximations are adopted for the scattering source terms from other energy groups and fission source. A nodal method using the conventional polynomial expansion and the sub-cell balances was also developed to be used for demonstrating the high accuracy of the new methods. Using the new methods, a multi-group eigenvalue problem has been solved as well as fixed source problems. The numerical test results of one-group problem show that the new method has third-order accuracy as mesh size is finely refined and it has much higher accuracies for large meshes than the diamond differencing method and the nodal method using sub-cell balances and polynomial expansion of angular flux. For multi-group problems including eigenvalue problem, it was demonstrated that the new method using the cubic polynomial approximation of the sources could produce very accurate solutions even with large mesh sizes. (author)
International Nuclear Information System (INIS)
Thermal reactor design calculations are being performed in India using the WIMS/D-4 multi group cross section library, obtained in late 60's, reflecting the status of the basic nuclear data and processing technology then available. Significant improvements in basic evaluated data files such as ENDF/B-IV to VI and JEF data files etc. have been made in the past four decades and the multigroup libraries have been updated world over using improved and comprehensive nuclear data processing code systems. A few of such updated multigroup cross sections in WIMS/D-4 format are available from KAERI and NEA data bank sources. This paper presents the analysis of a set of enriched UO2 and U-metal uniform critical lattice experiments. These include TRX(4), BAPL (3) and B and W (17) lattice, 64 enriched UO2 lattices complied in NEACRP-U-190 report, 56 enriched UO2 lattices and 61 U-metal lattices which were used for validating the WIMKAL-1988 library. Calculated reaction rate values from the participants of WIMS library update project (WLUP) are available for TRX, BAPL lattices. Integral data measured in the lattices of TRX, BAPL, B and W and NEACRP compilations are available in the open literature. Different calculational methods like J± and Pij, and resonance interpolation schemes were examined in the theoretical analysis. Possible shortcomings of the WIMS-D/4 multigroup cross section library currently being used are also identified. (author)
Daskalov, George M; Baker, R S; Rogers, D W O; Williamson, J F
2002-02-01
Our purpose in this work is to demonstrate that the efficiency of dose-rate computations in 125I brachytherapy, using multigroup discrete ordinates radiation transport simulations, can be significantly enhanced using broad energy group cross sections without a loss of accuracy. To this end, the DANTSYS multigroup discrete ordinates neutral particle transport code was used to estimate the absorbed dose-rate distributions around an 125I-model 6702 seed in two-dimensional (2-D) cylindrical R-Z geometry for four different problems spanning the geometries found in clinical practice. First, simulations with a high resolution 210 energy groups library were used to analyze the photon flux spectral distribution throughout this set of problems. These distributions were used to design an energy group structure consisting of three broad groups along with suitable weighting functions from which the three-group cross sections were derived. The accuracy of 2-D DANTSYS dose-rate calculations was benchmarked against parallel Monte Carlo simulations. Ray effects were remedied by using the DANTSYS internal first collision source algorithm. It is demonstrated that the 125I primary photon spectrum leads to inappropriate weighting functions. An accuracy of +/-5% is achieved in the four problem geometries considered using geometry-independent three-group libraries derived from either material-specific weighting functions or a single material-independent weighting function. Agreement between Monte Carlo and the three-group DANTSYS calculations, within three standard Monte Carlo deviations, is observed everywhere except for a limited region along the Z axis of rotational symmetry, where ray effects are difficult to mitigate. The three-group DANTSYS calculations are 10-13 times faster than ones with a 210-group cross section library for 125I dosimetry problems. Compared to 2-D EGS4 Monte Carlo calculations, the 3-group DANTSYS simulations are a 100-fold more efficient. Provided that these
International Nuclear Information System (INIS)
Our purpose in this work is to demonstrate that the efficiency of dose-rate computations in 125I brachytherapy, using multigroup discrete ordinates radiation transport simulations, can be significantly enhanced using broad energy group cross sections without a loss of accuracy. To this end, the DANTSYS multigroup discrete ordinates neutral particle transport code was used to estimate the absorbed dose-rate distributions around an 125I-model 6702 seed in two-dimensional (2-D) cylindrical R-Z geometry for four different problems spanning the geometries found in clinical practice. First, simulations with a high resolution 210 energy groups library were used to analyze the photon flux spectral distribution throughout this set of problems. These distributions were used to design an energy group structure consisting of three broad groups along with suitable weighting functions from which the three-group cross sections were derived. The accuracy of 2-D DANTSYS dose-rate calculations was benchmarked against parallel Monte Carlo simulations. Ray effects were remedied by using the DANTSYS internal first collision source algorithm. It is demonstrated that the 125I primary photon spectrum leads to inappropriate weighting functions. An accuracy of ±5% is achieved in the four problem geometries considered using geometry-independent three-group libraries derived from either material-specific weighting functions or a single material-independent weighting function. Agreement between Monte Carlo and the three-group DANTSYS calculations, within three standard Monte Carlo deviations, is observed everywhere except for a limited region along the Z axis of rotational symmetry, where ray effects are difficult to mitigate. The three-group DANTSYS calculations are 10-13 times faster than ones with a 210-group cross section library for 125I dosimetry problems. Compared to 2-D EGS4 Monte Carlo calculations, the 3-group DANTSYS simulations are a 100-fold more efficient. Provided that these
Multi-group SP3 approximation for simulation of a three-dimensional PWR rod ejection accident
International Nuclear Information System (INIS)
Highlights: • The multi-group SP3 method developed and implemented in PARCS for the MOX analysis. • The verifications were performed in 2D and 3D, 2G and MG, diffusion and transport, with and without feedback. • All results show consistency with the reference results obtained from the ANL PN transport code VARIANT for steady-state and transport calculations. • It was found that the SP3 angular approximation captures sufficient transport effects for both steady-state and transient, and provides essentially the same results as the VARIANT P5 method. • From the transient results of the full-core problem, it was noted that MG is more conservative than 2G, and P1 is more conservative than SP3. - Abstract: Previous researchers have shown that the simplified P3 (SP3) approximation is capable of providing sufficiently high accuracy for both static and transient simulations for reactor core analysis with considerably less computational expense than higher order transport methods such as the discrete ordinate or the full spherical harmonics methods. The objective of this paper is to provide a consistent comparison of two-group (2G) and multi-group (MG) diffusion and SP3 transport for rod ejection accident (REA) in a practical light water reactor (LWR) problem. The analysis is performed on two numerical benchmarks, a 3 × 3 assembly mini-core and a full pressurized water reactor (PWR) core. The calculations were performed using pin homogenized and assembly homogenized cross sections for a series of benchmarks of increasing difficulty, in two-dimensional (2D) and three-dimensional (3D), 2G and MG, diffusion and transport, as well as with and without feedback. All results show consistency with the reference results obtained from higher-order methods. It is demonstrated that the analyzed problems show small group-homogenization effects, but relatively significant transport effects which are satisfactorily addressed by the SP3 transport method. The sensitivity tests
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The neutron reflectometer is the most powerful and nondestructive tool to analyze the surface and buried interfaces in the layered films. Such films often have a close relation to the functional devices. Structural information in the vicinity of the interfaces is a key parameter in the field of the nanoscale science. (author)
International Nuclear Information System (INIS)
Principles of the method, complementarity with X or gamma rays, neutron energy, detection and applications on irradiated materials and for the industrial quality control are exposed. Examples of applications in pyrotechnics, plastics, fuels, lubricants, metallurgy are given. Techniques developed by the CEA for its own needs or for industry and description of facilities in the nuclear centers are reviewed
A nodal approach to the solution of the multi-group SP3 equations in trigonal geometry
International Nuclear Information System (INIS)
The neutronics model of the reactor dynamics code DYN3D developed for 3-D analyses of steady states and transients in light-water reactors (LWR) is being extended by a simplified P3 (SP3) neutron transport approach for hexagonal fuel assemblies with a view to new reactor types such as high-temperature reactors. (orig.)
Methods for absorbing neutrons
Guillen, Donna P.; Longhurst, Glen R.; Porter, Douglas L.; Parry, James R.
2012-07-24
A conduction cooled neutron absorber may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cmK. Apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber.
International Nuclear Information System (INIS)
The Myrrha project (1) at SCK-CEN, the Belgian nuclear research centre, intends to design and develop a prototype accelerator driven system. Such a system will enable, next to other application fields (Technological demonstration, integral experiments validation,...), the benchmarking of the codes applied to assess the performances of the ADS. In the present situation we coupled, at SCK.CEN, the high energy Monte Carlo code HETC to the DORT/TORT S-N Neutron transport codes to perform the neutronic calculations of the Myrrha project. The HETC code is used to compute the space and energy distribution of the primary spallation neutron source, also including all other particles involved. The high energy cascade is calculated down to 20 MeV neutrons. Whereas the neutrons below this energy limit are stored as primary particles (without any interaction in the spallation medium) in a multigroup energy structure and will be treated as a fixed neutron source in the S-N transport code. The neutron interaction cross-section library used in this step is based on the ENDF/B-IV nuclear data. It is a 27 energy group with 7 groups below the thermal cut-off and allowing the up-scattering and the anisotropic scattering up to P3. The neutron transport calculations of the sub-critical assembly are performed using the DORT code either in Keff or fixed source with multiplication modes. Quadrature sets of S8 and S16 were used during these calculations. This calculational scheme was validated on basis of Monte Carlo calculational results and experimental data. In this paper we present the global calculational scheme as we applied it to Myrrh a and the corresponding results. (Author) 14 refs
Determination of Neutron Spectra in Bulk Media by Time-Of-Flight
International Nuclear Information System (INIS)
The time-of-flight method has recently been applied to the measurement of fast neutron spectra in bulk media to study problems of computing neutron distributions in reactor and shielding geometries. The advent of the high intensity, short pulse electron linear accelerator makes possible the collection of data in reasonable time for high resolution deep penetration experiments. Several experiments have been completed on geometries designed to be calculable by specific computer codes. Of particular interest is the class of one-dimensional geometries consisting of a central neutron source surrounded by spherical shells of homogeneous media with the outer boundary either spherical or essentially at infinity. To evaluate the ability to calculate fast neutron spectra in high-A materials, leakage spectra from a few keV to a few MeV were measured for a 6-in diameter lead sphere with a central (γ-n) neutron source. Results are compared with multigroup transport theory using cross-sections obtained by normalizing statistical theory with cross-section data. Spectra were also obtained at the lead-graphite interface for the above sphere encased in a 5-in thick, spherical, graphite shell. As expected, comparison of theory with experiment indicates that slowing down by low-A media is much better understood than slowing down by high-A media. The deep penetration problem, e.g. neutron shielding calculations, imposes extreme demands on computational methods both on accuracy required for input cross-sections, and on computational detail required to describe large attenuations. To study this problem, spectra were measured from 0.5 to 14 MeV for a point source, essentially infinite medium of paraffin. Measurements were made at several angles and radii out to 45-cm penetration. Results are compared with multigroup transport theory. Measurements have also been performed on a two-dimensional experiment (cylindrical medium, point source on axis) on the spectrum of deep
International Nuclear Information System (INIS)
ONETRAN solves the one-dimensional multigroup transport equation in plane, cylindrical, spherical, and two-angle plane geometries. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective, periodic, white, albedo or inhomogeneous boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. The discrete ordinates approximation for the angular variable is used with the diamond (central) difference approximation for the angular extrapolation in curved geometries. A linear discontinuous finite element representation for the angular flux in each spatial mesh cell is used. Negative fluxes are eliminated by a local set-to-zero and correct algorithm. Standard inner (within-group) iteration cycles are accelerated by system rebalance, coarse-mesh rebalance, or Chebyshev acceleration. Outer iteration cycles are accelerated by coarse-mesh rebalance. Provision is made for creation of standard interface output files for S/sub N/ constants, inhomogeneous sources, angle-integrated fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and total or angular fluxes may be read. All binary operations are localized in subroutines REED and RITE. Flexible edit options, including restart capability are provided. ONETRAN is designed for use with CDC-7600 and IBM 360. (12 tables, 10 figures) (U.S.)
Mezzacappa, A; Bruenn, S W; Blondin, J M; Guidry, M W; Strayer, M R; Umar, A S
1996-01-01
We investigate neutrino-driven convection in core collapse supernovae and its ramifications for the explosion mechanism. We begin with an ``optimistic'' 15 solar mass precollapse model, which is representative of the class of stars with compact iron cores. This model is evolved through core collapse and bounce in one dimension using multigroup (neutrino-energy--dependent) flux-limited diffusion (MGFLD) neutrino transport and Lagrangian hydrodynamics, providing realistic initial conditions for the postbounce convection and evolution. Our two-dimensional simulation begins at 106 ms after bounce at a time when there is a well-developed gain region, and proceeds for 400 ms. We couple two-dimensional (PPM) hydrodynamics to one-dimensional MGFLD neutrino transport. At 225 ms after bounce we see large-scale convection behind the shock, characterized by high-entropy, mushroom-like, expanding upflows and dense, low-entropy, finger-like downflows. The upflows reach the shock and distort it from sphericity. The radial c...
Roberts, Luke F; Haas, Roland; O'Connor, Evan P; Diener, Peter; Schnetter, Erik
2016-01-01
We report on a set of long-term general-relativistic three-dimensional (3D) multi-group (energy-dependent) neutrino-radiation hydrodynamics simulations of core-collapse supernovae. We employ a full 3D two-moment scheme with the local M1 closure, three neutrino species, and 12 energy groups per species. With this, we follow the post-core-bounce evolution of the core of a nonrotating $27$-$M_\\odot$ progenitor in full unconstrained 3D and in octant symmetry for $\\gtrsim$$ 380\\,\\mathrm{ms}$. We find the development of an asymmetric runaway explosion in our unconstrained simulation. We test the resolution dependence of our results and, in agreement with previous work, find that low resolution artificially aids explosion and leads to an earlier runaway expansion of the shock. At low resolution, the octant and full 3D dynamics are qualitatively very similar, but at high resolution, only the full 3D simulation exhibits the onset of explosion.
Recent Advances in Neutron Physics
Feshbach, Herman; Sheldon, Eric
1977-01-01
Discusses new studies in neutron physics within the last decade, such as ultracold neutrons, neutron bottles, resonance behavior, subthreshold fission, doubly radiative capture, and neutron stars. (MLH)
《通天塔》中文化霸权的经济解读%An Economic Approach to Cultural Hegemony in Babel
Institute of Scientific and Technical Information of China (English)
熊欣
2014-01-01
From the film Babel directed by Alejandro González I árritu ,it is proved that people’s different economic status in the world results in the failure of the communication ,but not only due to the different languages .The power of the cultural psychology in a group ,which plays an important part in understanding each other ,is closely related to the respective social economic po-sitions and the possession of the wealth .The disequilibrium and the enormous differences of the development of the economic level is the fact of the world which only produces the inequality of the economic status and its development ,but not the quatily of cultures .As one kind of the su-perstructures ,the characteristics and its value orientation of culture are dependent on the econom-ic foundation .Dominance in economy absolutely leads to the prevalence of cultural hegemony .%伊纳里图执导的影片《通天塔》揭示的绝不仅仅是语言的隔膜造成的沟通的失败，还有其背后的根本原因---经济发展差异。人与人之间的交流沟通关乎其文化心理，而一个群体的文化心理必然与其所处的社会经济地位和对社会财富掌控的多少紧密相关。世界本就是一个经济发展水平极不平衡且具有广泛差异性的复合体，只有经济地位和经济发展水平的不平等，不存在文化的优劣高低。经济基础决定着作为上层建筑的文化的性质和价值取向，强势经济必然带来文化霸权的繁衍。
Neutron rich nuclei and neutron stars
Horowitz, C. J.
2013-01-01
The PREX experiment at Jefferson Laboratory measures the neutron radius of 208Pb with parity violating electron scattering in a way that is free from most strong interaction uncertainties. The 208Pb radius has important implications for neutron rich matter and the structure of neutron stars. We present first PREX results, describe future plans, and discuss a follow on measurement of the neutron radius of 48Ca. We review radio and X-ray observations of neutron star masses and radii. These cons...
The use of small MnSO4-baths for calibration of neutron sources
International Nuclear Information System (INIS)
The uncertainty of neutron source strength determination by means of the Mn55-activation in aqueous MnSO4-baths is principally determined by two kinds or errors: the uncertainties of the correction factors to be applied for all kind of fast, intermediate and thermal neutron losses by nuclear reactions and by leakage, and the experimental errors in determining the Mn55-concentration and finally the Mn56-saturation activation. With decreasing source strength the experimental error increases and determines often the uncertainty of the result. The situation can be improved by decreasing the bath size, which increases the specific activation. However, at the same time the neutron-loss correction and its uncertainty increases. Systematic experimental studies have been performed on the Mn55-activation in different sized spherical and a large size cylindrical bath, with varying Mn55-concentration, using AmBe-neutron sources with varying source strengths. The Mn56-activity is measured by a NaJ(TL) crystal dipped into the bath. The experimental results of bath activation were compared with computer data based on calculations with the one dimensional multigroup code ANISN. A quantitative error analysis was applied to determine the crucial uncertainties of the technique as a function of neutron source strength
International Nuclear Information System (INIS)
New assemblies and improved measuring techniques call for periodic review of the status of computation vs. experiment. It is appropriate to emphasize neutron-spectral characterizations because of the particularly elusive problems associated with absolute spectral-index measurement and the need for checks of computation beyond simple critical size. The ever-improving spectral-index measurements in conjunction with increasing precision, both of microscopic data for detector and assembly materials and of computational techniques, produce a gradual clarification of the characteristics of a family of fast-neutron critical assemblies. This family now includes unreflected and thick-uranium-reflected U233 in spherical geometry. Direct correlations among the experimental data will be presented to indicate the a priori possibilities for successful correlations with computation. Sensitivity of computed spectra and critical sizes to neutron-transport models (transport and linear approximations ) and arithmetic approximations (finite angular segmentations and multi-group representations) will be presented for several typical assemblies to help establish the necessary computational detail. Comparisons between experiment and prediction will include, in addition to spectral indices and critical sizes, neutron lifetimes and delayed-neutron fractions. (author)
International Nuclear Information System (INIS)
We present a numerical study about the application of two versions of a second-degree iterative method for the solution of the sparse linear systems arising in the discretization of the 3D multi-group time-dependent Neutron Diffusion Equation. In addition, we propose some modifications to them, as well as a study of well-known preconditioning techniques in order to improve their convergence and accuracy when they are applied to a sequence of solutions in time of a real nuclear core transient. This is important for studies of stability and security of nuclear reactors. (authors)
International Nuclear Information System (INIS)
It is the object of the present invention to provide a method of measuring neutron radiation which eliminates the use of powders as dosimeter target materials and reduces the requirement for repetitive weighing of dosimeters, for expensive radioisotopes as dosimeter target material, and for dosimeter housings (in many cases). The invention described is a method of measuring neutron radiation within a nuclear reactor consisting of placing one or more extruded sintered oxide wires comprising a dosimeter target oxide within the reactor and measuring the radioactivity induced in the wires by neutron radiation. These oxide wires consist of a dilution containing at least 0.1% by weight of the dosimeter target oxide in a diluent oxide. The diluent oxide is selected from a group consisting of Al2O3 and BeO. Almost any metal oxide may be used as a target oxide. The wires may be encapsulated within a co-extruded housing. These mixed oxide wires have been found to be sufficiently uniform for quantitative analyses. (JTA)
International Nuclear Information System (INIS)
The annual report on hand gives an overview of the research work carried out in the Laboratory for Neutron Scattering (LNS) of the ETH Zuerich in 1990. Using the method of neutron scattering, it is possible to examine in detail the static and dynamic properties of the condensed material. In accordance with the multidisciplined character of the method, the LNS has for years maintained a system of intensive co-operation with numerous institutes in the areas of biology, chemistry, solid-state physics, crystallography and materials research. In 1990 over 100 scientists from more than 40 research groups both at home and abroad took part in the experiments. It was again a pleasure to see the number of graduate students present, who were studying for a doctorate and who could be introduced into the neutron scattering during their stay at the LNS and thus were in the position to touch on central ways of looking at a problem in their dissertation using this modern experimental method of solid-state research. In addition to the numerous and interesting ways of formulating the questions to explain the structure, nowadays the scientific programme increasingly includes particularly topical studies in connection with high temperature-supraconductors and materials research
International Nuclear Information System (INIS)
The following topics are dealt with: Neutron scattering in contemporary research, neutron sources, symmetry of crystals, diffraction, nanostructures investigated by small-angle neutron scattering, the structure of macromolecules, spin dependent and magnetic scattering, structural analysis, neutron reflectometry, magnetic nanostructures, inelastic scattering, strongly correlated electrons, dynamics of macromolecules, applications of neutron scattering. (HSI)
International Nuclear Information System (INIS)
The following topics are dealt with: Neutron sources, symmetry of crystals, nanostructures investigated by small-angle neutron scattering, structure of macromolecules, spin dependent and magnetic scattering, structural analysis, neutron reflectometry, magnetic nanostructures, inelastic neutron scattering, strongly correlated electrons, polymer dynamics, applications of neutron scattering. (HSI)
Directory of Open Access Journals (Sweden)
Amaia Arizaleta
2012-06-01
Full Text Available Se comentan aquí una serie” de discursos redactados entre 1200 y 1250, todos ellos relacionados con el relato de la torre de Babel. Dichos textos dan fe de la existencia de un pensamiento común relativo a las lenguas y los pueblos en clérigos contemporáneos y de semejante alta cultura libresca, castellanos de origen o bien afectos a la causa de Castilla: Diego García, Rodrigo Jiménez de Rada y el anónimo de cuyo cálamo surgió el (Libro de Alexandre. Estos letrados, que participaron en permanencia de la lengua romance y la lengua latina, dieron prueba de su interés por la diversidad lingüística, e incluso propusieron algunas ideas innovadoras sobre la cuestión. Testigos y artífices de una cultura que ya no podía ser monolingüe, supieron escribir acerca del mito de la separación de las naciones.Il est ici question de quelques discours en rapport avec le récit de la tour de Babel qui furent composés entre 1200 et 1250. Leurs auteurs, Diego García, Rodrigo Jiménez de Rada et le poète anonyme auteur du (Libro de Alexandre, qui entretenaient des liens avec la cour et la chancellerie, semblent avoir partagé une pensée commune relative aux langues et aux peuples. Ces lettrés, d’origine castillane ou qui avaient épousé la cause castillane, s’intéressèrent à la diversité linguistique, et allèrent jusqu’à défendre certaines idées novatrices sur le fonctionnement d’une culture qui ne pouvait plus être monolingue.
SIXTUS-3, 3-D Nodal Neutron Diffusion Criticality in Hexagonal Geometry
International Nuclear Information System (INIS)
1 - Description of program or function: Three-dimensional, nodal, neutron diffusion theory criticality code in hexagonal geometry. 2 - Method of solution: Intra-nodal elementary solutions with an exact multigroup Eigenvalue spectrum are spanned on two irreducible symmetry bases for the C6 and C2 groups of rotations for lateral and axial components, respectively. They represent a general homogeneous solution which is augmented with the special heterogeneous one determined by the transversal leakage terms, and from it the multigroup nodal response matrix for the partial current symmetry components on the node interfaces is computed. The response of a node to the incoming current is propagated in the system in a node sweeping process and after a prescribed number of sweeps the multiplication factor and the response matrix is recalculated. The iterations are terminated when the criteria of convergence for the multiplication factor, fission source and flux are met. An acceleration algorithm based on a special version of the Lyusternyk-Wagner extrapolation scheme is employed. The program contains a number of EISPACK routines. 3 - Restrictions on the complexity of the problem: Number of materials should not exceed 255; however, this restriction can be easily removed
Spallation Neutron Source (SNS)
Federal Laboratory Consortium — The SNS at Oak Ridge National Laboratory is a next-generation spallation neutron source for neutron scattering that is currently the most powerful neutron source in...
International Nuclear Information System (INIS)
Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group (1-g) cross sections must be provided in advance. This paper focuses on generating accurate 1-g cross section values that are necessary for evaluation of nuclide densities as a function of burnup. The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires extensive computational efforts. The method presented here is based on the multi-group (MG) approach, in which pre-generated MG sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate 1-g cross sections requires their tabulation against the background cross-section (σ0) to account for the self-shielding effect. However, in previous studies, the model that was used to calculate σ0 was simplified by fixing Bell and Dancoff factors. This work demonstrates that 1-g values calculated under the previous simplified model may not agree with the tallied values. Therefore, the original background cross section model was extended by implicitly accounting for the Dancoff and bell factors. The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented into BGCore code system. The 1-g cross section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement (<0.05%) in the 1-g cross values was observed. The method dose not carry any additional computational burden and it is universally applicable to the analysis of thermal as well as fast reactor systems. (author)
International Nuclear Information System (INIS)
The following topics are dealt with: Neutron sources, neutron properties and elastic scattering, correlation functions measured by scattering experiments, symmetry of crystals, applications of neutron scattering, polarized-neutron scattering and polarization analysis, structural analysis, magnetic and lattice excitation studied by inelastic neutron scattering, macromolecules and self-assembly, dynamics of macromolecules, correlated electrons in complex transition-metal oxides, surfaces, interfaces, and thin films investigated by neutron reflectometry, nanomagnetism. (HSI)
Petrov, P. V.; Kolchevsky, N.N.
2013-01-01
Compound concave refractive lenses are used for focusing neutron beam. Investigations of spectral and focusing properties of a refractive neutron lens are presented. Resolution of the imaging system on the base of refractive neutron lenses depends on material properties and parameters of neutron source. Model of refractive neutron lens are proposed. Results of calculation diffraction resolution and focal depth of refractive neutron lens are discussed.
Neutron Capture Nucleosynthesis
Kiss, Miklos
2016-01-01
Heavy elements (beyond iron) are formed in neutron capture nucleosynthesis processes. We have proposed a simple unified model to investigate the neutron capture nucleosynthesis in arbitrary neutron density environment. We have also investigated what neutron density is required to reproduce the measured abundance of nuclei assuming equilibrium processes. We found both of these that the medium neutron density has a particularly important role at neutron capture nucleosynthesis. About these resu...
Energy Technology Data Exchange (ETDEWEB)
Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)
2010-07-01
The following topics are dealt with: Neutron sources, neutron properties and elastic scattering, correlation functions measured by scattering experiments, symmetry of crystals, applications of neutron scattering, polarized-neutron scattering and polarization analysis, structural analysis, magnetic and lattice excitation studied by inelastic neutron scattering, macromolecules and self-assembly, dynamics of macromolecules, correlated electrons in complex transition-metal oxides, surfaces, interfaces, and thin films investigated by neutron reflectometry, nanomagnetism. (HSI)
Neutron Spectra in Subcritical Graphite- and Beryllia-Moderated Assemblies
International Nuclear Information System (INIS)
A number of subcritical graphite- and beryllia-moderated assemblies fuelled with plutonium and highly enriched uranium plates have been constructed with ratios of moderator atoms to fissile atoms as low as 300. As part of a general study of these small under moderated systems in which fission takes place predominantly at energies above 0.1 eV , time-of-flight measurements have been carried out using a neutron chopper having a resolution of 4 μs/m. This adequately resolves the spectrum in the thermal region and in the region of the resonance in 239Pu at 0.296 eV. Calculations have been carried out using a multigroup transport code with scattering matrices for graphite and beryllia based on data obtained from the Chalk River scattering law experiments. A correct weighting over the experimental source area of the vector flux in the direction of the extracted beam produces a spectrum which can be compared directly with the observed spectrum. It has been found that in the region 0.4 to 1.0 eV the experimental spectra have fewer neutrons than are predicted and this is particularly noticeable in the BeO-moderated assemblies. The factors which could give rise to the discrepancy are discussed but it has not been possible to attribute it to any particular one. (author)
An adaptive finite element approach for neutron transport equation
International Nuclear Information System (INIS)
Highlights: → Using uniform grid solution gives high local residuals errors. → Element refinement in the region where the flux gradient is large improves accuracy of results. → It is not necessary to use high density element throughout problem domain. → The method provides great geometrical flexibility. → Implementation of different density of elements lowers computational cost. - Abstract: In this paper, we develop an adaptive element refinement strategy that progressively refines the elements in appropriate regions of domain to solve even-parity Boltzmann transport equation. A posteriori error approach has been used for checking the approximation solutions for various sizes of elements. The local balance of neutrons in elements is utilized as an error assessment. To implement the adaptive approach a new neutron transport code FEMPT, finite element modeling of particle transport, for arbitrary geometry has been developed. This code is based on even-parity spherical harmonics and finite element method. A variational formulation is implemented for the even-parity neutron transport equation for the general case of anisotropic scattering and sources. High order spherical harmonic functions expansion for angle and finite element method in space is used as trial function. This code can be used to solve the multi-group neutron transport equation in highly complex X-Y geometries with arbitrary boundary condition. Due to powerful element generator tools of FEMPT, the description of desired and complicated 2D geometry becomes quite convenient. The numerical results show that the locally adaptive element refinement approach enhances the accuracy of solution in comparison with uniform meshing approach.
International Nuclear Information System (INIS)
Highlights: • We compare the equal-probable cosine (EPC) method in NJOY and Gauss quadrature. • The EPC method has large discrepancy with analytic solutions. • Gauss quadrature does not suffer from the same problems. • We give quantitative comparisons between the methods for Legendre moments. - Abstract: As high-fidelity simulations become routine and computational modelers begin to ask questions about uncertainty in calculations, the understanding of uncertainties in nuclear data, including multigroup cross-sections to high scattering orders, is also becoming important. In this paper we look at how a widely-used data processing code, NJOY, processes thermal cross-section data. Using an alternative integration scheme, Gauss quadrature, for angular integration to generate thermal scattering cross sections from the thermal scattering laws for graphite and ZrHx, we observed discrepancies between these results and NJOY’s results using equal-probable cosines on high order Legendre moments. In order to find a reliable comparison between the methods in NJOY and alternatives, we derived novel analytical expressions for high order Legendre moments of the free gas scattering model. Such expressions are analytically tractable but complicated. Using these expressions we construct a semi-analytic benchmark for multigroup Legendre moments. By comparing the results among the benchmark, Gauss quadrature and NJOY, we found that Gauss quadrature can preserve comparable accuracy as NJOY for lower order Legendre moments for free gas scattering and outperforms NJOY in generating high order moments. Our findings indicate that for high-scattering moments, multigroup data could be an source of uncertainty in thermal reactor calculations
International Nuclear Information System (INIS)
A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (SN) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The SN discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the SN transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup SN eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)
Federal Laboratory Consortium — The Neutron Therapy Facility provides a moderate intensity, broad energy spectrum neutron beam that can be used for short term irradiations for radiobiology (cells)...
International Nuclear Information System (INIS)
Attenuation of neutrons and photons transmitted through grahite, iron, water and ordinary concrete assemblies were studied using gold foils for thermal neutron and an NE-213 organic scintillation detector with an (n-γ) discrimination technique for spectral measurements. Source neutrons and photons were produced by 52-MeV proton bombardment of a 21.4-mm-thick graphite target placed in front of the assembly. The distributions of the light output from the scintillator were unfolded by the revised FERDO code. These experimental results were used as benchmark data on neutron and photon penetration by neutrons energy above 15MeV. Multigroup Monte Carlo, one-dimensional ANISN and two-dimensional DOT-3.5 transport calculations were performed with the DLC-58/HELLO group cross sections to compare with the measurement and to evaluate the cross sections. The DOT code was also used for the estimation of room-scattered neutron and photon contribution to the measured spectra. The results of the ANISN calculation of neutrons and the three-dimensional Monte Carlo calculation agreed with the experimental values except for high energy neutrons transmitted through water and graphite. The agreement of both calculations was well within the accuracy of 7% in the measured attenuation coefficients. For photons, the ANISN calculation gave >20% overestimation of the attenuation coefficients in the case of deep penetration through the medium for which the photon mean-free-path is shorter than that of neutrons, such as in iron and concrete. The result of the DOT calculation of neutrons down to thermal energy agreed well with the gold foil measurement in the absolute value. (author)
International Nuclear Information System (INIS)
1 - Description of program or function: JENDL-3.3 based, 175 neutron-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group transport codes. Format: MATXS. Number of groups: 175 neutron, 42 gamma-ray. Nuclides: 337 nuclides contained in JENDL-3.3: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-Nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, S-32, S-33, S-34, S-36, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Sc-45, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-Nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Ge-70, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-79, Se-80, Se-82, Br-79, Br-81, Kr-78, Kr-80, Kr-82, Kr-83, Kr-84, Kr-85, Kr-86, Rb-85, Rb-87, Sr-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-91, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Tc-99, Ru-96, Ru-98, Ru-99, Ru-100, Ru-101, Ru-102, Ru-103, Ru-104, Ru-106, Rh-103, Rh-105, Pd-102, Pd-104, Pd-105, Pd-106, Pd-107, Pd-108, Pd-110, Ag-107, Ag-109, Ag-110m, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Sb-121, Sb-123, Sb-124, Sb-125, Te-120, Te-122, Te-123, Te-124, Te-125, Te-126, Te-127m, Te-128, Te-129m, Te-130, I-127, I-129, I-131, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-133, Xe-134, Xe-135, Xe-136, Cs-133, Cs-134, Cs-135, Cs-136, Cs-137, Ba-130, Ba-132, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Ba-140, La-138, La-139, Ce-140, Ce-141, Ce-142, Ce-144, Pr-141, Pr-143, Nd-142, Nd-143, Nd-144, Nd-145, Nd-146, Nd-147, Nd-148, Nd-150, Pm-147, Pm-148, Pm-148m, Pm-149, Sm-144, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Sm-153, Sm-154, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Eu
Subcriticality Determination on Reflected Graphite System by the Pulsed Neutron Method
International Nuclear Information System (INIS)
The usual concept of reactivity is based on the treatment of a persisting mode of neutron multiplication. The experimental methods for reactivity measurement in a highly subcritical system which are based on the measurement of the persisting mode have theoretical and experimental difficulties. The other reactivity measurement methods, which are based on that of the prompt mode (pulsed neutron experiment, Rossi-α method, etc.), have other difficulties, because the prompt neutron lifetime changes depending on subcriticality, and the relation with the method based on persisting mode is very complicated in a reflected system. The purpose of the present work is to establish another concept to avoid the above-mentioned problem, and to describe exactly the situation of neutron multiplication in a reflected subcritical system. The concept. is that the prompt neutron decay constant of fundamental mode a should be used directly as a measure of neu ron multiplication, because this quantity satisfies the following two sufficient, fundamental requirements: to be uniquely measurable free from space dependence and detector specification, and to be calculable from the fundamental Boltzmann equations. To prove that a does satisfy these requirements, a series of pulsed experiments was made at various subcritical states in a reflected graphite-moderated enriched fuel system (SHE). The experimental results confirmed that α00 has the same value at any point of core and reflector, by means of the separation of spatial harmonics, using bare and cadmium-covered BF3 counters. The state of SHE is changed by diminishing the core with or without a centre control rod and by adding distributed poison to the core or to the reflector. The multigroup subcritical kinetics of the reflected reactor have been extended, and a direct comparison between the experimental results from pulsed neutron experiments and the calculated values from multigroup treatment using P1 and S4 approximation with regard
Neutronics code VALE for two-dimensional triagonal (hexagonal) and three-dimensional geometries
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Fowler, T.B.
1981-08-01
This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes are used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180/sup 0/ rotational symmetry, and the rotational symmetry conditions for the 30/sup 0/, 60/sup 0/, and 120/sup 0/ triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source.
Monte Carlo neutron transport simulation of the Ghana Research Reactor-1
International Nuclear Information System (INIS)
Stochastic Monte Carlo neutron particle transport methods have been applied to successfully model in 3-D, the HEU-fueled Ghana Research Reactor-1 (GHARR-1), a commercial version of the Miniature Neutron Source Reactor (MNSR) using the MCNP version 4c3 particle transport code. The preliminary multigroup neutronic criticality calculations yielded a keff is contained in 1.00449 with a corresponding cold clean excess reactivity of 4.47mk (447pcm) compared with experimental values of keff is contained in 1.00402 and excess reactivity of 4.00mk (400pcm). The Monte Carlo simulations also show comparable results in the neutron fluxes in the HEU core and some regions of interest. The observed trends in the radial and axial flux distributions in the core, beryllium annular reflector and the water region in the top shim reflector tray were reproduced, indicating consistency of the results, accuracy of the model, precision of the MCNP transport code and the comparability of the Monte Carlo simulations. The results further illustrate the close agreement between stochastic transport theory and the experimental measurements conducted during off-site zero power cold tests. (author)
Neutron source for Neutron Capture Synovectomy
International Nuclear Information System (INIS)
Monte Carlo calculations were performed to obtain a thermal neutron field from a 239PuBe neutron source inside a cylindrical heterogeneous moderators for Neutron Capture Synovectomy. Studied moderators were light water and heavy water, graphite and heavy water, lucite and polyethylene and heavy water. The neutron spectrum of polyethylene and heavy water moderator was used to determine neutron spectra inside a knee model. In this model the elemental composition of synovium and synovial liquid was assumed like blood. Kerma factors for synovium and synovial liquid were calculated to compare with water Kerma factors, in this calculations the synovium was loaded with two different concentrations of Boron
International Nuclear Information System (INIS)
This review summarizes information on the following subjects: (1) physical processes of importance in neutron dosimetry; (2) biological effects of neutrons; (3) neutron sources; and (4) instruments and methods used in neutron dosimetry. Also, possible improvements in dosimetry instrumentation are outlined and discussed. (author)
2002-01-01
The Bonner Ball Neutron Detector measures neutron radiation. Neutrons are uncharged atomic particles that have the ability to penetrate living tissues, harming human beings in space. The Bonner Ball Neutron Detector is one of three radiation experiments during Expedition Two. The others are the Phantom Torso and Dosimetric Mapping.
Langlois, David
2001-01-01
Neutron stars are believed to contain (neutron and proton) superfluids. I will give a summary of a macroscopic description of the interior of neutron stars, in a formulation which is general relativistic. I will also present recent results on the oscillations of neutron stars, with superfluidity explicitly taken into account, which leads in particular to the existence of a new class of modes.
Neutrostriction in Neutron stars
Ignatovich, V. K.
2003-01-01
It is demonstrated that not only gravity, but also neutrostriction forces due to optical potential created by coherent elastic neutron-neutron scattering can hold a neutron star together. The latter forces can be stronger than gravitational ones. The effect of these forces on mass, radius and structure of the neutron star is estimated.
International Nuclear Information System (INIS)
The following topics are dealt with: Neutron sources, symmetry of crystals, diffraction, nanostructures investigated by small-angle neutron scattering, the structure of macromolecules, spin dependent and magnetic scattering, structural analysis, neutron reflectometry, magnetic nanostructures, inelastic scattering, strongly correlated electrons, dynamics of macromolecules, applications of neutron scattering. (HSI)
Energy Technology Data Exchange (ETDEWEB)
Brueckel, Thomas; Heger, Gernot; Richter, Dieter; Roth, Georg; Zorn, Reiner (eds.)
2010-07-01
The following topics are dealt with: Neutron sources, symmetry of crystals, diffraction, nanostructures investigated by small-angle neutron scattering, the structure of macromolecules, spin dependent and magnetic scattering, structural analysis, neutron reflectometry, magnetic nanostructures, inelastic scattering, strongly correlated electrons, dynamics of macromolecules, applications of neutron scattering. (HSI)
International Nuclear Information System (INIS)
The neutron standards are reviewed with emphasis on the evaluation for ENDFB-VI. Also discussed are the neutron spectrum of 252Cf spontaneous fission, activation cross sections for neutron flux measurement, and standards for neutron energies greater than 20 MeV. Recommendations are made for future work. 21 refs., 6 figs., 3 tabs
Higher order polynomial expansion nodal method for hexagonal core neutronics analysis
International Nuclear Information System (INIS)
A higher-order polynomial expansion nodal(PEN) method is newly formulated as a means to improve the accuracy of the conventional PEN method solutions to multi-group diffusion equations in hexagonal core geometry. The new method is applied to solving various hexagonal core neutronics benchmark problems. The computational accuracy of the higher order PEN method is then compared with that of the conventional PEN method, the analytic function expansion nodal (AFEN) method, and the ANC-H method. It is demonstrated that the higher order PEN method improves the accuracy of the conventional PEN method and that it compares very well with the other nodal methods like the AFEN and ANC-H methods in accuracy
CACTUS, a characteristics solution to the neutron transport equations in complicated geometries
International Nuclear Information System (INIS)
CACTUS has been written to solve the multigroup neutron transport equation in a general two-dimensional geometry. The method is based upon a characteristics formulation for the problem in which the transport equation is integrated explicitly along straight line tracks that are suitably distributed throughout the problem. Source distributions and scattering are assumed to be isotropic, but the only restriction on geometry is that the outer boundary should be rectangular. Within this rectangular boundary the user is free to build his problem geometry using any combination of intersecting straight lines and circular arcs. The theory of the method is described, followed by some details of a coding, a sensitivity study on the number of tracks required to integrate fluxes in a particular problem, a user's guide, and a few test cases. (author)
Measured and Predicted Variations in Fast Neutron Spectrum when Penetrating Laminated Fe-D2O
International Nuclear Information System (INIS)
Variations of the fast neutron spectrum in thin regions of alternating Fe and DO have been studied using threshold detectors (ln(n, n' ), S(n, p), Al(n, α)). The results have been compared to those calculated by two shielding codes (NRN and RASH D) of multigroup removal-diffusion type. The absolute fast spectrum calculated in our rather complicated configurations was found to agree with measurements within the same accuracy (a factor of two) as did the thermal flux. The calculated spectrum is slightly harder than the measured one, but the detailed variations (covering the range 1:5) in the form of the spectrum when penetrating Fe agree with observations to within 15-20 %. In and Al activities were found to be proportional to the integrated flux over 1 MeV throughout the whole configuration, while S showed the least proportionality
An approximate method to study the one-velocity neutron integral transport equation
International Nuclear Information System (INIS)
An approximate method to study the monokinetic linear transport equation is outlined, starting from its integral form, rather than the integro-differential one. The approximate solution may be deduced either analytically, in simple cases, or numerically by means of typical space discretization techniques, through a system of second-order differential equations, associated with proper boundary conditions. Both the system and the boundary conditions may be matched with the standard neutron diffusion multigroup ones, by means of a proper correspondence of the coefficients and of the unknowns. The slab and the radially-symmetric sphere are then analysed in detail. It is shown how, in the plane case, the present approximation is perfectly equivalent to the well-known discrete ordinate one. For curved geometries no such equivalence exists, and it is in these cases that the application of the method at hand looks promising, in order to avoid complications and numerical problems in practical applications. (author)
A portable, parallel, object-oriented Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
International Nuclear Information System (INIS)
MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms
Neutron Capture Nucleosynthesis
Kiss, Miklos
2016-01-01
Heavy elements (beyond iron) are formed in neutron capture nucleosynthesis processes. We have proposed a simple unified model to investigate the neutron capture nucleosynthesis in arbitrary neutron density environment. We have also investigated what neutron density is required to reproduce the measured abundance of nuclei assuming equilibrium processes. We found both of these that the medium neutron density has a particularly important role at neutron capture nucleosynthesis. About these results most of the nuclei can formed at medium neutron capture density environment e.g. in some kind of AGB stars. Besides these observations our model is capable to use educational purpose.