Science.gov

Sample records for diamant-2 multigroup neutron

  1. Solves the Multigroup Neutron Diffusion Equation

    Energy Science and Technology Software Center (ESTSC)

    1995-06-23

    GNOMER is a program which solves the multigroup neutron diffusion equation in 1D, 2D and 3D cartesian geometry. The program is designed to calculate the global core power distributions (with thermohydraulic feedbacks), as well as power distribution and homogenized cross sections over a fuel assembly.

  2. Multigroup neutron dose calculations for proton therapy

    SciTech Connect

    Kelsey Iv, Charles T; Prinja, Anil K

    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.

  3. Multigroup Complex Geometry Neutron Diffusion Code System.

    Energy Science and Technology Software Center (ESTSC)

    2002-12-18

    Version 01 SNAP-3D is based on SNAP2 and is a one- two- or three-dimensional multigroup diffusion code system. It is primarily intended for neutron diffusion calculations, but it can also carry out gamma-ray calculations if the diffusion approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. SNAP-3D can solve the multi-group neutron diffusion equations using finite difference methods in (x,y,z), (r,theta,z), (TRI,z), (HEX,z) or (spherical) coordinates.more » The one-dimensional slab and cylindrical geometries and the two-dimensional (x,y), (r,z), (r,theta), (HEX) and (TRI) 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. The problem classes are: 1) eigenvalue search for critical k-effective, 2) eigenvalue search for critical buckling, 3) eigenvalue search for critical time-constant, 4) fixed source problems in which the sources are functions of regions, 5) fixed source problems in which the sources are provided, on disc, for every mesh point and group.« less

  4. 3D Multigroup Sn Neutron Transport Code

    Energy Science and Technology Software Center (ESTSC)

    2001-02-14

    ATTILA is a 3D multigroup transport code with arbitrary order ansotropic scatter. The transport equation is solved in first order form using a tri-linear discontinuous spatial differencing on an arbitrary tetrahedral mesh. The overall solution technique is source iteration with DSA acceleration of the scattering source. Anisotropic boundary and internal sources may be entered in the form of spherical harmonics moments. Alpha and k eigenvalue problems are allowed, as well as fixed source problems. Forwardmore » and adjoint solutions are available. Reflective, vacumn, and source boundary conditions are available. ATTILA can perform charged particle transport calculations using slowing down (CSD) terms. ATTILA can also be used to peform infra-red steady-state calculations for radiative transfer purposes.« less

  5. 3D Multigroup Sn Neutron Transport Code

    SciTech Connect

    McGee, John; Wareing, Todd; Pautz, Shawn

    2001-02-14

    ATTILA is a 3D multigroup transport code with arbitrary order ansotropic scatter. The transport equation is solved in first order form using a tri-linear discontinuous spatial differencing on an arbitrary tetrahedral mesh. The overall solution technique is source iteration with DSA acceleration of the scattering source. Anisotropic boundary and internal sources may be entered in the form of spherical harmonics moments. Alpha and k eigenvalue problems are allowed, as well as fixed source problems. Forward and adjoint solutions are available. Reflective, vacumn, and source boundary conditions are available. ATTILA can perform charged particle transport calculations using slowing down (CSD) terms. ATTILA can also be used to peform infra-red steady-state calculations for radiative transfer purposes.

  6. A concurrent, multigroup, discrete ordinates model of neutron transport

    SciTech Connect

    Dorr, M.R.; Still, C.H.

    1993-10-22

    The authors present an algorithm for the concurrent solution of the linear system arising from a multigroup, discrete ordinates model of neutron transport. The target architectures consist of distributed memory computers ranging from workstation clusters to massively parallel computers. Based on an analysis of the memory requirement and floating point complexity of matrix-vector multiplication in the iterative solution of the linear system, the authors propose a data layout and communication strategy designed to achieve scalability with respect to all phase space variables. Numerical results are presented to demonstrate the performance of the algorithm on the nCUBE/2.

  7. Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry.

    Energy Science and Technology Software Center (ESTSC)

    1986-12-01

    Version 00 PALLAS-PL/SP solves multigroup time-independent one-dimensional neutron transport problems in plane or spherical geometry. The problems solved are subject to a variety of boundary conditions or a distributed source. General anisotropic scattering problems are treated for solving deep-penetration problems in which angle-dependent neutron spectra are calculated in detail.

  8. Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks.

    Energy Science and Technology Software Center (ESTSC)

    1994-02-07

    Version 01 GNOMER is a program which solves the multigroup neutron diffusion equation on coarse mesh in 1D, 2D, and 3D Cartesian geometry. The program is designed to calculate the global core power distributions (with thermohydraulic feedbacks) as well as power distributions and homogenized cross sections over a fuel assembly.

  9. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    SciTech Connect

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-01-01

    For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig.

  10. PHISICS multi-group transport neutronic capabilities for RELAP5

    SciTech Connect

    Epiney, A.; Rabiti, C.; Alfonsi, A.; Wang, Y.; Cogliati, J.; Strydom, G.

    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)

  11. A Multigroup Method for the Calculation of Neutron Fluence with a Source Term

    NASA Technical Reports Server (NTRS)

    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.

  12. Hybrid method of deterministic and probabilistic approaches for multigroup neutron transport problem

    SciTech Connect

    Lee, D.

    2012-07-01

    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)

  13. Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

    SciTech Connect

    Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

    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%.

  14. Converged accelerated finite difference scheme for the multigroup neutron diffusion equation

    SciTech Connect

    Terranova, N.; Mostacci, D.; Ganapol, B. D.

    2013-07-01

    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)

  15. Improved Convergence Rate of Multi-Group Scattering Moment Tallies for Monte Carlo Neutron Transport Codes

    NASA Astrophysics Data System (ADS)

    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

  16. Algorithm development and verification of UASCM for multi-dimension and multi-group neutron kinetics model

    SciTech Connect

    Si, S.

    2012-07-01

    The Universal Algorithm of Stiffness Confinement Method (UASCM) for neutron kinetics model of multi-dimensional and multi-group transport equations or diffusion equations has been developed. The numerical experiments based on transport theory code MGSNM and diffusion theory code MGNEM have demonstrated that the algorithm has sufficient accuracy and stability. (authors)

  17. Computer program VARI-QUIR 3 provides solution of steady-state, multigroup, two-dimensional neutron diffusion equations

    NASA Technical Reports Server (NTRS)

    Collier, G.

    1967-01-01

    Computer program VARI-QUIR 3 provides Gauss-Seidel type of solution with inner and outer iterations for steady-state, multigroup, two-dimensional neutron diffusion equations. The program has no restrictions on any of the input parameters such as the number of groups, regions, or materials.

  18. Release of the mtmg01ex NDI Neutron Multigroup Data Library

    SciTech Connect

    Gray, Mark Girard

    2013-02-04

    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.

  19. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    1996-12-19

    Version 03 The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes.

  20. VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system. Version 3

    SciTech Connect

    Shapiro, A.; Huria, H.C.; Cho, K.W.

    1991-12-01

    VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing to disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.

  1. VENTURE/PC manual: A multidimensional multigroup neutron diffusion code system

    SciTech Connect

    Shapiro, A.; Huria, H.C.; Cho, K.W. )

    1991-12-01

    VENTURE/PC is a recompilation of part of the Oak Ridge BOLD VENTURE code system, which will operate on an IBM PC or compatible computer. Neutron diffusion theory solutions are obtained for multidimensional, multigroup problems. This manual contains information associated with operating the code system. The purpose of the various modules used in the code system, and the input for these modules are discussed. The PC code structure is also given. Version 2 included several enhancements not given in the original version of the code. In particular, flux iterations can be done in core rather than by reading and writing to disk, for problems which allow sufficient memory for such in-core iterations. This speeds up the iteration process. Version 3 does not include any of the special processors used in the previous versions. These special processors utilized formatted input for various elements of the code system. All such input data is now entered through the Input Processor, which produces standard interface files for the various modules in the code system. In addition, a Standard Interface File Handbook is included in the documentation which is distributed with the code, to assist in developing the input for the Input Processor.

  2. MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1

    SciTech Connect

    Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis

    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 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.

  3. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    SciTech Connect

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    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.

  4. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    1995-06-01

    Version 04 The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup neutron and photon cross sections from ENDF/B evaluated nuclear data. This is the last NJOY-91 series. It uses the same module structure as the earlier versions and its graphics options depend on DISSPLA. This new release, designated NJOY91.119, includes bug fixes, improvements in several modules, and some new capabilities. Information on the changes is included inmore » the README file. A new test problem was added to test some ENDF-6 features, including Reich-Moore resonance reconstruction, energy-angle matrices in GROUPR, and energy-angle distributions in ACER. The 91.119 release is basically configured for UNIX.« less

  5. General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods.

    Energy Science and Technology Software Center (ESTSC)

    2008-02-29

    Version 00 (1) Problems to be solved: MVP/GMVP II can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, themore » following objects (BODIes) can be used. - BODIes with linear surfaces : half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism - BODIes with quadratic surface and linear surfaces : cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid - Arbitrary quadratic surface and torus The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. See Abstract for more detail.« less

  6. Procedure to Generate the MPACT Multigroup Library

    SciTech Connect

    Kim, Kang Seog

    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.

  7. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    SciTech Connect

    Yang, W. S.; Lee, C. H.

    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

  8. 137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table.

    Energy Science and Technology Software Center (ESTSC)

    1986-02-16

    Version 00 The basic function of MGCLIB is to generate effective neutron cross section sets in either 137 or 26 group structures for use in the discrete ordinates codes ANISN-JR or DOT 3.5 or in the Monte Carlo codes KENO-IV or MULTI-KENO.

  9. Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory.

    Energy Science and Technology Software Center (ESTSC)

    1985-10-10

    MARCOPOLO calculates the radial and axial diffusion coefficients in one-group and multi-group theory for a cylinderized cell (Wigner-Seitz theory) with several concentric zones according to the isotropic shock or linear anisotropic shock hypotheses.

  10. General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.

    Energy Science and Technology Software Center (ESTSC)

    1991-08-01

    Version: 00 The original MORSE code was a multipurpose neutron and gamma-ray transport Monte Carlo code. It was 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 could be obtained in either the forward or adjoint mode. Time dependence for both shielding and criticality problems is provided. General three-dimensional geometry could be used with an albedo option available atmore » any material surface. Isotropic or anisotropic scattering up to a P16 expansion of the angular distribution was allowed. MORSE-CG incorporated the Mathematical Applications, Inc. (MAGI) combinatorial geometry routines. MORSE-B modifies the Monte Carlo neutron and photon transport computer code MORSE-CG by adding routines which allow various flexible options.« less

  11. Multigroup Neutron/Gamma-Ray Direct Integration Transport Code System for Two-Dimensional Cylindrical Geometry.

    Energy Science and Technology Software Center (ESTSC)

    1980-10-15

    Version 00 PALLAS-2DCY-FX is a code for direct integration of the transport equation in two-dimensional (r,z) geometry. It solves the energy and angular-dependent Boltzmann transport equation with general anisotropic scattering in cylindrical geometry. Its principal applications are to neutron or gamma-ray transport problems in the forward mode. The code is particularly designed for and suited to the solution of deep penetration radiation transport problems with an external (fixed) source.

  12. A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format.

    Energy Science and Technology Software Center (ESTSC)

    1982-05-07

    XLACS-IIA calculates fine-group averaged neutron cross sections from ENDF data. Its primary purpose is to produce full range multigroup libraries for the XSDRN-PM program. It also serves this purpose in the AMPX system. Provisions are included for treating fast, resonance, and thermal ENDF/B data. Fine-group energy structures and expansion orders used to represent differential cross sections for XSDRN can be arbitrarily specified by the user. Cross sections can be averaged over an arbitrary user-supplied weightingmore » function or by any of several built-in weighting functions.« less

  13. General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability. We recommend C00474/ALLCP/02 MORSE-CGA.

    Energy Science and Technology Software Center (ESTSC)

    1991-05-01

    Version 00 MORSE-CGA was developed to add the capability of modelling rectangular lattices for nuclear reactor cores or for multipartitioned 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 obtainedmore » 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.« less

  14. Coupled multi-group neutron photon transport for the simulation of high-resolution gamma-ray spectroscopy applications

    SciTech Connect

    Burns, Kimberly A.

    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.

  15. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    1998-05-13

    Version 00 The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B-VI in the US, JEF-2.2 in Europe, JENDL-3.2 in Japan, BROND-2.2 in Russia), NJOY gives its users access to a wide variety of the most up-to-date nuclear data. NJOY provides comprehensivemore » capabilities for processing evaluated data, and it can serve applications ranging from continuous-energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photoatomic interactions, self shielding, probability tables, photon production, and high-energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color). More information on NJOY is available from the developer's home page at http://t2.lanl.gov. Follow the Tourbus section of the Tour area to find notes from the ICTP lectures held at Trieste in March 1998 on the ENDF format and on the NJOY code.« less

  16. Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.

    Energy Science and Technology Software Center (ESTSC)

    2000-03-28

    Version 00 The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B‑VI in the US, JEF‑2.2 in Europe, JENDL‑3.2 in Japan, BROND‑2.2 in Russia), NJOY gives its users access to a wide variety of the most up‑to‑date nuclear data. NJOY provides comprehensivemore » capabilities for processing evaluated data, and it can serve applications ranging from continuous‑energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photoatomic interactions, self shielding, probability tables, photon production, and high‑energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color).« less

  17. Multigroup calculations using VIM: A user's guide to ISOVIM

    SciTech Connect

    Blomquist, R.N.

    1992-09-01

    Monte Carlo calculations have long been used to benchmark more a mate approximate solution methods for reactor physics problems. The power of VIM (ref 1) lies partly in the detailed geometrical representations incorporating the (generally) curved surfaces of combinatorial geometry, and partly in the fine energy detail of pointwise cross sections which are independent of the neutron spectrum. When differences arise between Monte Carlo and deterministic calculations, the question arises, is the error in the multigroup cross sections, in the treatment of transport effects, or in the mesh-based treatment of space in the deterministic calculation The answers may not be obvious, but may be identified by combining the exact geometry capability of VIM with the multigroup formalism. We can now run VIM in a multigroup mode by producing special VIM Material files which contain point-wise data describing multigroup data with histograms. This forces VIM to solve the multigroup problem with only three small code modifications. P[sub N] scattering is simulated with the usual tabulated angular distributions with 20 equally-sized scattering angle cosine meshes. This document describes the VIM multigroup capability, the procedures for generating multigroup cross sections for VIM, and their use. The multigroup cross section generating code, ISOVIM, is described, and benchmark testing is documented.

  18. A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files.

    Energy Science and Technology Software Center (ESTSC)

    1990-11-20

    Version 00 REX2-87 is a computer code developed for the calculation of self-shielded multigroup average cross sections, and self-shielding factors for total, elastic, fission and capture processes from an ENDF/B formatted nuclear data file in which the tabulated cross sections follow linear interpolation throughout.

  19. MCNP: Multigroup/adjoint capabilities

    SciTech Connect

    Wagner, J.C.; Redmond, E.L. II; Palmtag, S.P.; Hendricks, J.S.

    1994-04-01

    This report discusses various aspects related to the use and validity of the general purpose Monte Carlo code MCNP for multigroup/adjoint calculations. The increased desire to perform comparisons between Monte Carlo and deterministic codes, along with the ever-present desire to increase the efficiency of large MCNP calculations has produced a greater user demand for the multigroup/adjoint capabilities. To more fully utilize these capabilities, we review the applications of the Monte Carlo multigroup/adjoint method, describe how to generate multigroup cross sections for MCNP with the auxiliary CRSRD code, describe how to use the multigroup/adjoint capability in MCNP, and provide examples and results indicating the effectiveness and validity of the MCNP multigroup/adjoint treatment. This information should assist users in taking advantage of the MCNP multigroup/adjoint capabilities.

  20. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    SciTech Connect

    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.

  1. The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

    SciTech Connect

    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 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.

  2. A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV.

    Energy Science and Technology Software Center (ESTSC)

    1987-05-20

    Version 00 The energy range of the library, from thermal to 800 MeV is relevant to the solution of shielding, nuclear heating, and other radiation protection problems connected with the accelerator neutron sources e.g. spallation target. The data contains 10 elements of shielding and biological importance. They can be easily implemented to the neutron transport codes like ANISN and DOT by using the activity option.

  3. One-Dimensional, Multigroup Cross Section and Design Sensitivity and Uncertainty Analysis Code System - Generalized Perturbation Theory.

    Energy Science and Technology Software Center (ESTSC)

    1981-02-02

    Version: 00 SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections (of standard multigroup cross-section sets) and for secondary energy distributions (SED's) of multigroup scattering matrices.

  4. Multigroup Constants fFle Based on ENDF/B IV.

    Energy Science and Technology Software Center (ESTSC)

    1980-08-13

    Version 00 The multigroup constants file (JIMCOF) has been prepared based on ENDF/V-IV. JIMCOF is composed of JIMCOF/F, the file for the fast neutron energy region; JIMCOF/T, the file for the thermal neutron energy region; and PROC, the processing program. The JIMCOF file contains multigroup constants of about 100 nuclides (68 groups in the fast neutron energy region and 50 groups in the thermal neutron energy region). Selection of individual data, as well as preparationmore » of library data from the DELIGHT code series, would be possible with the use of the PROC code.« less

  5. Multigroup calculations using VIM: A user`s guide to ISOVIM

    SciTech Connect

    Blomquist, R.N.

    1992-09-01

    Monte Carlo calculations have long been used to benchmark more a mate approximate solution methods for reactor physics problems. The power of VIM (ref 1) lies partly in the detailed geometrical representations incorporating the (generally) curved surfaces of combinatorial geometry, and partly in the fine energy detail of pointwise cross sections which are independent of the neutron spectrum. When differences arise between Monte Carlo and deterministic calculations, the question arises, is the error in the multigroup cross sections, in the treatment of transport effects, or in the mesh-based treatment of space in the deterministic calculation? The answers may not be obvious, but may be identified by combining the exact geometry capability of VIM with the multigroup formalism. We can now run VIM in a multigroup mode by producing special VIM Material files which contain point-wise data describing multigroup data with histograms. This forces VIM to solve the multigroup problem with only three small code modifications. P{sub N} scattering is simulated with the usual tabulated angular distributions with 20 equally-sized scattering angle cosine meshes. This document describes the VIM multigroup capability, the procedures for generating multigroup cross sections for VIM, and their use. The multigroup cross section generating code, ISOVIM, is described, and benchmark testing is documented.

  6. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    SciTech Connect

    Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

    1995-08-01

    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 {sup 6}Li, {sup 7}Li, 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).

  7. A Computer Program for the Reactivity and Kinetic Parameters for Two-Dimensional Triangular Geometry by Transport Perturbation Theory.

    Energy Science and Technology Software Center (ESTSC)

    1990-04-25

    Version 00 TPTRIA calculates reactivity, effective delayed neutron fractions and mean generation time for two-dimensional triangular geometry on the basis of neutron transport perturbation theory. DIAMANT2 (also designated as CCC-414), is a multigroup two-dimensional discrete ordinates transport code system for triangular and hexagonal geometry which calculates direct and adjoint angular fluxes.

  8. Improved convergence of Monte Carlo generated multi-group scattering moments

    SciTech Connect

    Nelson, A. G.; Martin, W. R.

    2013-07-01

    This paper introduces an improved method of obtaining multi-group scattering moments from a Monte Carlo neutron transport code for use in deterministic transport solvers. The new method increases the information obtained from scattering events and therefore has more useful convergence characteristics than the currently used analog techniques. A prototype of the improved method was implemented in the OpenMC Monte Carlo transport code to compare the accuracy and convergence characteristics of the new method. The prototype showed that accuracy was retained (or improved) while increasing the figure-of-merit for the generation of multi-group scattering moments. (authors)

  9. Transport Code for Regular Triangular Geometry

    Energy Science and Technology Software Center (ESTSC)

    1993-06-09

    DIAMANT2 solves the two-dimensional static multigroup neutron transport equation in planar regular triangular geometry. Both regular and adjoint, inhomogeneous and homogeneous problems subject to vacuum, reflective or input specified boundary flux conditions are solved. Anisotropy is allowed for the scattering source. Volume and surface sources are allowed for inhomogeneous problems.

  10. Multigroup Reactor Lattice Cell Calculation

    Energy Science and Technology Software Center (ESTSC)

    1990-03-01

    The Winfrith Improved Multigroup Scheme (WIMS), is a general code for reactor lattice cell calculations on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters, and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered themore » choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are available in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a succesor version of WIMS-D/4.« less

  11. The LAW Library -- A multigroup cross-section library for use in radioactive waste analysis calculations

    SciTech Connect

    Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.

    1994-08-01

    The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ({sup 14}N{sub 7}, {sup 15}N{sub 7}, {sup 16}O{sub 8}, {sup 154Eu}{sub 63}, and {sup 155}Eu{sub 63}). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections.

  12. New Methodologies for Generation of Multigroup Cross Sections for Shielding Applications

    NASA Astrophysics Data System (ADS)

    Arzu Alpan, F.; Haghighat, Alireza

    2003-06-01

    Coupled neutron and gamma multigroup (broad-group) libraries used for Light Water Reactor shielding and dosimetry commonly include 47-neutron and 20-gamma groups. These libraries are derived from the 199-neutron, 42-gamma fine-group VITAMIN-B6 library. In this paper, we introduce modifications to the generation procedure of the broad-group libraries. Among these modifications, we show that the fine-group structure and collapsing technique have the largest impact. We demonstrate that a more refined fine-group library and the bi-linear adjoint weighting collapsing technique can improve the accuracy of transport calculation results.

  13. Asymptotic, multigroup flux reconstruction and consistent discontinuity factors

    DOE PAGESBeta

    Trahan, Travis J.; Larsen, Edward W.

    2015-05-12

    Recent theoretical work has led to an asymptotically derived expression for reconstructing the neutron flux from lattice functions and multigroup diffusion solutions. The leading-order asymptotic term is the standard expression for flux reconstruction, i.e., it is the product of a shape function, obtained through a lattice calculation, and the multigroup diffusion solution. The first-order asymptotic correction term is significant only where the gradient of the diffusion solution is not small. Inclusion of this first-order correction term can significantly improve the accuracy of the reconstructed flux. One may define discontinuity factors (DFs) to make certain angular moments of the reconstructed fluxmore » continuous across interfaces between assemblies in 1-D. Indeed, the standard assembly discontinuity factors make the zeroth moment (scalar flux) of the reconstructed flux continuous. The inclusion of the correction term in the flux reconstruction provides an additional degree of freedom that can be used to make two angular moments of the reconstructed flux continuous across interfaces by using current DFs in addition to flux DFs. Thus, numerical results demonstrate that using flux and current DFs together can be more accurate than using only flux DFs, and that making the second angular moment continuous can be more accurate than making the zeroth moment continuous.« less

  14. Asymptotic, multigroup flux reconstruction and consistent discontinuity factors

    SciTech Connect

    Trahan, Travis J.; Larsen, Edward W.

    2015-05-12

    Recent theoretical work has led to an asymptotically derived expression for reconstructing the neutron flux from lattice functions and multigroup diffusion solutions. The leading-order asymptotic term is the standard expression for flux reconstruction, i.e., it is the product of a shape function, obtained through a lattice calculation, and the multigroup diffusion solution. The first-order asymptotic correction term is significant only where the gradient of the diffusion solution is not small. Inclusion of this first-order correction term can significantly improve the accuracy of the reconstructed flux. One may define discontinuity factors (DFs) to make certain angular moments of the reconstructed flux continuous across interfaces between assemblies in 1-D. Indeed, the standard assembly discontinuity factors make the zeroth moment (scalar flux) of the reconstructed flux continuous. The inclusion of the correction term in the flux reconstruction provides an additional degree of freedom that can be used to make two angular moments of the reconstructed flux continuous across interfaces by using current DFs in addition to flux DFs. Thus, numerical results demonstrate that using flux and current DFs together can be more accurate than using only flux DFs, and that making the second angular moment continuous can be more accurate than making the zeroth moment continuous.

  15. NASA-Lewis experiences with multigroup cross sections and shielding calculations

    NASA Technical Reports Server (NTRS)

    Lahti, G. P.

    1972-01-01

    The nuclear reactor shield analysis procedures employed at NASA-Lewis are described. Emphasis is placed on the generation, use, and testing of multigroup cross section data. Although coupled neutron and gamma ray cross section sets are useful in two dimensional Sn transport calculations, much insight has been gained from examination of uncoupled calculations. These have led to experimental and analytic studies of areas deemed to be of first order importance to reactor shield calculations. A discussion is given of problems encountered in using multigroup cross sections in the resolved resonance energy range. The addition to ENDF files of calculated and/or measured neutron-energy-dependent capture gamma ray spectra for shielding calculations is questioned for the resonance region. Anomalies inherent in two dimensional Sn transport calculations which may overwhelm any cross section discrepancies are illustrated.

  16. New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

    SciTech Connect

    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.

  17. Parallel computation of multigroup reactivity coefficient using iterative method

    NASA Astrophysics Data System (ADS)

    Susmikanti, Mike; Dewayatna, Winter

    2013-09-01

    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.

  18. Parallel computation of multigroup reactivity coefficient using iterative method

    SciTech Connect

    Susmikanti, Mike; Dewayatna, Winter

    2013-09-09

    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.

  19. A Note on Multigroup Comparisons Using SAS PROC CALIS

    ERIC Educational Resources Information Center

    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…

  20. Multigroup Free-atom Doppler-broadening Approximation. Theory

    SciTech Connect

    Gray, Mark Girard

    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.

  1. A Bayesian Approach for Multigroup Nonlinear Factor Analysis.

    ERIC Educational Resources Information Center

    Song, Xin-Yuan; Lee, Sik-Yum

    2002-01-01

    Developed a Bayesian approach for a general multigroup nonlinear factor analysis model that simultaneously obtains joint Bayesian estimates of the factor scores and the structural parameters subjected to some constraints across different groups. (SLD)

  2. Development of a new two-dimensional Cartesian geometry nodal multigroup discrete-ordinates method

    SciTech Connect

    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.

  3. New multigroup Monte Carlo scattering algorithm suitable for neutral- and charged-particle Boltzmann and Fokker-Planck calculations

    SciTech Connect

    Sloan, D.P.

    1983-05-01

    Morel (1981) has developed multigroup Legendre cross sections suitable for input to standard discrete ordinates transport codes for performing charged-particle Fokker-Planck calculations in one-dimensional slab and spherical geometries. Since the Monte Carlo neutron transport code, MORSE, uses the same multigroup cross section data that discrete ordinates codes use, it was natural to consider whether Fokker-Planck calculations could be performed with MORSE. In order to extend the unique three-dimensional forward or adjoint capability of MORSE to Fokker-Planck calculations, the MORSE code was modified to correctly treat the delta-function scattering of the energy operator, and a new set of physically acceptable cross sections was derived to model the angular operator. Morel (1979) has also developed multigroup Legendre cross sections suitable for input to standard discrete ordinates codes for performing electron Boltzmann calculations. These electron cross sections may be treated in MORSE with the same methods developed to treat the Fokker-Planck cross sections. The large magnitude of the elastic scattering cross section, however, severely increases the computation or run time. It is well-known that approximate elastic cross sections are easily obtained by applying the extended transport (or delta function) correction to the Legendre coefficients of the exact cross section. An exact method for performing the extended transport cross section correction produces cross sections which are physically acceptable. Sample calculations using electron cross sections have demonstrated this new technique to be very effective in decreasing the large magnitude of the cross sections.

  4. RZ calculations for self shielded multigroup cross sections

    SciTech Connect

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

    2006-07-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  5. Multigroup Confirmatory Factor Analysis: Locating the Invariant Referent Sets

    ERIC Educational Resources Information Center

    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…

  6. Application de la methode des sous-groupes au calcul Monte-Carlo multigroupe

    NASA Astrophysics Data System (ADS)

    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

  7. Nonlinear diffusion acceleration for the multigroup transport equation discretized with S{sub N} and continuous FEM with rattlesnake

    SciTech Connect

    Wang, Y.

    2013-07-01

    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 (S{sub N}) 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 RattleS{sub N}ake, a MOOSE-based application at INL. Convergence results are presented. (authors)

  8. Collision probability calculation and multigroup flux solvers using PVM

    SciTech Connect

    Qaddouri, A.; Roy, R.; Mayrand, M.; Goulard, B.

    1996-07-01

    Collision probability evaluation and flux computation are the most time-consuming aspects of applications based on the linearized time-independent transport equation. Parallelization for collision probability calculation and multigroup flux computation are investigated. Particular techniques pertinent to the two-step energy/space iterative process of solving a multigroup transport equation are described. The parallel performance is studied in cases where the cyclic tracking technique is used to integrate collision probability. Parallelization is achieved by distributing either different energy groups or different regions on a set of processors. These algorithms were tested on a four-processor IBM SP2 and an eight-processor SPARC 1000 as well as on networks of workstations using the public domain PVM library. Typical run times are provided for unit cell calculations.

  9. CASTRO: A NEW COMPRESSIBLE ASTROPHYSICAL SOLVER. III. MULTIGROUP RADIATION HYDRODYNAMICS

    SciTech Connect

    Zhang, W.; Almgren, A.; Bell, J.; Howell, L.; Burrows, A.; Dolence, J.

    2013-01-15

    We present a formulation for multigroup radiation hydrodynamics that is correct to order O(v/c) using the comoving-frame approach and the flux-limited diffusion approximation. We describe a numerical algorithm for solving the system, implemented in the compressible astrophysics code, CASTRO. CASTRO uses a Eulerian grid with block-structured adaptive mesh refinement based on a nested hierarchy of logically rectangular variable-sized grids with simultaneous refinement in both space and time. In our multigroup radiation solver, the system is split into three parts: one part that couples the radiation and fluid in a hyperbolic subsystem, another part that advects the radiation in frequency space, and a parabolic part that evolves radiation diffusion and source-sink terms. The hyperbolic subsystem and the frequency space advection are solved explicitly with high-order Godunov schemes, whereas the parabolic part is solved implicitly with a first-order backward Euler method. Our multigroup radiation solver works for both neutrino and photon radiation.

  10. 950809 Charged particle transport updated multi-group diffusion

    SciTech Connect

    Corman, E.G.; Perkins, S.T.; Dairiki, N.T.

    1995-09-01

    In 1974, a charged particle transport scheme was introduced which utilized a multi-group diffusion method for the spatial transport and slowing down of energetic ions in a hot plasma. In this treatment a diffusion coefficient was used which was flux-limited to provide, hopefully, some degree of accuracy when the slowing down of an energetic charged particle is dominated by Coulomb collisions with thermal ions and electrons in a plasma medium. An advantage of this method was a very fast, memory-contained program for calculating the behavior of energetic charged particles which resulted in smoothly varying particle number densities and energy depositions. The main limitation of the original multi-group charged particle diffusion scheme is its constraint to a basic ten group structure; the same ten group structure for each of the five energetic ions tracked. This is regarded as a severe limitation, inasmuch as more groups would be desired to simulate more accurately the corresponding Monte Carlo results of energies deposited over spatial zones from a charged particle source. More generally, it seems preferable to have a different group structure for each particle type since they are created at inherently different energies. In this paper, the basic theory and multi-group description will be given. This is followed by the specific techniques that were used to solve the resultant equations. Finally, the modifications that were made to the cross section data as well as the methods used for energy and momentum deposition are described.

  11. CASTRO: A New Compressible Astrophysical Solver. III. Multigroup Radiation Hydrodynamics

    NASA Astrophysics Data System (ADS)

    Zhang, W.; Howell, L.; Almgren, A.; Burrows, A.; Dolence, J.; Bell, J.

    2013-01-01

    We present a formulation for multigroup radiation hydrodynamics that is correct to order O(v/c) using the comoving-frame approach and the flux-limited diffusion approximation. We describe a numerical algorithm for solving the system, implemented in the compressible astrophysics code, CASTRO. CASTRO uses a Eulerian grid with block-structured adaptive mesh refinement based on a nested hierarchy of logically rectangular variable-sized grids with simultaneous refinement in both space and time. In our multigroup radiation solver, the system is split into three parts: one part that couples the radiation and fluid in a hyperbolic subsystem, another part that advects the radiation in frequency space, and a parabolic part that evolves radiation diffusion and source-sink terms. The hyperbolic subsystem and the frequency space advection are solved explicitly with high-order Godunov schemes, whereas the parabolic part is solved implicitly with a first-order backward Euler method. Our multigroup radiation solver works for both neutrino and photon radiation.

  12. Hybrid method of deterministic and probabilistic approaches for continuous energy neutron transport problem

    SciTech Connect

    Lee, H.; Lee, D.

    2013-07-01

    This paper presents a new hybrid method of continuous energy Monte Carlo (MC) and multi-group Method of Characteristics (MOC). For a continuous energy neutron transport analysis, the hybrid method employs a continuous energy MC for resonance energy range to treat the resonances accurately and a multi-group MOC for high and low energy ranges for efficiency. Numerical test with a model problem confirms that the hybrid method can produce consistent results with the reference continuous energy MC-only calculation as well as multi-group MOC-only calculation. (authors)

  13. A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections.

    Energy Science and Technology Software Center (ESTSC)

    1983-03-23

    Version 01/02 The code reads multigroup cross sections from a compatible data file and collapses user-selected reaction cross sections to any few-group structure using one of a variety of user neutron flux spectrum options given below: Option Flux description 1 Built-in function including Maxwellian, fission, fusion and slowing-down regions and requiring user-specified parameters and energy-region boundaries. 2 Set of log-log flux-energy interpolation points read from input cross-section data file. 3 Set of log-log flux-energy interpolationmore » points read from user-supplied card input. 4 - 6 Histogram flux values read from user-supplied card input in arbitrary group structure in units of flux-per unit-energy, flux-per-unit lethargy, or integral group flux. LAFPX-E may be used to collapse any set of multigroup reaction cross sections furnished in the required format. However, the code was developed for, and is furnished with, a library of 154-group fission-product cross sections processed from ENDF/B-IV with a typical light water reactor (LWR) flux spectrum and temperature. Four-group radiative capture cross sections produced for LWR calculations are tabulated in the code documentation and are incorporated in the EPRI-CINDER data library, RSIC Code Package CCC-309.« less

  14. a New ENDF/B-VII.0 Based Multigroup Cross-Section Library for Reactor Dosimetry

    NASA Astrophysics Data System (ADS)

    Alpan, F. A.; Anderson, S. L.

    2009-08-01

    The latest of the ENDF/B libraries, ENDF/B-VII.0 was released in December 2006. In this paper, the ENDF/B-VII.O evaluations were used in generating a new coupled neutron/gamma multigroup library having the same group structure of VITAMIN-B6, i.e., the 199-neutron, 42-gamma group library. The new library was generated utilizing NJOY99.259 for pre-processing and the AMPX modules for post-processing of cross sections. An ENDF/B-VI.3 based VITAMIN-B6-like library was also generated. The fine-group libraries and the ENDF/B-VI.3 based 47-neutron, 20-gamma group BUGLE-96 library were used with the discrete ordinates code DORT to obtain a three-dimensional synthesized flux distribution from r, r-θ, and r-z models for a standard Westinghouse 3-loop design reactor. Reaction rates were calculated for ex-vessel neutron dosimetry containing 63Cu(n,α)60Co, 46Ti(n,p)46Sc, 54Fe(n,P)54Mn, 58Ni(n,P)58Co, 238U(n,f)137Cs, 237Np(n,f)137Cs, and 59Co(n,γ)60Co (bare and cadmium covered) reactions. Results were compared to measurements. In comparing the 199-neutron, 42-gamma group ENDF/B-VI.3 and ENDF/B-VII.O libraries, it was observed that the ENDF/B-VI.3 based library results were in better agreement with measurements. There is a maximum difference of 7% (for the 63Cu(n,α)60Co reaction rate calculation) between ENDF/B-VI.3 and ENDF/B-VII.O. Differences between ENDF/B-VI.3 and ENDF/B-VII.O libraries are due to 16O, 1H, 90Zr, 91Zr, 92Zr, 238U, and 239Pu evaluations. Both ENDF/B-VI.3 and ENDF/B-VII.O library calculated reaction rates are within 20% of measurement and meet the criterion specified in the U. S. Nuclear Regulatory Commission Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."

  15. Multigroup Free-atom Doppler-broadening Approximation. Experiment

    SciTech Connect

    Gray, Mark Girard

    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 1H, 56Fe, and 235U 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.

  16. Multi-Group Formulation of the Temperature-Dependent Resonance Scattering Model and its Impact on Reactor Core Parameters

    SciTech Connect

    Ghrayeb, Shadi Z.; Ougouag, Abderrafi M.; Ouisloumen, Mohamed; Ivanov, Kostadin N.

    2014-01-01

    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 accounts for it 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 the results done on this topic to date.

  17. BUGLE-96: A revised multigroup cross section library for LWR applications based on ENDF/B-VI Release 3

    SciTech Connect

    White, J.E.; Ingersoll, D.T.; Slater, C.O.; Roussin, R.W.

    1996-05-01

    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.

  18. Status of the MORSE multigroup Monte Carlo radiation transport code

    SciTech Connect

    Emmett, M.B.

    1993-06-01

    There are two versions of the MORSE multigroup Monte Carlo radiation transport computer code system at Oak Ridge National Laboratory. MORSE-CGA is the most well-known and has undergone extensive use for many years. MORSE-SGC was originally developed in about 1980 in order to restructure the cross-section handling and thereby save storage. However, with the advent of new computer systems having much larger storage capacity, that aspect of SGC has become unnecessary. Both versions use data from multigroup cross-section libraries, although in somewhat different formats. MORSE-SGC is the version of MORSE that is part of the SCALE system, but it can also be run stand-alone. Both CGA and SGC use the Multiple Array System (MARS) geometry package. In the last six months the main focus of the work on these two versions has been on making them operational on workstations, in particular, the IBM RISC 6000 family. A new version of SCALE for workstations is being released to the Radiation Shielding Information Center (RSIC). MORSE-CGA, Version 2.0, is also being released to RSIC. Both SGC and CGA have undergone other revisions recently. This paper reports on the current status of the MORSE code system.

  19. ATR neutron spectral characterization

    SciTech Connect

    Rogers, J.W.; Anderl, R.A.

    1995-11-01

    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.

  20. Using SAS PROC TCALIS for multigroup structural equation modelling with mean structures.

    PubMed

    Gu, Fei; Wu, Wei

    2011-11-01

    Multigroup structural equation modelling (SEM) is a technique frequently used to evaluate measurement invariance in social and behavioural science research. Before version 9.2, SAS was incapable of handling multigroup SEM. However, this limitation is resolved in PROC TCALIS in SAS 9.2. For the purpose of illustration, this paper provides a step-by-step guide to programming the tests of measurement invariance and partial invariance using PROC TCALIS for multigroup SEM with mean structures. Fit indices and parameter estimates are validated, thus providing an alternative tool for researchers conducting both applied and simulated studies. Other new features (e.g., different types of modelling languages and estimation methods) and limitations (e.g., ordered-categorical SEM and multilevel SEM) of the TCALIS procedure are also briefly discussed. PMID:21973099

  1. Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry.

    Energy Science and Technology Software Center (ESTSC)

    1991-01-25

    Version 00 TPHEX calculates the multigroup neutron flux distribution in an assembly of hexagonal cells using a transmission probability (interface current) method. It is primarily intended for calculations on hexagonal LWR fuel assemblies but can be used for other purposes subject to the qualifications mentioned in Restrictions/Limitations.

  2. 66 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV.

    Energy Science and Technology Software Center (ESTSC)

    1995-12-12

    Version 00 For a variety of applications (accelerator shielding, the use of neutrons in radiotherapy, radiation damage studies, etc.) It is necessary to carry out transport calculations involving medium-energy neutrons. HILO86R multigroup cross sections are in the form needed for the CCC-254/ANISN-ORNL and CCC-543/TORT-DORT discrete ordinates codes and in the CCC-474/MORSE-CGA Monte Carlo code.

  3. VELM61 and VELM22: Multigroup cross-section libraries for sodium-cooled reactor shield analysis

    SciTech Connect

    Fu, C.Y.; Ingersoll, D.T.

    1987-04-01

    Two coupled neutron and photon multigroup cross-section libraries, derived from ENDF/B-V nuclear data, are described. The energy group structures, 61n/23..gamma.. and 22n/10..gamma.., are subsets of the Vitamin-E 174n/38..gamma.. 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.

  4. A multilevel method for coupling the neutron kinetics and heat transfer equations

    SciTech Connect

    Tamang, A.; Anistratov, D. Y.

    2013-07-01

    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 {alpha}-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)

  5. Testing for Two-Way Interactions in the Multigroup Common Factor Model

    ERIC Educational Resources Information Center

    van Smeden, Maarten; Hessen, David J.

    2013-01-01

    In this article, a 2-way multigroup common factor model (MG-CFM) is presented. The MG-CFM can be used to estimate interaction effects between 2 grouping variables on 1 or more hypothesized latent variables. For testing the significance of such interactions, a likelihood ratio test is presented. In a simulation study, the robustness of the…

  6. A discretization of the multigroup PN radiative transfer equation on general meshes

    NASA Astrophysics Data System (ADS)

    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.

  7. Psychometric Properties of Multigroup Ethnic Identity Measure (MEIM) Scores with Australian Adolescents from Diverse Ethnocultural Groups

    ERIC Educational Resources Information Center

    Dandy, Justine; Durkin, Kevin; McEvoy, Peter; Barber, Bonnie L.; Houghton, Stephen

    2008-01-01

    The present study investigated the reliability and factor structure of scores on a 12-item version of Phinney's multigroup ethnic identity measure with an Australian sample from diverse cultural backgrounds. Participants were 485 students aged between 10 and 15 years. The results generally supported the reliability of the ethnic identity scale…

  8. 1,2,3-D Diffusion Depletion Multi-Group

    Energy Science and Technology Software Center (ESTSC)

    1992-04-20

    CITATION is designed to solve problems using the finite difference representation of neutron diffusion theory, treating up to three space dimensions with arbitrary group to group scattering. X-y-z, theta-r-z, hexagonal z, and triagonal z geometries may be treated. Depletion problems may be solved and fuel managed for multi-cycle analysis. Extensive first order perturbation results may be obtained given microscopic data and nuclide concentrations. Statics problems may be solved and perturbation results obtained with microscopic data.

  9. A conservative multi-group approach to the Boltzmann equations for reactive gas mixtures

    NASA Astrophysics Data System (ADS)

    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.

  10. Honor Thy Parents: An Ethnic Multigroup Analysis of Filial Responsibility, Health Perceptions, and Caregiving Decisions.

    PubMed

    Santoro, Maya S; Van Liew, Charles; Holloway, Breanna; McKinnon, Symone; Little, Timothy; Cronan, Terry A

    2016-08-01

    The present study explores patterns of parity and disparity in the effect of filial responsibility on health-related evaluations and caregiving decisions. Participants who identified as White, Black, Hispanic, or Asian/Pacific Islander read a vignette about an older man needing medical care. They were asked to imagine that they were the man's son and answer questions regarding their likelihood of hiring a health care advocate (HCA) for services related to the father's care. A multigroup (ethnicity) path analysis was performed, and an intercept invariant multigroup model fits the data best. Direct and indirect effect estimation showed that filial responsibility mediated the relationship between both the perceived severity of the father's medical condition and the perceived need for medical assistance and the likelihood of hiring an HCA only for White and Hispanic participants, albeit differently. The findings demonstrate that culture and ethnicity affect health evaluations and caregiving decision making. PMID:26282571

  11. A multigroup radiation diffusion test problem: Comparison of code results with analytic solution

    SciTech Connect

    Shestakov, A I; Harte, J A; Bolstad, J H; Offner, S R

    2006-12-21

    We consider a 1D, slab-symmetric test problem for the multigroup radiation diffusion and matter energy balance equations. The test simulates diffusion of energy from a hot central region. Opacities vary with the cube of the frequency and radiation emission is given by a Wien spectrum. We compare results from two LLNL codes, Raptor and Lasnex, with tabular data that define the analytic solution.

  12. EXTENSION OF THE 1D FOUR-GROUP ANALYTIC NODAL METHOD TO FULL MULTIGROUP

    SciTech Connect

    B. D. Ganapol; D. W. Nigg

    2008-09-01

    In the mid 80’s, a four-group/two-region, entirely analytical 1D nodal benchmark appeared. It was readily acknowledged that this special case was as far as one could go in terms of group number and still achieve an analytical solution. In this work, we show that by decomposing the solution to the multigroup diffusion equation into homogeneous and particular solutions, extension to any number of groups is a relatively straightforward exercise using the mathematics of linear algebra.

  13. Evaluated Nuclear Data Library for Transport Calculations Involving Incident Neutrons and Protons of Energy Up to 100 MeV.

    Energy Science and Technology Software Center (ESTSC)

    1993-08-09

    Version 00 This data base was developed for use in Monte Carlo or discrete ordinate transport codes, for example, the general Monte Carlo code MCNP. Various modules of the NJOY processing code system have been enhanced to permit processing of the ENDF/B-VI formatted evaluations into both continuous-energy and multi-group format. The transport data files for all 18 projectile-plus-target systems have been processed through NJOY, and coupled multi-particle, multi-group transport libraries for MCNP now exist. Inmore » addition, pointwise MCNP libraries to 100 MeV for incident neutrons have been prepared for the nine targets. The production version of the MCNP code is being modified to handle the new pointwise libraries. The production version of MCNP already supports the use of coupled multi-group libraries.« less

  14. Anisotropic Elastic Resonance Scattering model for the Neutron Transport equation

    SciTech Connect

    Mohamed Ouisloumen; Abderrafi M. Ougouag; Shadi Z. Ghrayeb

    2014-11-24

    The resonance scattering transfer cross-section has been reformulated to account for anisotropic scattering in the center-of-mass of the neutron-nucleus system. The main innovation over previous implementations is the relaxation of the ubiquitous assumption of isotropic scattering in the center-of-mass and the actual effective use of scattering angle distributions from evaluated nuclear data files in the computation of the angular moments of the resonant scattering kernels. The formulas for the high order anisotropic moments in the laboratory system are also derived. A multi-group numerical formulation is derived and implemented into a module incorporated within the NJOY nuclear data processing code. An ultra-fine energy mesh cross section library was generated using these new theoretical models and then was used for fuel assembly calculations with the PARAGON lattice physics code. The results obtained indicate a strong effect of this new model on reactivity, multi-group fluxes and isotopic inventory during depletion.

  15. Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor

    SciTech Connect

    Wemple, C.A.

    1995-05-01

    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%.

  16. Global analysis on a class of multi-group SEIR model with latency and relapse.

    PubMed

    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

  17. Simulate-HEX - The multi-group diffusion equation in hexagonal-z geometry

    SciTech Connect

    Lindahl, S. O.

    2013-07-01

    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)

  18. Geospatial Data Fusion and Multigroup Decision Support for Surface Water Quality Management

    NASA Astrophysics Data System (ADS)

    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

  19. Conformity Between LR0 Mock-Ups and Vvers Npp Rpv Neutron Flux Attenuation

    NASA Astrophysics Data System (ADS)

    Belousov, Sergey; Ilieva, Krassimira; Kirilova, Desislava

    2009-08-01

    The conformity of the mock-up results and those for reactor pressure vessel (RPV) of nuclear power plants (NPP) has been evaluated in order to qualify if the mock-ups data could be used for benchmark's purpose only, or/and for simulating of the NPP irradiation conditions. Neutron transport through the vessel has been calculated by the three-dimensional discrete ordinate code TORT with problem oriented multigroup energy neutron cross-section library BGL. Neutron flux/fluence and spectrum shape represented by normalized group neutron fluxes in the multigroup energy structure, for neutrons with energy above 0.5 MeV, have been used for conformity analysis. It has been demonstrated that the relative difference of the attenuation factor as well as the group neutron fluxes did not exceed 10% at all considered positions for VVER-440. For VVER-1000, it has been obtained the same consistency, except for the location behind the RPV. The neutron flux attenuation behind the RPV is 18% higher than the mock-up attenuation. It has been shown that this difference arises from the dissimilarity of the biological shielding. The obtained results have demonstrated that the VVERs' mock-ups are appropriate for simulating the NPP irradiation conditions. The mock-up results for VVER-1000 have to be applied more carefully i.e. taking into account the existing peculiarity of the biological shielding and RPV attenuation azimuthal dependence.

  20. The Group-Level Consequences of Sexual Conflict in Multigroup Populations

    PubMed Central

    Eldakar, Omar Tonsi; Gallup, Andrew C.

    2011-01-01

    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. PMID:22039491

  1. Consistent Multigroup Theory Enabling Accurate Course-Group Simulation of Gen IV Reactors

    SciTech Connect

    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.

  2. A stable 1D multigroup high-order low-order method

    DOE PAGESBeta

    Yee, Ben Chung; Wollaber, Allan Benton; Haut, Terry Scot; Park, HyeongKae

    2016-07-13

    The high-order low-order (HOLO) method is a recently developed moment-based acceleration scheme for solving time-dependent thermal radiative transfer problems, and has been shown to exhibit orders of magnitude speedups over traditional time-stepping schemes. However, a linear stability analysis by Haut et al. (2015 Haut, T. S., Lowrie, R. B., Park, H., Rauenzahn, R. M., Wollaber, A. B. (2015). A linear stability analysis of the multigroup High-Order Low-Order (HOLO) method. In Proceedings of the Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method; Nashville, TN, April 19–23, 2015. American Nuclear Society.)more » revealed that the current formulation of the multigroup HOLO method was unstable in certain parameter regions. Since then, we have replaced the intensity-weighted opacity in the first angular moment equation of the low-order (LO) system with the Rosseland opacity. Furthermore, this results in a modified HOLO method (HOLO-R) that is significantly more stable.« less

  3. Ethnic Identity Assessment among Inner-City African American Children: Evaluating the Applicability of the Multigroup Ethnic Identity Measure.

    ERIC Educational Resources Information Center

    Reese, Le'Roy E.; Vera, Elizabeth M.; Paikoff, Roberta L.

    1998-01-01

    Examined the reliability of the Multigroup Ethnic Identity Scale (MEIM) (J. Phinney, 1992) in measuring feelings of ethnic affirmation and belongings, ethnic behaviors, and ethnic knowledge with 118 inner-city African American children aged 8 to 12 years. Results of reliability analyses are mixed and indicate that constructs of ethnic behavior and…

  4. The Power to Detect Sex Differences in IQ Test Scores Using Multi-Group Covariance and Means Structure Analyses

    ERIC Educational Resources Information Center

    Molenaar, Dylan; Dolan, Conor V.; Wicherts, Jelle M.

    2009-01-01

    Research into sex differences in general intelligence, g, has resulted in two opposite views. In the first view, a g-difference is nonexistent, while in the second view, g is associated with a male advantage. Past research using Multi-Group Covariance and Mean Structure Analysis (MG-CMSA) found no sex difference in g. This failure raised the…

  5. An Improved Neutron Transport Algorithm for Space Radiation

    NASA Technical Reports Server (NTRS)

    Heinbockel, John H.; Clowdsley, Martha S.; Wilson, John W.

    2000-01-01

    A low-energy neutron transport algorithm for use in space radiation protection is developed. The algorithm is based upon a multigroup analysis of the straight-ahead Boltzmann equation by using a mean value theorem for integrals. This analysis is accomplished by solving a realistic but simplified neutron transport test problem. The test problem is analyzed by using numerical and analytical procedures to obtain an accurate solution within specified error bounds. Results from the test problem are then used for determining mean values associated with rescattering terms that are associated with a multigroup solution of the straight-ahead Boltzmann equation. The algorithm is then coupled to the Langley HZETRN code through the evaporation source term. Evaluation of the neutron fluence generated by the solar particle event of February 23, 1956, for a water and an aluminum-water shield-target configuration is then compared with LAHET and MCNPX Monte Carlo code calculations for the same shield-target configuration. The algorithm developed showed a great improvement in results over the unmodified HZETRN solution. In addition, a two-directional solution of the evaporation source showed even further improvement of the fluence near the front of the water target where diffusion from the front surface is important.

  6. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  7. Recent validation experience with multigroup cross-section libraries and scale

    SciTech Connect

    Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.

    1995-12-01

    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.

  8. Global dynamics of a novel multi-group model for computer worms

    NASA Astrophysics Data System (ADS)

    Gong, Yong-Wang; Song, Yu-Rong; Jiang, Guo-Ping

    2013-04-01

    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.

  9. Analyzing average and conditional effects with multigroup multilevel structural equation models

    PubMed Central

    Mayer, Axel; Nagengast, Benjamin; Fletcher, John; Steyer, Rolf

    2014-01-01

    Conventionally, multilevel analysis of covariance (ML-ANCOVA) has been the recommended approach for analyzing treatment effects in quasi-experimental multilevel designs with treatment application at the cluster-level. In this paper, we introduce the generalized ML-ANCOVA with linear effect functions that identifies average and conditional treatment effects in the presence of treatment-covariate interactions. We show how the generalized ML-ANCOVA model can be estimated with multigroup multilevel structural equation models that offer considerable advantages compared to traditional ML-ANCOVA. The proposed model takes into account measurement error in the covariates, sampling error in contextual covariates, treatment-covariate interactions, and stochastic predictors. We illustrate the implementation of ML-ANCOVA with an example from educational effectiveness research where we estimate average and conditional effects of early transition to secondary schooling on reading comprehension. PMID:24795668

  10. Inference for low- and high-dimensional multigroup repeated measures designs with unequal covariance matrices.

    PubMed

    Happ, Martin; Harrar, Solomon W; Bathke, Arne C

    2016-07-01

    We propose tests for main and simple treatment effects, time effects, as well as treatment by time interactions in possibly high-dimensional multigroup repeated measures designs. The proposed inference procedures extend the work by Brunner et al. (2012) from two to several treatment groups and remain valid for unbalanced data and under unequal covariance matrices. In addition to showing consistency when sample size and dimension tend to infinity at the same rate, we provide finite sample approximations and evaluate their performance in a simulation study, demonstrating better maintenance of the nominal α-level than the popular Box-Greenhouse-Geisser and Huynh-Feldt methods, and a gain in power for informatively increasing dimension. Application is illustrated using electroencephalography (EEG) data from a neurological study involving patients with Alzheimer's disease and other cognitive impairments. PMID:26700536

  11. Ethnic Residential Segregation: A Multilevel, Multigroup, Multiscale Approach Exemplified by London in 2011.

    PubMed

    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

  12. Reactor Statics Module, RS-9: Multigroup Diffusion Program Using an Exponential Acceleration Technique.

    ERIC Educational Resources Information Center

    Macek, Victor C.

    The nine Reactor Statics Modules are designed to introduce students to the use of numerical methods and digital computers for calculation of neutron flux distributions in space and energy which are needed to calculate criticality, power distribution, and fuel burnup for both slow neutron and fast neutron fission reactors. The last module, RS-9,…

  13. Transport analysis of measured neutron energy spectra in a graphite stack with a collimated deuterium-tritium neutron beam

    SciTech Connect

    Tsechanski, A.; Ofek, R.; Goldfeld, A.; Shani, G.

    1989-02-01

    The Ben-Gurion University measurements of neutron energy spectra in a graphite stack, resulting from the scattering of 14.7-MeV neutrons streaming through a 6-cm-diam collimator in a 121-cm-thick paraffin wall, have been used as a benchmark for the compatability and accuracy of discrete ordinates, P/sub n/, and transport calculations and as a tool for fusion reactor neutronics. The transport analysis has been carried out with the DOT 4.2 discrete ordinates code and with cross sections processed with the NJOY code. Most of the parameters affecting the accuracy of the flux and L system scattering cross sections in the P/sub n/ approximation, the quadrature set employed, and the energy multigroup structure. First, a spectrum calculated with DOT 4.2, with a detector located on the axis of the system, was compared with a spectrum calculated with the MCNP Monte Carlo code, which was a preliminary verification of the DOT 4.2 results. Both calculated spectra were in good agreement. Next, the DOT 4.2 calculations were compared with the measured spectra. The comparison showed that the discrepancies between the measurements and the calculations increase as the distance between the detector and the system axis increases. This trend indicates that when the flux is determined mainly by multiple scatterings, a more divided multigroup structure should be employed.

  14. Local Multi-Grouped Binary Descriptor With Ring-Based Pooling Configuration and Optimization.

    PubMed

    Gao, Yongqiang; Huang, Weilin; Qiao, Yu

    2015-12-01

    Local binary descriptors are attracting increasingly attention due to their great advantages in computational speed, which are able to achieve real-time performance in numerous image/vision applications. Various methods have been proposed to learn data-dependent binary descriptors. However, most existing binary descriptors aim overly at computational simplicity at the expense of significant information loss which causes ambiguity in similarity measure using Hamming distance. In this paper, by considering multiple features might share complementary information, we present a novel local binary descriptor, referred as ring-based multi-grouped descriptor (RMGD), to successfully bridge the performance gap between current binary and floated-point descriptors. Our contributions are twofold. First, we introduce a new pooling configuration based on spatial ring-region sampling, allowing for involving binary tests on the full set of pairwise regions with different shapes, scales, and distances. This leads to a more meaningful description than the existing methods which normally apply a limited set of pooling configurations. Then, an extended Adaboost is proposed for an efficient bit selection by emphasizing high variance and low correlation, achieving a highly compact representation. Second, the RMGD is computed from multiple image properties where binary strings are extracted. We cast multi-grouped features integration as rankSVM or sparse support vector machine learning problem, so that different features can compensate strongly for each other, which is the key to discriminativeness and robustness. The performance of the RMGD was evaluated on a number of publicly available benchmarks, where the RMGD outperforms the state-of-the-art binary descriptors significantly. PMID:26292340

  15. Spatial corrections for pulsed-neutron reactivity measurements.

    SciTech Connect

    Cao, Y.; Lee, J.; Nuclear Engineering Division; Univ. of Michigan

    2010-07-01

    For pulsed-neutron experiments performed in a subcritical reactor, the reactivity obtained from the area-ratio method is sensitive to detector positions. The spatial effects are induced by the presence of both the prompt neutron harmonics and the delayed neutron harmonics in the reactor. The traditional kinetics distortion factor is only limited to correcting the spatial effects caused by the fundamental prompt-{alpha} mode. In this paper, we derive spatial correction factors fp and fd to account for spatial effects induced by the prompt neutron harmonics and the delayed neutron harmonics, respectively. Our numerical simulations with the FX2-TH time-dependent multigroup diffusion code indicate that the high-order prompt neutron harmonics lead to significant spatial effects and cannot be neglected in calculating the spatial correction factors. The prompt spatial correction factor fp can be simply determined by the ratio of the normalized detector responses corresponding to the fundamental k-mode and the prompt neutron flux integrated over the pulse period. Thus, it is convenient to calculate and provides physically intuitive explanations on the spatial dependence of reactivity measured in the MUSE-4 experiments: overestimation of the subcriticality in regions close to the external neutron source and underestimation of the subcriticality away from the source but within the fuel region.

  16. Axial expansion methods for solution of the multi-dimensional neutron diffusion equation

    SciTech Connect

    Beaklini Filho, J.F.

    1984-01-01

    The feasibility and practical implementation of axial expansion methods for the solution of the multi-dimensional multigroup neutron diffusion (MGD) equations is investigated. The theoretical examination which is applicable to the general MGD equations in arbitrary geometry includes the derivation of a new weak (reduced) form of the MGD equations by expanding the axial component of the neutron flux in a series of known trial functions and utilizing the Galerkin weighting. A general two-group albedo boundary condition is included in the weak form as a natural boundary condition. The application of different types of trial functions is presented.

  17. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  18. User's manual for FENAT: a two-dimensional multigroup diffusion theory Finite Element Neutral Atom Transport code

    SciTech Connect

    Hasan, M.Z.

    1986-07-01

    FENAT solves the two-dimensional energy dependent diffusion equation in Cartesian (X-Y) and cylindrical/toroidal (R-Z) coordinates. The boundary conditions allowed are: vacuum, reflection, albedo and surface source. The energy variable is treated by multigroup method. The resulting multigroup diffusion equation is solved by finite element Galerkin's method with triangular element discretization of the spatial domain. The algebraic matrix equation is solved by the direct method of Crout variation of Gauss' elimination. Dynamic memory allocation has been used so that the maximum problem size is limited by the size of active core storage of the machine. When necessary, the global matrix is stored in a binary disk file. FENAT is particularly suitable for the transport of neutral atoms in fusion plasmas.

  19. Second order time evolution of the multigroup diffusion and P{sub 1} equations for radiation transport

    SciTech Connect

    Olson, Gordon L.

    2011-08-20

    Highlights: {yields} An existing multigroup transport algorithm is extended to be second-order in time. {yields} A new algorithm is presented that does not require a grey acceleration solution. {yields} The two algorithms are tested with 2D, multi-material problems. {yields} 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 P{sub 1} forms of the transport equation. Two dimensional, multi-material test problems are used to compare the solution methods.

  20. FTR Set 500: a multigroup cross-section set for FTR analysis

    SciTech Connect

    Mann, F M

    1982-02-01

    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.

  1. Multi-Group Reductions of LTE Air Plasma Radiative Transfer in Cylindrical Geometries

    NASA Technical Reports Server (NTRS)

    Scoggins, James; Magin, Thierry Edouard Bertran; Wray, Alan; Mansour, Nagi N.

    2013-01-01

    Air plasma radiation in Local Thermodynamic Equilibrium (LTE) within cylindrical geometries is studied with an application towards modeling the radiative transfer inside arc-constrictors, a central component of constricted-arc arc jets. A detailed database of spectral absorption coefficients for LTE air is formulated using the NEQAIR code developed at NASA Ames Research Center. The database stores calculated absorption coefficients for 1,051,755 wavelengths between 0.04 µm and 200 µm over a wide temperature (500K to 15 000K) and pressure (0.1 atm to 10.0 atm) range. The multi-group method for spectral reduction is studied by generating a range of reductions including pure binning and banding reductions from the detailed absorption coefficient database. The accuracy of each reduction is compared to line-by-line calculations for cylindrical temperature profiles resembling typical profiles found in arc-constrictors. It is found that a reduction of only 1000 groups is sufficient to accurately model the LTE air radiation over a large temperature and pressure range. In addition to the reduction comparison, the cylindrical-slab formulation is compared with the finite-volume method for the numerical integration of the radiative flux inside cylinders with varying length. It is determined that cylindrical-slabs can be used to accurately model most arc-constrictors due to their high length to radius ratios.

  2. Stability analysis of multi-group deterministic and stochastic epidemic models with vaccination rate

    NASA Astrophysics Data System (ADS)

    Wang, Zhi-Gang; Gao, Rui-Mei; Fan, Xiao-Ming; Han, Qi-Xing

    2014-09-01

    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.

  3. Review of uncertainty files and improved multigroup cross section files for FENDL

    NASA Astrophysics Data System (ADS)

    Ganesan, S.

    1994-03-01

    The IAEA Nuclear Data Section, in co-operation with several national nuclear data centers and research groups, is creating an internationally available Fusion Evaluated Nuclear Data Library (FENDL), which will serve as a comprehensive source of processed and tested nuclear data tailored to the requirements of the Engineering and Development Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project and other fusion-related development projects. The FENDL project of the International Atomic Energy Agency has the task of coordination with the goal of assembling, processing and testing a comprehensive, fusion-relevant Fusion Evaluated Nuclear Data Library with unrestricted international distribution. The present report contains the summary of the IAEA Advisory Group Meeting on 'Review of Uncertainty Files and Improved Multigroup Cross Section Files for FENDL', held during 8-12 November 1993 at the Tokai Research Establishment, JAERI, Japan, organized in cooperation with the Japan Atomic Energy Research Institute. The report presents the current status of the FENDL activity and the future work plans in the form of conclusions and recommendations of the four Working Groups of the Advisory Group Meeting on: (1) experimental and calculational benchmarks; (2) preparation processed libraries for FENDL/ITER; (3) specifying procedures for improving FENDL; and (4) selection of activation libraries for FENDL.

  4. Performance testing of the upgraded uranium isotopics multi-group analysis code MGAU

    NASA Astrophysics Data System (ADS)

    Berlizov, A. N.; Gunnink, R.; Zsigrai, J.; Nguyen, C. T.; Tryshyn, V. V.

    2007-06-01

    The paper describes recent developments of the MGAU (Multi-Group Analysis for Uranium) method, which resulted in the creation of an upgraded version 4.0 of the MGAU code. The major improvements concerned the procedure of the intrinsic efficiency calibration, particularly in the 120-205 keV region. The results of the tests carried out with the use of certified reference uranium isotopic materials SRM 969 and CRM 146 showed a significantly improved performance of the upgraded MGAU code for the accurate characterization of 235U and 234U abundances. The relative systematic biases of the abundances measured were evaluated not to exceed 1% and 3% over the concentration intervals 0.32-93.2 and 0.002-0.98 mass% for 235U and 234U, respectively. The influence of an up to 4-5 mm steel equivalent absorber and sample thickness on the measurement results was found to be much smaller than in previous versions of the code.

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

    DOE PAGESBeta

    Olson, Gordon L.

    2015-09-24

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

  6. Global dynamics of a general class of multi-group epidemic models with latency and relapse.

    PubMed

    Feng, Xiaomei; Teng, Zhidong; Zhang, Fengqin

    2015-02-01

    A multi-group model is proposed to describe a general relapse phenomenon of infectious diseases in heterogeneous populations. In each group, the population is divided into susceptible, exposed, infectious, and recovered subclasses. A general nonlinear incidence rate is used in the model. The results show that the global dynamics are completely determined by the basic reproduction number R0. In particular, a matrix-theoretic method is used to prove the global stability of the disease-free equilibrium when R0 ≤ 1, while a new combinatorial identity (Theorem 3.3 in Shuai and van den Driessche) in graph theory is applied to prove the global stability of the endemic equilibrium when R0 > 1. We would like to mention that by applying the new combinatorial identity, a graph of 3n (or 2n+m) vertices can be converted into a graph of n vertices in order to deal with the global stability of the endemic equilibrium in this paper. PMID:25811334

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

    SciTech Connect

    Olson, Gordon L.

    2015-09-24

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

  8. Neutron diffusion in graphite poisoned with 1/v and non-1/v absorbers

    SciTech Connect

    Malik, U.; Kothari, L.S.; Kumar, A.

    1982-05-01

    Neutron diffusion in graphite containing 1/v and non-1/v absorbers has been studied in the diffusion theory approximation using a multigroup (30-group) approach and the neutron scattering kernel proposed earlier by the authors. It is observed that, in this case as in the case of water investigated earlier, the behavior of neutrons in graphite poisoned with gadolinium is different from that in graphite poisoned with samarium or cadmium. To explain the reason for this difference, a hypothetical model for the energy variation of the absorption cross section has been constructed that closely resembles samarium in one limit and goes over to gadolinium in the other. The effect of varying the concentration of non-1/v absorbers on the flux of sub-Bragg and epicold neutrons has been studied for this model, and some interesting results are obtained.

  9. Effect of /sup 235/U concentration on fast neutron space and time eigenvalues and eigenfunctions in uranium

    SciTech Connect

    Mohan, R.; Ahmed, F.; Kothari, L.S.

    1985-05-01

    The results of a detailed study of a fast neutron diffusion length and pulsed problem in depleted and enriched subcritical uranium assemblies (0.2 to 4% /sup 235/U) are reported. The multigroup space- and time-dependent equations are solved using the eigenfunction expansion method. The effect of /sup 235/U concentration on space (diffusion length problem) and time (pulsed problem) eigenvalues and eigenfunctions, particularly on the ''discrete'' eigenvalue and eigenfunction, is discussed. The approach to equilibrium (both in space and in time) of fast neutrons changes with changing /sup 235/U concentration.

  10. Comparison of the 3-D Deterministic Neutron Transport Code Attila® To Measure Data, MCNP And MCNPX For The Advanced Test Reactor

    SciTech Connect

    D. Scott Lucas; D. S. Lucas

    2005-09-01

    An LDRD (Laboratory Directed Research and Development) project is underway at the Idaho National Laboratory (INL) to apply the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the development of Attila models for ATR, capabilities of Attila, the generation and use of different cross-section libraries, and comparisons to ATR data, MCNP, MCNPX and future applications.

  11. Neutron shielding analysis for remote handled transuranic waste containers in facility casks at the Waste Isolation Pilot Plant

    SciTech Connect

    Livingston, J.V.; Disney, R.K.

    1984-04-01

    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.

  12. Atmospheric neutrons

    NASA Technical Reports Server (NTRS)

    Korff, S. A.; Mendell, R. B.; Merker, M.; Light, E. S.; Verschell, H. J.; Sandie, W. S.

    1979-01-01

    Contributions to fast neutron measurements in the atmosphere are outlined. The results of a calculation to determine the production, distribution and final disappearance of atmospheric neutrons over the entire spectrum are presented. An attempt is made to answer questions that relate to processes such as neutron escape from the atmosphere and C-14 production. In addition, since variations of secondary neutrons can be related to variations in the primary radiation, comment on the modulation of both radiation components is made.

  13. Neutron dosimetry

    DOEpatents

    Quinby, Thomas C.

    1976-07-27

    A method of measuring neutron radiation within a nuclear reactor is provided. A sintered oxide wire is disposed within the reactor and exposed to neutron radiation. The induced radioactivity is measured to provide an indication of the neutron energy and flux within the reactor.

  14. Neutron guide

    DOEpatents

    Greene, Geoffrey L.

    1999-01-01

    A neutron guide in which lengths of cylindrical glass tubing have rectangular glass plates properly dimensioned to allow insertion into the cylindrical glass tubing so that a sealed geometrically precise polygonal cross-section is formed in the cylindrical glass tubing. The neutron guide provides easier alignment between adjacent sections than do the neutron guides of the prior art.

  15. Radiation Transport for Explosive Outflows: A Multigroup Hybrid Monte Carlo Method

    NASA Astrophysics Data System (ADS)

    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.

  16. Symmetry breaking in the opinion dynamics of a multi-group project organization

    NASA Astrophysics Data System (ADS)

    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.

  17. The Multigroup Ethnic Identity Measure-Revised: Measurement invariance across racial and ethnic groups

    PubMed Central

    Brown, Susan D.; Unger Hu, Kirsten A.; Mevi, Ashley A.; Hedderson, Monique M.; Shan, Jun; Quesenberry, Charles P.; Ferrara, Assiamira

    2014-01-01

    The Multigroup Ethnic Identity Measure-Revised (MEIM-R), a brief instrument assessing affiliation with one’s ethnic group, is a promising advance in the ethnic identity literature. However, equivalency of its measurement properties across specific racial and ethnic groups should be confirmed before using it in diverse samples. We examined a) the psychometric properties of the MEIM-R including factor structure, measurement invariance, and internal consistency reliability, and b) levels of and differences in ethnic identity across multiple racial and ethnic groups and subgroups. Asian (n = 630), Black/African American (n = 58), Hispanic (n = 240), multiethnic (n = 160), and White (n = 375) women completed the MEIM-R as part of the “Gestational diabetes’ Effect on Moms” diabetes prevention trial in the Kaiser Permanente Northern California health care setting (N = 1,463; M age 32.5 years, SD = 4.9). Multiple-groups confirmatory factor analyses provided provisional evidence of measurement invariance, i.e., an equal, correlated two-factor structure, equal factor loadings, and equal item intercepts across racial and ethnic groups. Latent factor means for the two MEIM-R subscales, exploration and commitment, differed across groups; effect sizes ranging from small to large generally supported the notion of ethnic identity as more salient among people of color. Pending replication, good psychometric properties in this large and diverse sample of women support the future use of the MEIM-R. Preliminary evidence of measurement invariance suggests that the MEIM-R could be used to measure and compare ethnic identity across multiple racial and ethnic groups. PMID:24188656

  18. The spectral green's function nodal method for multigroup slab-geometry fixed-source S{sub N} problems with anisotropic scattering

    SciTech Connect

    Menezes, W. A.; Filho, H. A.; Barros, R. C.

    2013-07-01

    A generalization of the spectral Green's function (SGF) method is developed for multigroup, fixed-source, slab-geometry discrete ordinates (S{sub N}) 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 S{sub N} 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, presented here is 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. (authors)

  19. Specifications for the development of BUGLE-93: An ENDF/B-VI multigroup cross section library for LWR shielding and pressure vessel dosimetry

    SciTech Connect

    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.

  20. Variational Determination of the Neutron Integral Transport Equation Eigenvalues Using Space Asymptotic Trial Functions

    NASA Astrophysics Data System (ADS)

    Colombo, V.; Ravetto, P.; Sumini, M.

    1988-08-01

    An approximate determination of the critical eigenvalue of the neutron transport equation in integral form, within both the one speed and energy multigroup models, for a homogeneous medium, is achieved by means of a variational technique. The space asymptotic solutions for both the direct and adjoint problems are used as trial functions. A variational procedure is also developed and numerically exploited within the Fourier transformed domain, where noticeable theoretical features can be demonstrated. It is evidenced that excellent results can be obtained with little computational effort, and a set of critical calculations in plane geometry is presented and discussed.

  1. Variational determination of the neutron integral transport equation eigenvalues using space asymptotic trial functions

    SciTech Connect

    Colombo, V.; Ravetto, P.; Sumini, M.

    1988-08-01

    An approximate determination of the critical eigenvalue of the neutron transport equation in integral form, within both the one speed and energy multigroup models, for a homogeneous medium, is achieved by means of a variational technique. The space asymptotic solutions for both the direct and adjoint problems are used as trial functions. A variational procedure is also developed and numerically exploited within the Fourier transformed domain, where noticeable theoretical features can be demonstrated. It is evidenced that excellent results can be obtained with little computational effort, and a set of critical calculations in plane geometry is presented and discussed. copyright 1988 Academic Press, Inc.

  2. Neutron detector

    DOEpatents

    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.

  3. Multi-group acculturation orientations in a changing context: Palestinian Christian Arab adolescents in Israel after the lost decade.

    PubMed

    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

  4. Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.

    Energy Science and Technology Software Center (ESTSC)

    1987-09-18

    Version 00 TRISTAN solves the three-dimensional, fixed-source, Boltzmann transport equation for neutrons or gamma rays in rectangular geometry. The code can solve an adjoint problem as well as a usual transport problem. TRISTAN is a suitable tool to analyze radiation shielding problems such as streaming and deep penetration problems.

  5. Micromegas neutron beam monitor neutronics.

    PubMed

    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

  6. Multi-Population Invariance with Dichotomous Measures: Combining Multi-Group and MIMIC Methodologies in Evaluating the General Aptitude Test in the Arabic Language

    ERIC Educational Resources Information Center

    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…

  7. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  8. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  9. Neutron source

    DOEpatents

    Cason, J.L. Jr.; Shaw, C.B.

    1975-10-21

    A neutron source which is particularly useful for neutron radiography consists of a vessel containing a moderating media of relatively low moderating ratio, a flux trap including a moderating media of relatively high moderating ratio at the center of the vessel, a shell of depleted uranium dioxide surrounding the moderating media of relatively high moderating ratio, a plurality of guide tubes each containing a movable source of neutrons surrounding the flux trap, a neutron shield surrounding one part of each guide tube, and at least one collimator extending from the flux trap to the exterior of the neutron source. The shell of depleted uranium dioxide has a window provided with depleted uranium dioxide shutters for each collimator. Reflectors are provided above and below the flux trap and on the guide tubes away from the flux trap.

  10. Neutron tubes

    DOEpatents

    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.

  11. A multigroup diffusion solver using pseudo transient continuation for a radiation-hydrodynamic code with patch-based AMR

    SciTech Connect

    Shestakov, Aleksei I. Offner, Stella S.R.

    2008-01-10

    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 Dirichlet 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 and demonstrates

  12. A Multigroup diffusion Solver Using Pseudo Transient Continuation for a Radiaiton-Hydrodynamic Code with Patch-Based AMR

    SciTech Connect

    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 and demonstrates

  13. A multigroup diffusion solver using pseudo transient continuation for a radiation-hydrodynamic code with patch-based AMR

    NASA Astrophysics Data System (ADS)

    Shestakov, Aleksei I.; Offner, Stella S. R.

    2008-01-01

    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 (Ψtc). We analyze the magnitude of the Ψtc 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 Dirichlet 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 Ψtc 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 and demonstrates the

  14. A Multigroup diffusion solver using pseudo transient continuation for a radiation-hydrodynamic code with patch-based AMR

    SciTech Connect

    Shestakov, A I; Offner, S R

    2006-09-21

    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 and demonstrates

  15. A New Code for Proto-neutron Star Evolution

    NASA Astrophysics Data System (ADS)

    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.

  16. A NEW CODE FOR PROTO-NEUTRON STAR EVOLUTION

    SciTech Connect

    Roberts, L. F.

    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.

  17. Simulation of Space Shuttle neutron measurements with FLUKA.

    PubMed

    Pinsky, L; Carminati, F; Ferrari, A

    2001-06-01

    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. PMID:11855415

  18. NEUTRON SOURCE

    DOEpatents

    Bernander, N.K. et al.

    1960-10-18

    An apparatus is described for producing neutrons through target bombardment with deuterons. Deuterium gas is ionized by electron bombardment and the deuteron ions are accelerated through a magnetic field to collimate them into a continuous high intensity beam. The ion beam is directed against a deuteron pervious metal target of substantially the same nnaterial throughout to embed the deuterous therein and react them to produce neutrons. A large quantity of neutrons is produced in this manner due to the increased energy and quantity of ions bombarding the target.

  19. Two-Dimensional x-y and r-z Geometry Multigroup Transport Code System for Large Toroidal Reactors.

    Energy Science and Technology Software Center (ESTSC)

    1980-06-16

    Version: 00 Although TRIDENT-CTR is a follow-on code to TRIDENT, it has incorporated several features that make it significantly different. It can handle a wide range of irregular geometric domains in both x-y and r-z geometries. However, it was principally designed to solve shielding and blanket problems for large toroidal reactors. TRIDENT-CTR is a two-dimensional, x-y and r-z geometry, multigroup, neutral particle transport code. The use of triangular finite elements gives it the geometric flexibilitymore » to cope with the nonorthogonal shapes of many toroidal designs. The code is capable of handling a wide variety of problems having irregular domains in both x-y and r-z geometries.« less

  20. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    SciTech Connect

    Díez, C.J.; Cabellos, O.; Martínez, J.S.

    2015-01-15

    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.

  1. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    NASA Astrophysics Data System (ADS)

    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.

  2. Thermal neutron detection system

    DOEpatents

    Peurrung, Anthony J.; Stromswold, David C.

    2000-01-01

    According to the present invention, a system for measuring a thermal neutron emission from a neutron source, has a reflector/moderator proximate the neutron source that reflects and moderates neutrons from the neutron source. The reflector/moderator further directs thermal neutrons toward an unmoderated thermal neutron detector.

  3. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  4. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  5. NEUTRON SOURCES

    DOEpatents

    Richmond, J.L.; Wells, C.E.

    1963-01-15

    A neutron source is obtained without employing any separate beryllia receptacle, as was formerly required. The new method is safer and faster, and affords a source with both improved yield and symmetry of neutron emission. A Be container is used to hold and react with Pu. This container has a thin isolating layer that does not obstruct the desired Pu--Be reaction and obviates procedures previously employed to disassemble and remove a beryllia receptacle. (AEC)

  6. Neutron range spectrometer

    DOEpatents

    Manglos, S.H.

    1988-03-10

    A neutron range spectrometer and method for determining the neutron energy spectrum of a neutron emitting source are disclosed. Neutrons from the source are colliminated along a collimation axis and a position sensitive neutron counter is disposed in the path of the collimated neutron beam. The counter determines positions along the collimation axis of interactions between the neutrons in the neutron beam and a neutron-absorbing material in the counter. From the interaction positions, a computer analyzes the data and determines the neutron energy spectrum of the neutron beam. The counter is preferably shielded and a suitable neutron-absorbing material is He-3. 1 fig.

  7. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  8. FOREWORD: Neutron metrology Neutron metrology

    NASA Astrophysics Data System (ADS)

    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

  9. FOREWORD: Neutron metrology Neutron metrology

    NASA Astrophysics Data System (ADS)

    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

  10. Neutronic calculations for the conversion to LEU of a research reactor core

    SciTech Connect

    Varvayanni, M.; Catsaros, N.; Stakakis, E.; Grigoriadis, D.

    2008-07-15

    For a five-year transitional period the Greek Research Reactor (GRR-1) was operating with a mixed core, containing both Low Enrichment (LEU) and High Enrichment (HEU) Uranium MTR- type fuel assemblies. The neutronic study of the GRR-1 conversion to LEU has been performed using a code system comprising the core-analysis code CITATION-LDI2 and the cell-calculation modules XSDRNPM and NITAWL-II of the SCALE code. A conceptual LEU core configuration was defined and analyzed with respect to the three dimensional multi-group neutron fluxes, the power distribution, the control-rod worth and the compliance with pre-defined Operation Limiting Conditions. Perturbation calculations and reactivity feedback computations were also carried out to provide input to a subsequent thermal-hydraulic study. (author)

  11. An Improved Multigroup Monte Carlo Criticality Code System for Cross Section Processing.

    Energy Science and Technology Software Center (ESTSC)

    1982-11-23

    Version 00 KENO-IV1 is an improvement and extension of KENO2 which was contributed by Oak Ridge National Laboratory and written for the IBM 360 computer. It is flexibly dimensioned, utilizes free-form input, and offers more geometry options than KENO. KENO-IV solves nuclear criticality eigenvalue problems. The results calculated by KENO-IV include k-effective, lifetime and generation time, energy-dependent leakages and absorptions, energy- and region-dependent fluxes and region-dependent fission densities. Criticality searches can be made on unitmore » dimensions or on the number of units in an array. KENO-IV/CRC has several added features which include a neutron balance edit, PICTURE3 routines to check the input geometry, and a random number sequencing subroutine written in Fortran for the Cray-1 computer.« less

  12. An Improved Multigroup Monte Carlo Criticality Code System for Cross Section Processing.

    SciTech Connect

    1982-11-23

    Version 00 KENO-IV1 is an improvement and extension of KENO2 which was contributed by Oak Ridge National Laboratory and written for the IBM 360 computer. It is flexibly dimensioned, utilizes free-form input, and offers more geometry options than KENO. KENO-IV solves nuclear criticality eigenvalue problems. The results calculated by KENO-IV include k-effective, lifetime and generation time, energy-dependent leakages and absorptions, energy- and region-dependent fluxes and region-dependent fission densities. Criticality searches can be made on unit dimensions or on the number of units in an array. KENO-IV/CRC has several added features which include a neutron balance edit, PICTURE3 routines to check the input geometry, and a random number sequencing subroutine written in Fortran for the Cray-1 computer.

  13. RADSAT Benchmarks for Prompt Gamma Neutron Activation Analysis Measurements

    SciTech Connect

    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.

  14. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  15. Neutron therapy of cancer

    NASA Technical Reports Server (NTRS)

    Frigerio, N. A.; Nellans, H. N.; Shaw, M. J.

    1969-01-01

    Reports relate applications of neutrons to the problem of cancer therapy. The biochemical and biophysical aspects of fast-neutron therapy, neutron-capture and neutron-conversion therapy with intermediate-range neutrons are presented. Also included is a computer program for neutron-gamma radiobiology.

  16. Calculation of the Local Neutronic Parameters for CANDU Fuel Bundles Using Transport Methods

    SciTech Connect

    Balaceanu, Victoria; Rizoiu, Andrei; Hristea, Viorel

    2006-07-01

    For a realistic neutronic evaluation of the CANDU reactor core it is important to accurately perform the local neutronic parameters (i.e. multigroup macroscopic cross sections for the core materials) calculation. This means using codes that allow a good geometric representation of the CANDU fuel bundle and then solving the transport equation. The paper reported here intends to study in detail the local behavior for two types of CANDU fuel, NU{sub 3}7 (Natural Uranium, 37 elements) and SEU{sub 4}3 (Slightly Enriched Uranium, 43 elements, with 1.1 wt% enrichment). The considered fuel types represent fresh and used bundles. The two types of CANDU super-cells are reference NU{sub 3}7, perturbed NU{sub 3}7, reference SEU{sub 4}3 and perturbed SEU{sub 4}3. The perturbed super-cells contain a Mechanical Control Absorber (a very strong reactivity device). For reaching the proposed objective a methodology is used based on WIMS and PIJXYZ codes. WIMS is a standard lattice-cell code, based on transport theory and it is used for producing fuel cell multigroup macroscopic cross sections. For obtaining the fine local neutronic parameters in the CANDU super-cells (k-eff values, local MCA reactivity worth, flux distributions and reaction rates), the PIJXYZ code is used. PIJXYZ is a 3D integral transport code using the first collision probability method and it has been developed for CANDU cell geometry. It is consistent with WIMS lattice-cell calculations and allows a good geometrical representation of the CANDU bundle in three dimensions. The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON code. This comparison shows a good agreement between these results. (authors)

  17. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  18. NEUTRON SOURCE

    DOEpatents

    Foster, J.S. Jr.

    1960-04-19

    A compact electronic device capable of providing short time high density outputs of neutrons is described. The device of the invention includes an evacuated vacuum housing adapted to be supplied with a deuterium, tritium, or other atmosphere and means for establishing an electrical discharge along a path through the gas. An energized solenoid is arranged to constrain the ionized gas (plasma) along the path. An anode bearing adsorbed or adherent target material is arranged to enclose the constrained plasma. To produce neutrons a high voltage is applied from appropriate supply means between the plasma and anode to accelerate ions from the plasma to impinge upcn the target material, e.g., comprising deuterium.

  19. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1960-09-27

    A unit assembly is described for a neutronic reactor comprising a tube and plurality of spaced parallel sandwiches in the tube extending lengthwise thereof, each sandwich including a middle plate having a central opening for plutonium and other openings for fertile material at opposite ends of the plate.

  20. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  1. Methods for absorbing neutrons

    DOEpatents

    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.

  2. Multigroup Confirmatory Factor Analysis of the Cognitive Dysfunction Questionnaire: instrument refinement and measurement invariance across age and sex.

    PubMed

    Vestergren, Peter; Rönnlund, Michael; Nyberg, Lars; Nilsson, Lars-Göran

    2012-10-01

    The study adopted Confirmatory Factor Analysis (CFA) to investigate the factorial structure and reduce the number of items of the Cognitive Dysfunction Questionnaire (CDQ). The analyses were based on data for a total of 1,115 participants from population based samples (mean age: 63.0 ± 14.5 years, range: 25-95) randomly split into a refinement (N = 569) and a cross-validation (N = 546) sample. Equivalence of the measurement and structural portions of the refined model was demonstrated across the refinement and cross-validation samples. Among competing models the best fitting and parsimonious model had a hierarchical factor structure with five first-order and one second-order general factor. For the final version of the CDQ, 20 items within five domains were selected (Procedural actions, Semantic word knowledge, Face recognition, Temporal orientation, and Spatial navigation). Internal consistency reliabilities were adequate for the total scale and for the subscales. Multigroup CFAs indicated measurement invariance across age and sex up to the scalar level. Finally, higher levels of cognitive dysfunction as reflected by CDQ scores were predicted by advancing age, fewer years of education, and with deficits in general cognitive functioning as reflected by scores on the Mini-Mental State Examination. In conclusion, the CDQ appears to be psychometrically sound and shows the expected relationships with variables known to be associated with cognitive dysfunction and dementia. Future studies should apply it among clinical groups to further test its usefulness. PMID:22962857

  3. Neutron reflecting supermirror structure

    DOEpatents

    Wood, James L.

    1992-01-01

    An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources.

  4. Neutron reflecting supermirror structure

    DOEpatents

    Wood, J.L.

    1992-12-01

    An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. 2 figs.

  5. Recent Advances in Neutron Physics

    ERIC Educational Resources Information Center

    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)

  6. Dosimetric modeling of the microselectron high-dose rate 192Ir source by the multigroup discrete ordinates method.

    PubMed

    Daskalov, G M; Baker, R S; Rogers, D W; Williamson, J F

    2000-10-01

    The DANTSYS multigroup discrete ordinates computer code is applied to quantitatively estimate the absorbed dose rate distributions in the vicinity of a microSelectron 192Ir high-dose-rate (HDR) source in two-dimensional cylindrical R-Z geometry. The source is modeled in a cylindrical water phantom of diameter 20 cm and height 20 cm. The results are also used for evaluation of the Task Group 43 (TG-43) dosimetric quantities. The DANTSYS accuracy is estimated by direct comparisons with corresponding Monte Carlo results. Our 210-group photon cross section library developed previously, together with angular quadratures consisting of 36 (S16) to 210 (S40) directions and associated weights per octant, are used in the DANTSYS simulations. Strong ray effects are observed but are significantly mitigated through the use of DANTSYS's stochastic ray-tracing first collision source algorithm. The DANTSYS simulations closely approximate Monte Carlo estimates of both direct dose calculations and TG-43 dosimetric quantities. The discrepancies with S20 angular quadrature (55 directions and weights per octant) or higher are shown to be less than +/- 5% (about 2.5 standard deviations of Monte Carlo calculations) everywhere except for limited regions along the Z axis of rotational symmetry, where technical limitations in the DANTSYS first collision source implementation makes adequate suppression of ray effects difficult to achieve. The efficiency of DANTSYS simulations is compared with that of the EGS4 Monte Carlo code. It is demonstrated that even with the 210-group cross section library, DANTSYS achieves two-fold efficiency gains using the the S20 quadrature set. The potential of discrete ordinates method for further efficiency improvements is also discussed. PMID:11099199

  7. High energy neutron radiography

    SciTech Connect

    Gavron, A.; Morley, K.; Morris, C.; Seestrom, S.; Ullmann, J.; Yates, G.; Zumbro, J.

    1996-06-01

    High-energy spallation neutron sources are now being considered in the US and elsewhere as a replacement for neutron beams produced by reactors. High-energy and high intensity neutron beams, produced by unmoderated spallation sources, open potential new vistas of neutron radiography. The authors discuss the basic advantages and disadvantages of high-energy neutron radiography, and consider some experimental results obtained at the Weapons Neutron Research (WNR) facility at Los Alamos.

  8. NEUTRON SOURCE

    DOEpatents

    Reardon, W.A.; Lennox, D.H.; Nobles, R.G.

    1959-01-13

    A neutron source of the antimony--beryllium type is presented. The source is comprised of a solid mass of beryllium having a cylindrical recess extending therein and a cylinder containing antimony-124 slidably disposed within the cylindrical recess. The antimony cylinder is encased in aluminum. A berylliunn plug is removably inserted in the open end of the cylindrical recess to completely enclose the antimony cylinder in bsryllium. The plug and antimony cylinder are each provided with a stud on their upper ends to facilitate handling remotely.

  9. Calculations of neutron spectra after neutron neutron scattering

    NASA Astrophysics Data System (ADS)

    Crawford, B. E.; Stephenson, S. L.; Howell, C. R.; Mitchell, G. E.; Tornow, W.; Furman, W. I.; Lychagin, E. V.; Muzichka, A. Yu; Nekhaev, G. V.; Strelkov, A. V.; Sharapov, E. I.; Shvetsov, V. N.

    2004-09-01

    A direct neutron-neutron scattering length, ann, measurement with the goal of 3% accuracy (0.5 fm) is under preparation at the aperiodic pulsed reactor YAGUAR. A direct measurement of ann will not only help resolve conflicting results of ann by indirect means, but also in comparison to the proton-proton scattering length, app, shed light on the charge-symmetry of the nuclear force. We discuss in detail the analysis of the nn-scattering data in terms of a simple analytical expression. We also discuss calibration measurements using the time-of-flight spectra of neutrons scattered on He and Ar gases and the neutron activation technique. In particular, we calculate the neutron velocity and time-of-flight spectra after scattering neutrons on neutrons and after scattering neutrons on He and Ar atoms for the proposed experimental geometry, using a realistic neutron flux spectrum—Maxwellian plus epithermal tail. The shape of the neutron spectrum after scattering is appreciably different from the initial spectrum, due to collisions between thermal-thermal and thermal-epithermal neutrons. At the same time, the integral over the Maxwellian part of the realistic scattering spectrum differs by only about 6 per cent from that of a pure Maxwellian nn-scattering spectrum.

  10. Development and Testing of the VITAMIN-B7/BUGLE-B7 Coupled Neutron-Gamma Multigroup Cross-Section Libraries

    SciTech Connect

    Risner, Joel M; Wiarda, Dorothea; Miller, Thomas Martin; Peplow, Douglas E.; Patton, Bruce W; Dunn, Michael E; Parks, Benjamin T

    2011-01-01

    The U.S. Nuclear Regulatory Commission s Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the Evaluated Nuclear Data File (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries was accomplished using diagnostic checks in AMPX, unit tests for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in LWR shielding applications, and meet the calculational uncertainty criterion in Regulatory Guide 1.190.

  11. Development and testing of the VITAMIN-B7/BUGLE-B7 coupled neutron-gamma multigroup cross-section libraries

    SciTech Connect

    Risner, J.M.; Wiarda, D.; Miller, T.M.; Peplow, D.E.; Patton, B.W.; Dunn, M.E.; Parks, B.T.

    2011-07-01

    The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the evaluated nuclear data file (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI.3 data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII.0. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries were accomplished using diagnostic checks in AMPX, 'unit tests' for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in RPV fluence calculations and meet the calculational uncertainty criterion in Regulatory Guide 1.190. (authors)

  12. Optimization of the Direct Discrete Method Using the Solution of the Adjoint Equation and its Application in the Multi-Group Neutron Diffusion Equation

    SciTech Connect

    Ayyoubzadeh, Seyed Mohsen; Vosoughi, Naser

    2011-09-14

    Obtaining the set of algebraic equations that directly correspond to a physical phenomenon has been viable in the recent direct discrete method (DDM). Although this method may find its roots in physical and geometrical considerations, there are still some degrees of freedom that one may suspect optimize-able. Here we have used the information embedded in the corresponding adjoint equation to form a local functional, which in turn by its minimization, yield suitable dual mesh positioning.

  13. Monte Carlo Multigroup Neutron Code System for the Solution of Criticality Problems. We recommend C00474/ALLCP/02 MORSE-CGA.

    Energy Science and Technology Software Center (ESTSC)

    1983-04-13

    Version 00 MORSE-C is based on the original ORNL versions of CCC-127/MORSE and CCC-261/MORSE-L but is restricted to criticality problems. Continued efforts in criticality safety calculations led to the development of techniques which resulted in improvements in energy resolution of cross sections, upscatter in the thermal region, and a better cross section library. Only time-independent problems are treated in the packaged version.

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

    SciTech Connect

    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.

  15. Neutron standard data

    SciTech Connect

    Peelle, R.; Conde, H.

    1988-01-01

    The neutron standards are reviewed with emphasis on the evaluation for ENDFB-VI. Also discussed are the neutron spectrum of /sup 252/Cf 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.

  16. Borner Ball Neutron Detector

    NASA Technical Reports Server (NTRS)

    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.

  17. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  18. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  19. Direct Measurement of Neutron-Neutron Scattering

    SciTech Connect

    Sharapov, E.I.; Furman, W.I.; Lychagin, W.I.; Muzichka, G.V.; Nekhaev, G.V.; Safronov, Yu.V.; Shvetsov, V.N.; Strelkov, A.V.; Bowman, C.D.; Crawford, B.E.; Stephenson, S.L.; Howell, C.R.; Tornow, W.; Levakov, B.G.; Litvin, V.I.; Lyzhin, A.E.; Magda, E.P.; Mitchell, G.E.

    2003-08-26

    In order to resolve long-standing discrepancies in indirect measurements of the neutron-neutron scattering length ann and contribute to solving the problem of the charge symmetry of the nuclear force, the collaboration DIANNA (Direct Investigation of ann Association) plans to measure the neutron-neutron scattering cross section {sigma}nn. The key issue of our approach is the use of the through-channel in the Russia reactor YAGUAR with a peak neutron flux of 10{sup 18} /cm2/s. The proposed experimental setup is described. Results of calculations are presented to connect {sigma}nn with the nn-collision detector count rate and the neutron flux density in the reactor channel. Measurements of the thermal neutron fields inside polyethylene converters show excellent prospects for the realization of the direct nn-experiment.

  20. Neutron reflecting supermirror structure

    DOEpatents

    Wood, James L.

    1992-01-01

    An improved neutron reflecting supermirror structure comprising a plurality of stacked sets of bilayers of neutron reflecting materials. The improved neutron reflecting supermirror structure is adapted to provide extremely good performance at high incidence angles, i.e. up to four time the critical angle of standard neutron mirror structures. The reflection of neutrons striking the supermirror structure at a high critical angle provides enhanced neutron throughput, and hence more efficient and economical use of neutron sources. One layer of each set of bilayers consist of titanium, and the second layer of each set of bilayers consist of an alloy of nickel with carbon interstitially present in the nickel alloy.

  1. A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

    SciTech Connect

    Lee, S.R.; Cummings, J.C.; Nolen, S.D. |

    1997-05-01

    We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and {alpha}-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 {alpha}-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed.

  2. Neutrons in cancer therapy

    NASA Astrophysics Data System (ADS)

    Allen, Barry J.

    1995-03-01

    The role of neutrons in the management of cancer has a long history. However, it is only in recent years that neutrons are beginning to find an accepted place as an efficacious radiation modality. Fast neutron therapy is already well established for the treatment of certain cancers, and clinical trials are ongoing. Californium neutron sources are being used in brachytherapy. Boron neutron capture therapy has been well tested with thermal neutrons and epithermal neutron dose escalation studies are about to commence in the USA and Europe. Possibilities of neutron induced auger electron therapy are also discussed. With respect to chemotherapy, prompt neutron capture analysis is being used to study the dose optimization of chemotherapy in the management of breast cancer. The rationales behind these applications of neutrons in the management of cancer are examined.

  3. Neutronics design

    SciTech Connect

    Moir, R.

    1984-10-01

    Initial scoping calculations were done by Lee at LLNL with the TART code and ENDL data to determine the tritium breeding potential of this blanket type. A radially zoned cylindrical nucleonics model was used and is described. Results, local (100% blanket coverage) T and M vs Be zone thickness, are shown. The tritium breeding ratio, T, is seen to vary between 0.5 with no Be to 1.7 with a 60-cm Be zone. Correspondingly, energy multiplication, M, varies between 1.1 and 1.4. The effects of less than 100% blanket coverage on T is shown. For example, if the effective coverage is only 80, a 15-cm Be zone is needed for T = 1.01 compared to 10 cm at full coverage. Higher T can be achieved, of course, by increasing the Be zone thickness. Another possibly attractive use of the excess neutrons generated in Be is for higher M. While this was not the objective here it is clearly possible to include material in the blanket with significantly higher Q's than 4.8 MeV for the Li6(n,t) reaction. Also enriching the Li in Li6 can increase T.

  4. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  5. Neutron range spectrometer

    DOEpatents

    Manglos, Stephen H.

    1989-06-06

    A neutron range spectrometer and method for determining the neutron energy spectrum of a neutron emitting source are disclosed. Neutrons from the source are collimnated along a collimation axis and a position sensitive neutron counter is disposed in the path of the collimated neutron beam. The counter determines positions along the collimation axis of interactions between the neutrons in the neutron beam and a neutron-absorbing material in the counter. From the interaction positions, a computer analyzes the data and determines the neutron energy spectrum of the neutron beam. The counter is preferably shielded and a suitable neutron-absorbing material is He-3. The computer solves the following equation in the analysis: ##EQU1## where: N(x).DELTA.x=the number of neutron interactions measured between a position x and x+.DELTA.x, A.sub.i (E.sub.i).DELTA.E.sub.i =the number of incident neutrons with energy between E.sub.i and E.sub.i +.DELTA.E.sub.i, and C=C(E.sub.i)=N .sigma.(E.sub.i) where N=the number density of absorbing atoms in the position sensitive counter means and .sigma. (E.sub.i)=the average cross section of the absorbing interaction between E.sub.i and E.sub.i +.DELTA.E.sub.i.

  6. Organic metal neutron detector

    DOEpatents

    Butler, M.A.; Ginley, D.S.

    1984-11-21

    A device for detection of neutrons comprises: as an active neutron sensing element, a conductive organic polymer having an electrical conductivity and a cross-section for said neutrons whereby a detectable change in said conductivity is caused by impingement of said neutrons on the conductive organic polymer which is responsive to a property of said polymer which is altered by impingement of said neutrons on the polymer; and means for associating a change in said alterable property with the presence of neutrons at the location of said device.

  7. Neutron streak camera

    DOEpatents

    Wang, C.L.

    1981-05-14

    Apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon.

  8. Neutron streak camera

    DOEpatents

    Wang, C.L.

    1983-09-13

    Disclosed is an apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon. 4 figs.

  9. Neutron streak camera

    DOEpatents

    Wang, Ching L.

    1983-09-13

    Apparatus for improved sensitivity and time resolution of a neutron measurement. The detector is provided with an electrode assembly having a neutron sensitive cathode which emits relatively low energy secondary electrons. The neutron sensitive cathode has a large surface area which provides increased sensitivity by intercepting a greater number of neutrons. The cathode is also curved to compensate for differences in transit time of the neutrons emanating from the point source. The slower speeds of the secondary electrons emitted from a certain portion of the cathode are matched to the transit times of the neutrons impinging thereupon.

  10. Layered semiconductor neutron detectors

    DOEpatents

    Mao, Samuel S; Perry, Dale L

    2013-12-10

    Room temperature operating solid state hand held neutron detectors integrate one or more relatively thin layers of a high neutron interaction cross-section element or materials with semiconductor detectors. The high neutron interaction cross-section element (e.g., Gd, B or Li) or materials comprising at least one high neutron interaction cross-section element can be in the form of unstructured layers or micro- or nano-structured arrays. Such architecture provides high efficiency neutron detector devices by capturing substantially more carriers produced from high energy .alpha.-particles or .gamma.-photons generated by neutron interaction.

  11. Time dependent Monte Carlo calculations of the ORELA (Oak Ridge Electron Linear Accelerator) target neutron spectrum

    SciTech Connect

    Cramer, S.N.; Perey, F.G.

    1990-01-01

    The time dependent spectrum of neutrons in the water-moderated Oak Ridge Electron Linear Accelerator (ORELA) target has been calculated using a modified version of the MORSE multi-group Monte Carlo code with an analytic hydrogen scattering model. Distributions of effective neutron distance traversed in the target are estimated with a time and energy dependent algorithm from the leakage normal to the target face. These data are used in the resonance shaped analyses of time-of-flight cross section measurements to account for the experimental resolution function. The 20 MeV--10 eV energy range is adequately represented in the MORSE code by the 174 group VITAMIN-E cross section library with a P{sub 5} expansion. An approximate representation of the ORELA positron source facility, recently installed near the target, has been included in the calculations to determine any perturbations the positron source might create in the computed neutron distributions from the target. A series of coupled Monte Carlo calculations was performed from the target to the positron source and back to the target using a next-event estimation surface source for each step. The principal effect of the positron source was found to be an increase in the distance for the lower energy neutron spectra, producing no real change in the distributions where the ORELA source is utilized for experiments. Different configurations for the target were investigated in order to simulate the placement of a shadow bar in the neutron beam. These beam configurations included neutrons escaping from : (1) the central tantalum plates only, (2) the entire target with the tantalum plates blocked out, and (3) only a small area from the water. Comparisons of the current data with previous calculations having a less detailed model of the tantalum plates have been satisfactory. 10 refs.

  12. Neutronic Reactor Design to Reduce Neutron Loss

    DOEpatents

    Miles, F. T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall. The wall is surrounded by successive layers of pure fertile material and moderator containing fertile material. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. Since the steel has a smaller capture cross section for the fast neutrons, greater nunnbers of neutrons will pass into the blanket, thereby increasing the over-all efficiency of the reactor. (AEC)

  13. NEUTRONIC REACTOR DESIGN TO REDUCE NEUTRON LOSS

    DOEpatents

    Mills, F.T.

    1961-05-01

    A nuclear reactor construction is described in which an unmoderated layer of the fissionable material is inserted between the moderated portion of the reactor core and the core container steel wall which is surrounded by successive layers of pure fertile material and fertile material having moderator. The unmoderated layer of the fissionable material will insure that a greater portion of fast neutrons will pass through the steel wall than would thermal neutrons. As the steel has a smaller capture cross-section for the fast neutrons, then greater numbers of the neutrons will pass into the blanket thereby increasing the over-all efficiency of the reactor.

  14. Neutron anatomy

    SciTech Connect

    Bacon, G.E.

    1994-12-31

    The familiar extremes of crystalline material are single-crystals and random powders. In between these two extremes are polycrystalline aggregates, not randomly arranged but possessing some preferred orientation and this is the form taken by constructional materials, be they steel girders or the bones of a human or animal skeleton. The details of the preferred orientation determine the ability of the material to withstand stress in any direction. In the case of bone the crucial factor is the orientation of the c-axes of the mineral content - the crystals of the hexagonal hydroxyapatite - and this can readily be determined by neutron diffraction. In particular it can be measured over the volume of a piece of bone, utilizing distances ranging from 1mm to 10mm. The major practical problem is to avoid the intense incoherent scattering from the hydrogen in the accompanying collagen; this can best be achieved by heat-treatment and it is demonstrated that this does not affect the underlying apatite. These studies of bone give leading anatomical information on the life and activities of humans and animals - including, for example, the life history of the human femur, the locomotion of sheep, the fracture of the legs of racehorses and the life-styles of Neolithic tribes. We conclude that the material is placed economically in the bone to withstand the expected stresses of life and the environment. The experimental results are presented in terms of the magnitude of the 0002 apatite reflection. It so happens that for a random powder the 0002, 1121 reflections, which are neighboring lines in the powder pattern, are approximately equal in intensity. The latter reflection, being of manifold multiplicity, is scarcely affected by preferred orientation so that the numerical value of the 0002/1121 ratio serves quite accurately as a quantitative measure of the degree of orientation of the c-axes in any chosen direction for a sample of bone.

  15. Neutron noise calculations in a hexagonal geometry and comparison with analytical solutions

    SciTech Connect

    Tran, H. N.; Demaziere, C.

    2012-07-01

    This paper presents the development of a neutronic and kinetic solver for hexagonal geometries. The tool is developed based on the diffusion theory with multi-energy groups and multi-groups of delayed neutron precursors allowing the solutions of forward and adjoint problems of static and dynamic states, and is applicable to both thermal and fast systems with hexagonal geometries. In the dynamic problems, the small stationary fluctuations of macroscopic cross sections are considered as noise sources, and then the induced first order noise is calculated fully in the frequency domain. Numerical algorithms for solving the static and noise equations are implemented with a spatial discretization based on finite differences and a power iterative solution. A coarse mesh finite difference method has been adopted for speeding up the convergence. Since no other numerical tool could calculate frequency-dependent noise in hexagonal geometry, validation calculations have been performed and benchmarked to analytical solutions based on a 2-D homogeneous system with two-energy groups and one-group of delayed neutron precursor, in which point-like perturbations of thermal absorption cross section at central and non-central positions are considered as noise sources. (authors)

  16. Ultrafast neutron detector

    DOEpatents

    Wang, C.L.

    1985-06-19

    A neutron detector of very high temporal resolution is described. It may be used to measure distributions of neutrons produced by fusion reactions that persist for times as short as about 50 picoseconds.

  17. Dose equivalent neutron dosimeter

    DOEpatents

    Griffith, Richard V.; Hankins, Dale E.; Tomasino, Luigi; Gomaa, Mohamed A. M.

    1983-01-01

    A neutron dosimeter is disclosed which provides a single measurements indicating the amount of potential biological damage resulting from the neutron exposure of the wearer, for a wide range of neutron energies. The dosimeter includes a detecting sheet of track etch detecting material such as a carbonate plastic, for detecting higher energy neutrons, and a radiator layer containing conversion material such as .sup.6 Li and .sup.10 B lying adjacent to the detecting sheet for converting moderate energy neutrons to alpha particles that produce tracks in the adjacent detecting sheet. The density of conversion material in the radiator layer is of an amount which is chosen so that the density of tracks produced in the detecting sheet is proportional to the biological damage done by neutrons, regardless of whether the tracks are produced as the result of moderate energy neutrons striking the radiator layer or as the result of higher energy neutrons striking the sheet of track etch material.

  18. Neutron dose equivalent meter

    DOEpatents

    Olsher, Richard H.; Hsu, Hsiao-Hua; Casson, William H.; Vasilik, Dennis G.; Kleck, Jeffrey H.; Beverding, Anthony

    1996-01-01

    A neutron dose equivalent detector for measuring neutron dose capable of accurately responding to neutron energies according to published fluence to dose curves. The neutron dose equivalent meter has an inner sphere of polyethylene, with a middle shell overlying the inner sphere, the middle shell comprising RTV.RTM. silicone (organosiloxane) loaded with boron. An outer shell overlies the middle shell and comprises polyethylene loaded with tungsten. The neutron dose equivalent meter defines a channel through the outer shell, the middle shell, and the inner sphere for accepting a neutron counter tube. The outer shell is loaded with tungsten to provide neutron generation, increasing the neutron dose equivalent meter's response sensitivity above 8 MeV.

  19. ULTRASONIC NEUTRON DOSIMETER

    DOEpatents

    Truell, R.; de Klerk, J.; Levy, P.W.

    1960-02-23

    A neutron dosimeter is described which utilizes ultrasonic waves in the megacycle region for determination of the extent of neutron damage in a borosilicate glass through ultrasonic wave velocity and attenuation measurements before and after damage.

  20. On neutron surface waves

    SciTech Connect

    Ignatovich, V. K.

    2009-01-15

    It is shown that neutron surface waves do not exist. The difference between the neutron wave mechanics and the wave physics of electromagnetic and acoustic processes, which allows the existence of surface waves, is analyzed.

  1. Tungsten thermal neutron dosimeter

    NASA Technical Reports Server (NTRS)

    Ball, L. L.; Richardson, P. J.; Sheibley, D. W.

    1969-01-01

    Tungsten-185 activity, which is produced by neutron activation of tungsten-184, determines thermal neutron flux. Radiochemical separation methods and counting techniques for irradiated tungsten provide accurate determination of the radiation exposure.

  2. Pulsed-neutron monochromator

    DOEpatents

    Mook, H.A. Jr.

    1984-01-01

    In one aspect, the invention is an improved pulsed-neutron monochromator of the vibrated-crystal type. The monochromator is designed to provide neutron pulses which are characterized both by short duration and high density. A row of neutron-reflecting crystals is disposed in a neutron beam to reflect neutrons onto a common target. The crystals in the row define progressively larger neutron-scattering angles and are vibrated sequentially in descending order with respect to the size of their scattering angles, thus generating neutron pulses which arrive simultaneously at the target. Transducers are coupled to one end of the crystals to vibrate them in an essentially non-resonant mode. The transducers propagate transverse waves in the crystal which progress longitudinally therein. The waves are absorbed at the undriven ends of the crystals by damping material mounted thereon. In another aspect, the invention is a method for generating neutron pulses characterized by high intensity and short duration.

  3. Pulsed-neutron monochromator

    DOEpatents

    Mook, Jr., Herbert A.

    1985-01-01

    In one aspect, the invention is an improved pulsed-neutron monochromator of the vibrated-crystal type. The monochromator is designed to provide neutron pulses which are characterized both by short duration and high density. A row of neutron-reflecting crystals is disposed in a neutron beam to reflect neutrons onto a common target. The crystals in the row define progressively larger neutron-scattering angles and are vibrated sequentially in descending order with respect to the size of their scattering angles, thus generating neutron pulses which arrive simultaneously at the target. Transducers are coupled to one end of the crystals to vibrate them in an essentially non-resonant mode. The transducers propagate transverse waves in the crystal which progress longitudinally therein. The wave are absorbed at the undriven ends of the crystals by damping material mounted thereon. In another aspect, the invention is a method for generating neutron pulses characterized by high intensity and short duration.

  4. Intense fusion neutron sources

    NASA Astrophysics Data System (ADS)

    Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.

    2010-04-01

    The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 1015-1021 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 1020 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

  5. Dibaryons in neutron stars

    NASA Technical Reports Server (NTRS)

    Olinto, Angela V.; Haensel, Pawel; Frieman, Joshua A.

    1991-01-01

    The effects are studied of H-dibaryons on the structure of neutron stars. It was found that H particles could be present in neutron stars for a wide range of dibaryon masses. The appearance of dibaryons softens the equations of state, lowers the maximum neutron star mass, and affects the transport properties of dense matter. The parameter space is constrained for dibaryons by requiring that a 1.44 solar mass neutron star be gravitationally stable.

  6. 500 MHz neutron detector

    SciTech Connect

    Yen, Yi-Fen; Bowman, J.D.; Matsuda, Y.

    1993-12-01

    A {sup 10}B-loaded scintillation detector was built for neutron transmission measurements at the Los Alamos Neutron Scattering Center. The efficiency of the detector is nearly 100% for neutron energies from 0 to 1 keV. The neutron moderation time in the scintillator is about 250 ns and is energy independent. The detector and data processing system are designed to handle an instantaneous rate as high as 500 MHz. The active area of the detector is 40 cm in diameter.

  7. Using Multigroup-Multiphase Latent State-Trait Models to Study Treatment-Induced Changes in Intra-Individual State Variability: An Application to Smokers' Affect

    PubMed Central

    Geiser, Christian; Griffin, Daniel; Shiffman, Saul

    2016-01-01

    Sometimes, researchers are interested in whether an intervention, experimental manipulation, or other treatment causes changes in intra-individual state variability. The authors show how multigroup-multiphase latent state-trait (MG-MP-LST) models can be used to examine treatment effects with regard to both mean differences and differences in state variability. The approach is illustrated based on a randomized controlled trial in which N = 338 smokers were randomly assigned to nicotine replacement therapy (NRT) vs. placebo prior to quitting smoking. We found that post quitting, smokers in both the NRT and placebo group had significantly reduced intra-individual affect state variability with respect to the affect items calm and content relative to the pre-quitting phase. This reduction in state variability did not differ between the NRT and placebo groups, indicating that quitting smoking may lead to a stabilization of individuals' affect states regardless of whether or not individuals receive NRT. PMID:27499744

  8. Examining the psychometric validity of the Multigroup Ethnic Identity Measure-Revised (MEIM-R) in a community sample of African American and European American adults.

    PubMed

    Chakawa, Ayanda; Butler, Robert C; Shapiro, Steven K

    2015-10-01

    The purpose of this study was to examine the psychometric properties of the Multigroup Ethnic Identity Measure-Revised (MEIM-R), focusing on a sample drawn from a geographic region in the United States that has not been included in previously published research on the MEIM-R. Data were obtained from a community-based sample of 105 African American (AA) and 91 European American (EA) adults located in the state of Alabama. The MEIM-R was best represented by two constructs-exploration and commitment. AA adults reported higher levels of racial/ethnic identity exploration and commitment than EA adults. Differential item functioning was found among 1 of the exploration items. The current study provides additional support for the structural validity of the MEIM-R. Further research on the invariance of responses to the MEIM-R across a variety of sociodemographic factors is still necessary. PMID:25642783

  9. Neutron removal cross section as a measure of neutron skin

    SciTech Connect

    Fang, D. Q.; Ma, Y. G.; Cai, X. Z.; Tian, W. D.; Wang, H. W.

    2010-04-15

    We study the relation between neutron removal cross section (sigma{sub -N}) and neutron skin thickness for finite neutron-rich nuclei using the statistical abrasion ablation model. Different sizes of neutron skin are obtained by adjusting the diffuseness parameter of neutrons in the Fermi distribution. It is demonstrated that there is a good linear correlation between sigma{sub -N} and the neutron skin thickness for neutron-rich nuclei. Further analysis suggests that the relative increase of neutron removal cross section could be used as a quantitative measure for neutron skin thickness in neutron-rich nuclei.

  10. Arsenic activation neutron detector

    DOEpatents

    Jacobs, Eddy L.

    1981-01-01

    A detector of bursts of neutrons from a deuterium-deuteron reaction includes a quantity of arsenic adjacent a gamma detector such as a scintillator and photomultiplier tube. The arsenic is activated by the 2.5 Mev neutrons to release gamma radiation which is detected to give a quantitative representation of detected neutrons.

  11. Arsenic activation neutron detector

    DOEpatents

    Jacobs, E.L.

    1980-01-28

    A detector of bursts of neutrons from a deuterium-deuteron reaction includes a quantity of arsenic adjacent a gamma detector such as a scintillator and photomultiplier tube. The arsenic is activated by the 2.5-MeV neutrons to release gamma radiation which is detected to give a quantitative representation of detected neutrons.

  12. Advanced neutron absorber materials

    DOEpatents

    Branagan, Daniel J.; Smolik, Galen R.

    2000-01-01

    A neutron absorbing material and method utilizing rare earth elements such as gadolinium, europium and samarium to form metallic glasses and/or noble base nano/microcrystalline materials, the neutron absorbing material having a combination of superior neutron capture cross sections coupled with enhanced resistance to corrosion, oxidation and leaching.

  13. Prototype Neutron Energy Spectrometer

    SciTech Connect

    Stephen Mitchell, Sanjoy Mukhopadhyay, Richard Maurer, Ronald Wolff

    2010-06-16

    The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production [ship effect], [a, n] reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.

  14. Comparison of a 3-D multi-group SN particle transport code with Monte Carlo for intracavitary brachytherapy of the cervix uteri.

    PubMed

    Gifford, Kent A; Wareing, Todd A; Failla, Gregory; Horton, John L; Eifel, Patricia J; Mourtada, Firas

    2010-01-01

    A patient dose distribution was calculated by a 3D multi-group S N particle transport code for intracavitary brachytherapy of the cervix uteri and compared to previously published Monte Carlo results. A Cs-137 LDR intracavitary brachytherapy CT data set was chosen from our clinical database. MCNPX version 2.5.c, was used to calculate the dose distribution. A 3D multi-group S N particle transport code, Attila version 6.1.1 was used to simulate the same patient. Each patient applicator was built in SolidWorks, a mechanical design package, and then assembled with a coordinate transformation and rotation for the patient. The SolidWorks exported applicator geometry was imported into Attila for calculation. Dose matrices were overlaid on the patient CT data set. Dose volume histograms and point doses were compared. The MCNPX calculation required 14.8 hours, whereas the Attila calculation required 22.2 minutes on a 1.8 GHz AMD Opteron CPU. Agreement between Attila and MCNPX dose calculations at the ICRU 38 points was within +/- 3%. Calculated doses to the 2 cc and 5 cc volumes of highest dose differed by not more than +/- 1.1% between the two codes. Dose and DVH overlays agreed well qualitatively. Attila can calculate dose accurately and efficiently for this Cs-137 CT-based patient geometry. Our data showed that a three-group cross-section set is adequate for Cs-137 computations. Future work is aimed at implementing an optimized version of Attila for radiotherapy calculations. PMID:20160682

  15. Organic metal neutron detector

    DOEpatents

    Butler, Michael A.; Ginley, David S.

    1987-01-01

    A device for detecting neutrons comprises a layer of conductive polymer sandwiched between electrodes, which may be covered on each face with a neutron transmissive insulating material layer. Conventional electrodes are used for a non-imaging integrating total neutron fluence-measuring embodiment, while wire grids are used in an imaging version of the device. The change in conductivity of the polymer after exposure to a neutron flux is determined in either case to provide the desired data. Alternatively, the exposed conductive polymer layer may be treated with a chemical reagent which selectively binds to the sites altered by neutrons to produce an image of the flux detected.

  16. Neutron activation analysis system

    DOEpatents

    Taylor, M.C.; Rhodes, J.R.

    1973-12-25

    A neutron activation analysis system for monitoring a generally fluid media, such as slurries, solutions, and fluidized powders, including two separate conduit loops for circulating fluid samples within the range of radiation sources and detectors is described. Associated with the first loop is a neutron source that emits s high flux of slow and thermal neutrons. The second loop employs a fast neutron source, the flux from which is substantially free of thermal neutrons. Adjacent to both loops are gamma counters for spectrographic determination of the fluid constituents. Other gsmma sources and detectors are arranged across a portion of each loop for deterMining the fluid density. (Official Gazette)

  17. Grazing incidence neutron optics

    NASA Technical Reports Server (NTRS)

    Gubarev, Mikhail V. (Inventor); Ramsey, Brian D. (Inventor); Engelhaupt, Darell E. (Inventor)

    2012-01-01

    Neutron optics based on the two-reflection geometries are capable of controlling beams of long wavelength neutrons with low angular divergence. The preferred mirror fabrication technique is a replication process with electroform nickel replication process being preferable. In the preliminary demonstration test an electroform nickel optics gave the neutron current density gain at the focal spot of the mirror at least 8 for neutron wavelengths in the range from 6 to 20 .ANG.. The replication techniques can be also be used to fabricate neutron beam controlling guides.

  18. Grazing Incidence Neutron Optics

    NASA Technical Reports Server (NTRS)

    Gubarev, Mikhail V. (Inventor); Ramsey, Brian D. (Inventor); Engelhaupt, Darell E. (Inventor)

    2013-01-01

    Neutron optics based on the two-reflection geometries are capable of controlling beams of long wavelength neutrons with low angular divergence. The preferred mirror fabrication technique is a replication process with electroform nickel replication process being preferable. In the preliminary demonstration test an electroform nickel optics gave the neutron current density gain at the focal spot of the mirror at least 8 for neutron wavelengths in the range from 6 to 20.ANG.. The replication techniques can be also be used to fabricate neutron beam controlling guides.

  19. PERSONNEL NEUTRON DOSIMETER

    DOEpatents

    Fitzgerald, J.J.; Detwiler, C.G. Jr.

    1960-05-24

    A description is given of a personnel neutron dosimeter capable of indicating the complete spectrum of the neutron dose received as well as the dose for each neutron energy range therein. The device consists of three sets of indium foils supported in an aluminum case. The first set consists of three foils of indium, the second set consists of a similar set of indium foils sandwiched between layers of cadmium, whereas the third set is similar to the second set but is sandwiched between layers of polyethylene. By analysis of all the foils the neutron spectrum and the total dose from neutrons of all energy levels can be ascertained.

  20. High energy neutron dosimeter

    DOEpatents

    Rai, K.S.F.

    1994-01-11

    A device for measuring dose equivalents in neutron radiation fields is described. The device includes nested symmetrical hemispheres (forming spheres) of different neutron moderating materials that allow the measurement of dose equivalents from 0.025 eV to past 1 GeV. The layers of moderating material surround a spherical neutron counter. The neutron counter is connected by an electrical cable to an electrical sensing means which interprets the signal from the neutron counter in the center of the moderating spheres. The spherical shape of the device allows for accurate measurement of dose equivalents regardless of its positioning. 2 figures.

  1. High energy neutron dosimeter

    DOEpatents

    Sun, Rai Ko S.F.

    1994-01-01

    A device for measuring dose equivalents in neutron radiation fields. The device includes nested symmetrical hemispheres (forming spheres) of different neutron moderating materials that allow the measurement of dose equivalents from 0.025 eV to past 1 GeV. The layers of moderating material surround a spherical neutron counter. The neutron counter is connected by an electrical cable to an electrical sensing means which interprets the signal from the neutron counter in the center of the moderating spheres. The spherical shape of the device allows for accurate measurement of dose equivalents regardless of its positioning.

  2. Neutron scatter camera

    DOEpatents

    Mascarenhas, Nicholas; Marleau, Peter; Brennan, James S.; Krenz, Kevin D.

    2010-06-22

    An instrument that will directly image the fast fission neutrons from a special nuclear material source has been described. This instrument can improve the signal to background compared to non imaging neutron detection techniques by a factor given by ratio of the angular resolution window to 4.pi.. In addition to being a neutron imager, this instrument will also be an excellent neutron spectrometer, and will be able to differentiate between different types of neutron sources (e.g. fission, alpha-n, cosmic ray, and D-D or D-T fusion). Moreover, the instrument is able to pinpoint the source location.

  3. Precision Neutron Polarimetry for Neutron Beta Decay

    PubMed Central

    Penttila, S. I.; Bowman, J. D.

    2005-01-01

    The abBA collaboration is developing a new type of field-expansion spectrometer for a measurement of the three correlation coefficients a, A, and B and the shape parameter b. The measurement of A and B requires precision neutron polarimetry. We will polarize a pulsed cold neutron beam from the SNS using a 3He neutron spin filter. The well-known polarizing cross section for n-3He has a 1/v dependence, where v is the neutron velocity, which is used to determine the absolute beam polarization through a time-of-flight (TOF) measurement. We show that by measuring the TOF dependence of A and B, the coefficients and the neutron polarization can be determined with a small loss of the statistical precision and with negligible systematic error. We conclude that it is possible to determine the neutron polarization averaged over a long run in the neutron beta decay experiment with a statistical error less than 10−4. We discuss various sources of systematic uncertainty in the measurement of A and B and conclude that the fractional systematic errors are less than 2 × 10−4. PMID:27308142

  4. Improvements and New Findings in Monte Carlo Method with Complex-valued Weights for Neutron Leakage-corrected Assembly Calculations

    NASA Astrophysics Data System (ADS)

    Yamamoto, Toshihiro

    2014-06-01

    The author of this paper recently proposed a Monte Carlo calculation algorithm to solve a complex transport equation with complex-valued weights. The algorithm enables one to generate neutron leakage-corrected group constants and anisotropic diffusion coefficients for a unit fuel pin cell or assembly. The group constants are subsequently used for multi-group deterministic core calculations. The technique, however, had some limitations in applying itself to general problems. Some improvements have been done in this paper. The reflective boundary condition has newly become available. It has been found that a cumbersome weight cancellation of fission sources with positive and negative weights can be omitted in general fuel assembly geometries. A homogenization method of diffusion coefficients for a fuel assembly has been proposed.

  5. NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

    DOEpatents

    Young, G.J.

    1959-06-30

    The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

  6. Neutron beam design, development, and performance for neutron capture therapy

    SciTech Connect

    Harling, O.K.; Bernard, J.A. ); Zamenhof, R.G. )

    1990-01-01

    The report presents topics presented at a workshop on neutron beams and neutron capture therapy. Topics include: neutron beam design; reactor-based neutron beams; accelerator-based neutron beams; and dosimetry and treatment planning. Individual projects are processed separately for the databases. (CBS)

  7. Parallel computation safety analysis irradiation targets fission product molybdenum in neutronic aspect using the successive over-relaxation algorithm

    NASA Astrophysics Data System (ADS)

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-01

    One of the research activities in support of commercial radioisotope production program is a safety research on target FPM (Fission Product Molybdenum) irradiation. FPM targets form a tube made of stainless steel which contains nuclear-grade high-enrichment uranium. The FPM irradiation tube is intended to obtain fission products. Fission materials such as Mo99 used widely the form of kits in the medical world. The neutronics problem is solved using first-order perturbation theory derived from the diffusion equation for four groups. In contrast, Mo isotopes have longer half-lives, about 3 days (66 hours), so the delivery of radioisotopes to consumer centers and storage is possible though still limited. The production of this isotope potentially gives significant economic value. The criticality and flux in multigroup diffusion model was calculated for various irradiation positions and uranium contents. This model involves complex computation, with large and sparse matrix system. Several parallel algorithms have been developed for the sparse and large matrix solution. In this paper, a successive over-relaxation (SOR) algorithm was implemented for the calculation of reactivity coefficients which can be done in parallel. Previous works performed reactivity calculations serially with Gauss-Seidel iteratives. The parallel method can be used to solve multigroup diffusion equation system and calculate the criticality and reactivity coefficients. In this research a computer code was developed to exploit parallel processing to perform reactivity calculations which were to be used in safety analysis. The parallel processing in the multicore computer system allows the calculation to be performed more quickly. This code was applied for the safety limits calculation of irradiated FPM targets containing highly enriched uranium. The results of calculations neutron show that for uranium contents of 1.7676 g and 6.1866 g (× 106 cm-1) in a tube, their delta reactivities are the still

  8. Precision neutron polarimetry for neutron beta decay

    SciTech Connect

    Penttila, S. I.; Bowman, J. D.

    2004-01-01

    The abBA collaboration is developing a new type of field-expansion spectrometer for measurement of the three correlation coefficients a, A, and B and shape parameter b. The measurement of A and B requires precision neutron polarimetry. We will polarize a pulsed cold neutron beam from SNS using a {sup 3}He neutron spin filter. The well-known polarizing cross section for n-{sup 3}He has 1/v dependence, which is used to determine the absolute beam polarization through a time-of-flight (TOF) measurement. We show that measuring the TOF dependence of A and B, the coefficients and the neutron polarization can be determined with small loss of statistical precision and negligible systematic error. We conclude that it is possible to determine the neutron polarization averaged over a run in the neutron beta decay experiment to better than 10{sup -3}. We discuss various sources of systematic uncertainties in the measurement of A and B and conclude that they are less than 10{sup -4}.

  9. Neutron radiography using neutron imaging plate.

    PubMed

    Chankow, Nares; Punnachaiya, Suvit; Wonglee, Sarinrat

    2010-01-01

    The aims of this research are to study properties of a neutron imaging plate (NIP) and to test it for use in nondestructive testing (NDT) of materials. The experiments were carried out by using a BAS-ND 2040 Fuji NIP and a neutron beam from the Thai Research Reactor TRR-1/M1. The neutron intensity and Cd ratio at the specimen position were approximately 9x10(5) ns/cm(2) s and 100 respectively. It was found that the photostimulated luminescence (PSL) readout of the imaging plate was directly proportional to the exposure time and approximately 40 times faster than the conventional NR using Gd converter screen/X-ray film technique. The sensitivities of the imaging plate to slow neutron and to Ir-192 gamma-rays were found to be approximately 4.2x10(-3) PSL/mm(2) per neutron and 6.7x10(-5) PSL/mm(2) per gamma-ray photon respectively. Finally, some specimens containing light elements were selected to be radiographed with neutrons using the NIP and the Gd converter screen/X-ray film technique. The image quality obtained from the two recording media was found to be comparable. PMID:19828321

  10. Neutron metrology laboratory facility simulation.

    PubMed

    Pereira, Mariana; Salgado, Ana P; Filho, Aidano S; Pereira, Walsan W; Patrão, Karla C S; Fonseca, Evaldo S

    2014-10-01

    The Neutron Low Scattering Laboratory in Brazil has been completely rebuilt. Evaluation of air attenuation parameters and neutron component scattering in the room was done using Monte Carlo simulation code. Neutron fields produced by referenced neutron source were used to calculate neutron scattering and air attenuation. PMID:24864318