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Sample records for ggc-4 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. 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)

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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…

  18. 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.

  19. 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)

  20. 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.

  1. 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.

  2. 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)

  3. 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…

  4. 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

  5. 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)

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. 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

  12. 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."

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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.

  2. 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)

  3. 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.

  4. 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…

  5. 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…

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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%.

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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…

  2. 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…

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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,…

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  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. 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)

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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

  4. 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…

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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/.

  2. 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)

  3. 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)

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. 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)

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  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. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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

  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. 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

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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)

  5. 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

  6. 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

  7. 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}.

  8. Neutron chopper development at LANSCE

    SciTech Connect

    Nutter, M.; Lewis, L.; Tepper, S.; Silver, R.N.; Heffner, R.H.

    1985-01-01

    Progress is reported on neutron chopper systems for the Los Alamos Neutron Scattering Center pulsed spallation neutron source. This includes the development of 600+ Hz active magnetic bearing neutron chopper and a high speed control system designed to operate with the Proton Storage Ring to phase the chopper to the neutron source. 5 refs., 3 figs.

  9. 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

  10. Neutron sources and applications

    SciTech Connect

    Price, D.L.; Rush, J.J.

    1994-01-01

    Review of Neutron Sources and Applications was held at Oak Brook, Illinois, during September 8--10, 1992. This review involved some 70 national and international experts in different areas of neutron research, sources, and applications. Separate working groups were asked to (1) review the current status of advanced research reactors and spallation sources; and (2) provide an update on scientific, technological, and medical applications, including neutron scattering research in a number of disciplines, isotope production, materials irradiation, and other important uses of neutron sources such as materials analysis and fundamental neutron physics. This report summarizes the findings and conclusions of the different working groups involved in the review, and contains some of the best current expertise on neutron sources and applications.

  11. Neutron Imaging and Applications

    SciTech Connect

    Anderson, Ian S; McGreevy, Robert L; Bilheux, Hassina Z

    2009-04-01

    Neutron Imaging and Applications offers an introduction to the basics of neutron beam production and instrumentation in addition to the wide scope of techniques that provide unique imaging capabilities over a broad and diverse range of applications. An instructional overview of neutron sources, optics and detectors, allows readers to delve more deeply into the discussions of radiography, tomography, phase contrast imaging and prospective applications using advanced neutron holography techniques and polarized beams. A section devoted to overviews in a growing range of applications describes imaging of fuel cells and hydrogen storage devices for a robust hydrogen economy; new directions in material science and engineering; the investigation of precious artifacts of cultural heritage importance; determination of plant physiology and growth processes; imaging of biological tissues and macromolecules, and the practical elements of neutron imaging for homeland security and contraband detection. Written by key experts in the field, researchers and engineers involved with imaging technologies will find Neutron Imaging and Applications a valuable reference.

  12. Frascati neutron generator (FNG)

    NASA Astrophysics Data System (ADS)

    Martone, M.; Angelone, M.; Pillon, Mario

    1995-03-01

    The 14 MeV neutron generator (FNG), in operation at the ENEA Energy Center of Frascati, Italy, is described. It produces up to 1 X 1011 neutrons per second and consists essentially of a deuterium-ion accelerator, a beam transport system, and a target of titanium tritide, where neutrons are produced by the T(d,n)4He fusion reactions. An application of FNG in the context of research activity on controlled thermonuclear fusion research is also briefly described.

  13. The Advanced Neutron Source

    SciTech Connect

    Hayter, J.B.

    1989-01-01

    The Advanced Neutron Source (ANS) is a new user experimental facility planned to be operational at Oak Ridge in the late 1990's. The centerpiece of the ANS will be a steady-state research reactor of unprecedented thermal neutron flux ({phi}{sub th} {approx} 9{center dot}10{sup 19} m{sup -2}{center dot}s{sup -1}) accompanied by extensive and comprehensive equipment and facilities for neutron-based research. 5 refs., 5 figs.

  14. Introduction to neutron stars

    SciTech Connect

    Lattimer, James M.

    2015-02-24

    Neutron stars contain the densest form of matter in the present universe. General relativity and causality set important constraints to their compactness. In addition, analytic GR solutions are useful in understanding the relationships that exist among the maximum mass, radii, moments of inertia, and tidal Love numbers of neutron stars, all of which are accessible to observation. Some of these relations are independent of the underlying dense matter equation of state, while others are very sensitive to the equation of state. Recent observations of neutron stars from pulsar timing, quiescent X-ray emission from binaries, and Type I X-ray bursts can set important constraints on the structure of neutron stars and the underlying equation of state. In addition, measurements of thermal radiation from neutron stars has uncovered the possible existence of neutron and proton superfluidity/superconductivity in the core of a neutron star, as well as offering powerful evidence that typical neutron stars have significant crusts. These observations impose constraints on the existence of strange quark matter stars, and limit the possibility that abundant deconfined quark matter or hyperons exist in the cores of neutron stars.

  15. Neutron detection technique

    SciTech Connect

    Oblath, N.S.; Poon, A.W.P.

    2000-09-14

    The Sudbury Neutrino Observatory (SNO) has the ability to measure the total flux of all active flavors of neutrinos using the neutral current reaction, whose signature is a neutron. By comparing the rates of the neutral current reaction to the charged current reaction, which only detects electron neutrinos, one can test the neutrino oscillation hypothesis independent of solar models. It is necessary to understand the neutron detection efficiency of the detector to make use of the neutral current reaction. This report demonstrates a coincidence technique to identify neutrons emitted from the {sup 252}Cf neutron calibration source. The source releases on average four neutrons when a {sup 252}Cf nucleus spontaneously fissions. Each neutron is detected as a separate event when the neutron is captured by a deuteron, releasing a gamma ray of approximately 6.25 MeV. This gamma ray is in turn detected by the photomultiplier tube (PMT) array. By investigating the time and spatial separation between neutron-like events, it is possible to obtain a pure sample of neutrons for calibration study. Preliminary results of the technique applied to two calibration runs are presented.

  16. Code System to Calculate Neutron and Gamma-Ray Skyshine Doses Using the Integral Line-Beam Method.

    Energy Science and Technology Software Center (ESTSC)

    2000-11-16

    Version 03 This package includes the SKYNEUT 1.1, SKYDOSE 2.3, MCSKY 2.3 and SKYCONES 1.1 codes plus the DLC-188/SKYDATA library to form a comprehensive system for calculating skyshine doses. See the author's web site for related information: http://athena.mne.ksu.edu/~jks/ SKYNEUT evaluates the neutron and neutron-induced secondary gamma-ray skyshine doses from an isotropic, point, neutron source collimated by three simple geometries: an open silo, a vertical black (perfectly absorbing) wall, and a rectangular building. The source maymore » emit monoenergetic neutrons or neutrons with an arbitrary multigroup spectrum of energies. SKYDOSE evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated by three simple geometries: (1) a source in a silo, (2) a source behind an infinitely long, vertical, black wall, and (3) a source in a rectangular building. In all three geometries an optional overhead slab shield may be specified. MCSKY evaluates the gamma-ray skyshine dose from an isotropic, monoenergetic, point gamma-photon source collimated into either a vertical cone (i.e., silo geometry) or into a vertically oriented structure with an N-sided polygon cross section. An overhead laminate shield composed of two different materials is assumed, although shield thicknesses of zero may be specified to model an unshielded SKYSHINE source. SKYCONES evaluates the skyshine doses produced by a point neutron or gamma-photon source emitting, into the atmosphere, radiation that is collimated into an upward conical annulus between two arbitrary polar angles. The source is assumed to be axially (azimuthally) symmetric about a vertical axis through the source and can have an arbitrary polyenergetic spectrum. Nested contiguous annular cones can thus be used to represent the energy and polar-angle dependence of a skyshine source emitting radiation into the atmosphere.« less

  17. Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II.

    Energy Science and Technology Software Center (ESTSC)

    2008-11-21

    Version 00 SELFS-3 corrects for the influence of the self-shielding effect in neutron spectrum determinations by means of the multifoil activation method. It is used in combination with the SAND-II program for unfolding the responses of an irradiated set of activation detectors in 620 groups. The program SELFS can calculate a corrected 620 group cross section data set for specified reactions used in the SAND-II library, and for specified foil thicknesses. This procedure requires nomore » additional assumption on the shape of the neutron spectrum and on other experimental conditions, but only some foil characteristics (reaction type, material composition, foil thickness). Application of this procedure is possible when multigroup unfolding programs are used with suitably small energy intervals. This code system was developed in the 1970’s at Reactor Centrum Nederland, Petten, The Netherlands, and was contributed to RSICC through the NEA Data Bank. No changes were made to the package when it was released by RSICC in 2008. Modifications will be required to run SELFS-3 on current computer systems.« less

  18. Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II.

    SciTech Connect

    KONDO, IKUO

    2008-11-21

    Version 00 SELFS-3 corrects for the influence of the self-shielding effect in neutron spectrum determinations by means of the multifoil activation method. It is used in combination with the SAND-II program for unfolding the responses of an irradiated set of activation detectors in 620 groups. The program SELFS can calculate a corrected 620 group cross section data set for specified reactions used in the SAND-II library, and for specified foil thicknesses. This procedure requires no additional assumption on the shape of the neutron spectrum and on other experimental conditions, but only some foil characteristics (reaction type, material composition, foil thickness). Application of this procedure is possible when multigroup unfolding programs are used with suitably small energy intervals. This code system was developed in the 1970’s at Reactor Centrum Nederland, Petten, The Netherlands, and was contributed to RSICC through the NEA Data Bank. No changes were made to the package when it was released by RSICC in 2008. Modifications will be required to run SELFS-3 on current computer systems.

  19. GPU-accelerated 3D neutron diffusion code based on finite difference method

    SciTech Connect

    Xu, Q.; Yu, G.; Wang, K.

    2012-07-01

    Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)

  20. Neutron field characteristics of Ciemat's Neutron Standards Laboratory.

    PubMed

    Guzman-Garcia, Karen A; Mendez-Villafañe, Roberto; Vega-Carrillo, Hector Rene

    2015-06-01

    Monte Carlo calculations were carried out to characterize the neutron field produced by the calibration neutron sources of the Neutron Standards Laboratory at the Research Center for Energy, Environment, and Technology (CIEMAT) in Spain. For (241)AmBe and (252)Cf neutron sources, the neutron spectra, the ambient dose equivalent rates and the total neutron fluence rates were estimated. In the calibration hall, there are several items that modify the neutron field. To evaluate their effects different cases were used, from point-like source in vacuum up to the full model. Additionally, using the full model, the neutron spectra were estimated to different distances along the bench; with these spectra, the total neutron fluence and the ambient dose equivalent rates were calculated. The hall walls induce the largest changes in the neutron spectra and the respective integral quantities. The free-field neutron spectrum is modified due the room return effect. PMID:25468287

  1. Multigroup calculation of criticality and power distribution in a two-pass fast spectrum cermet-fueled reactor

    SciTech Connect

    Anghaie, S.; Feller, G.J. ); Peery, S.D.; Parsley, R.C. )

    1992-01-01

    The advanced propulsion group at Pratt Whitney has developed a nuclear thermal rocket concept, the XNR2000, for use on lunear, Mars, and deep-space planetary missions. The XNR2000 engine is powered by a fast spectrum cermet-fueled nuclear reactor that heats up hydrogen propellant to a maximum of 2850 K. An expander cycle is used to deliver 12 kg/s hydrogen to the core, producing 25,000 lb[sub f] thrust at 944 s of specific impulse. The reactor comprises a beryllium-reflected outer annulus core and an inner core with the hydrogen propellant entering from the bottom of the outer core and exiting from the bottom part of the inner core to the thrust chamber. Both the outer and inner cores are loaded with prismatic cermet fuel elements. The baseline XNR2000 reactor core consists of 90 fuel elements in the outer core and 61 in the inner core, arranged in the pattern. This paper focuses on the neutronic analysis of the baseline XNR2000 reactor.

  2. Transport analysis of measured neutron leakage spectra from spheres as tests of evaluated high energy cross sections

    NASA Technical Reports Server (NTRS)

    Bogart, D. D.; Shook, D. F.; Fieno, D.

    1973-01-01

    Integral tests of evaluated ENDF/B high-energy cross sections have been made by comparing measured and calculated neutron leakage flux spectra from spheres of various materials. An Am-Be (alpha,n) source was used to provide fast neutrons at the center of the test spheres of Be, CH2, Pb, Nb, Mo, Ta, and W. The absolute leakage flux spectra were measured in the energy range 0.5 to 12 MeV using a calibrated NE213 liquid scintillator neutron spectrometer. Absolute calculations of the spectra were made using version 3 ENDF/B cross sections and an S sub n discrete ordinates multigroup transport code. Generally excellent agreement was obtained for Be, CH2, Pb, and Mo, and good agreement was observed for Nb although discrepancies were observed for some energy ranges. Poor comparative results, obtained for Ta and W, are attributed to unsatisfactory nonelastic cross sections. The experimental sphere leakage flux spectra are tabulated and serve as possible benchmarks for these elements against which reevaluated cross sections may be tested.

  3. A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry Applications

    SciTech Connect

    Alpan, F.A.

    2011-07-01

    A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and {sup 58}Ni(n,{gamma}) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contribution and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within {+-} 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190. (authors)

  4. Neutron filters for producing monoenergetic neutron beams

    SciTech Connect

    Harvey, J.A.; Hill, N.W.; Harvey, J.R.

    1982-01-01

    Neutron transmission measurements have been made on high-purity, highly-enriched samples of /sup 58/Ni (99.9%), /sup 60/Ni (99.7%), /sup 64/Zn (97.9%) and /sup 184/W (94.5%) to measure their neutron windows and to assess their potential usefulness for producing monoenergetic beams of intermediate energies from a reactor. Transmission measurements on the Los Alamos Sc filter (44.26 cm Sc and 1.0 cm Ti) have been made to determine the characteristics of the transmitted neutron beam and to measure the total cross section of Sc at the 2.0 keV minimum. When corrected for the Ti and impurities, a value of 0.35 +- 0.03 b was obtained for this minimum.

  5. Neutron capture therapies

    SciTech Connect

    Yanch, J.C.; Shefer, R.E.; Klinkowstein, R.E.

    1999-11-02

    In one embodiment there is provided an application of the {sup 10}B(n,{alpha}){sup 7}Li nuclear reaction or other neutron capture reactions for the treatment of rheumatoid arthritis. This application, called Boron Neutron Capture Synovectomy (BNCS), requires substantially altered demands on neutron beam design than for instance treatment of deep seated tumors. Considerations for neutron beam design for the treatment of arthritic joints via BNCS are provided for, and comparisons with the design requirements for Boron Neutron Capture Therapy (BNCT) of tumors are made. In addition, exemplary moderator/reflector assemblies are provided which produce intense, high-quality neutron beams based on (p,n) accelerator-based reactions. In another embodiment there is provided the use of deuteron-based charged particle reactions to be used as sources for epithermal or thermal neutron beams for neutron capture therapies. Many d,n reactions (e.g. using deuterium, tritium or beryllium targets) are very prolific at relatively low deuteron energies.

  6. Neutron capture therapies

    SciTech Connect

    Yanch, Jacquelyn C.; Shefer, Ruth E.; Klinkowstein, Robert E.

    1999-01-01

    In one embodiment there is provided an application of the .sup.10 B(n,.alpha.).sup.7 Li nuclear reaction or other neutron capture reactions for the treatment of rheumatoid arthritis. This application, called Boron Neutron Capture Synovectomy (BNCS), requires substantially altered demands on neutron beam design than for instance treatment of deep seated tumors. Considerations for neutron beam design for the treatment of arthritic joints via BNCS are provided for, and comparisons with the design requirements for Boron Neutron Capture Therapy (BNCT) of tumors are made. In addition, exemplary moderator/reflector assemblies are provided which produce intense, high-quality neutron beams based on (p,n) accelerator-based reactions. In another embodiment there is provided the use of deuteron-based charged particle reactions to be used as sources for epithermal or thermal neutron beams for neutron capture therapies. Many d,n reactions (e.g. using deuterium, tritium or beryllium targets) are very prolific at relatively low deuteron energies.

  7. Neutron radiographic viewing system

    NASA Technical Reports Server (NTRS)

    Leysath, W.; Brown, R. L.

    1972-01-01

    Neutron radiographic viewing system consisting of camera head and control processor is developed for use in nondestructive testing applications. Camera head consists of neutron-sensitive image intensifier system, power supply, and SEC vidicon camera head. Both systems, with their optics, are housed on test mount.

  8. Neutron star models

    NASA Technical Reports Server (NTRS)

    Canuto, V.; Bowers, R. L.

    1981-01-01

    The current state of neutron star structure calculations is reviewed. Uncertainties in the equation of state for matter at and above nuclear density remain. The role of the delta resonance, pion condensates, and quark matter is reviewed. It is found that recent models yield stable neutron star masses which are consistent with observational estimates.

  9. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Untermyer, S.; Hutter, E.

    1959-08-01

    This patent relates to "shadow" control of a nuclear reactor. The control means comprises a plurality ot elongated rods disposed adjacent and parallel to each other, The morphology and effects of gases generated within sections of neutron absorbing materials and equal length sections of neutron permeable materials together with means for longitudinally pcsitioning the rcds relative to each other.

  10. Compact neutron generator

    DOEpatents

    Leung, Ka-Ngo; Lou, Tak Pui

    2005-03-22

    A compact neutron generator has at its outer circumference a toroidal shaped plasma chamber in which a tritium (or other) plasma is generated. A RF antenna is wrapped around the plasma chamber. A plurality of tritium ion beamlets are extracted through spaced extraction apertures of a plasma electrode on the inner surface of the toroidal plasma chamber and directed inwardly toward the center of neutron generator. The beamlets pass through spaced acceleration and focusing electrodes to a neutron generating target at the center of neutron generator. The target is typically made of titanium tubing. Water is flowed through the tubing for cooling. The beam can be pulsed rapidly to achieve ultrashort neutron bursts. The target may be moved rapidly up and down so that the average power deposited on the surface of the target may be kept at a reasonable level. The neutron generator can produce fast neutrons from a T-T reaction which can be used for luggage and cargo interrogation applications. A luggage or cargo inspection system has a pulsed T-T neutron generator or source at the center, surrounded by associated gamma detectors and other components for identifying explosives or other contraband.

  11. Gender and Acceptance of E-Learning: A Multi-Group Analysis Based on a Structural Equation Model among College Students in Chile and Spain

    PubMed Central

    2015-01-01

    The scope of this study was to evaluate whether the adoption of e-learning in two universities, and in particular, the relationship between the perception of external control and perceived ease of use, is different because of gender differences. The study was carried out with participating students in two different universities, one in Chile and one in Spain. The Technology Acceptance Model was used as a theoretical framework for the study. A multi-group analysis method in partial least squares was employed to relate differences between groups. The four main conclusions of the study are: (1) a version of the Technology Acceptance Model has been successfully used to explain the process of adoption of e-learning at an undergraduate level of study; (2) the finding of a strong and significant relationship between perception of external control and perception of ease of use of the e-learning platform; (3) a significant relationship between perceived enjoyment and perceived ease of use and between results demonstrability and perceived usefulness is found; (4) the study indicates a few statistically significant differences between males and females when adopting an e-learning platform, according to the tested model. PMID:26465895

  12. School belongingness and mental health functioning across the primary-secondary transition in a mainstream sample: multi-group cross-lagged analyses.

    PubMed

    Vaz, Sharmila; Falkmer, Marita; Parsons, Richard; Passmore, Anne Elizabeth; Parkin, Timothy; Falkmer, Torbjörn

    2014-01-01

    The relationship between school belongingness and mental health functioning before and after the primary-secondary school transition has not been previously investigated in students with and without disabilities. This study used a prospective longitudinal design to test the bi-directional relationships between these constructs, by surveying 266 students with and without disabilities and their parents, 6-months before and after the transition to secondary school. Cross-lagged multi-group analyses found student perception of belongingness in the final year of primary school to contribute to change in their mental health functioning a year later. The beneficial longitudinal effects of school belongingness on subsequent mental health functioning were evident in all student subgroups; even after accounting for prior mental health scores and the cross-time stability in mental health functioning and school belongingness scores. Findings of the current study substantiate the role of school contextual influences on early adolescent mental health functioning. They highlight the importance for primary and secondary schools to assess students' school belongingness and mental health functioning and transfer these records as part of the transition process, so that appropriate scaffolds are in place to support those in need. Longer term longitudinal studies are needed to increase the understanding of the temporal sequencing between school belongingness and mental health functioning of all mainstream students. PMID:24967580

  13. Gender and Acceptance of E-Learning: A Multi-Group Analysis Based on a Structural Equation Model among College Students in Chile and Spain.

    PubMed

    Ramírez-Correa, Patricio E; Arenas-Gaitán, Jorge; Rondán-Cataluña, F Javier

    2015-01-01

    The scope of this study was to evaluate whether the adoption of e-learning in two universities, and in particular, the relationship between the perception of external control and perceived ease of use, is different because of gender differences. The study was carried out with participating students in two different universities, one in Chile and one in Spain. The Technology Acceptance Model was used as a theoretical framework for the study. A multi-group analysis method in partial least squares was employed to relate differences between groups. The four main conclusions of the study are: (1) a version of the Technology Acceptance Model has been successfully used to explain the process of adoption of e-learning at an undergraduate level of study; (2) the finding of a strong and significant relationship between perception of external control and perception of ease of use of the e-learning platform; (3) a significant relationship between perceived enjoyment and perceived ease of use and between results demonstrability and perceived usefulness is found; (4) the study indicates a few statistically significant differences between males and females when adopting an e-learning platform, according to the tested model. PMID:26465895

  14. Optimal design of HTS magnets for a modular toroid-type 2.5 MJ SMES using multi-grouped particle swarm optimization

    NASA Astrophysics Data System (ADS)

    Lee, S. Y.; Kwak, S. Y.; Seo, J. H.; Lee, S. Y.; Park, S. H.; Kim, W. S.; Lee, J. K.; Bae, J. H.; Kim, S. H.; Sim, K. D.; Seong, K. C.; Jung, H. K.; Choi, K.; Hahn, S.

    2009-10-01

    Superconducting magnetic energy storage (SMES) is one of the promising power system applications of superconducting technology and has been actively researched and developed worldwide. Generally, there are three types of SMES-solenoid, multiple solenoid, and toroid. Among these types, toroid type seems to require more wires than solenoid type and multiple solenoid type at the same operating current. However toroid type reduces normal field in the wire and stray field dramatically because magnetic field is confined inside the coil. So, the total length of wire in the toroid type can be reduced in comparison with that in the solenoid type by increasing operating current. In this paper, a 2.5 MJ class SMES with HTS magnets of single solenoid, multiple solenoid and modular toroid type were optimized using a recently developed multi-modal optimization technique named multi-grouped particle swarm optimization (MGPSO). The objective of the optimization was to minimize the total length of HTS superconductor wires satisfying some equality and inequality constraints. The stored energy and constraints were calculated using 3D magnetic field analysis techniques and an automatic tetrahedral mesh generator. Optimized results were verified by 3D finite element method (FEM).

  15. Neutrons against cancer

    NASA Astrophysics Data System (ADS)

    Dovbnya, A. N.; Kuplennikov, E. L.; Kandybey, S. S.; Krasiljnikov, V. V.

    2014-09-01

    The review is devoted to the analysis and generalization of the research carried out during recent years in industrially advanced countries on the use of fast, epithermal, and thermal neutrons for therapy of malignant tumors. Basic facilities for neutron production used for cancer treatment are presented. Optimal parameters of therapeutic beams are described. Techniques using neutrons of different energy regions are discussed. Results and medical treatment efficiency are given. Comparison of the current state of neutron therapy of tumors and alternative treatments with beams of protons and carbon ions has been conducted. Main attention is given to the possibility of the practical use of accumulated experience of application of neutron beams for cancer therapy.

  16. Simulated workplace neutron fields

    NASA Astrophysics Data System (ADS)

    Lacoste, V.; Taylor, G.; Röttger, S.

    2011-12-01

    The use of simulated workplace neutron fields, which aim at replicating radiation fields at practical workplaces, is an alternative solution for the calibration of neutron dosemeters. They offer more appropriate calibration coefficients when the mean fluence-to-dose equivalent conversion coefficients of the simulated and practical fields are comparable. Intensive Monte Carlo modelling work has become quite indispensable for the design and/or the characterization of the produced mixed neutron/photon fields, and the use of Bonner sphere systems and proton recoil spectrometers is also mandatory for a reliable experimental determination of the neutron fluence energy distribution over the whole energy range. The establishment of a calibration capability with a simulated workplace neutron field is not an easy task; to date only few facilities are available as standard calibration fields.

  17. NEUTRON SHIELDING STRUCTURE

    DOEpatents

    Mattingly, J.T.

    1962-09-25

    A lightweight neutron shielding structure comprises a honeycomb core which is filled with a neutron absorbing powder. The honeycomb core is faced with parallel planar facing sheets to form a lightweight rigid unit. Suitable absorber powders are selected from among the following: B, B/sub 4/C, B/sub 2/O/ sub 3/, CaB/sub 6/, Li/sub 2/CO3, LiOH, LiBO/sub 2/, Li/s ub 2/O. The facing sheets are constructed of a neutron moderating material, so that fast neutrons will be moderated while traversing the facing sheets, and ultimately be absorbed by the absorber powder in the honeycomb. Beryllium is a preferred moderator material for use in the facing sheets. The advantage of the structure is that it combines the rigidity and light weight of a honeycomb construction with the neutron absorption properties of boron and lithium. (AEC)

  18. Precision Neutron Polarimetry

    NASA Astrophysics Data System (ADS)

    Sharma, Monisha; Barron-Palos, L.; Bowman, J. D.; Chupp, T. E.; Crawford, C.; Danagoulian, A.; Klein, A.; Penttila, S. I.; Salas-Bacci, A. F.; Wilburn, W. S.

    2008-04-01

    Proposed PANDA and abBA experiments aim to measure the correlation coefficients in the polarized neutron beta decay at the SNS. The goal of these experiments is 0.1% measurement which will require neutron polarimetry at 0.1% level. The FnPB neutron beam will be polarized either using a ^3He spin filter or a supermirror polarizer and the neutron polarization will be measured using a ^3He spin filter. Experiment to establish the accuracy to which neutron polarization can be determined using ^3He spin fliters was performed at Los Alamos National Laboratory in Summer 2007 and the analysis is in progress. The details of the experiment and the results will be presented.

  19. Pocked surface neutron detector

    DOEpatents

    McGregor, Douglas; Klann, Raymond

    2003-04-08

    The detection efficiency, or sensitivity, of a neutron detector material such as of Si, SiC, amorphous Si, GaAs, or diamond is substantially increased by forming one or more cavities, or holes, in its surface. A neutron reactive material such as of elemental, or any compound of, .sup.10 B, .sup.6 Li, .sup.6 LiF, U, or Gd is deposited on the surface of the detector material so as to be disposed within the cavities therein. The portions of the neutron reactive material extending into the detector material substantially increase the probability of an energetic neutron reaction product in the form of a charged particle being directed into and detected by the neutron detector material.

  20. Neutron stars - General review

    NASA Technical Reports Server (NTRS)

    Cameron, A. G. W.; Canuto, V.

    1974-01-01

    A review is presented of those properties of neutron stars upon which there is general agreement and of those areas which currently remain in doubt. Developments in theoretical physics of neutron star interiors are summarized with particular attention devoted to hyperon interactions and the structure of interior layers. Determination of energy states and the composition of matter is described for successive layers, beginning with the surface and proceeding through the central region into the core. Problems encountered in determining the behavior of matter in the ultra-high density regime are discussed, and the effects of the magnetic field of a neutron star are evaluated along with the behavior of atomic structures in the field. The evolution of a neutron star is outlined with discussion centering on carbon detonation, cooling, vibrational damping, rotation, and pulsar glitches. The role of neutron stars in cosmic-ray propagation is considered.

  1. The Neutron Structure Function

    NASA Astrophysics Data System (ADS)

    Holt, Roy

    2013-10-01

    Knowledge of the neutron structure function is important for testing models of the nucleon, for a complete understanding of deep inelastic scattering (DIS) from nuclei, and for high energy experiments. As there exist no free neutron targets, neutron structure functions have been determined from deep inelastic scattering from the deuteron. Unfortunately, the short-range part of the deuteron wave function becomes important in extracting the neutron structure function at very high Bjorken x. New methods have been devised for Jefferson Lab experiments to mitigate this problem. The BONUS experiment involves tagging spectator neutrons in the deuteron, while the MARATHON experiment minimizes nuclear structure effects by a comparison of DIS from 3H and 3He. A summary of the status and future plans will be presented. This work supported by the U. S. Department of Energy, Office of Nuclear Physics, under contract DE-AC02-06CH11357.

  2. Pulsed neutron detector

    DOEpatents

    Robertson, deceased, J. Craig; Rowland, Mark S.

    1989-03-21

    A pulsed neutron detector and system for detecting low intensity fast neutron pulses has a body of beryllium adjacent a body of hydrogenous material the latter of which acts as a beta particle detector, scintillator, and moderator. The fast neutrons (defined as having En>1.5 MeV) react in the beryllium and the hydrogenous material to produce larger numbers of slow neutrons than would be generated in the beryllium itself and which in the beryllium generate hellium-6 which decays and yields beta particles. The beta particles reach the hydrogenous material which scintillates to yield light of intensity related to the number of fast neutrons. A photomultiplier adjacent the hydrogenous material (scintillator) senses the light emission from the scintillator. Utilization means, such as a summing device, sums the pulses from the photo-multiplier for monitoring or other purposes.

  3. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    SciTech Connect

    Woo Y. Yoon; David W. Nigg

    2008-09-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  4. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    SciTech Connect

    Woo Y. Yoon; David W. Nigg

    2009-08-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  5. Optical polarizing neutron devices designed for pulsed neutron sources

    SciTech Connect

    Takeda, M.; Kurahashi, K.; Endoh, Y.; Itoh, S.

    1997-09-01

    We have designed two polarizing neutron devices for pulsed cold neutrons. The devices have been tested at the pulsed neutron source at the Booster Synchrotron Utilization Facility of the National Laboratory for High Energy Physics. These two devices proved to have a practical use for experiments to investigate condensed matter physics using pulsed cold polarized neutrons.

  6. Neutron Decay Array for beta-delayed neutron Decay Studies

    NASA Astrophysics Data System (ADS)

    Lorusso, Giuseppe; Pereira, J.; Hosmer, P.; Kern, L.; Kratz, K.; Montes, F.; Reeder, P.; Santi, P.; Schatz, H.; Schertz, F.; Wör, A.

    The Neutron Emission Ratio Observer (NERO), has been constructed for use at the National Superconducting Cyclotron Laboratory to work in conjunction with the NSCL Beta Counting System BCS [1] in order to detect β-delayed neutrons. The design of the detector provides high and flat efficiency for a wide range of neutron energies, as well as a low neutron background.

  7. Modular Code and Data System for Fast Reactor Neutronics Analyses

    SciTech Connect

    RIMPAULT, G.

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&D organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected in the

  8. Modular Code and Data System for Fast Reactor Neutronics Analyses

    Energy Science and Technology Software Center (ESTSC)

    2008-06-30

    Version 00. The European Reactor ANalysis Optimized calculation System, ERANOS, has been developed and validated with the aim of providing a suitable basis for reliable neutronic calculations of current as well as advanced fast reactor cores. It consists of data libraries, deterministic codes and calculation procedures which have been developed within the European Collaboration on Fast Reactors over the past 20 years or so, in order to answer the needs of both industrial and R&Dmore » organisations. The whole system counts roughly 250 functions and 3000 subroutines totalling 450000 lines of FORTRAN-77 and ESOPE instructions. ERANOS is written using the ALOS software which requires only standard FORTRAN compilers and includes advanced programming features. A modular structure was adopted for easier evolution and incorporation of new functionalities. Blocks of data (SETs) can be created or used by the modules themselves or by the user via the LU control language. Programming, and dynamic memory allocation, are performed by means of the ESOPE language. External temporary storage and permanent storage capabilities are provided by the GEMAT and ARCHIVE functions, respectively. ESOPE, LU, GEMAT and ARCHIVE are all part of the ALOS software. This modular structure allows different modules to be linked together in procedures corresponding to recommended calculation routes ranging from fast-running and moderately-accurate 'routine' procedures to slow-running but highly-accurate 'reference' procedures. The main contents of the ERANOS-2.0 package are: nuclear data libraries (multigroup cross-sections from the JEF-2.2 evaluated nuclear data file, and other specific data files), a cell and lattice code (ECCO), reactor flux solvers (diffusion, Sn transport, nodal variational transport), a burn-up module, various processing modules (material and neutron balance, breeding gains,...), tools related to perturbation theory and sensitivity analysis, core follow-up modules (connected

  9. Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System.

    Energy Science and Technology Software Center (ESTSC)

    2013-06-24

    Version 07 TART2012 is a coupled neutron-photon Monte Carlo transport code designed to use three-dimensional (3-D) combinatorial geometry. Neutron and/or photon sources as well as neutron induced photon production can be tracked. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART2012 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared tomore » other similar codes. Use of the entire system can save you a great deal of time and energy. TART2012 extends the general utility of the code to even more areas of application than available in previous releases by concentrating on improving the physics, particularly with regard to improved treatment of neutron fission, resonance self-shielding, molecular binding, and extending input options used by the code. Several utilities are included for creating input files and displaying TART results and data. TART2012 uses the latest ENDF/B-VI, Release 8, data. New for TART2012 is the use of continuous energy neutron cross sections, in addition to its traditional multigroup cross sections. For neutron interaction, the data are derived using ENDF-ENDL2005 and include both continuous energy cross sections and 700 group neutron data derived using a combination of ENDF/B-VI, Release 8, and ENDL data. The 700 group structure extends from 10-5 eV up to 1 GeV. Presently nuclear data are only available up to 20 MeV, so that only 616 of the groups are currently used. For photon interaction, 701 point photon data were derived using the Livermore EPDL97 file. The new 701 point structure extends from 100 eV up to 1 GeV, and is currently used over this entire energy range. TART2012 completely supersedes all older versions of TART, and it is strongly recommended that one use only the most recent version of TART2012 and its data files. Check author’s homepage for related information: http

  10. Neutron Laue macromolecular crystallography

    SciTech Connect

    Meilleur, Flora; Myles, Dean A A; Blakeley, M. P.

    2006-01-01

    Recent progress in neutron protein crystallography such as the use of the Laue technique and improved neutron optics and detector technologies have dramatically improved the speed and precision with which neutron protein structures can now be determined. These studies are providing unique and complementary insights on hydrogen and hydration in protein crystal structures that are not available from X-ray structures alone. Parallel improvements in modern molecular biology now allow fully (per)deuterated protein samples to be produced for neutron scattering that essentially eradicate the large--and ultimately limiting--hydrogen incoherent scattering background that has hampered such studies in the past. High quality neutron data can now be collected to near atomic resolution ({approx}2.0 Angstroms) for proteins of up to {approx}50 kDa molecular weight using crystals of volume {approx}0.1 mm3 on the Laue diffractometer at ILL. The ability to flash-cool and collect high resolution neutron data from protein crystals at cryogenic temperature (15 K) has opened the way for kinetic crystallography on freeze trapped systems. Current instrument developments now promise to reduce crystal volume requirements by a further order of magnitude, making neutron protein crystallography a more accessible and routine technique.

  11. Colloquium: The neutron lifetime

    SciTech Connect

    Greene, Geoffrey L; Wietfeldt, F

    2011-01-01

    The decay of the free neutron into a proton, electron, and antineutrino is the prototype semileptonic weak decay and is the simplest example of nuclear beta decay. It played a key role in the early Universe as it determined the ratio of neutrons to protons during the era of primordial light element nucleosynthesis. Neutron decay is physically related to important processes in solar physics and neutrino detection. The mean neutron lifetime has been the subject of more than 20 major experiments done, using a variety of methods, between 1950 and the present. The most precise recent measurements have stated accuracies approaching 0.1%, but are not in good agreement as they differ by as much as 5 sigma using quoted uncertainties. The history of neutron lifetime measurements is reviewed and the different methods used are described, giving important examples of each. The discrepancies and some systematic issues in the experiments that may be responsible are discussed, and it is shown by means of global averages that the neutron lifetime is likely to lie in the range of 880 884 s. Plans and prospects for future experiments are considered that will address these systematic issues and improve our knowledge of the neutron lifetime.

  12. The neutron channeling phenomenon.

    PubMed

    Khanouchi, A; Sabir, A; Boulkheir, M; Ichaoui, R; Ghassoun, J; Jehouani, A

    1997-01-01

    Shields, used for protection against radiation, are often pierced with vacuum channels for passing cables and other instruments for measurements. The neutron transmission through these shields is an unavoidable phenomenon. In this work we study and discuss the effect of channels on neutron transmission through shields. We consider an infinite homogeneous slab, with a fixed thickness (20 lambda, with lambda the mean free path of the neutron in the slab), which contains a vacuum channel. This slab is irradiated with an infinite source of neutrons on the left side and on the other side (right side) many detectors with windows equal to 2 lambda are placed in order to evaluate the neutron transmission probabilities (Khanouchi, A., Aboubekr, A., Ghassoun, J. and Jehouani, A. (1994) Rencontre Nationale des Jeunes Chercheurs en Physique. Casa Blanca Maroc; Khanouchi, A., Sabir, A., Ghassoun, J. and Jehouani, A. (1995) Premier Congré International des Intéractions Rayonnements Matière. Eljadida Maroc). The neutron history within the slab is simulated by the Monte Carlo method (Booth, T. E. and Hendricks, J. S. (1994) Nuclear Technology 5) and using the exponential biasing technique in order to improve the Monte Carlo calculation (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500-504; Jehouani, A., Ghassoun, J. and Aboubker, A. (1994) In Proceedings of the 6th International Symposium on Radiation Physics, Rabat, Morocco). Then different geometries of the vacuum channel have been studied. For each geometry we have determined the detector response and calculated the neutron transmission probability for different detector positions. This neutron transmission probability presents a peak for the detectors placed in front of the vacuum channel. This study allowed us to clearly identify the neutron channeling phenomenon. One application of our study is to detect vacuum defects in materials. PMID:9463884

  13. Ultrashort pulsed neutron source.

    PubMed

    Pomerantz, I; McCary, E; Meadows, A R; Arefiev, A; Bernstein, A C; Chester, C; Cortez, J; Donovan, M E; Dyer, G; Gaul, E W; Hamilton, D; Kuk, D; Lestrade, A C; Wang, C; Ditmire, T; Hegelich, B M

    2014-10-31

    We report on a novel compact laser-driven neutron source with an unprecedented short pulse duration (<50  ps) and high peak flux (>10(18)  n/cm(2)/s), an order of magnitude higher than any existing source. In our experiments, high-energy electron jets are generated from thin (<3  μm) plastic targets irradiated by a petawatt laser. These intense electron beams are employed to generate neutrons from a metal converter. Our method opens venues for enhancing neutron radiography contrast and for creating astrophysical conditions of heavy element synthesis in the laboratory. PMID:25396373

  14. Ultrashort Pulsed Neutron Source

    NASA Astrophysics Data System (ADS)

    Pomerantz, I.; McCary, E.; Meadows, A. R.; Arefiev, A.; Bernstein, A. C.; Chester, C.; Cortez, J.; Donovan, M. E.; Dyer, G.; Gaul, E. W.; Hamilton, D.; Kuk, D.; Lestrade, A. C.; Wang, C.; Ditmire, T.; Hegelich, B. M.

    2014-10-01

    We report on a novel compact laser-driven neutron source with an unprecedented short pulse duration (<50 ps ) and high peak flux (>1018 n /cm2/s ), an order of magnitude higher than any existing source. In our experiments, high-energy electron jets are generated from thin (<3 μ m ) plastic targets irradiated by a petawatt laser. These intense electron beams are employed to generate neutrons from a metal converter. Our method opens venues for enhancing neutron radiography contrast and for creating astrophysical conditions of heavy element synthesis in the laboratory.

  15. METHOD OF PRODUCING NEUTRONS

    DOEpatents

    Imhoff, D.H.; Harker, W.H.

    1964-01-14

    This patent relates to a method of producing neutrons in which there is produced a heated plasma containing heavy hydrogen isotope ions wherein heated ions are injected and confined in an elongated axially symmetric magnetic field having at least one magnetic field gradient region. In accordance with the method herein, the amplitude of the field and gradients are varied at an oscillatory periodic frequency to effect confinement by providing proper ratios of rotational to axial velocity components in the motion of said particles. The energetic neutrons may then be used as in a blanket zone containing a moderator and a source fissionable material to produce heat and thermal neutron fissionable materials. (AEC)

  16. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  17. FABRICATION OF NEUTRON SOURCES

    DOEpatents

    Birden, J.H.

    1959-04-21

    A method is presented for preparing a neutron source from polonium-210 and substances, such as beryllium and boron, characterized by emission of neutrons upon exposure to alpha particles from the polonium. According to the invention, a source is prepared by placing powdered beryllium and a platinum foil electroplated with polonium-2;.0 in a beryllium container. The container is sealed and then heated by induction to a temperature of 450 to 1100 deg C to volatilize the polonium off the foil into the powder. The heating step is terminated upon detection of a maximum in the neutron flux level.

  18. Coupled moderator neutronics

    SciTech Connect

    Russell, G.J.; Pitcher, E.J.; Ferguson, P.D.

    1995-12-01

    Optimizing the neutronic performance of a coupled-moderator system for a Long-Pulse Spallation Source is a new and challenging area for the spallation target-system designer. For optimal performance of a neutron source, it is essential to have good communication with instrument scientists to obtain proper design criteria and continued interaction with mechanical, thermal-hydraulic, and materials engineers to attain a practical design. A good comprehension of the basics of coupled-moderator neutronics will aid in the proper design of a target system for a Long-Pulse Spallation Source.

  19. Switchable radioactive neutron source device

    DOEpatents

    Stanford, G.S.; Rhodes, E.A.; Devolpi, A.; Boyar, R.E.

    1987-11-06

    This invention is a switchable neutron generating apparatus comprised of a pair of plates, the first plate having an alpha emitter section on it and the second plate having a target material portion on it which generates neutrons when its nuclei absorb an alpha particle. In operation, the alpha portion of the first plate is aligned with the neutron portion of the second plate to produce neutrons and brought out of alignment to cease production of neutrons. 3 figs.

  20. Switchable radioactive neutron source device

    DOEpatents

    Boyar, Robert E.; DeVolpi, Alexander; Stanford, George S.; Rhodes, Edgar A.

    1989-01-01

    This invention is a switchable neutron generating apparatus comprised of a pair of plates, the first plate having an alpha emitter section on it and the second plate having a target material portion on it which generates neutrons when its nuclei absorb an alpha particle. In operation, the alpha portion of the first plate is aligned with the neutron portion of the second plate to produce neutrons and brought out of alignment to cease production of neutrons.

  1. Isotope-Identifying neutron reflectometry

    SciTech Connect

    Nikitenko, Yu. V. Petrenko, A. V.; Gundorin, N. A.; Gledenov, Yu. M.; Aksenov, V. L.

    2015-07-15

    The possibilities of an isotope-indentifying study of layered structures in different regimes of a neutron wave field are considered. The detection of specularly reflected neutrons and secondary radiation (caused by neutron capture) in the form of charged particles, γ quanta, and nuclear fission fragments, as well as neutrons spin-flipped in a noncollinear magnetic field and on nuclei of elements with spin, makes it possible to implement isotope-indentifying neutron reflectometry.

  2. Magnetized Neutron Stars

    NASA Astrophysics Data System (ADS)

    Liebling, Steven; Anderson, Matthew; Hirschmann, Eric; Lehner, Luis; Motl, Patrick; Neilsen, David; Palenzuela, Carlos; Tohline, Joel

    2008-04-01

    Magnetized neutron stars, whether considered individually or within compact binary systems, demonstrate a number of interesting dynamical effects. Using a distributed adaptive mesh refinement (AMR) code, we evolve such stars and study their dynamics.

  3. Precision Polarization of Neutrons

    NASA Astrophysics Data System (ADS)

    Martin, Elise; Barron-Palos, Libertad; Couture, Aaron; Crawford, Christopher; Chupp, Tim; Danagoulian, Areg; Estes, Mary; Hona, Binita; Jones, Gordon; Klein, Andi; Penttila, Seppo; Sharma, Monisha; Wilburn, Scott

    2009-05-01

    Determining polarization of a cold neutron beam to high precision is required for the next generation neutron decay correlation experiments at the SNS, such as the proposed abBA and PANDA experiments. Precision polarimetry measurements were conducted at Los Alamos National Laboratory with the goal of determining the beam polarization to the level of 10-3 or better. The cold neutrons from FP12 were polarized using optically polarized ^3He gas as a spin filter, which has a highly spin-dependent absorption cross section. A second ^ 3He spin filter was used to analyze the neutron polarization after passing through a resonant RF spin rotator. A discussion of the experiment and results will be given.

  4. Neutron Detector Waveform Digitization

    NASA Astrophysics Data System (ADS)

    Toebbe, Jonathan; Gray, Fred; Grafil, Elliot; Greife, Uwe

    2010-11-01

    In the frame of a DoE Office of Nuclear Energy funded collaboration to design a next generation neutron elastic and inelastic scattering experiment, the Colorado School of Mines/Regis University group is responsible for developing and testing neutron detectors, pulse shape discrimination and read-out methods. This contribution will describe the test setup based on an n-ToF neutron selection using a ^244Cm-^13C source and the Regis Digitizer. Results on pulse shape discrimination from waveform digitization will be compared to other commercially available discrimination methods. We will also present our efforts to explore different types of algorithm for extraction of neutron assignment probabilities from the collected waveforms.

  5. Neutron personnel dosimetry

    SciTech Connect

    Griffith, R.V.

    1981-06-16

    The current state-of-the-art in neutron personnel dosimetry is reviewed. Topics covered include dosimetry needs and alternatives, current dosimetry approaches, personnel monitoring devices, calibration strategies, and future developments. (ACR)

  6. Directionally positionable neutron beam

    SciTech Connect

    Bumgardner, H.M.; Dance, W.E.

    1981-11-10

    Disclosed is apparatus for forming and directionally positioning a neutron beam. The apparatus includes an enclosed housing rotatable about a first axis with a neutron source axially positionable on the axis of rotation of the enclosed housing but rotationally fixed with respect to the housing. The rotatable housing is carried by a vertically positionable arm carried on a mobile transport. A collimator is supported by the rotatable housing and projects into the housing to orientationally position its inlet window at an adjustably fixed axial and radial spacing from the neutron source so that rotation of the enclosed housing causes the inlet window to rotate about a circle which is a fixed axial distance from the neutron source and has the axis of rotation of the housing as its center.

  7. Shifting scintillator neutron detector

    SciTech Connect

    Clonts, Lloyd G; Cooper, Ronald G; Crow, Jr., Morris Lowell; Hannah, Bruce W; Hodges, Jason P; Richards, John D; Riedel, Richard A

    2014-03-04

    Provided are sensors and methods for detecting thermal neutrons. Provided is an apparatus having a scintillator for absorbing a neutron, the scintillator having a back side for discharging a scintillation light of a first wavelength in response to the absorbed neutron, an array of wavelength-shifting fibers proximate to the back side of the scintillator for shifting the scintillation light of the first wavelength to light of a second wavelength, the wavelength-shifting fibers being disposed in a two-dimensional pattern and defining a plurality of scattering plane pixels where the wavelength-shifting fibers overlap, a plurality of photomultiplier tubes, in coded optical communication with the wavelength-shifting fibers, for converting the light of the second wavelength to an electronic signal, and a processor for processing the electronic signal to identify one of the plurality of scattering plane pixels as indicative of a position within the scintillator where the neutron was absorbed.

  8. Personnel neutron dosimetry

    SciTech Connect

    Hankins, D.

    1982-04-01

    This edited transcript of a presentation on personnel neutron discusses the accuracy of present dosimetry practices, requirements, calibration, dosemeter types, quality factors, operational problems, and dosimetry for a criticality accident. 32 figs. (ACR)

  9. Neutron resonance averaging

    SciTech Connect

    Chrien, R.E.

    1986-10-01

    The principles of resonance averaging as applied to neutron capture reactions are described. Several illustrations of resonance averaging to problems of nuclear structure and the distribution of radiative strength in nuclei are provided. 30 refs., 12 figs.

  10. Cooling of neutron stars

    NASA Technical Reports Server (NTRS)

    Pethick, C. J.

    1992-01-01

    It is at present impossible to predict the interior constitution of neutron stars based on theory and results from laboratory studies. It has been proposed that it is possible to obtain information on neutron star interiors by studying thermal radiation from their surfaces, because neutrino emission rates, and hence the temperature of the central part of a neutron star, depend on the properties of dense matter. The theory predicts that neutron stars cool relatively slowly if their cores are made up of nucleons, and cool faster if the matter is in an exotic state, such as a pion condensate, a kaon condensate, or quark matter. This view has recently been questioned by the discovery of a number of other processes that could lead to copious neutrino emission and rapid cooling.

  11. Cylindrical neutron generator

    DOEpatents

    Leung, Ka-Ngo

    2009-12-29

    A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360.degree. and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

  12. Cylindrical neutron generator

    DOEpatents

    Leung, Ka-Ngo

    2008-04-22

    A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360.degree. and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

  13. Cylindrical neutron generator

    DOEpatents

    Leung, Ka-Ngo

    2005-06-14

    A cylindrical neutron generator is formed with a coaxial RF-driven plasma ion source and target. A deuterium (or deuterium and tritium) plasma is produced by RF excitation in a cylindrical plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator, separated by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360.degree. and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. The plasma generator may be in the center and the neutron target on the outside, or the plasma generator may be on the outside and the target on the inside. In a nested configuration, several concentric targets and plasma generating regions are nested to increase the neutron flux.

  14. NEUTRONIC REACTOR FUEL COMPOSITION

    DOEpatents

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  15. Neutron phase spin echo

    NASA Astrophysics Data System (ADS)

    Piegsa, Florian M.; Hautle, Patrick; Schanzer, Christian

    2016-04-01

    A novel neutron spin resonance technique is presented based on the well-known neutron spin echo method. In a first proof-of-principle measurement using a monochromatic neutron beam, it is demonstrated that relative velocity changes of down to a precision of 4 ×10-7 can be resolved, corresponding to an energy resolution of better than 3 neV. Currently, the sensitivity is only limited by counting statistics and not by systematic effects. An improvement by another two orders of magnitude can be achieved with a dedicated setup, allowing energy resolutions in the 10 peV regime. The new technique is ideally suited for investigations in the field of precision fundamental neutron physics, but will also be beneficial in scattering applications.

  16. Determination of the Neutron Lifetime Using Magnetically Trapped Neutrons

    PubMed Central

    Dzhosyuk, S. N.; Copete, A.; Doyle, J. M.; Yang, L.; Coakley, K. J.; Golub, R.; Korobkina, E.; Kreft, T.; Lamoreaux, S. K.; Thompson, A. K.; Yang, G. L.; Huffman, P. R.

    2005-01-01

    We report progress on an experiment to measure the neutron lifetime using magnetically trapped neutrons. Neutrons are loaded into a 1.1 T deep superconducting Ioffe-type trap by scattering 0.89 nm neutrons in isotopically pure superfluid 4He. Neutron decays are detected in real time using the scintillation light produced in the helium by the beta-decay electrons. The measured trap lifetime at a helium temperature of 300 mK and with no ameliorative magnetic ramping is substantially shorter than the free neutron lifetime. This is attributed to the presence of neutrons with energies higher than the magnetic potential of the trap. Magnetic field ramping is implemented to eliminate these neutrons, resulting in an 833−63+74s trap lifetime, consistent with the currently accepted value of the free neutron lifetime. PMID:27308147

  17. Pulsed spallation Neutron Sources

    SciTech Connect

    Carpenter, J.M.

    1994-12-31

    This paper reviews the early history of pulsed spallation neutron source development at Argonne and provides an overview of existing sources world wide. A number of proposals for machines more powerful than currently exist are under development, which are briefly described. The author reviews the status of the Intense Pulsed Neutron Source, its instrumentation, and its user program, and provides a few examples of applications in fundamental condensed matter physics, materials science and technology.

  18. NEUTRON FLUX INTENSITY DETECTION

    DOEpatents

    Russell, J.T.

    1964-04-21

    A method of measuring the instantaneous intensity of neutron flux in the core of a nuclear reactor is described. A target gas capable of being transmuted by neutron bombardment to a product having a resonance absorption line nt a particular microwave frequency is passed through the core of the reactor. Frequency-modulated microwave energy is passed through the target gas and the attenuation of the energy due to the formation of the transmuted product is measured. (AEC)

  19. Pulsed spallation neutron sources

    SciTech Connect

    Carpenter, J.M.

    1996-05-01

    This paper reviews the early history of pulsed spallation neutron source development ar Argonne and provides an overview of existing sources world wide. A number of proposals for machines more powerful than currently exist are under development, which are briefly described. The author reviews the status of the Intense Pulsed Neutron Source, its instrumentation, and its user program, and provide a few examples of applications in fundamental condensed matter physics, materials science and technology.

  20. FABRICATION OF NEUTRON SOURCES

    DOEpatents

    Birden, J.H.

    1959-01-20

    A method is presented for preparing a more efficient neutron source comprising inserting in a container a quantity of Po-210, inserting B powder coated with either Ag, Pt, or Ni. The container is sealed and then slowly heated to about 450 C to volatilize the Po and effect combination of the coated powder with the Po. The neutron flux emitted by the unit is moritored and the heating step is terminated when the flux reaches a maximum or selected level.

  1. Matter accreting neutron stars

    NASA Technical Reports Server (NTRS)

    Meszaros, P.

    1981-01-01

    Some of the fundamental neutron star parameters, such as the mass and the magnetic field strength, were experimentally determined in accreting neutron star systems. Some of the relevant data and the models used to derive useful information from them, are reviewed concentrating mainly on X-ray pulsars. The latest advances in our understanding of the radiation mechanisms and the transfer in the strongly magnetized polar cap regions are discussed.

  2. Multigroup confirmatory factor analysis of U.S. and Italian children's performance on the PASS theory of intelligence as measured by the Cognitive Assessment System.

    PubMed

    Naglieri, Jack A; Taddei, Stefano; Williams, Kevin M

    2013-03-01

    This study examined Italian and U.S. children's performance on the English and Italian versions, respectively, of the Cognitive Assessment System (CAS; Naglieri & Conway, 2009; Naglieri & Das, 1997), a test based on a neurocognitive theory of intelligence entitled PASS (Planning, Attention, Simultaneous, and Successive; Naglieri & Das, 1997; Naglieri & Otero, 2011). CAS subtest, PASS scales, and Full Scale scores for Italian (N=809) and U.S. (N=1,174) samples, matched by age and gender, were examined. Multigroup confirmatory factor analysis results supported the configural invariance of the CAS factor structure between Italians and Americans for the 5- to 7-year-old (root-mean-square error of approximation [RMSEA]=.038; 90% confidence interval [CI]=.033, .043; comparative fit index [CFI]=.96) and 8- to 18-year-old (RMSEA=.036; 90% CI=.028, .043; CFI=.97) age groups. The Full Scale standard scores (using the U.S. norms) for the Italian (100.9) and U.S. (100.5) samples were nearly identical. The scores between the samples for the PASS scales were very similar, except for the Attention Scale (d=0.26), where the Italian sample's mean score was slightly higher. Negligible mean differences were found for 9 of the 13 subtest scores, 3 showed small d-ratios (2 in favor of the Italian sample), and 1 was large (in favor of the U.S. sample), but some differences in subtest variances were found. These findings suggest that the PASS theory, as measured by CAS, yields similar mean scores and showed factorial invariance for these samples of Italian and American children, who differ on cultural and linguistic characteristics. PMID:22984802

  3. Neutron scattering in Australia

    SciTech Connect

    Knott, R.B.

    1994-12-31

    Neutron scattering techniques have been part of the Australian scientific research community for the past three decades. The High Flux Australian Reactor (HIFAR) is a multi-use facility of modest performance that provides the only neutron source in the country suitable for neutron scattering. The limitations of HIFAR have been recognized and recently a Government initiated inquiry sought to evaluate the future needs of a neutron source. In essence, the inquiry suggested that a delay of several years would enable a number of key issues to be resolved, and therefore a more appropriate decision made. In the meantime, use of the present source is being optimized, and where necessary research is being undertaken at major overseas neutron facilities either on a formal or informal basis. Australia has, at present, a formal agreement with the Rutherford Appleton Laboratory (UK) for access to the spallation source ISIS. Various aspects of neutron scattering have been implemented on HIFAR, including investigations of the structure of biological relevant molecules. One aspect of these investigations will be presented. Preliminary results from a study of the interaction of the immunosuppressant drug, cyclosporin-A, with reconstituted membranes suggest that the hydrophobic drug interdigitated with lipid chains.

  4. Coded source neutron imaging

    SciTech Connect

    Bingham, Philip R; Santos-Villalobos, Hector J

    2011-01-01

    Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100 m) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100um and 10um aperture hole diameters show resolutions matching the hole diameters.

  5. Neutron Nucleic Acid Crystallography.

    PubMed

    Chatake, Toshiyuki

    2016-01-01

    The hydration shells surrounding nucleic acids and hydrogen-bonding networks involving water molecules and nucleic acids are essential interactions for the structural stability and function of nucleic acids. Water molecules in the hydration shells influence various conformations of DNA and RNA by specific hydrogen-bonding networks, which often contribute to the chemical reactivity and molecular recognition of nucleic acids. However, X-ray crystallography could not provide a complete description of structural information with respect to hydrogen bonds. Indeed, X-ray crystallography is a powerful tool for determining the locations of water molecules, i.e., the location of the oxygen atom of H2O; however, it is very difficult to determine the orientation of the water molecules, i.e., the orientation of the two hydrogen atoms of H2O, because X-ray scattering from the hydrogen atom is very small.Neutron crystallography is a specialized tool for determining the positions of hydrogen atoms. Neutrons are not diffracted by electrons, but are diffracted by atomic nuclei; accordingly, neutron scattering lengths of hydrogen and its isotopes are comparable to those of non-hydrogen atoms. Therefore, neutron crystallography can determine both of the locations and orientations of water molecules. This chapter describes the current status of neutron nucleic acid crystallographic research as well as the basic principles of neutron diffraction experiments performed on nucleic acid crystals: materials, crystallization, diffraction experiments, and structure determination. PMID:26227050

  6. MAGNETIC NEUTRON SCATTERING

    SciTech Connect

    ZALIZNYAK,I.A.; LEE,S.H.

    2004-07-30

    Much of our understanding of the atomic-scale magnetic structure and the dynamical properties of solids and liquids was gained from neutron-scattering studies. Elastic and inelastic neutron spectroscopy provided physicists with an unprecedented, detailed access to spin structures, magnetic-excitation spectra, soft-modes and critical dynamics at magnetic-phase transitions, which is unrivaled by other experimental techniques. Because the neutron has no electric charge, it is an ideal weakly interacting and highly penetrating probe of matter's inner structure and dynamics. Unlike techniques using photon electric fields or charged particles (e.g., electrons, muons) that significantly modify the local electronic environment, neutron spectroscopy allows determination of a material's intrinsic, unperturbed physical properties. The method is not sensitive to extraneous charges, electric fields, and the imperfection of surface layers. Because the neutron is a highly penetrating and non-destructive probe, neutron spectroscopy can probe the microscopic properties of bulk materials (not just their surface layers) and study samples embedded in complex environments, such as cryostats, magnets, and pressure cells, which are essential for understanding the physical origins of magnetic phenomena. Neutron scattering is arguably the most powerful and versatile experimental tool for studying the microscopic properties of the magnetic materials. The magnitude of the cross-section of the neutron magnetic scattering is similar to the cross-section of nuclear scattering by short-range nuclear forces, and is large enough to provide measurable scattering by the ordered magnetic structures and electron spin fluctuations. In the half-a-century or so that has passed since neutron beams with sufficient intensity for scattering applications became available with the advent of the nuclear reactors, they have became indispensable tools for studying a variety of important areas of modern science

  7. Neutron dosimetry in boron neutron capture therapy

    SciTech Connect

    Fairchild, R.G.; Miola, U.J.; Ettinger, K.V.

    1981-01-01

    The recent development of various borated compounds and the utilization of one of these (Na/sub 2/B/sub 12/H/sub 11/SH) to treat brain tumors in clinical studies in Japan has renewed interest in neutron capture therapy. In these procedures thermal neutrons interact with /sup 10/B in boron containing cells through the /sup 10/B(n,..cap alpha..)/sup 7/Li reaction producing charged particles with a maximum range of approx. 10..mu..m in tissue. Borated analogs of chlorpromazine, porphyrin, thiouracil and deoxyuridine promise improved tumor uptake and blood clearance. The therapy beam from the Medical Research Reactor in Brookhaven contains neutrons from a modified and filtered fission spectrum and dosimetric consequences of the use of the above mentioned compounds in conjunction with thermal and epithermal fluxes are discussed in the paper. One of the important problems of radiation dosimetry in capture therapy is determination of the flux profile and, hence, the dose profile in the brain. This has been achieved by constructing a brain phantom made of TE plastic. The lyoluminescence technique provides a convenient way of monitoring the neutron flux distributions; the detectors for this purpose utilize /sup 6/Li and /sup 10/B compounds. Such compounds have been synthesized specially for the purpose of dosimetry of thermal and epithermal beams. In addition, standard lyoluminescent phosphors, like glutamine, could be used to determine the collisional component of the dose as well as the contribution of the /sup 14/N(n,p)/sup 14/C reaction. Measurements of thermal flux were compared with calculations and with measurements done with activation foils.

  8. Final evaluation of characterizing pipe-over-pack containers using high efficiency neutron counters

    SciTech Connect

    Carson, Pete; Stanfield, Sean B; Wachter, Joe; Cramer, Doug; Harvill, Joe

    2009-01-01

    Nondestructive assay (NDA) measurements of Transuranic (TRU) waste at Los Alamos National Laboratory (LANL) packed in Pipe-over-Pack Containers (POC) contain a number of complexities. The POC is highly attenuating to both gamma rays and neutrons which presents a difficult waste matrix for correct quantification of material in the container. Currently there are a number ofPOC containers at LANL that require evaluation for shipment to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. Updated data has been evaluated that finalizes the evaluation of characterizing Pipe-Over-Pack Containers. Currently at LANL, a single instrument has been used to explore the appropriateness of both passive neutron and quantitative gamma ray methods for measuring POC's. The passive neutron approach uses the Reals coincidence count rate to establish plutonium mass and other parameters of interest for TRU waste. The quantitative gamma ray method assumes a homogeneous distribution of radioactive source material with the surrounding material throughout the drum volume. Drums are assayed with a calibration based on the known density of the matrix. Both methods are supplemented by a simultaneous isotopic measurement using Multi-Group Analysis (MGA) to determine the plutonium isotopic composition. If MGA fails to provide a viable isotopic result Fixed Energy Response function Analysis with Multiple efficiencies (FRAM) has been used to replace the MGA results. Acceptable Knowledge (AK) may also be used in certain instances. This report will discuss the two methods in detail. Included in the discussion will be descriptions of the setup parameters and calibration techniques for the instrument. A number of test measurements have been performed to compare HENC data with certified historical data. Empty POCs loaded with known sources have also been measured to determine the viability of the technique. A comparison between calorimetry data, historical measurements and HENC data will also be

  9. Fundamental neutron physics at LANSCE

    SciTech Connect

    Greene, G.

    1995-10-01

    Modern neutron sources and science share a common origin in mid-20th-century scientific investigations concerned with the study of the fundamental interactions between elementary particles. Since the time of that common origin, neutron science and the study of elementary particles have evolved into quite disparate disciplines. The neutron became recognized as a powerful tool for studying condensed matter with modern neutron sources being primarily used (and justified) as tools for neutron scattering and materials science research. The study of elementary particles has, of course, led to the development of rather different tools and is now dominated by activities performed at extremely high energies. Notwithstanding this trend, the study of fundamental interactions using neutrons has continued and remains a vigorous activity at many contemporary neutron sources. This research, like neutron scattering research, has benefited enormously by the development of modern high-flux neutron facilities. Future sources, particularly high-power spallation sources, offer exciting possibilities for continuing this research.

  10. Novel neutron detectors

    NASA Astrophysics Data System (ADS)

    Burgett, Eric Anthony

    A new set of thermal neutron detectors has been developed as a near term 3He tube replacement. The zinc oxide scintillator is an ultrafast scintillator which can be doped to have performance equal to or superior to 3He tubes. Originally investigated in the early 1950s, this room temperature semiconductor has been evaluated as a thermal neutron scintillator. Zinc oxide can be doped with different nuclei to tune the band gap, improve optical clarity, and improve the thermal neutron detection efficiency. The effects of various dopant effects on the scintillation properties, materials properties, and crystal growth parameters have been analyzed. Two different growth modalities were investigated: bulk melt grown materials as well as thin film scintillators grown by metalorganic chemical vapor deposition (MOCVD). MOCVD has shown significant advantages including precise thickness control, high dopant incorporation, and epitaxial coatings of neutron target nuclei. Detector designs were modeled and simulated to design an improved thermal neutron detector using doped ZnO layers, conformal coatings and light collection improvements including Bragg reflectors and photonic crystal structures. The detectors have been tested for crystalline quality by XRD and FTIR spectroscopy, for scintillation efficiency by photo-luminescence spectroscopy, and for neutron detection efficiency by alpha and neutron radiation tests. Lastly, a novel method for improving light collection efficiency has been investigated, the creation of a photonic crystal scintillator. Here, the flow of optical light photons is controlled through an engineered structure created with the scintillator materials. This work has resulted in a novel radiation detection material for the near term replacement of 3He tubes with performance characteristics equal to or superior to that of 3He.

  11. Laser generated neutron source for neutron resonance spectroscopy

    SciTech Connect

    Higginson, D. P.; Bartal, T.; McNaney, J. M.; Swift, D. C.; Hey, D. S.; Le Pape, S.; Mackinnon, A.; Kodama, R.; Tanaka, K. A.; Mariscal, D.; Beg, F. N.; Nakamura, H.; Nakanii, N.

    2010-10-15

    A neutron source for neutron resonance spectroscopy has been developed using high-intensity, short-pulse lasers. This technique will allow robust measurement of interior ion temperature of laser-shocked materials and provide insight into material equation of state. The neutron generation technique uses laser-accelerated protons to create neutrons in LiF through (p,n) reactions. The incident proton beam has been diagnosed using radiochromic film. This distribution is used as the input for a (p,n) neutron prediction code which is validated with experimentally measured neutron yields. The calculation infers a total fluence of 1.8x10{sup 9} neutrons, which are expected to be sufficient for neutron resonance spectroscopy temperature measurements.

  12. The Fundamental Neutron Physics Beamline at the Spallation Neutron Source

    PubMed Central

    Greene, Geoffrey; Cianciolo, Vince; Koehler, Paul; Allen, Richard; Snow, William Michael; Huffman, Paul; Gould, Chris; Bowman, David; Cooper, Martin; Doyle, John

    2005-01-01

    The Spallation Neutron Source (SNS), currently under construction at Oak Ridge National Laboratory with an anticipated start-up in early 2006, will provide the most intense pulsed beams of cold neutrons in the world. At a projected power of 1.4 MW, the time averaged fluxes and fluences of the SNS will approach those of high flux reactors. One of the flight paths on the cold, coupled moderator will be devoted to fundamental neutron physics. The fundamental neutron physics beamline is anticipated to include two beam-lines; a broad band cold beam, and a monochromatic beam of 0.89 nm neutrons for ultracold neutron (UCN) experiments. The fundamental neutron physics beamline will be operated as a user facility with experiment selection based on a peer reviewed proposal process. An initial program of five experiments in neutron decay, hadronic weak interaction and time reversal symmetry violation have been proposed. PMID:27308112

  13. A multitask neutron beam line for spallation neutron sources

    NASA Astrophysics Data System (ADS)

    Pietropaolo, A.; Festa, G.; Grazzi, F.; Barzagli, E.; Scherillo, A.; Schooneveld, E. M.; Civita, F.

    2011-08-01

    Here we present a new concept for a time-of-flight neutron scattering instrument allowing for simultaneous application of three different techniques: time-of-flight neutron diffraction, neutron resonance capture analysis and Bragg edge transmission analysis. The instrument can provide average resolution neutron radiography too. The potential of the proposed concept was explored by implementing the necessary equipment on INES (Italian Neutron Experimental Station) at the ISIS spallation neutron source (UK). The results obtained show the effectiveness of the proposed instrument to acquire relevant quantitative information in a non-invasive way on a historical metallurgical sample, namely a Japanese hand guard (tsuba). The aforementioned neutron techniques simultaneously exploited the extended neutron energy range available from 10 meV to 1 keV. This allowed a fully satisfactory characterization of the sample in terms of metal components and their combination in different phases, and forging and assembling methods.

  14. Imaging with cold neutrons

    NASA Astrophysics Data System (ADS)

    Lehmann, E. H.; Kaestner, A.; Josic, L.; Hartmann, S.; Mannes, D.

    2011-09-01

    Neutrons for imaging purposes are provided mainly from thermal beam lines at suitable facilities around the world. The access to cold neutrons is presently limited to very few places only. However, many challenging options for imaging with cold neutrons have been found out, given by the interaction behavior of the observed materials with neutrons in the cold energy range (3-10 Å). For absorbing materials, the interaction probability increases proportionally with the wavelength with the consequence of more contrast but less transmission with cold neutrons. Many materials are predominantly scattering neutrons, in particular most of crystalline structural materials. In these cases, cold neutrons play an important role by covering the energy range of the most important Bragg edges given by the lattice planes of the crystallites. This particular behavior can be used for at least two important aspects—choosing the right energy of the initial beam enables to have a material more or less transparent, and a direct macroscopic visualization of the crystalline structure and its change in a manufacturing process. Since 2006, PSI operates its second beam line for neutron imaging, where cold neutrons are provided from a liquid deuterium cold source (operated at 25 K). It has been designed to cover the most current aspects in neutron imaging research with the help of high flexibility. This has been done with changeable inlet apertures, a turbine based velocity selector, two beam positions and variable detector systems, satisfying the demands of the individual investigation. The most important detection system was found to be a micro-tomography system that enables studies in the presently best spatial resolution. In this case, the high contrast from the sample interaction process and the high detection probability for the cold neutrons combines in an ideal combination for the best possible performance. Recently, it was found out that the energy selective studies might become a

  15. The fast neutron fluence and the activation detector activity calculations using the effective source method and the adjoint function

    SciTech Connect

    Hep, J.; Konecna, A.; Krysl, V.; Smutny, V.

    2011-07-01

    This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weighting is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with few group method calculation using the active core calculation code MOBY-DICK. The follow-up neutron transport calculation was performed using the neutron transport code TORT to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code TORT and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV

  16. The ambiguous neutron

    NASA Astrophysics Data System (ADS)

    Hawes, Joan L.

    1980-09-01

    The ways in which a neutron may be described suggest that it is a particle; is a wave; has no electric charge; has a spin magnetic moment similar to that of an electron and a proton; is a stable fundamental unit of matter; and has a halflife of approximately 12 min. These are only some of the seemingly ambiguous properties of a very remarkable entity. Mostly-the machinations of wave mechanics notwithstanding-there seems little doubt that the neutron is imagined to be a particle. It is probably regarded as a very small, round, invisible object which has no electric charge and resides in the atomic nucleus. Indeed, the fact that without it stable nuclei cannot exist seems paradoxically allied to the statement that neither can radioactive ones. Again, a certain ambiguity is evident in the notion that any electrically neutral entity can show magnetic properties. And, if it is the force effects of the neutron that underline its role as a fundamental building brick of matter, how does it exert these forces and remain uncharged? Many of the solutions to these and other questions and propositions about the neutron are of relatively recent history; some still remain hidden-the precise nature of the neutron's forces of interaction for example. But the search to understanding lies in the same realm of patient experimental and theoretical enquiry that embodied its initial discovery by James Chadwick in 1932.

  17. Origin of Neutron Stars

    NASA Astrophysics Data System (ADS)

    Brecher, K.

    1999-12-01

    The origin of the concept of neutron stars can be traced to two brief, incredibly insightful publications. Work on the earlier paper by Lev Landau (Phys. Z. Sowjetunion, 1, 285, 1932) actually predated the discovery of neutrons. Nonetheless, Landau arrived at the notion of a collapsed star with the density of a nucleus (really a "nucleus star") and demonstrated (at about the same time as, and independent of, Chandrasekhar) that there is an upper mass limit for dense stellar objects of about 1.5 solar masses. Perhaps even more remarkable is the abstract of a talk presented at the December 1933 meeting of the American Physical Society published by Walter Baade and Fritz Zwicky in 1934 (Phys. Rev. 45, 138). It followed the discovery of the neutron by just over a year. Their report, which was about the same length as the present abstract: (1) invented the concept and word supernova; (2) suggested that cosmic rays are produced by supernovae; and (3) in the authors own words, proposed "with all reserve ... the view that supernovae represent the transitions from ordinary stars to neutron stars (italics), which in their final stages consist of extremely closely packed neutrons." The abstract by Baade and Zwicky probably contains the highest density of new, important (and correct) ideas in high energy astrophysics ever published in a single paper. In this talk, we will discuss some of the facts and myths surrounding these two publications.

  18. Apollo 16 neutron stratigraphy.

    NASA Technical Reports Server (NTRS)

    Russ, G. P., III

    1973-01-01

    The Apollo 16 soils have the largest low-energy neutron fluences yet observed in lunar samples. Variations in the isotopic ratios Gd-158/Gd-157 and Sm-150/Sm-149 (up to 1.9 and 2.0%, respectively) indicate that the low-energy neutron fluence in the Apollo 16 drill stem increases with depth throughout the section sampled. Such a variation implies that accretion has been the dominant regolith 'gardening' process at this location. The data may be fit by a model of continuous accretion of pre-irradiated material or by models involving as few as two slabs of material in which the first slab could have been deposited as long as 1 b.y. ago. The ratio of the number of neutrons captured per atom by Sm to the number captured per atom by Gd is lower than in previously measured lunar samples, which implies a lower energy neutron spectrum at this site. The variation of this ratio with chemical composition is qualitatively similar to that predicted by Lingenfelter et al. (1972). Variations are observed in the ratio Gd-152/Gd-160 which are fluence-correlated and probably result from neutron capture by Eu-151.

  19. Twisting Neutron Waves

    NASA Astrophysics Data System (ADS)

    Pushin, Dmitry

    Most waves encountered in nature can be given a ``twist'', so that their phase winds around an axis parallel to the direction of wave propagation. Such waves are said to possess orbital angular momentum (OAM). For quantum particles such as photons, atoms, and electrons, this corresponds to the particle wavefunction having angular momentum of Lℏ along its propagation axis. Controlled generation and detection of OAM states of photons began in the 1990s, sparking considerable interest in applications of OAM in light and matter waves. OAM states of photons have found diverse applications such as broadband data multiplexing, massive quantum entanglement, optical trapping, microscopy, quantum state determination and teleportation, and interferometry. OAM states of electron beams have been used to rotate nanoparticles, determine the chirality of crystals and for magnetic microscopy. Here I discuss the first demonstration of OAM control of neutrons. Using neutron interferometry with a spatially incoherent input beam, we show the addition and conservation of quantum angular momenta, entanglement between quantum path and OAM degrees of freedom. Neutron-based quantum information science heretofore limited to spin, path, and energy degrees of freedom, now has access to another quantized variable, and OAM modalities of light, x-ray, and electron beams are extended to a massive, penetrating neutral particle. The methods of neutron phase imprinting demonstrated here expand the toolbox available for development of phase-sensitive techniques of neutron imaging. Financial support provided by the NSERC Create and Discovery programs, CERC and the NIST Quantum Information Program is acknowledged.

  20. Hybrid superconducting neutron detectors

    SciTech Connect

    Merlo, V.; Lucci, M.; Ottaviani, I.; Salvato, M.; Cirillo, M.; Scherillo, A.; Celentano, G.; Pietropaolo, A.

    2015-03-16

    A neutron detection concept is presented that is based on superconductive niobium (Nb) strips coated by a boron (B) layer. The working principle of the detector relies on the nuclear reaction, {sup 10}B + n → α + {sup 7}Li, with α and Li ions generating a hot spot on the current-biased Nb strip which in turn induces a superconducting-normal state transition. The latter is recognized as a voltage signal which is the evidence of the incident neutron. The above described detection principle has been experimentally assessed and verified by irradiating the samples with a pulsed neutron beam at the ISIS spallation neutron source (UK). It is found that the boron coated superconducting strips, kept at a temperature T = 8 K and current-biased below the critical current I{sub c}, are driven into the normal state upon thermal neutron irradiation. As a result of the transition, voltage pulses in excess of 40 mV are measured while the bias current can be properly modulated to bring the strip back to the superconducting state, thus resetting the detector. Measurements on the counting rate of the device are presented and the basic physical features of the detector are discussed.

  1. SUSANS With Polarized Neutrons

    PubMed Central

    Wagh, Apoorva G.; Rakhecha, Veer Chand; Strobl, Makus; Treimer, Wolfgang

    2005-01-01

    Super Ultra-Small Angle Neutron Scattering (SUSANS) studies over wave vector transfers of 10–4 nm–1 to 10–3 nm–1 afford information on micrometer-size agglomerates in samples. Using a right-angled magnetic air prism, we have achieved a separation of ≈10 arcsec between ≈2 arcsec wide up- and down-spin peaks of 0.54 nm neutrons. The SUSANS instrument has thus been equipped with the polarized neutron option. The samples are placed in a uniform vertical field of 8.8 × 104 A/m (1.1 kOe). Several magnetic alloy ribbon samples broaden the up-spin neutron peak significantly over the ±1.3 × 10–3 nm–1 range, while leaving the down-spin peak essentially unaltered. Fourier transforms of these SUSANS spectra corrected for the instrument resolution, yield micrometer-range pair distribution functions for up- and down-spin neutrons as well as the nuclear and magnetic scattering length density distributions in the samples. PMID:27308127

  2. Hybrid superconducting neutron detectors

    NASA Astrophysics Data System (ADS)

    Merlo, V.; Salvato, M.; Cirillo, M.; Lucci, M.; Ottaviani, I.; Scherillo, A.; Celentano, G.; Pietropaolo, A.

    2015-03-01

    A neutron detection concept is presented that is based on superconductive niobium (Nb) strips coated by a boron (B) layer. The working principle of the detector relies on the nuclear reaction, 10B + n → α + 7Li, with α and Li ions generating a hot spot on the current-biased Nb strip which in turn induces a superconducting-normal state transition. The latter is recognized as a voltage signal which is the evidence of the incident neutron. The above described detection principle has been experimentally assessed and verified by irradiating the samples with a pulsed neutron beam at the ISIS spallation neutron source (UK). It is found that the boron coated superconducting strips, kept at a temperature T = 8 K and current-biased below the critical current Ic, are driven into the normal state upon thermal neutron irradiation. As a result of the transition, voltage pulses in excess of 40 mV are measured while the bias current can be properly modulated to bring the strip back to the superconducting state, thus resetting the detector. Measurements on the counting rate of the device are presented and the basic physical features of the detector are discussed.

  3. The coupling of the neutron transport application RATTLESNAKE to the nuclear fuels performance application BISON under the MOOSE framework

    SciTech Connect

    Gleicher, Frederick N.; Williamson, Richard L.; Ortensi, Javier; Wang, Yaqi; Spencer, Benjamin W.; Novascone, Stephen R.; Hales, Jason D.; Martineau, Richard C.

    2014-10-01

    The MOOSE neutron transport application RATTLESNAKE was coupled to the fuels performance application BISON to provide a higher fidelity tool for fuel performance simulation. This project is motivated by the desire to couple a high fidelity core analysis program (based on the self-adjoint angular flux equations) to a high fidelity fuel performance program, both of which can simulate on unstructured meshes. RATTLESNAKE solves self-adjoint angular flux transport equation and provides a sub-pin level resolution of the multigroup neutron flux with resonance treatment during burnup or a fast transient. BISON solves the coupled thermomechanical equations for the fuel on a sub-millimeter scale. Both applications are able to solve their respective systems on aligned and unaligned unstructured finite element meshes. The power density and local burnup was transferred from RATTLESNAKE to BISON with the MOOSE Multiapp transfer system. Multiple depletion cases were run with one-way data transfer from RATTLESNAKE to BISON. The eigenvalues are shown to agree well with values obtained from the lattice physics code DRAGON. The one-way data transfer of power density is shown to agree with the power density obtained from an internal Lassman-style model in BISON.

  4. Neutron beam characterization at the Neutron Radiography Reactor (NRAD)

    SciTech Connect

    Imel, G.R.; Urbatsch, T.; Pruett, D.P.; Ross, J.R.

    1990-01-01

    The Neutron Radiography Reactor (NRAD) is a 250-kW TRIGA Reactor operated by Argonne National Laboratory and is located near Idaho Falls, Idaho. The reactor and its facilities regarding radiography are detailed in another paper at this conference; this paper summarizes neutron flux measurements and calculations that have been performed to better understand and potentially improve the neutronics characteristics of the reactor.

  5. Neutronic reactor construction

    DOEpatents

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  6. Hyperons in Neutron Stars

    NASA Astrophysics Data System (ADS)

    Vidaña, Isaac

    2016-01-01

    In this work I briefly review some of the effects of hyperons on the properties of neutron and proto-neutron stars. In particular, I revise the problem of the strong softening of the EoS, and the consequent reduction of the maximum mass, induced by the presence of hyperons, a puzzle which has become more intringuing and difficult to solve because of the recent measurements of the unusually high masses of the millisecond pulsars PSR J1903+0327 (1.667 ± 0.021M⊙), PSR J1614-2230 (1.97 ± 0.04M⊙), and PSR J0348+0432 (2.01 ± 0.04M⊙). Some of the solutions proposed to tackle this problem are discussed. Finally, I re-examine also the role of hyperons on the cooling properties of newly born neutron stars and on the so-called r-mode instability.

  7. Ultrafast neutron detector

    DOEpatents

    Wang, Ching L.

    1987-01-01

    The invention comprises a neutron detector (50) of very high temporal resolution that is particularly well suited for measuring the fusion reaction neutrons produced by laser-driven inertial confinement fusion targets. The detector comprises a biased two-conductor traveling-wave transmission line (54, 56, 58, 68) having a uranium cathode (60) and a phosphor anode (62) as respective parts of the two conductors. A charge line and Auston switch assembly (70, 72, 74) launch an electric field pulse along the transmission line. Neutrons striking the uranium cathode at a location where the field pulse is passing, are enabled to strike the phosphor anode and produce light that is recorded on photographic film (64). The transmission line may be variously configured to achieve specific experimental goals.

  8. NEUTRONIC REACTOR CORE INSTRUMENT

    DOEpatents

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  9. METHOD OF PRODUCING NEUTRONS

    DOEpatents

    Imhoff, D.H.; Harker, W.H.

    1964-02-01

    A method for producing neutrons is described in which there is employed a confinement zone defined between longitudinally spaced localized gradient regions of an elongated magnetic field. Changed particles and neutralizing electrons, more specifically deuterons and tritons and neutralizng electrons, are injected into the confinement field from ion sources located outside the field. The rotational energy of the parrticles is increased at the gradients by imposing an oscillating transverse electrical field thereacross. The imposition of such oscillating transverse electrical fields improves the reflection capability of such gradient fielda so that the reactive particles are retained more effectively within the zone. With the attainment of appropriate densities of plasma particles and provided that such particles are at a sufficiently high temperature, neutron-producing reactions ensue and large quantities of neutrons emerge from the containment zone. (AEC)

  10. Personnel electronic neutron dosimeter

    DOEpatents

    Falk, Roger B.; Tyree, William H.

    1984-12-18

    A personnel electronic dosimeter includes a neutron-proton and neutron-alpha converter for providing an electrical signal having a magnitude proportional to the energy of a detected proton or alpha particle produced from the converter, a pulse generator circuit for generating a pulse having a duration controlled by the weighed effect of the amplitude of the electrical signal, an oscillator enabled by the pulse for generating a train of clock pulses for a time dependent upon the pulse length, a counter for counting the clock pulses, and an indicator for providing a direct reading and aural alarm when the count indicates that the wearer has been exposed to a selected level of neutron dose equivalent.

  11. Personnel electronic neutron dosimeter

    DOEpatents

    Falk, R.B.; Tyree, W.H.

    1982-03-03

    A personnel electronic dosimeter includes a neutron-proton and neutron-alpha converter for providing an electrical signal having a magnitude proportional to the energy of a detected proton or alpha particle produced from the converter, a pulse generator circuit for generating a pulse having a duration controlled by the weighed effect of the amplitude of the electrical signal, an oscillator enabled by the pulse for generating a train of clock pulses for a time dependent upon the pulse length, a counter for counting the clock pulses, and an indicator for providing a direct reading and aural alarm when the count indicates that the wearer has been exposed to a selected level of neutron dose equivalent.

  12. Spherical neutron generator

    DOEpatents

    Leung, Ka-Ngo

    2006-11-21

    A spherical neutron generator is formed with a small spherical target and a spherical shell RF-driven plasma ion source surrounding the target. A deuterium (or deuterium and tritium) ion plasma is produced by RF excitation in the plasma ion source using an RF antenna. The plasma generation region is a spherical shell between an outer chamber and an inner extraction electrode. A spherical neutron generating target is at the center of the chamber and is biased negatively with respect to the extraction electrode which contains many holes. Ions passing through the holes in the extraction electrode are focused onto the target which produces neutrons by D-D or D-T reactions.

  13. Hyperons and neutron stars

    SciTech Connect

    Vidaña, Isaac

    2015-02-24

    In this lecture I will briefly review some of the effects of hyperons on the properties of neutron and proto-neutron stars. In particular, I will revise the problem of the strong softening of the EoS, and the consequent reduction of the maximum mass, induced by the presence of hyperons, a puzzle which has become more intringuing and difficult to solve due the recent measurements of the unusually high masses of the millisecond pulsars PSR J1903+0327 (1.667±0.021M{sub ⊙}), PSR J1614–2230 (1.97±0.04M{sub ⊙}), and PSR J0348+0432 (2.01±0.04M{sub ⊙}). Finally, I will also examine the role of hyperons on the cooling properties of newly born neutron stars and on the so-called r-mode instability.

  14. Neutron Star Compared to Manhattan

    NASA Video Gallery

    A pulsar is a neutron star, the crushed core of a star that has exploded. Neutron stars crush half a million times more mass than Earth into a sphere no larger than Manhattan, as animated in this s...

  15. POLARIZED NEUTRONS IN RHIC

    SciTech Connect

    COURANT,E.D.

    1998-04-27

    There does not appear to be any obvious way to accelerate neutrons, polarized or otherwise, to high energies by themselves. To investigate the behavior of polarized neutrons the authors therefore have to obtain them by accelerating them as components of heavier nuclei, and then sorting out the contribution of the neutrons in the analysis of the reactions produced by the heavy ion beams. The best neutron carriers for this purpose are probably {sup 3}He nuclei and deuterons. A polarized deuteron is primarily a combination of a proton and a neutron with their spins pointing in the same direction; in the {sup 3}He nucleus the spins of the two protons are opposite and the net spin (and magnetic moment) is almost the same as that of a free neutron. Polarized ions other than protons may be accelerated, stored and collided in a ring such as RHIC provided the techniques proposed for polarized proton operation can be adapted (or replaced by other strategies) for these ions. To accelerate polarized particles in a ring, one must make provisions for overcoming the depolarizing resonances that occur at certain energies. These resonances arise when the spin tune (ratio of spin precession frequency to orbit frequency) resonates with a component present in the horizontal field. The horizontal field oscillates with the vertical motion of the particles (due to vertical focusing); its frequency spectrum is dominated by the vertical oscillation frequency and its modulation by the periodic structure of the accelerator ring. In addition, the magnet imperfections that distort the closed orbit vertically contain all integral Fourier harmonics of the orbit frequency.

  16. Corrosion resistant neutron absorbing coatings

    SciTech Connect

    Choi, Jor-Shan; Farmer, Joseph C.; Lee, Chuck K.; Walker, Jeffrey; Russell, Paige; Kirkwood, Jon; Yang, Nancy; Champagne, Victor

    2012-05-29

    A method of forming a corrosion resistant neutron absorbing coating comprising the steps of spray or deposition or sputtering or welding processing to form a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material. Also a corrosion resistant neutron absorbing coating comprising a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material.

  17. Corrosion resistant neutron absorbing coatings

    SciTech Connect

    Choi, Jor-Shan; Farmer, Joseph C; Lee, Chuck K; Walker, Jeffrey; Russell, Paige; Kirkwood, Jon; Yang, Nancy; Champagne, Victor

    2013-11-12

    A method of forming a corrosion resistant neutron absorbing coating comprising the steps of spray or deposition or sputtering or welding processing to form a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material. Also a corrosion resistant neutron absorbing coating comprising a composite material made of a spray or deposition or sputtering or welding material, and a neutron absorbing material.

  18. Neutron densities from muon capture

    NASA Astrophysics Data System (ADS)

    Huan Ching, Chiang; Oset, Eulogio

    1991-10-01

    We show that, because of Pauli blocking and renormalization of the weak currents in nuclei, the muon capture rates are rather sensitive to the neutron distributions. We also show that, because of intrinsic theoretical uncertainties, neutron radia cannot be determined with precision but some reasonable limits can be given. However, the ratio of capture rates in different isotopes serves to determine the neutron radii of the isotopes provided the neutron density distribution for one of them is known.

  19. The neutron star zoo

    NASA Astrophysics Data System (ADS)

    Harding, Alice K.

    2013-12-01

    Neutron stars are a very diverse population, both in their observational and their physical properties. They prefer to radiate most of their energy at X-ray and gamma-ray wavelengths. But whether their emission is powered by rotation, accretion, heat, magnetic fields or nuclear reactions, they are all different species of the same animal whose magnetic field evolution and interior composition remain a mystery. This article will broadly review the properties of inhabitants of the neutron star zoo, with emphasis on their high-energy emission.

  20. GUIDE FOR POLARIZED NEUTRONS

    DOEpatents

    Sailor, V.L.; Aichroth, R.W.

    1962-12-01

    The plane of polarization of a beam of polarized neutrons is changed by this invention, and the plane can be flipped back and forth quicitly in two directions in a trouble-free manner. The invention comprises a guide having a plurality of oppositely directed magnets forming a gap for the neutron beam and the gaps are spaced longitudinally in a spiral along the beam at small stepped angles. When it is desired to flip the plane of polarization the magnets are suitably rotated to change the direction of the spiral of the gaps. (AEC)

  1. Simplified fast neutron dosimeter

    DOEpatents

    Sohrabi, Mehdi

    1979-01-01

    Direct fast-neutron-induced recoil and alpha particle tracks in polycarbonate films may be enlarged for direct visual observation and automated counting procedures employing electrochemical etching techniques. Electrochemical etching is, for example, carried out in a 28% KOH solution at room temperature by applying a 2000 V peak-to-peak voltage at 1 kHz frequency. Such recoil particle amplification can be used for the detection of wide neutron dose ranges from 1 mrad. to 1000 rads. or higher, if desired.

  2. Neutrinos from neutron stars

    NASA Technical Reports Server (NTRS)

    Helfand, D. J.

    1979-01-01

    A calculation of the flux of ultra-high energy neutrinos from galactic neutron stars is presented. The calculation is used to determine the number of point sources detectable at the sensitivity threshold of a proposed deep underwater muon and neutrino detector array. The detector array would have a point source detection threshold of about 100 eV/sq cm-sec. Analysis of neutrino luminosities and the number of detectable sources suggests that the deep underwater detector may make a few discoveries. In particular, a suspected neutron star in the Cyg X-3 source seems a promising target for the deep underwater array.

  3. Neutron Imaging Camera

    NASA Technical Reports Server (NTRS)

    Hunter, Stanley; deNolfo, G. A.; Barbier, L. M.; Link, J. T.; Son, S.; Floyd, S. R.; Guardala, N.; Skopec, M.; Stark, B.

    2008-01-01

    The Neutron Imaging Camera (NIC) is based on the Three-dimensional Track Imager (3DTI) technology developed at GSFC for gamma-ray astrophysics applications. The 3-DTI, a large volume time-projection chamber, provides accurate, approximately 0.4 mm resolution, 3-D tracking of charged particles. The incident direction of fast neutrons, En > 0.5 MeV, are reconstructed from the momenta and energies of the proton and triton fragments resulting from (sup 3)He(n,p) (sup 3)H interactions in the 3-DTI volume. The performance of the NIC from laboratory and accelerator tests is presented.

  4. FAST NEUTRONIC REACTOR

    DOEpatents

    Snell, A.H.

    1957-12-01

    This patent relates to a reactor and process for carrying out a controlled fast neutron chain reaction. A cubical reactive mass, weighing at least 920 metric tons, of uranium metal containing predominantly U/sup 238/ and having a U/sup 235/ content of at least 7.63% is assembled and the maximum neutron reproduction ratio is limited to not substantially over 1.01 by insertion and removal of a varying amount of boron, the reactive mass being substantially freed of moderator.

  5. NEUTRONIC REACTOR CONTROL ELEMENT

    DOEpatents

    Newson, H.W.

    1960-09-13

    A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

  6. Lunar neutron stratigraphy.

    NASA Technical Reports Server (NTRS)

    Russ, G. P., III; Burnett, D. S.; Wasserburg, G. J.

    1972-01-01

    Study of the isotopic composition of gadolinium and samarium in four soil and seven drill stem samples returned by the Apollo 15 mission. The results show the possibility to date sedimentary processes on the lunar surface for time scales of around 100 million years because of the particular dependence of neutron capture reactions on depth. The neutron flux has a distinct peak as a function of depth. This peak appears to lie below the level of shallow cratering for time scales of less than one billion years and consequently forms a readily identified marker layer of both depth and time.

  7. Fast neutron dosimetry

    SciTech Connect

    DeLuca, P.M. Jr.; Pearson, D.W.

    1992-01-01

    This progress report concentrates on two major areas of dosimetry research: measurement of fast neutron kerma factors for several elements for monochromatic and white spectrum neutron fields and determination of the response of thermoluminescent phosphors to various ultra-soft X-ray energies and beta-rays. Dr. Zhixin Zhou from the Shanghai Institute of Radiation Medicine, People's Republic of China brought with him special expertise in the fabrication and use of ultra-thin TLD materials. Such materials are not available in the USA. The rather unique properties of these materials were investigated during this grant period.

  8. New electronically black neutron detectors

    SciTech Connect

    Drake, D.M.; Feldman, W.C.; Hurlbut, C.

    1986-03-01

    Two neutron detectors are described that can function in a continuous radiation background. Both detectors identify neutrons by recording a proton recoil pulse followed by a characteristic capture pulse. This peculiar signature indicates that the neutron has lost all its energy in the scintillator. Resolutions and efficiencies have been measured for both detectors.

  9. Neutron recognition in the LAND detector for large neutron multiplicity

    NASA Astrophysics Data System (ADS)

    Pawłowski, P.; Brzychczyk, J.; Leifels, Y.; Trautmann, W.; Adrich, P.; Aumann, T.; Bacri, C. O.; Barczyk, T.; Bassini, R.; Bianchin, S.; Boiano, C.; Boretzky, K.; Boudard, A.; Chbihi, A.; Cibor, J.; Czech, B.; De Napoli, M.; Ducret, J.-E.; Emling, H.; Frankland, J. D.; Gorbinet, T.; Hellström, M.; Henzlova, D.; Hlavac, S.; Immè, J.; Iori, I.; Johansson, H.; Kezzar, K.; Kupny, S.; Lafriakh, A.; Le Fèvre, A.; Le Gentil, E.; Leray, S.; Łukasik, J.; Lühning, J.; Lynch, W. G.; Lynen, U.; Majka, Z.; Mocko, M.; Müller, W. F. J.; Mykulyak, A.; Orth, H.; Otte, A. N.; Palit, R.; Panebianco, S.; Pullia, A.; Raciti, G.; Rapisarda, E.; Rossi, D.; Salsac, M.-D.; Sann, H.; Schwarz, C.; Simon, H.; Sfienti, C.; Sümmerer, K.; Tsang, M. B.; Verde, G.; Veselsky, M.; Volant, C.; Wallace, M.; Weick, H.; Wiechula, J.; Wieloch, A.; Zwiegliński, B.

    2012-12-01

    The performance of the LAND neutron detector is studied. Using an event-mixing technique based on one-neutron data obtained in the S107 experiment at the GSI laboratory, we test the efficiency of various analytic tools used to determine the multiplicity and kinematic properties of detected neutrons. A new algorithm developed recently for recognizing neutron showers from spectator decays in the ALADIN experiment S254 is described in detail. Its performance is assessed in comparison with other methods. The properties of the observed neutron events are used to estimate the detection efficiency of LAND in this experiment.

  10. Neutron-induced peaks in Ge detectors from evaporation neutrons

    NASA Astrophysics Data System (ADS)

    Gete, E.; Measday, D. F.; Moftah, B. A.; Saliba, M. A.; Stocki, T. J.

    1997-02-01

    We have studied the peak shapes at 596 and 691 keV resulting from fast neutron interactions inside germanium detectors. We have used neutrons from a 252Cf source, as well as from the 28Si(μ -, nv), and 209Bi(π -, xn) reactions to compare the peaks and to check for a dependence of peak shape on the incoming neutron energy. In our investigation, no difference between these three measurements has been observed. In a comparison of these peak shapes with other studies, we found similar results to ours except for those measurements using monoenergetic neutrons in which a significant variation with neutron energy has been observed.

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

    SciTech Connect

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

    1997-10-01

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

  12. A comparison of acceleration methods for solving the neutron transport k-eigenvalue problem

    SciTech Connect

    Willert, Jeffrey; Park, H.; Knoll, D.A.

    2014-10-01

    Over the past several years a number of papers have been written describing modern techniques for numerically computing the dominant eigenvalue of the neutron transport criticality problem. These methods fall into two distinct categories. The first category of methods rewrite the multi-group k-eigenvalue problem as a nonlinear system of equations and solve the resulting system using either a Jacobian-Free Newton–Krylov (JFNK) method or Nonlinear Krylov Acceleration (NKA), a variant of Anderson Acceleration. These methods are generally successful in significantly reducing the number of transport sweeps required to compute the dominant eigenvalue. The second category of methods utilize Moment-Based Acceleration (or High-Order/Low-Order (HOLO) Acceleration). These methods solve a sequence of modified diffusion eigenvalue problems whose solutions converge to the solution of the original transport eigenvalue problem. This second class of methods is, in our experience, always superior to the first, as most of the computational work is eliminated by the acceleration from the LO diffusion system. In this paper, we review each of these methods. Our computational results support our claim that the choice of which nonlinear solver to use, JFNK or NKA, should be secondary. The primary computational savings result from the implementation of a HOLO algorithm. We display computational results for a series of challenging multi-dimensional test problems.

  13. Advanced nodal neutron diffusion method with space-dependent cross sections: ILLICO-VX

    SciTech Connect

    Rajic, H.L.; Ougouag, A.M.

    1987-01-01

    Advanced transverse integrated nodal methods for neutron diffusion developed since the 1970s require that node- or assembly-homogenized cross sections be known. The underlying structural heterogeneity can be accurately accounted for in homogenization procedures by the use of heterogeneity or discontinuity factors. Other (milder) types of heterogeneity, burnup-induced or due to thermal-hydraulic feedback, can be resolved by explicitly accounting for the spatial variations of material properties. This can be done during the nodal computations via nonlinear iterations. The new method has been implemented in the code ILLICO-VX (ILLICO variable cross-section method). Numerous numerical tests were performed. As expected, the convergence rate of ILLICO-VX is lower than that of ILLICO, requiring approx. 30% more outer iterations per k/sub eff/ computation. The methodology has also been implemented as the NOMAD-VX option of the NOMAD, multicycle, multigroup, two- and three-dimensional nodal diffusion depletion code. The burnup-induced heterogeneities (space dependence of cross sections) are calculated during the burnup steps.

  14. A comparison of acceleration methods for solving the neutron transport k-eigenvalue problem

    NASA Astrophysics Data System (ADS)

    Willert, Jeffrey; Park, H.; Knoll, D. A.

    2014-10-01

    Over the past several years a number of papers have been written describing modern techniques for numerically computing the dominant eigenvalue of the neutron transport criticality problem. These methods fall into two distinct categories. The first category of methods rewrite the multi-group k-eigenvalue problem as a nonlinear system of equations and solve the resulting system using either a Jacobian-Free Newton-Krylov (JFNK) method or Nonlinear Krylov Acceleration (NKA), a variant of Anderson Acceleration. These methods are generally successful in significantly reducing the number of transport sweeps required to compute the dominant eigenvalue. The second category of methods utilize Moment-Based Acceleration (or High-Order/Low-Order (HOLO) Acceleration). These methods solve a sequence of modified diffusion eigenvalue problems whose solutions converge to the solution of the original transport eigenvalue problem. This second class of methods is, in our experience, always superior to the first, as most of the computational work is eliminated by the acceleration from the LO diffusion system. In this paper, we review each of these methods. Our computational results support our claim that the choice of which nonlinear solver to use, JFNK or NKA, should be secondary. The primary computational savings result from the implementation of a HOLO algorithm. We display computational results for a series of challenging multi-dimensional test problems.

  15. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    SciTech Connect

    Woo Y. Yoon; David W. Nigg

    2011-09-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B3 or B1 zero-dimensional approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constants may be output in any of several standard formats including INL format, ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional (1-D) discrete-ordinate transport code, is incorporated into COMBINE7.1. As an option, the 167 fine-group constants generated by zero-dimensional COMBINE portion in the program can be

  16. Neutron analysis of spent fuel storage installation using parallel computing and advance discrete ordinates and Monte Carlo techniques.

    PubMed

    Shedlock, Daniel; Haghighat, Alireza

    2005-01-01

    In the United States, the Nuclear Waste Policy Act of 1982 mandated centralised storage of spent nuclear fuel by 1988. However, the Yucca Mountain project is currently scheduled to start accepting spent nuclear fuel in 2010. Since many nuclear power plants were only designed for -10 y of spent fuel pool storage, > 35 plants have been forced into alternate means of spent fuel storage. In order to continue operation and make room in spent fuel pools, nuclear generators are turning towards independent spent fuel storage installations (ISFSIs). Typical vertical concrete ISFSIs are -6.1 m high and 3.3 m in diameter. The inherently large system, and the presence of thick concrete shields result in difficulties for both Monte Carlo (MC) and discrete ordinates (SN) calculations. MC calculations require significant variance reduction and multiple runs to obtain a detailed dose distribution. SN models need a large number of spatial meshes to accurately model the geometry and high quadrature orders to reduce ray effects, therefore, requiring significant amounts of computer memory and time. The use of various differencing schemes is needed to account for radial heterogeneity in material cross sections and densities. Two P3, S12, discrete ordinate, PENTRAN (parallel environment neutral-particle TRANsport) models were analysed and different MC models compared. A multigroup MCNP model was developed for direct comparison to the SN models. The biased A3MCNP (automated adjoint accelerated MCNP) and unbiased (MCNP) continuous energy MC models were developed to assess the adequacy of the CASK multigroup (22 neutron, 18 gamma) cross sections. The PENTRAN SN results are in close agreement (5%) with the multigroup MC results; however, they differ by -20-30% from the continuous-energy MC predictions. This large difference can be attributed to the expected difference between multigroup and continuous energy cross sections, and the fact that the CASK library is based on the old ENDF

  17. Neutron reflectometry: Filling Δq with neutrons

    NASA Astrophysics Data System (ADS)

    Pleshanov, N. K.

    2016-06-01

    Luminosity of the reflectometer is defined as the neutron flux incident onto the sample surface for measurements made with a given momentum transfer resolution Δq. The filling of Δq with neutrons near a certain q depends not only on the source luminance and the source-sample tract transmittance, but also on the neutron beam tailoring. The correct choice of the working wavelength and measurements with optimum neutron beam parameters increase luminosity in several times. New optimization criteria for neutron reflectometers are suggested. Standard schemes of the reflectivity measurement with monochromatic and white beams are re-examined. Optimization of reflectivity measurements generally requires numerical calculations. Analytically, its potential is demonstrated by considering thermalized neutron beams. Such innovations as velocity selector on the basis of aperiodic multilayers, small angle Soller collimator with traps for neutrons reflected from the channel walls and fan beam time-of-flight technique are proposed to further increase the luminosity of reflectometers.

  18. The tokamak as a neutron source

    SciTech Connect

    Hendel, H.W.; Jassby, D.L.

    1989-11-01

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs.

  19. Surface Mounted Neutron Generators

    NASA Astrophysics Data System (ADS)

    Elizondo-Decanini, Juan M.

    2012-10-01

    A deuterium-tritium (DT) base reaction pulsed neutron generator packaged in a flat computer chip shape of 1.54 cm (0.600 in) wide by 3.175 cm (1.25 in) length and 0.3 cm (0.120 in) thick has been successfully demonstrated to produce 14 MeV neutrons at a rate of 10^9 neutrons per second. The neutron generator is based on a deuterium ion beam accelerated to impact a tritium loaded target. The accelerating voltage is in the 15 to 20 kV in a 3 mm (0.120 in) gap, the ion beam is shaped by using a lens design to produce a flat ion beam that conforms to the flat rectangular target. The ion source is a simple surface mounted deuterium filled titanium film with a fused gap that operates at a current-voltage design to release the deuterium during a pulse length of about 1 μs. We present the general description of the working prototypes, which we have labeled the ``NEUTRISTOR.''[4pt] Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration. Work funded by the LDRD office.

  20. NEUTRONIC REACTOR SHIELDING

    DOEpatents

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  1. Neutronic reactor thermal shield

    DOEpatents

    Wende, Charles W. J.

    1976-06-15

    1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

  2. Optimization of neutron source

    SciTech Connect

    Hooper, E.B.

    1993-11-09

    I consider here the optimization of the two component neutron source, allowing beam species and energy to vary. A simple model is developed, based on the earlier publications, that permits the optimum to be obtained simply. The two component plasma, with one species of hot ion (D{sup +} or T{sup +}) and the complementary species of cold ion, is easy to analyze in the case of a spatially uniform cold plasma, as to good approximation the total number of hot ions is important but not their spatial distribution. Consequently, the optimization can ignore spatial effects. The problem of a plasma with both types of hot ions and cold ions is rather more difficult, as the neutron production by hot-hot interactions is sensitive to their spatial distributions. Consequently, consideration of this problem will be delayed to a future memorandum. The basic model is that used in the published articles on the two-component, beam-plasma mirror source. I integrate the Fokker-Planck equation analytically, obtaining good agreement with previous numerical results. This simplifies the optimization, by providing a functional form for the neutron production. The primary result is expressed in terms of the power efficiency: watts of neutrons/watts of primary power. The latter includes the positive ion neutralization efficiency. At 150 keV, the present model obtains an efficiency of 0.66%, compared with 0.53% of the earlier calculation.

  3. Optical neutron polarizers

    SciTech Connect

    Hayter, J.B.

    1990-01-01

    A neutron wave will be refracted by an appropriately varying potential. Optical neutron polarizers use spatially varying, spin- dependent potentials to refract neutrons of opposite spin states into different directions, so that an unpolarized beam will be split into two beams of complementary polarization by such a device. This paper will concentrate on two methods of producing spin-dependent potentials which are particularly well-suited to polarizing cold neutron beams, namely thin-film structures and field-gradient techniques. Thin-film optical devices, such as supermirror multilayer structures, are usually designed to deviate only one spin-state, so that they offer the possibility of making insertion (transmission) polarizers. Very good supermirrors may now be designed and fabricated, but it is not always straightforward to design mirror-based devices which are useful in real (divergent beam) applications, and some practical configurations will be discussed. Field-gradient devices, which are usually based on multipolar magnets, have tended to be too expensive for general use, but this may change with new developments in superconductivity. Dipolar and hexapolar configurations will be considered, with emphasis on the focusing characteristics of the latter. 21 refs., 7 figs.

  4. Small Angle Neutron Scattering

    SciTech Connect

    Urban, Volker S

    2012-01-01

    Small Angle Neutron Scattering (SANS) probes structural details at the nanometer scale in a non-destructive way. This article gives an introduction to scientists who have no prior small-angle scattering knowledge, but who seek a technique that allows elucidating structural information in challenging situations that thwart approaches by other methods. SANS is applicable to a wide variety of materials including metals and alloys, ceramics, concrete, glasses, polymers, composites and biological materials. Isotope and magnetic interactions provide unique methods for labeling and contrast variation to highlight specific structural features of interest. In situ studies of a material s responses to temperature, pressure, shear, magnetic and electric fields, etc., are feasible as a result of the high penetrating power of neutrons. SANS provides statistical information on significant structural features averaged over the probed sample volume, and one can use SANS to quantify with high precision the structural details that are observed, for example, in electron microscopy. Neutron scattering is non-destructive; there is no need to cut specimens into thin sections, and neutrons penetrate deeply, providing information on the bulk material, free from surface effects. The basic principles of a SANS experiment are fairly simple, but the measurement, analysis and interpretation of small angle scattering data involves theoretical concepts that are unique to the technique and that are not widely known. This article includes a concise description of the basics, as well as practical know-how that is essential for a successful SANS experiment.

  5. Neutron-image intensifier

    NASA Technical Reports Server (NTRS)

    Berger, H.

    1970-01-01

    Electronic intensifier tube with a demagnification ratio of 9-1 enhances the usefulness of neutron-radiographic techniques. A television signal can be obtained by optical coupling of a small-output phosphor-light image to a television camera.

  6. Neutron Absorbing Alloys

    DOEpatents

    Mizia, Ronald E.; Shaber, Eric L.; DuPont, John N.; Robino, Charles V.; Williams, David B.

    2004-05-04

    The present invention is drawn to new classes of advanced neutron absorbing structural materials for use in spent nuclear fuel applications requiring structural strength, weldability, and long term corrosion resistance. Particularly, an austenitic stainless steel alloy containing gadolinium and less than 5% of a ferrite content is disclosed. Additionally, a nickel-based alloy containing gadolinium and greater than 50% nickel is also disclosed.

  7. NEUTRONIC REACTOR CONTROL ELEMENT

    DOEpatents

    Beaver, R.J.; Leitten, C.F. Jr.

    1962-04-17

    A boron-10 containing reactor control element wherein the boron-10 is dispersed in a matrix material is describeri. The concentration of boron-10 in the matrix varies transversely across the element from a minimum at the surface to a maximum at the center of the element, prior to exposure to neutrons. (AEC)

  8. Neutronic Reactor Structure

    DOEpatents

    Vernon, H. C.; Weinberg, A. M.

    1961-05-30

    The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

  9. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  10. Neutron multiplicity analysis tool

    SciTech Connect

    Stewart, Scott L

    2010-01-01

    I describe the capabilities of the EXCOM (EXcel based COincidence and Multiplicity) calculation tool which is used to analyze experimental data or simulated neutron multiplicity data. The input to the program is the count-rate data (including the multiplicity distribution) for a measurement, the isotopic composition of the sample and relevant dates. The program carries out deadtime correction and background subtraction and then performs a number of analyses. These are: passive calibration curve, known alpha and multiplicity analysis. The latter is done with both the point model and with the weighted point model. In the current application EXCOM carries out the rapid analysis of Monte Carlo calculated quantities and allows the user to determine the magnitude of sample perturbations that lead to systematic errors. Neutron multiplicity counting is an assay method used in the analysis of plutonium for safeguards applications. It is widely used in nuclear material accountancy by international (IAEA) and national inspectors. The method uses the measurement of the correlations in a pulse train to extract information on the spontaneous fission rate in the presence of neutrons from ({alpha},n) reactions and induced fission. The measurement is relatively simple to perform and gives results very quickly ({le} 1 hour). By contrast, destructive analysis techniques are extremely costly and time consuming (several days). By improving the achievable accuracy of neutron multiplicity counting, a nondestructive analysis technique, it could be possible to reduce the use of destructive analysis measurements required in safeguards applications. The accuracy of a neutron multiplicity measurement can be affected by a number of variables such as density, isotopic composition, chemical composition and moisture in the material. In order to determine the magnitude of these effects on the measured plutonium mass a calculational tool, EXCOM, has been produced using VBA within Excel. This

  11. Neutron proton crystallography station (PCS)

    SciTech Connect

    Fisher, Zoe; Kovalevsky, Andrey; Johnson, Hannah; Mustyakimov, Marat

    2009-01-01

    The PCS (Protein Crystallography Station) at Los Alamos Neutron Science Center (LANSCE) is a unique facility in the USA that is designed and optimized for detecting and collecting neutron diffraction data from macromolecular crystals. PCS utilizes the 20 Hz spallation neutron source at LANSCE to enable time-of-flight measurements using 0.6-7.0 {angstrom} neutrons. This increases the neutron flux on the sample by using a wavelength range that is optimal for studying macromolecular crystal structures. The diagram below show a schematic of PCS and photos of the detector and instrument cave.

  12. Dose-equivalent neutron dosimeter

    DOEpatents

    Griffith, R.V.; Hankins, D.E.; Tomasino, L.; Gomaa, M.A.M.

    1981-01-07

    A neutron dosimeter is disclosed which provides a single measurement 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 contaning 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.

  13. Accelerator based epithermal neutron source

    NASA Astrophysics Data System (ADS)

    Taskaev, S. Yu.

    2015-11-01

    We review the current status of the development of accelerator sources of epithermal neutrons for boron neutron capture therapy (BNCT), a promising method of malignant tumor treatment. Particular attention is given to the source of epithermal neutrons on the basis of a new type of charged particle accelerator: tandem accelerator with vacuum insulation and lithium neutron-producing target. It is also shown that the accelerator with specialized targets makes it possible to generate fast and monoenergetic neutrons, resonance and monoenergetic gamma-rays, alpha-particles, and positrons.

  14. Solid state neutron detector array

    DOEpatents

    Seidel, J.G.; Ruddy, F.H.; Brandt, C.D.; Dulloo, A.R.; Lott, R.G.; Sirianni, E.; Wilson, R.O.

    1999-08-17

    A neutron detector array is capable of measuring a wide range of neutron fluxes. The array includes multiple semiconductor neutron detectors. Each detector has a semiconductor active region that is resistant to radiation damage. In one embodiment, the array preferably has a relatively small size, making it possible to place the array in confined locations. The ability of the array to detect a wide range of neutron fluxes is highly advantageous for many applications such as detecting neutron flux during start up, ramp up and full power of nuclear reactors. 7 figs.

  15. Solid state neutron detector array

    DOEpatents

    Seidel, John G.; Ruddy, Frank H.; Brandt, Charles D.; Dulloo, Abdul R.; Lott, Randy G.; Sirianni, Ernest; Wilson, Randall O.

    1999-01-01

    A neutron detector array is capable of measuring a wide range of neutron fluxes. The array includes multiple semiconductor neutron detectors. Each detector has a semiconductor active region that is resistant to radiation damage. In one embodiment, the array preferably has a relatively small size, making it possible to place the array in confined locations. The ability of the array to detect a wide range of neutron fluxes is highly advantageous for many applications such as detecting neutron flux during start up, ramp up and full power of nuclear reactors.

  16. Fast and thermal neutron radiography

    NASA Astrophysics Data System (ADS)

    Cremer, Jay T.; Piestrup, Melvin A.; Wu, Xizeng

    2005-09-01

    There is a need for high brightness neutron sources that are portable, relatively inexpensive, and capable of neutron radiography in short imaging times. Fast and thermal neutron radiography is as an excellent method to penetrate high-density, high-Z objects, thick objects and image its interior contents, especially hydrogen-based materials. In this paper we model the expected imaging performance characteristics and limitations of fast and thermal radiography systems employing a Rose Model based transfer analysis. For fast neutron detection plastic fiber array scintllators or liquid scintillator filled capillary arrays are employed for fast neutron detection, and 6Li doped ZnS(Cu) phosphors are employed for thermal neutron detection. These simulations can provide guidance in the design, construction, and testing of neutron imaging systems. In particular we determined for a range of slab thickness, the range of thicknesses of embedded cracks (air-filled or filled with material such as water) which can be detected and imaged.

  17. Coated Fiber Neutron Detector Test

    SciTech Connect

    Lintereur, Azaree T.; Ely, James H.; Kouzes, Richard T.; Stromswold, David C.

    2009-10-23

    Radiation portal monitors used for interdiction of illicit materials at borders include highly sensitive neutron detection systems. The main reason for having neutron detection capability is to detect fission neutrons from plutonium. The currently deployed radiation portal monitors (RPMs) from Ludlum and Science Applications International Corporation (SAIC) use neutron detectors based upon 3He-filled gas proportional counters, which are the most common large neutron detector. There is a declining supply of 3He in the world, and thus, methods to reduce the use of this gas in RPMs with minimal changes to the current system designs and sensitivity to cargo-borne neutrons are being investigated. Reported here are the results of tests of the 6Li/ZnS(Ag)-coated non-scintillating plastic fibers option. This testing measured the required performance for neutron detection efficiency and gamma ray rejection capabilities of a system manufactured by Innovative American Technology (IAT).

  18. Observation of Neutron Skyshine from an Accelerator Based Neutron Source

    SciTech Connect

    Franklyn, C. B.

    2011-12-13

    A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >10{sup 11} n{center_dot}s{sup -1}. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.

  19. Observation of Neutron Skyshine from an Accelerator Based Neutron Source

    NASA Astrophysics Data System (ADS)

    Franklyn, C. B.

    2011-12-01

    A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >1011 nṡs-1. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.

  20. Cryogenic Neutron Spectrometer Development

    SciTech Connect

    Niedermayr, T; Hau, I D; Friedrich, S; Burger, A; Roy, U N; Bell, Z W

    2006-03-08

    Cryogenic microcalorimeter detectors operating at temperatures around {approx}0.1 K have been developed for the last two decades, driven mostly by the need for ultra-high energy resolution (<0.1%) in X-ray astrophysics and dark matter searches [1]. The Advanced Detector Group at Lawrence Livermore National Laboratory has developed different cryogenic detector technologies for applications ranging from X-ray astrophysics to nuclear science and non-proliferation. In particular, we have adapted cryogenic detector technologies for ultra-high energy resolution gamma-spectroscopy [2] and, more recently, fast-neutron spectroscopy [3]. Microcalorimeters are essentially ultra-sensitive thermometers that measure the energy of the radiation from the increase in temperature upon absorption. They consist of a sensitive superconducting thermometer operated at the transition between its superconducting and its normal state, where its resistance changes very rapidly with temperature such that even the minute energies deposited by single radiation quanta are sufficient to be detectable with high precision. The energy resolution of microcalorimeters is fundamentally limited by thermal fluctuations to {Delta}E{sub FWHM} {approx} 2.355 (k{sub B}T{sup 2}C{sub abs}){sup 1/2}, and thus allows an energy below 1 keV for neutron spectrometers for an operating temperature of T {approx} 0.1 K . The {Delta}E{sub FWHM} does not depend on the energy of the incident photon or particle. This expression is equivalent to the familiar (F{var_epsilon}E{sub {gamma}}){sup 1/2} considering that an absorber at temperature T contains a total energy C{sub abs}T, and the associated fluctuation are due to variations in uncorrelated (F=1) phonons ({var_epsilon} = k{sub B}T) dominated by the background energy C{sub abs}T >> E{gamma}. The rationale behind developing a cryogenic neutron spectrometer is the very high energy resolution combined with the high efficiency. Additionally, the response function is simple

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

    SciTech Connect

    Susmikanti, Mike; Dewayatna, Winter; Sulistyo, Yos

    2014-09-30

    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 Mo{sup 99} 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 (× 10{sup 6} cm{sup −1}) in a tube, their delta

  2. Direct Fast-Neutron Detection

    SciTech Connect

    DC Stromswold; AJ Peurrung; RR Hansen; PL Reeder

    2000-01-18

    Direct fast-neutron detection is the detection of fast neutrons before they are moderated to thermal energy. We have investigated two approaches for using proton-recoil in plastic scintillators to detect fast neutrons and distinguish them from gamma-ray interactions. Both approaches use the difference in travel speed between neutrons and gamma rays as the basis for separating the types of events. In the first method, we examined the pulses generated during scattering in a plastic scintillator to see if they provide a means for distinguishing fast-neutron events from gamma-ray events. The slower speed of neutrons compared to gamma rays results in the production of broader pulses when neutrons scatter several times within a plastic scintillator. In contrast, gamma-ray interactions should produce narrow pulses, even if multiple scattering takes place, because the time between successive scattering is small. Experiments using a fast scintillator confirmed the presence of broader pulses from neutrons than from gamma rays. However, the difference in pulse widths between neutrons and gamma rays using the best commercially available scintillators was not sufficiently large to provide a practical means for distinguishing fast neutrons and gamma rays on a pulse-by-pulse basis. A faster scintillator is needed, and that scintillator might become available in the literature. Results of the pulse-width studies were presented in a previous report (peurrung et al. 1998), and they are only summarized here.

  3. Materials and neutronic research at the Low Energy Neutron Source

    NASA Astrophysics Data System (ADS)

    Baxter, David V.

    2016-04-01

    In the decade since the Low Energy Neutron Source (LENS) at Indiana University Center for Exploration of Energy and Matter (CEEM) produced its first neutrons, the facility has made important contributions to the international neutron scattering community. LENS employs a 13MeV proton beam at up to 4kW beam power onto one of two Be targets to produce neutrons for research in fields ranging from radiation effects in electronics to studies of the structure of fluids confined in nanoporous materials. The neutron source design at the heart of LENS facilitates relatively rapid hands-on access to most of its components which provides a foundation for a research program in experimental neutronics and affords numerous opportunities for novel educational experiences. We describe in some detail a number of the unique capabilities of this facility.

  4. SELF-REACTIVATING NEUTRON SOURCE FOR A NEUTRONIC REACTOR

    DOEpatents

    Newson, H.W.

    1959-02-01

    Reactors of the type employing beryllium in a reflector region around the active portion and to a neutron source for use therewith are discussed. The neutron source is comprised or a quantity of antimony permanently incorporated in, and as an integral part of, the reactor in or near the beryllium reflector region. During operation of the reactor the natural occurring antimony isotope of atomic weight 123 absorbs neutrons and is thereby transformed to the antimony isotope of atomic weight 124, which is radioactive and emits gamma rays. The gamma rays react with the beryllium to produce neutrons. The beryllium and antimony thus cooperate to produce a built in neutron source which is automatically reactivated by the operation of the reactor itself and which is of sufficient strength to maintain the slow neutron flux at a sufficiently high level to be reliably measured during periods when the reactor is shut down.

  5. Neutron Scattering Stiudies

    SciTech Connect

    Kegel, Gunter H.R.; Egan, James J

    2007-04-18

    This project covers four principal areas of research: Elastic and inelastic neutron scattering studies in odd-A terbium, thulium and other highly deformed nuclei near A=160 with special regard to interband transitions and to the investigation of the direct-interaction versus the compound-nucleus excitation process in these nuclei. Examination of new, fast photomultiplier tubes suitable for use in a miniaturized neutron-time-of-flight spectrometer. Measurement of certain inelastic cross sections of 238U. Determination of the multiplicity of prompt fission gamma rays in even-A fissile actinides. Energies and mean lives of fission isomers produced by fast fission of even-Z, even-A actinides. Study of the mean life of 7Be in different host matrices and its possible astro-physical significance.

  6. FAST NEUTRON SPECTROMETER

    DOEpatents

    Davis, F.J.; Hurst, G.S.; Reinhardt, P.W.

    1959-08-18

    An improved proton recoil spectrometer for determining the energy spectrum of a fast neutron beam is described. Instead of discriminating against and thereby"throwing away" the many recoil protons other than those traveling parallel to the neutron beam axis as do conventional spectrometers, this device utilizes protons scattered over a very wide solid angle. An ovoidal gas-filled recoil chamber is coated on the inside with a scintillator. The ovoidal shape of the sensitive portion of the wall defining the chamber conforms to the envelope of the range of the proton recoils from the radiator disposed within the chamber. A photomultiplier monitors the output of the scintillator, and a counter counts the pulses caused by protons of energy just sufficient to reach the scintillator.

  7. Neutron Imaging Camera

    NASA Technical Reports Server (NTRS)

    Hunter, Stanley D.; DeNolfo, Georgia; Floyd, Sam; Krizmanic, John; Link, Jason; Son, Seunghee; Guardala, Noel; Skopec, Marlene; Stark, Robert

    2008-01-01

    We describe the Neutron Imaging Camera (NIC) being developed for DTRA applications by NASA/GSFC and NSWC/Carderock. The NIC is based on the Three-dimensional Track Imager (3-DTI) technology developed at GSFC for gamma-ray astrophysics applications. The 3-DTI, a large volume time-projection chamber, provides accurate, approximately 0.4 mm resolution. 3-D tracking of charged particles. The incident direction of fast neutrons, E(sub N) > 0.5 MeV. arc reconstructed from the momenta and energies of the proton and triton fragments resulting from 3He(n,p)3H interactions in the 3-DTI volume. We present angular and energy resolution performance of the NIC derived from accelerator tests.

  8. COMPOSITE NEUTRONIC REACTOR

    DOEpatents

    Menke, J.R.

    1963-06-11

    This patent relates to a reactor having a core which comprises an inner active region and an outer active region, each region separately having a k effective less than one and a k infinity greater than one. The inner and outer regions in combination have a k effective at least equal to one and each region contributes substantially to the k effective of the reactor core. The inner region has a low moderator to fuel ratio such that the majority of fissions occurring therein are induced by neutrons having energies greater than thermal. The outer region has a high moderator to fuel ratio such that the majority of fissions occurring therein are induced by thermal neutrons. (AEC)

  9. Neutron beam measurement dosimetry

    SciTech Connect

    Amaro, C.R.

    1995-11-01

    This report describes animal dosimetry studies and phantom measurements. During 1994, 12 dogs were irradiated at BMRR as part of a 4 fraction dose tolerance study. The animals were first infused with BSH and irradiated daily for 4 consecutive days. BNL irradiated 2 beagles as part of their dose tolerance study using BPA fructose. In addition, a dog at WSU was irradiated at BMRR after an infusion of BPA fructose. During 1994, the INEL BNCT dosimetry team measured neutron flux and gamma dose profiles in two phantoms exposed to the epithermal neutron beam at the BMRR. These measurements were performed as a preparatory step to the commencement of human clinical trials in progress at the BMRR.

  10. Short pulse neutron generator

    DOEpatents

    Elizondo-Decanini, Juan M.

    2016-08-02

    Short pulse neutron generators are described herein. In a general embodiment, the short pulse neutron generator includes a Blumlein structure. The Blumlein structure includes a first conductive plate, a second conductive plate, a third conductive plate, at least one of an inductor or a resistor, a switch, and a dielectric material. The first conductive plate is positioned relative to the second conductive plate such that a gap separates these plates. A vacuum chamber is positioned in the gap, and an ion source is positioned to emit ions in the vacuum chamber. The third conductive plate is electrically grounded, and the switch is operable to electrically connect and disconnect the second conductive plate and the third conductive plate. The at least one of the resistor or the inductor is coupled to the first conductive plate and the second conductive plate.

  11. Superdense neutron matter

    NASA Technical Reports Server (NTRS)

    Canuto, V.; Datta, B.; Kalman, G.

    1978-01-01

    A relativistic theory of high-density matter is presented which takes into account the short-range interaction due to the exchange of spin-2 mesons. An equation of state is derived and used to compute neutron-star properties. The prediction of the theory for the values of maximum mass and moment of inertia for a stable neutron star are 1.75 solar masses and 1.68 by 10 to the 45th power g-sq cm, in very good agreement with the presently known observational bounds. The corresponding radius is found to be 10.7 km. It is found that the inclusion of the spin-2 interaction reduces the disagreement between the relativistic and nonrelativistic theories in their predictions of masses and moments of inertia.

  12. Planets Around Neutron Stars

    NASA Technical Reports Server (NTRS)

    Wolszczan, Alexander; Kulkarni, Shrinivas R; Anderson, Stuart B.

    2003-01-01

    The objective of this proposal was to continue investigations of neutron star planetary systems in an effort to describe and understand their origin, orbital dynamics, basic physical properties and their relationship to planets around normal stars. This research represents an important element of the process of constraining the physics of planet formation around various types of stars. The research goals of this project included long-term timing measurements of the planets pulsar, PSR B1257+12, to search for more planets around it and to study the dynamics of the whole system, and sensitive searches for millisecond pulsars to detect further examples of old, rapidly spinning neutron stars with planetary systems. The instrumentation used in our project included the 305-m Arecibo antenna with the Penn State Pulsar Machine (PSPM), the 100-m Green Bank Telescope with the Berkeley- Caltech Pulsar Machine (BCPM), and the 100-m Effelsberg and 64-m Parkes telescopes equipped with the observatory supplied backend hardware.

  13. Porous material neutron detector

    DOEpatents

    Diawara, Yacouba; Kocsis, Menyhert

    2012-04-10

    A neutron detector employs a porous material layer including pores between nanoparticles. The composition of the nanoparticles is selected to cause emission of electrons upon detection of a neutron. The nanoparticles have a maximum dimension that is in the range from 0.1 micron to 1 millimeter, and can be sintered with pores thereamongst. A passing radiation generates electrons at one or more nanoparticles, some of which are scattered into a pore and directed toward a direction opposite to the applied electrical field. These electrons travel through the pore and collide with additional nanoparticles, which generate more electrons. The electrons are amplified in a cascade reaction that occurs along the pores behind the initial detection point. An electron amplification device may be placed behind the porous material layer to further amplify the electrons exiting the porous material layer.

  14. High intensity, pulsed thermal neutron source

    DOEpatents

    Carpenter, J.M.

    1973-12-11

    This invention relates to a high intensity, pulsed thermal neutron source comprising a neutron-producing source which emits pulses of fast neutrons, a moderator block adjacent to the last neutron source, a reflector block which encases the fast neutron source and the moderator block and has a thermal neutron exit port extending therethrough from the moderator block, and a neutron energy- dependent decoupling reflector liner covering the interior surfaces of the thermal neutron exit port and surrounding all surfaces of the moderator block except the surface viewed by the thermal neutron exit port. (Official Gazette)

  15. First neutron generation in the BINP accelerator based neutron source.

    PubMed

    Bayanov, B; Burdakov, A; Chudaev, V; Ivanov, A; Konstantinov, S; Kuznetsov, A; Makarov, A; Malyshkin, G; Mekler, K; Sorokin, I; Sulyaev, Yu; Taskaev, S

    2009-07-01

    Pilot innovative facility for neutron capture therapy was built at Budker Institute of Nuclear Physics, Novosibirsk. This facility is based on a compact vacuum insulation tandem accelerator designed to produce proton current up to 10 mA. Epithermal neutrons are proposed to be generated by 1.915 MeV protons bombarding a lithium target using (7)Li(p,n)(7)Be threshold reaction. The results of the first experiments on neutron generation are reported and discussed. PMID:19375928

  16. European Neutron Activation System.

    Energy Science and Technology Software Center (ESTSC)

    2013-01-11

    Version 03 EASY-2010 (European Activation System) consists of a wide range of codes, data and documentation all aimed at satisfying the objective of calculating the response of materials irradiated in a neutron flux. The main difference from the previous version is the upper energy limit, which has increased from 20 to 60 MeV. It is designed to investigate both fusion devices and accelerator based materials test facilities that will act as intense sources of high-energymore » neutrons causing significant activation of the surrounding materials. The very general nature of the calculational method and the data libraries means that it is applicable (with some reservations) to all situations (e.g. fission reactors or neutron sources) where materials are exposed to neutrons below 60 MeV. EASY can be divided into two parts: data and code development tools and user tools and data. The former are required to develop the latter, but EASY users only need to be able to use the inventory code FISPACT and be aware of the contents of the EAF library (the data source). The complete EASY package contains the FISPACT-2007 inventory code, the EAF-2003, EAF-2005, EAF-2007 and EAF-2010 libraries, and the EASY User Interface for the Window version. The activation package EASY-2010 is the result of significant development to extend the upper energy range from 20 to 60 MeV so that it is capable of being used for IFMIF calculations. The EAF-2010 library contains 66,256 reactions, almost five times more than in EAF-2003 (12,617). Deuteron-induced and proton-induced cross section libraries are also included, and can be used with EASY to enable calculations of the activation due to deuterons and proton [2].« less

  17. MONDE: MOmentum Neutron DEtector

    NASA Astrophysics Data System (ADS)

    Santa Rita, P.; Acosta, L.; Favela, F.; Huerta, A.; Ortiz, M. E.; Policroniades, R.; Chávez, E.

    2016-07-01

    MONDE is a large area neutron momentum detector, consisting of a 70x160x5 cm3 plastic scintillator slab surrounded by 16 photomultiplier tubes, standard NIM signal processing electronics and a CAMAC data acquisition system. In this work we present data from a characterization run using an external trigger. For that purpose, coincident gamma rays from a 60Co radioactive source were used together with a NaI external detector. First results with an "external" trigger are presented.

  18. Gravitoastronomy with neutron stars

    NASA Astrophysics Data System (ADS)

    Woan, Graham

    2004-09-01

    Recent advances in gravitational wave detectors mean that we can start to make astrophysically important statements about the physics of neutron stars based on observed upper limits to their gravitational luminosity. Here we consider statements we can already make about a selection of known radio pulsars, based on data from the LIGO and GEO600 detectors, and look forward to what could be learned from the first detections.

  19. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Dreffin, R.S.

    1959-12-15

    A control means for a nuclear reactor is described. Particularly a device extending into the active portion of the reactor consisting of two hollow elements coaxially disposed and forming a channel therebetween, the cross sectional area of the channel increasing from each extremity of the device towards the center thereof. An element of neutron absorbing material is slidably positionable within the inner hollow element and a fluid reactor poison is introduced into the channel defined by the two hollow elements.

  20. Neutron instrumentation for biology

    SciTech Connect

    Mason, S.A.

    1994-12-31

    In the October 1994 round of proposals at the ILL, the external biology review sub- committee was asked to allocate neutron beam time to a wide range of experiments, on almost half the total number of scheduled neutron instruments: on 3 diffractometers, on 3 small angle scattering instruments, and on some 6 inelastic scattering spectrometers. In the 3.5 years since the temporary reactor shutdown, the ILL`s management structure has been optimized, budgets and staff have been trimmed, the ILL reactor has been re-built, and many of the instruments up-graded, many powerful (mainly Unix) workstations have been introduced, and the neighboring European Synchrotron Radiation Facility has established itself as the leading synchrotron radiation source and has started its official user program. The ILL reactor remains the world`s most intense dedicated neutron source. In this challenging context, it is of interest to review briefly the park of ILL instruments used to study the structure and energetics of small and large biological systems. A brief summary will be made of each class of experiments actually proposed in the latest ILL proposal round.

  1. Neutron logging tool

    SciTech Connect

    Taylor, J.A.; Taylor, K.G.

    1987-02-03

    A method is described of logging earth formations traversed by a well bore and utilizing a logging tool having a neutron source and a short spaced and a long spaced thermal neutron detector which produce an independent response as a function of depth of the logging tool in a well bore. The method comprises: moving the logging tool through a well bore to locate a section of the earth formations which has minimum porosity and obtaining measurement responses from each of the long and short spaced detectors; normalizing the responses of the long and short spaced detectors by matching the sensitivity of response of the long spaced detector to the sensitivity of response of the short spaced detector for an earth formation which has minimum porosity so that the normalized responses track one another in an earth formation which has minimum porosity; and moving the tool over the length of the well bore to be surveyed while recording the normalized responses of the long and short spaced neutron detectors as a function of depth.

  2. NEUTRONIC REACTOR CORE

    DOEpatents

    Thomson, W.B.; Corbin, A. Jr.

    1961-07-18

    An improved core for a gas-cooled power reactor which admits gas coolant at high temperatures while affording strong integral supporting structure and efficient moderation of neutrons is described. The multiplicities of fuel elements constituting the critical amassment of fissionable material are supported and confined by a matrix of metallic structure which is interspersed therebetween. Thermal insulation is interposed between substantially all of the metallic matrix and the fuel elements; the insulation then defines the principal conduit system for conducting the coolant gas in heat-transfer relationship with the fuel elements. The metallic matrix itseif comprises a system of ducts through which an externally-cooled hydrogeneous liquid, such as water, is circulated to serve as the principal neutron moderant for the core and conjointly as the principal coolant for the insulated metallic structure. In this way, use of substantially neutron transparent metals, such as aluminum, becomes possible for the supporting structure, despite the high temperatures of the proximate gas. The Aircraft Nuclear Propulsion program's "R-1" reactor design is a preferred embodiment.

  3. Proton recoil scintillator neutron rem meter

    DOEpatents

    Olsher, Richard H.; Seagraves, David T.

    2003-01-01

    A neutron rem meter utilizing proton recoil and thermal neutron scintillators to provide neutron detection and dose measurement. In using both fast scintillators and a thermal neutron scintillator the meter provides a wide range of sensitivity, uniform directional response, and uniform dose response. The scintillators output light to a photomultiplier tube that produces an electrical signal to an external neutron counter.

  4. Neutron Activation Analysis: Techniques and Applications

    SciTech Connect

    MacLellan, Ryan

    2011-04-27

    The role of neutron activation analysis in low-energy low-background experimentsis discussed in terms of comparible methods. Radiochemical neutron activation analysis is introduce. The procedure of instrumental neutron activation analysis is detailed especially with respect to the measurement of trace amounts of natural radioactivity. The determination of reactor neutron spectrum parameters required for neutron activation analysis is also presented.

  5. Fast neutron imaging device and method

    DOEpatents

    Popov, Vladimir; Degtiarenko, Pavel; Musatov, Igor V.

    2014-02-11

    A fast neutron imaging apparatus and method of constructing fast neutron radiography images, the apparatus including a neutron source and a detector that provides event-by-event acquisition of position and energy deposition, and optionally timing and pulse shape for each individual neutron event detected by the detector. The method for constructing fast neutron radiography images utilizes the apparatus of the invention.

  6. Prompt Fission Neutron Energy Spectra Induced by Fast Neutrons

    NASA Astrophysics Data System (ADS)

    Staples, Parrish Alan

    Prompt fission neutron energy spectra for ^{235}U and ^{239 }Pu have been measured for fission neutron energies greater than the energy of the incident neutrons inducing fission. The measurements were undertaken to investigate the shape dependence of the fission neutron spectra upon both the incident neutron energy and the mass of the nucleus undergoing fission. Measurements were made for both nuclides at the following incident neutron energies; 0.50 MeV, 1.50 MeV, 2.50 MeV and 3.50 MeV. The data are presented either as relative yields or as ratios of a measured spectrum to the ^{235}U spectrum at 0.50 MeV. Incident neutrons were produced by the ^7Li(p,n)^7Be reaction using a pulsed, bunched proton beam from the 5.5 MV Van de Graaff accelerator at the University of Massachusetts Lowell Pinanski Energy Center. The neutrons were detected by a thin liquid scintillator with good time resolution capabilities; time-of-flight techniques were used for neutron energy determination; in addition pulse-shape-discrimination was used to reduce gamma-ray background levels. The measurements are compared to calculations based on the Los Alamos Model of Madland and Nix to test its predictive capabilities. The data are fit by the Watt equation to determine the mean energy of the spectra, and to facilitate comparison of the results to previous measurements. The data are also compared directly to previous measurements.

  7. A National Spallation Neutron Source for neutron scattering

    SciTech Connect

    Appleton, B.R.

    1996-10-01

    The National Spallation Neutron Source is a collaborative project or perform the conceptual design for a next generation neutron source for the Department of Energy. This paper reviews the need and justification for a new neutron source, the origins and structure of the collaboration formed to address this need, and the community input leading up to the current design approach. A reference design is presented for an accelerator based spallation neutron source that would begin operation at about 1 megawatt of power but designed so that it could be upgraded to significantly higher powers in the future. The technology approach, status, and progress on the conceptual design to date are presented.

  8. Neutron-fragment and Neutron-neutron Correlations in Low-energy Fission

    NASA Astrophysics Data System (ADS)

    Lestone, J. P.

    2016-01-01

    A computational method has been developed to simulate neutron emission from thermal-neutron induced fission of 235U and from spontaneous fission of 252Cf. Measured pre-emission mass-yield curves, average total kinetic energies and their variances, both as functions of mass split, are used to obtain a representation of the distribution of fragment velocities. Measured average neutron multiplicities as a function of mass split and their dependence on total kinetic energy are used. Simulations can be made to reproduce measured factorial moments of neutron-multiplicity distributions with only minor empirical adjustments to some experimental inputs. The neutron-emission spectra in the rest-frame of the fragments are highly constrained by ENDF/B-VII.1 prompt-fission neutron-spectra evaluations. The n-f correlation measurements of Vorobyev et al. (2010) are consistent with predictions where all neutrons are assumed to be evaporated isotropically from the rest frame of fully accelerated fragments. Measured n-f and n-n correlations of others are a little weaker than the predictions presented here. These weaker correlations could be used to infer a weak scission-neutron source. However, the effect of neutron scattering on the experimental results must be studied in detail before moving away from a null hypothesis that all neutrons are evaporated from the fragments.

  9. Static Response of Neutron Matter.

    PubMed

    Buraczynski, Mateusz; Gezerlis, Alexandros

    2016-04-15

    We generalize the problem of strongly interacting neutron matter by adding a periodic external modulation. This allows us to study from first principles a neutron system that is extended and inhomogeneous, with connections to the physics of both neutron-star crusts and neutron-rich nuclei. We carry out fully nonperturbative microscopic quantum Monte Carlo calculations of the energy of neutron matter at different densities, as well as different strengths and periodicities of the external potential. In order to remove systematic errors, we examine finite-size effects and the impact of the wave function ansatz. We also make contact with energy-density functional theories of nuclei and disentangle isovector gradient contributions from bulk properties. Finally, we calculate the static density-density linear response function of neutron matter and compare it with the response of other physical systems. PMID:27127963

  10. Static Response of Neutron Matter

    NASA Astrophysics Data System (ADS)

    Buraczynski, Mateusz; Gezerlis, Alexandros

    2016-04-01

    We generalize the problem of strongly interacting neutron matter by adding a periodic external modulation. This allows us to study from first principles a neutron system that is extended and inhomogeneous, with connections to the physics of both neutron-star crusts and neutron-rich nuclei. We carry out fully nonperturbative microscopic quantum Monte Carlo calculations of the energy of neutron matter at different densities, as well as different strengths and periodicities of the external potential. In order to remove systematic errors, we examine finite-size effects and the impact of the wave function ansatz. We also make contact with energy-density functional theories of nuclei and disentangle isovector gradient contributions from bulk properties. Finally, we calculate the static density-density linear response function of neutron matter and compare it with the response of other physical systems.

  11. NEUTRON MEASURING METHOD AND APPARATUS

    DOEpatents

    Seaborg, G.T.; Friedlander, G.; Gofman, J.W.

    1958-07-29

    A fast neutron fission detecting apparatus is described consisting of a source of fast neutrons, an ion chamber containing air, two electrodes within the ion chamber in confronting spaced relationship, a high voltage potential placed across the electrodes, a shield placed about the source, and a suitable pulse annplifier and recording system in the electrode circuit to record the impulse due to fissions in a sannple material. The sample material is coated onto the active surface of the disc electrode and shielding means of a material having high neutron capture capabilities for thermal neutrons are provided in the vicinity of the electrodes and about the ion chamber so as to absorb slow neutrons of thermal energy to effectively prevent their diffusing back to the sample and causing an error in the measurement of fast neutron fissions.

  12. The physics of neutron stars.

    PubMed

    Lattimer, J M; Prakash, M

    2004-04-23

    Neutron stars are some of the densest manifestations of massive objects in the universe. They are ideal astrophysical laboratories for testing theories of dense matter physics and provide connections among nuclear physics, particle physics, and astrophysics. Neutron stars may exhibit conditions and phenomena not observed elsewhere, such as hyperon-dominated matter, deconfined quark matter, superfluidity and superconductivity with critical temperatures near 10(10) kelvin, opaqueness to neutrinos, and magnetic fields in excess of 10(13) Gauss. Here, we describe the formation, structure, internal composition, and evolution of neutron stars. Observations that include studies of pulsars in binary systems, thermal emission from isolated neutron stars, glitches from pulsars, and quasi-periodic oscillations from accreting neutron stars provide information about neutron star masses, radii, temperatures, ages, and internal compositions. PMID:15105490

  13. Neutron scattering and absorption properties

    SciTech Connect

    Holden, N.E.

    1993-12-01

    The Table in this report presents an evaluated set of values for the experimental quantities, which characterize the properties for scattering and absorption of neutrons. The neutron cross section is given for room temperature neutrons, 20.43{degree}C, corresponds to a thermal neutron energy of 0.0253 electron volts (eV) or a neutron velocity of 2200 meters/second. The neutron resonance integral is defined over the energy range from 0.5 eV to 0.1 {times} 10{sup 6} eV, or 0.1 MeV. A list of the major references used is given below. The literature cutoff data is October 1993. Uncertainties are given in parentheses. Parentheses with two or more numbers indicate values to the excited states(s) and to the ground state of the product nucleus.

  14. Epithermal neutron instrumentation at ISIS

    NASA Astrophysics Data System (ADS)

    Gorini, G.; Festa, G.; Andreani, C.

    2014-12-01

    The advent of pulsed neutron sources makes available high epithermal neutron fluxes (in the energy range between 500 meV and 100 eV). New dedicated instrumentation, such as Resonance Detectors, was developed at ISIS spallation neutron source in the last years to apply the specific properties of this kind of neutron beam to the study of condensed matter. New detection strategies like Filter Difference method and Foil Cycling Technique were also developed in parallel to the detector improvement at the VESUVIO beamline. Recently, epithermal neutron beams were also used at the INES beamline to study elemental and isotopic composition of materials, with special application to cultural heritage studies. In this paper we review a series of epithermal neutron instrumentation developed at ISIS, their evolution over time and main results obtained.

  15. APPARATUS FOR CONTROLLING NEUTRONIC REACTORS

    DOEpatents

    Dietrich, J.R.; Harrer, J.M.

    1958-09-16

    A device is described for rapidly cortrolling the reactivity of an active portion of a reactor. The inveniion consists of coaxially disposed members each having circumferenital sections of material having dlfferent neutron absorbing characteristics and means fur moving the members rotatably and translatably relative to each other within the active portion to vary the neutron flux therein. The angular and translational movements of any member change the neutron flux shadowing effect of that member upon the other member.

  16. Neutron diffraction on pulsed sources

    NASA Astrophysics Data System (ADS)

    Aksenov, V. L.; Balagurov, A. M.

    2016-03-01

    The current capabilities of and major scientific problems solved by time-of-flight neutron diffraction are reviewed. The reasons for the rapid development of the method over the last two decades have been mainly the emergence of third-generation pulsed sources with a megawatt time-averaged power and advances in neutron optical devices and detector systems. The paper discusses some historical aspects of time-of-flight neutron diffraction and examines the contribution to this method from F L Shapiro, the centennial of whose birth was celebrated in 2015. The state of the art with respect to neutron sources for studies on extracted beams is reviewed in a special section.

  17. NEUTRON ABSORPTION AND SHIELDING DEVICE

    DOEpatents

    Axelrad, I.R.

    1960-06-21

    A neutron absorption and shielding device is described which is adapted for mounting in a radiation shielding wall surrounding a radioactive area through which instrumentation leads and the like may safely pass without permitting gamma or neutron radiation to pass to the exterior. The shielding device comprises a container having at least one nonrectilinear tube or passageway means extending therethrough, which is adapted to contain instrumentation leads or the like, a layer of a substance capable of absorbing gamma rays, and a solid resinous composition adapted to attenuate fast-moving neutrons and capture slow- moving or thermal neutrons.

  18. ACCELERATOR BASED CONTINUOUS NEUTRON SOURCE.

    SciTech Connect

    SHAPIRO,S.M.; RUGGIERO,A.G.; LUDEWIG,H.

    2003-03-25

    Until the last decade, most neutron experiments have been performed at steady-state, reactor-based sources. Recently, however, pulsed spallation sources have been shown to be very useful in a wide range of neutron studies. A major review of neutron sources in the US was conducted by a committee chaired by Nobel laureate Prof. W. Kohn: ''Neutron Sources for America's Future-BESAC Panel on Neutron Sources 1/93''. This distinguished panel concluded that steady state and pulsed sources are complementary and that the nation has need for both to maintain a balanced neutron research program. The report recommended that both a new reactor and a spallation source be built. This complementarity is recognized worldwide. The conclusion of this report is that a new continuous neutron source is needed for the second decade of the 20 year plan to replace aging US research reactors and close the US neutron gap. it is based on spallation production of neutrons using a high power continuous superconducting linac to generate protons impinging on a heavy metal target. There do not appear to be any major technical challenges to the building of such a facility since a continuous spallation source has been operating in Switzerland for several years.

  19. NEUTRON IMAGING, RADIOGRAPHY AND TOMOGRAPHY.

    SciTech Connect

    SMITH,G.C.

    2002-03-01

    Neutrons are an invaluable probe in a wide range of scientific, medical and commercial endeavors. Many of these applications require the recording of an image of the neutron signal, either in one-dimension or in two-dimensions. We summarize the reactions of neutrons with the most important elements that are used for their detection. A description is then given of the major techniques used in neutron imaging, with emphasis on the detection media and position readout principle. Important characteristics such as position resolution, linearity, counting rate capability and sensitivity to gamma-background are discussed. Finally, the application of a subset of these instruments in radiology and tomography is described.

  20. Portable neutron spectrometer and dosimeter

    DOEpatents

    Waechter, D.A.; Erkkila, B.H.; Vasilik, D.G.

    The disclosure relates to a battery operated neutron spectrometer/dosimeter utilizing a microprocessor, a built-in tissue equivalent LET neutron detector, and a 128-channel pulse height analyzer with integral liquid crystal display. The apparatus calculates doses and dose rates from neutrons incident on the detector and displays a spectrum of rad or rem as a function of keV per micron of equivalent tissue and also calculates and displays accumulated dose in millirads and millirem as well as neutron dose rates in millirads per hour and millirem per hour.

  1. Portable neutron spectrometer and dosimeter

    DOEpatents

    Waechter, David A.; Erkkila, Bruce H.; Vasilik, Dennis G.

    1985-01-01

    The disclosure relates to a battery operated neutron spectrometer/dosimeter utilizing a microprocessor, a built-in tissue equivalent LET neutron detector, and a 128-channel pulse height analyzer with integral liquid crystal display. The apparatus calculates doses and dose rates from neutrons incident on the detector and displays a spectrum of rad or rem as a function of keV per micron of equivalent tissue and also calculates and displays accumulated dose in millirads and millirem as well as neutron dose rates in millirads per hour and millirem per hour.

  2. Enrico Fermi's Discovery of Neutron-Induced Artificial Radioactivity: Neutrons and Neutron Sources

    NASA Astrophysics Data System (ADS)

    Guerra, Francesco; Leone, Matteo; Robotti, Nadia

    2006-09-01

    We reconstruct and analyze the path leading from James Chadwick’s discovery of the neutron in February 1932 through Frédéric Joliot and Irène Curie’s discovery of artificial radioactivity in January 1934 to Enrico Fermi’s discovery of neutron-induced artificial radioactivity in March 1934. We show, in particular, that Fermi’s innovative construction and use of radon-beryllium neutron sources permitted him to make his discovery.

  3. Slow neutron leakage spectra from spallation neutron sources

    SciTech Connect

    Das, S.G.; Carpenter, J.M.; Prael, R.E.

    1980-02-01

    An efficient technique is described for Monte Carlo simulation of neutron beam spectra from target-moderator-reflector assemblies typical of pulsed spallation neutron sources. The technique involves the scoring of the transport-theoretical probability that a neutron will emerge from the moderator surface in the direction of interest, at each collision. An angle-biasing probability is also introduced which further enhances efficiency in simple problems. These modifications were introduced into the VIM low energy neutron transport code, representing the spatial and energy distributions of the source neutrons approximately as those of evaporation neutrons generated through the spallation process by protons of various energies. The intensity of slow neutrons leaking from various reflected moderators was studied for various neutron source arrangements. These include computations relating to early measurements on a mockup-assembly, a brief survey of moderator materials and sizes, and a survey of the effects of varying source and moderator configurations with a practical, liquid metal cooled uranium source Wing and slab, i.e., tangential and radial moderator arrangements, and Be vs CH/sub 2/ reflectors are compared. Results are also presented for several complicated geometries which more closely represent realistic arrangements for a practical source, and for a subcritical fission multiplier such as might be driven by an electron linac. An adaptation of the code was developed to enable time dependent calculations, and investigated the effects of the reflector, decoupling and void liner materials on the pulse shape.

  4. Neutron radiography and neutron-induced autoradiography for the classroom

    SciTech Connect

    Aderhold, H.C. )

    1992-01-01

    The Cornell 500-kW MARK II TRIGA reactor at the Ward Laboratory of Nuclear Engineering has been used to illustrate the application of neutron radiography (NR) and neutron-induced autoradiography (NIAR) for solving problems in engineering as well as problems in art history. The applications are described in the paper.

  5. Neutron sources for a neutron capture therapy facility

    SciTech Connect

    Lennox, A.J.

    1993-04-01

    Recent advances in the development of boron pharmaceuticals have reopened the possibility of using epithermal neutrons to treat brain tumors containing boron-10. This paper summarizes the approaches being used to generate the neutron sources and identifies specific areas where more research and development are needed.

  6. Neutron fan beam source for neutron radiography purpose

    SciTech Connect

    Le Tourneur, P.; Bach, P.; Dance, W. E.

    1999-06-10

    The development of the DIANE neutron radiography system included a sealed-tube neutron generator for this purpose and the optimization of the system's neutron beam quality in terms of divergence and useful thermal neutron yield for each 14-MeV neutron produced. Following this development, the concept of a DIANE fan beam source is herewith introduced. The goal which drives this design is one of economy: by simply increasing the aperture dimension of a conventional DIANE beam in one plane of its collimator axis to a small-angle, fan-shaped output, the useful beam area for neutron radiography would be substantially increased. Thus with the same source, the throughput, or number of objects under examination at any given time, would be augmented significantly. Presented here are the design of this thermal neutron source and the initial Monte Carlo calculations. Taking into account the experience with the conventional DIANE neutron radiography system, these result are discussed and the potential of and interest in such a fan-beam source are explored.

  7. Neutron Multiplicity Measurements With 3He Alternative: Straw Neutron Detectors

    DOE PAGESBeta

    Mukhopadhyay, Sanjoy; Wolff, Ronald S.; Meade, John A.; Detweiler, Ryan; Maurer, Richard J.; Mitchell, Stephen E.; Guss, Paul P.; Lacy, Jeffrey L.; Sun, Liang; Athanasiades, Athanasios

    2015-01-27

    Counting neutrons emitted by special nuclear material (SNM) and differentiating them from the background neutrons of various origins is the most effective passive means of detecting SNM. Unfortunately, neutron detection, counting, and partitioning in a maritime environment are complex due to the presence of high-multiplicity spallation neutrons (commonly known as “ship effect”) and to the complicated nature of the neutron scattering in that environment. In this study, a prototype neutron detector was built using 10B as the converter in a special form factor called “straws” that would address the above problems by looking into the details of multiplicity distributions ofmore » neutrons originating from a fissioning source. This paper describes the straw neutron multiplicity counter (NMC) and assesses the performance with those of a commercially available fission meter. The prototype straw neutron detector provides a large-area, efficient, lightweight, more granular (than fission meter) neutron-responsive detection surface (to facilitate imaging) to enhance the ease of application of fission meters. Presented here are the results of preliminary investigations, modeling, and engineering considerations leading to the construction of this prototype. This design is capable of multiplicity and Feynman variance measurements. This prototype may lead to a near-term solution to the crisis that has arisen from the global scarcity of 3He by offering a viable alternative to fission meters. This paper describes the work performed during a 2-year site-directed research and development (SDRD) project that incorporated straw detectors for neutron multiplicity counting. The NMC is a two-panel detector system. We used 10B (in the form of enriched boron carbide: 10B4C) for neutron detection instead of 3He. In the first year, the project worked with a panel of straw neutron detectors, investigated its characteristics, and developed a data acquisition (DAQ) system to collect

  8. Neutron multiplicity ,easurements With 3He alternative: Straw neutron detectors

    DOE PAGESBeta

    Mukhopadhyay, Sanjoy; Wolff, Ronald S.; Meade, John A.; Detweiler, Ryan; Maurer, Richard J.; Mitchell, Stephen E.; Guss, Paul P.; Lacy, Jeffrey L.; Sun, Liang; Athanasiades, Athanasios

    2015-01-27

    Counting neutrons emitted by special nuclear material (SNM) and differentiating them from the background neutrons of various origins is the most effective passive means of detecting SNM. Unfortunately, neutron detection, counting, and partitioning in a maritime environment are complex due to the presence of high-multiplicity spallation neutrons (commonly known as “ship effect”) and to the complicated nature of the neutron scattering in that environment. In this study, a prototype neutron detector was built using 10B as the converter in a special form factor called “straws” that would address the above problems by looking into the details of multiplicity distributions ofmore » neutrons originating from a fissioning source. This paper describes the straw neutron multiplicity counter (NMC) and assesses the performance with those of a commercially available fission meter. The prototype straw neutron detector provides a large-area, efficient, lightweight, more granular (than fission meter) neutron-responsive detection surface (to facilitate imaging) to enhance the ease of application of fission meters. Presented here are the results of preliminary investigations, modeling, and engineering considerations leading to the construction of this prototype. This design is capable of multiplicity and Feynman variance measurements. This prototype may lead to a near-term solution to the crisis that has arisen from the global scarcity of 3He by offering a viable alternative to fission meters. This paper describes the work performed during a 2-year site-directed research and development (SDRD) project that incorporated straw detectors for neutron multiplicity counting. The NMC is a two-panel detector system. We used 10B (in the form of enriched boron carbide: 10B4C) for neutron detection instead of 3He. In the first year, the project worked with a panel of straw neutron detectors, investigated its characteristics, and developed a data acquisition (DAQ) system to collect

  9. Fast Neutron Sensitivity with HPGe

    SciTech Connect

    Seifert, Allen; Hensley, Walter K.; Siciliano, Edward R.; Pitts, W. K.

    2008-01-22

    In addition to being excellent gamma-ray detectors, germanium detectors are also sensitive to fast neutrons. Incident neutrons undergo inelastic scattering {Ge(n,n')Ge*} off germanium nuclei and the resulting excited states emit gamma rays or conversion electrons. The response of a standard 140% high-purity germanium (HPGe) detector with a bismuth germanate (BGO) anti-coincidence shield was measured for several neutron sources to characterize the ability of the HPGe detector to detect fast neutrons. For a sensitivity calculation performed using the characteristic fast neutron response peak that occurs at 692 keV, the 140% germanium detector system exhibited a sensitivity of ~175 counts / kg of WGPumetal in 1000 seconds at a source-detector distance of 1 meter with 4 in. of lead shielding between source and detector. Theoretical work also indicates that it might be possible to use the shape of the fast-neutron inelastic scattering signatures (specifically, the end-point energy of the long high energy tail of the resulting asymmetric peak) to gain additional information about the energy distribution of the incident neutron spectrum. However, the experimentally observed end-point energies appear to be almost identical for each of the fast neutron sources counted. Detailed MCNP calculations show that the neutron energy distributions impingent on the detector for these sources are very similar in this experimental configuration, due to neutron scattering in a lead shield (placed between the neutron source and HPGe detector to reduce the gamma ray flux), the BGO anti-coincidence detector, and the concrete floor.

  10. AFCI-2.0 Neutron Cross Section Covariance Library

    SciTech Connect

    Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

    2011-03-01

    The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural

  11. Accelerator based epithermal neutron source for neutron capture therapy

    SciTech Connect

    Brugger, R.; Kunze, J.

    1991-05-01

    Several investigators have suggested that a charged particle accelerator with light element reactions might be able to produce enough epithermal neutrons to be useful in Neutron Capture Therapy. The reaction choice so far has been the Li(p,n) reaction with protons up to 2.5 MeV. A moderator around the target would reduce the faster neutrons down to the epithermal energy region. The goals of the present research are: identify better reactions; improve the moderators; and find better combinations of 1 and 2. The target is to achieve, at the patient location, an epithermal neutron current of greater than 10{sup 9}n/cm{sup 2}sec, with a dose to tissue from the neutrons alone of less than 10{sup {minus}10} rads/n and a dose from the gamma rays in the beam of less than 10{sup {minus}10} rads/n.

  12. Neutron beam imaging at neutron spectrometers at Dhruva

    SciTech Connect

    Desai, Shraddha S.; Rao, Mala N.

    2012-06-05

    A low efficiency, 2-Dimensional Position Sensitive Neutron Detector based on delay line position encoding is developed. It is designed to handle beam flux of 10{sup 6}-10{sup 7} n/cm{sup 2}/s and for monitoring intensity profiles of neutron beams. The present detector can be mounted in transmission mode, as the hardware allows maximum neutron transmission in sensitive region. Position resolution of 1.2 mm in X and Y directions, is obtained. Online monitoring of beam images and intensity profile of various neutron scattering spectrometers at Dhruva are presented. It shows better dynamic range of intensity over commercial neutron camera and is also time effective over the traditionally used photographic method.

  13. Small angle neutron scattering

    NASA Astrophysics Data System (ADS)

    Cousin, Fabrice

    2015-10-01

    Small Angle Neutron Scattering (SANS) is a technique that enables to probe the 3-D structure of materials on a typical size range lying from ˜ 1 nm up to ˜ a few 100 nm, the obtained information being statistically averaged on a sample whose volume is ˜ 1 cm3. This very rich technique enables to make a full structural characterization of a given object of nanometric dimensions (radius of gyration, shape, volume or mass, fractal dimension, specific area…) through the determination of the form factor as well as the determination of the way objects are organized within in a continuous media, and therefore to describe interactions between them, through the determination of the structure factor. The specific properties of neutrons (possibility of tuning the scattering intensity by using the isotopic substitution, sensitivity to magnetism, negligible absorption, low energy of the incident neutrons) make it particularly interesting in the fields of soft matter, biophysics, magnetic materials and metallurgy. In particular, the contrast variation methods allow to extract some informations that cannot be obtained by any other experimental techniques. This course is divided in two parts. The first one is devoted to the description of the principle of SANS: basics (formalism, coherent scattering/incoherent scattering, notion of elementary scatterer), form factor analysis (I(q→0), Guinier regime, intermediate regime, Porod regime, polydisperse system), structure factor analysis (2nd Virial coefficient, integral equations, characterization of aggregates), and contrast variation methods (how to create contrast in an homogeneous system, matching in ternary systems, extrapolation to zero concentration, Zero Averaged Contrast). It is illustrated by some representative examples. The second one describes the experimental aspects of SANS to guide user in its future experiments: description of SANS spectrometer, resolution of the spectrometer, optimization of spectrometer

  14. Neutron quality factor

    SciTech Connect

    1995-06-01

    Both the International Commission on Radiological Protection (ICRP) and the National Council on Radiation Protection and Measurements (NCRP) have recommended that the radiation quality weighting factor for neutrons (Q{sub n}, or the corresponding new modifying factor, w{sub R}) be increased by a value of two for most radiation protection practices. This means an increase in the recommended value for Q{sub n} from a nominal value of 10 to a nominal value of 20. This increase may be interpreted to mean that the biological effectiveness of neutrons is two times greater than previously thought. A decision to increase the value of Q{sub n} will have a major impact on the regulations and radiation protection programs of Federal agencies responsible for the protection of radiation workers. Therefore, the purposes of this report are: (1) to examine the general concept of {open_quotes}quality factor{close_quotes} (Q) in radiation protection and the rationale for the selection of specific values of Q{sub n}; and (2) to make such recommendations to the Federal agencies, as appropriate. This report is not intended to be an exhaustive review of the scientific literature on the biological effects of neutrons, with the aim of defending a particular value for Q{sub n}. Rather, the working group examined the technical issues surrounding the current recommendations of scientific advisory bodies on this matter, with the aim of determining if these recommendations should be adopted by the Federal agencies. Ultimately, the group concluded that there was no compelling basis for a change in Q{sub n}. The report was prepared by Federal scientists working under the auspices of the Science Panel of the Committee on Interagency Radiation Research and Policy Coordination (CIRRPC).

  15. NEUTRONIC REACTOR CONSTRUCTION

    DOEpatents

    Vernon, H.C.; Goett, J.J.

    1958-09-01

    A cover device is described for the fuel element receiving tube of a neutronic reactor of the heterogeneous, water cooled type wherein said tubes are arranged in a moderator with their longitudinal axes vertical. The cover is provided with means to support a rod-type fuel element from the bottom thereof and means to lock the cover in place, the latter being adapted for remote operation. This cover device is easily removable and seals the opening in the upper end of the fuel tube against leakage of coolant.

  16. Advanced Neutron Spectrometer

    NASA Technical Reports Server (NTRS)

    Christl, Mark; Dobson, Chris; Norwood, Joseph; Kayatin, Matthew; Apple, Jeff; Gibson, Brian; Dietz, Kurt; Benson, Carl; Smith, Dennis; Howard, David; Rodriquez, Miguel; Watts, John; Sabra, Mohammed; Kuznetsov, Evgeny

    2013-01-01

    Energetic neutron measurements remain a challenge for space science investigations and radiation monitoring for human exploration beyond LEO. We are investigating a new composite scintillator design that uses Li6 glass scintillator embedded in a PVT block. A comparison between Li6 and Boron 10 loaded scintillators are being studied to assess the advantages and shortcomings of these two techniques. We present the details of the new Li6 design and results from the comparison of the B10 and Li6 techniques during exposures in a mixed radiation field produced by high energy protons interacting in a target material.

  17. High power neutron production targets

    SciTech Connect

    Wender, S.

    1996-06-01

    The author describes issues of concern in the design of targets and associated systems for high power neutron production facilities. The facilities include uses for neutron scattering, accelerator driven transmutation, accelerator production of tritium, short pulse spallation sources, and long pulse spallation sources. Each of these applications requires a source with different design needs and consequently different implementation in practise.

  18. Personnel neutron monitoring in space

    NASA Technical Reports Server (NTRS)

    Schaefer, H. J.

    1978-01-01

    A brief review is presented of available information on the galactic neutron spectrum. An examination is made of the difficulties encountered in the determination of the dose equivalent of neutron recoil protons in the presence of a substantially larger background of trapped and star-produced protons as well as other ionizing particles in space.

  19. Neutron Technologies for Bioenergy Research

    SciTech Connect

    Langan, Paul

    2012-01-01

    Neutron scattering is a powerful technique that can be used to probe the structures and dynamics of complex systems. It can provide a fundamental understanding of the processes involved in the production of biofuels from lignocellulosic biomass. A variety of neutron scattering technologies are available to elucidate both the organization and deconstruction of this complex composite material and the associations and morphology of the component polymers and the enzymes acting on them, across multiple length scales ranging from Angstroms to micrometers and time scales from microseconds to picoseconds. Unlike most other experimental techniques, neutron scattering is uniquely sensitive to hydrogen (and its isotope deuterium), an atom abundantly present throughout biomass and a key effector in many biological, chemical, and industrial processes for producing biofuels. Sensitivity to hydrogen, the ability to replace hydrogen with deuterium to alter scattering levels, the fact that neutrons cause little or no direct radiation damage, and the ability of neutrons to exchange thermal energies with materials, provide neutron scattering technologies with unique capabilities for bioenergy research. Further, neutrons are highly penetrating, making it possible to employ sample environments that are not suitable for other techniques. The true power of neutron scattering is realized when it is combined with computer simulation and modeling and contrast variation techniques enabled through selective deuterium labeling.

  20. Fission fragment driven neutron source

    DOEpatents

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  1. Neutrons for technology and science

    SciTech Connect

    Aeppli, G.

    1995-10-01

    We reviewed recent work using neutrons generated at nuclear reactors an accelerator-based spallation sources. Provided that large new sources become available, neutron beams will continue to have as great an impact on technology and science as in the past.

  2. Measuring neutron spectra in radiotherapy using the nested neutron spectrometer

    SciTech Connect

    Maglieri, Robert Evans, Michael; Seuntjens, Jan; Kildea, John; Licea, Angel

    2015-11-15

    Purpose: Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. Methods: The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation–maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. Results: The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors’ measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. Conclusions: The NNS may

  3. Grand unification of neutron stars

    PubMed Central

    Kaspi, Victoria M.

    2010-01-01

    The last decade has shown us that the observational properties of neutron stars are remarkably diverse. From magnetars to rotating radio transients, from radio pulsars to isolated neutron stars, from central compact objects to millisecond pulsars, observational manifestations of neutron stars are surprisingly varied, with most properties totally unpredicted. The challenge is to establish an overarching physical theory of neutron stars and their birth properties that can explain this great diversity. Here I survey the disparate neutron stars classes, describe their properties, and highlight results made possible by the Chandra X-Ray Observatory, in celebration of its 10th anniversary. Finally, I describe the current status of efforts at physical “grand unification” of this wealth of observational phenomena, and comment on possibilities for Chandra’s next decade in this field. PMID:20404205

  4. Neutron protein crystallography in JAERI

    NASA Astrophysics Data System (ADS)

    Tanaka, I.

    2004-07-01

    Neutron diffraction provides an experimental method of directly locating hy- drogen atoms in proteins. After developing an original neutron detector (neutron imaging plate) and a novel practical neutron monochromator (elastically bent perfect Si monochro- mator), BIX-type diffractometers which were equipped with these tools were e+/-ciently constructed at JRR-3 in Japan Atomic Energy Research Institute (JAERI), Japan and they have finished many protein crystallographic measurements and interesting results have come one after another. At the same time a method of growing large protein single crystals and a database of hydrogen and hydration have also been developed. In the near future, a pulsed neutron diffractometer for biological macromolecules has been proposed at J-PARC in JAERI.

  5. Methods for Neutron Spectrometry

    DOE R&D Accomplishments Database

    Brockhouse, Bertram N.

    1961-01-09

    The appropriate theories and the general philosophy of methods of measurement and treatment of data neutron spectrometry are discussed. Methods of analysis of results for liquids using the Van Hove formulation, and for crystals using the Born-von Karman theory, are reviewed. The most useful of the available methods of measurement are considered to be the crystal spectrometer methods and the pulsed monoenergetic beam/time-of-flight method. Pulsed-beam spectrometers have the advantage of higher counting rates than crystal spectrometers, especially in view of the fact that simultaneous measurements in several counters at different angles of scattering are possible in pulsed-beam spectrometers. The crystal spectrometer permits several valuable new types of specialized experiments to be performed, especially energy distribution measurements at constant momentum transfer. The Chalk River triple-axis crystal-spectrometer is discussed, with reference to its use in making the specialized experiments. The Chalk River rotating crystal (pulsed-beam) spectrometer is described, and a comparison of this type instrument with other pulsed-beam spectrometers is made. A partial outline of the theory of operation of rotating-crystal spectrometers is presented. The use of quartz-crystal filters for fast neutron elimination and for order elimination is discussed. (auth)

  6. Fast neutron environments.

    SciTech Connect

    Buchheit, Thomas Edward; Kotula, Paul Gabriel; Lu, Ping; Brewer, Luke N.; Goods, Steven Howard; Foiles, Stephen Martin; Puskar, Joseph David; Hattar, Khalid Mikhiel; Doyle, Barney Lee; Boyce, Brad Lee; Clark, Blythe G.

    2011-10-01

    The goal of this LDRD project is to develop a rapid first-order experimental procedure for the testing of advanced cladding materials that may be considered for generation IV nuclear reactors. In order to investigate this, a technique was developed to expose the coupons of potential materials to high displacement damage at elevated temperatures to simulate the neutron environment expected in Generation IV reactors. This was completed through a high temperature high-energy heavy-ion implantation. The mechanical properties of the ion irradiated region were tested by either micropillar compression or nanoindentation to determine the local properties, as a function of the implantation dose and exposure temperature. In order to directly compare the microstructural evolution and property degradation from the accelerated testing and classical neutron testing, 316L, 409, and 420 stainless steels were tested. In addition, two sets of diffusion couples from 316L and HT9 stainless steels with various refractory metals. This study has shown that if the ion irradiation size scale is taken into consideration when developing and analyzing the mechanical property data, significant insight into the structural properties of the potential cladding materials can be gained in about a week.

  7. Neutron activation for ITER

    SciTech Connect

    Barnes, C.W.; Loughlin, M.J.; Nishitani, Takeo

    1996-04-29

    There are three primary goals for the Neutron Activation system for ITER: maintain a robust relative measure of fusion power with stability and high dynamic range (7 orders of magnitude); allow an absolute calibration of fusion power (energy); and provide a flexible and reliable system for materials testing. The nature of the activation technique is such that stability and high dynamic range can be intrinsic properties of the system. It has also been the technique that demonstrated (on JET and TFTR) the highest accuracy neutron measurements in DT operation. Since the gamma-ray detectors are not located on the tokamak and are therefore amenable to accurate characterization, and if material foils are placed very close to the ITER plasma with minimum scattering or attenuation, high overall accuracy in the fusion energy production (7--10%) should be achievable on ITER. In the paper, a conceptual design is presented. A system is shown to be capable of meeting these three goals, also detailed design issues remain to be solved.

  8. Modeling Binary Neutron Stars

    NASA Astrophysics Data System (ADS)

    Park, Conner; Read, Jocelyn; Flynn, Eric; Lockett-Ruiz, Veronica

    2016-03-01

    Gravitational waves, predicted by Einstein's Theory of Relativity, are a new frontier in astronomical observation we can use to observe phenomena in the universe. Laser Interferometer Gravitational wave Observatory (LIGO) is currently searching for gravitational wave signals, and requires accurate predictions in order to best extract astronomical signals from all other sources of fluctuations. The focus of my research is in increasing the accuracy of Post-Newtonian models of binary neutron star coalescence to match the computationally expensive Numerical models. Numerical simulations can take months to compute a couple of milliseconds of signal whereas the Post-Newtonian can generate similar signals in seconds. However the Post-Newtonian model is an approximation, e.g. the Taylor T4 Post-Newtonian model assumes that the two bodies in the binary neutron star system are point charges. To increase the effectiveness of the approximation, I added in tidal effects, resonance frequencies, and a windowing function. Using these observed effects from simulations significantly increases the Post-Newtonian model's similarity to the Numerical signal.

  9. Neutron Star Mass Distribution in Binaries

    NASA Astrophysics Data System (ADS)

    Lee, Chang-Hwan; Kim, Young-Min

    2016-05-01

    Massive neutron stars with ∼ 2Mʘ have been observed in neutron star-white dwarf binaries. On the other hand, well-measured neutron star masses in double-neutron-star binaries are still consistent with the limit of 1.5Mʘ. These observations raised questions on the neutron star equations of state and the neutron star binary evolution processes. In this presentation, a hypothesis of super-Eddington accretion and its implications are discussed. We argue that a 2Mʘ neutron star is an outcome of the super-Eddington accretion during the evolution of neutron star-white dwarf binary progenitors. We also suggest the possibility of the existence of new type of neutron star binary which consists of a typical neutron star and a massive compact companion (high-mass neutron star or black hole) with M ≥ 2Mʘ.

  10. Characterisation of neutron fields at Cernavoda NPP.

    PubMed

    Cauwels, Vanessa; Vanhavere, Filip; Dumitrescu, Dorin; Chirosca, Alecsandru; Hager, Luke; Million, Marc; Bartz, James

    2013-04-01

    Near a nuclear reactor or a fuel container, mixed neutron/gamma fields are very common, necessitating routine neutron dosimetry. Accurate neutron dosimetry is complicated by the fact that the neutron effective dose is strongly dependent on the neutron energy and the direction distribution of the neutron fluence. Neutron field characterisation is indispensable if one wants to obtain a reliable estimate for the neutron dose. A measurement campaign at CANDU nuclear power plant located in Cernavoda, Romania, was set up to characterise the neutron fields in four different locations and to investigate the behaviour of different neutron personal dosemeters. This investigation intends to assist in choosing a suitable neutron dosimetry system at this nuclear power plant. PMID:22874895

  11. 70 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set.

    Energy Science and Technology Software Center (ESTSC)

    1984-10-29

    Version 00 These multigroup cross sections are used in fast reactor calculations. The benchmark calculations for the 23 fast critical assemblies used in the benchmark tests of JFS-2 were performed with one-dimensional diffusion theory by using the JFS-3-J2 set.

  12. Neutron tube design study for boron neutron capture therapy application

    SciTech Connect

    Verbeke, J.M.; Lee, Y.; Leung, K.N.; Vujic, J.; Williams, M.D.; Wu, L.K.; Zahir, N.

    1999-05-06

    Radio-frequency (RF) driven ion sources are being developed in Lawrence Berkeley National Laboratory (LBNL) for sealed-accelerator-tube neutron generator application. By using a 5-cm-diameter RF-driven multicusp source H{sup +} yields over 95% have been achieved. These experimental findings will enable one to develop compact neutron generators based on the D-D or D-T fusion reactions. In this new neutron generator, the ion source, the accelerator and the target are all housed in a sealed metal container without external pumping. Recent moderator design simulation studies have shown that 14 MeV neutrons could be moderated to therapeutically useful energy ranges for boron neutron capture therapy (BNCT). The dose near the center of the brain with optimized moderators is about 65% higher than the dose obtained from a typical neutron spectrum produced by the Brookhaven Medical Research Reactor (BMRR), and is comparable to the dose obtained by other accelerator-based neutron sources. With a 120 keV and 1 A deuteron beam, a treatment time of {approx}35 minutes is estimated for BNCT.

  13. Search for Neutron Anti-Neutron Oscillation using Cold Neutron Beams with Focusing Optics

    NASA Astrophysics Data System (ADS)

    Shimizu, Hirohiko; NNBar Collaboration

    2014-09-01

    The electric charge of neutrons is experimentally known as less than 10-21 e and considered as exactly zero and the transition between neutron and anti-neutron is allowed in terms of the conservation of the electric charge but is considered forbidden according to the empirical conservation law of the baryon number. On the other hand, the existence of physical processes which violates the conservation of the baryon number is required in the Sakharov's conditions to explain the baryon assymmetry in the big-bang cosmology. The search for the neutron antineutron (n n) oscillation offers information the baryon number violation with the Δ (B - L) = 2 complementary to the attempts with Δ (B - L) = 0 . The sensitivity to the n n oscillation has been improved by searching for the instability of nuclei via n n oscillation in large-scale deep-underground experiments, which are now limited by the background. On the other hand, the improvement of accelerator-driven neutron sources and transport optics of slow neutron beams have introduced new possibility to improve the sensitivity to n n by orders of magnitude. In this paper, we discuss the experimental sensitivity to n n oscillation with accelerator-based neutron sources and neutron focusing optics.

  14. Accelerators and Neutron Capture Therapy

    NASA Astrophysics Data System (ADS)

    Burlon, A. A.; Kreiner, A. J.; Valda, A.

    2002-08-01

    Within the frame of Accelerator Based Boron Neutron Capture Therapy (AB-BNCT), the 7Li (p,n) 7Be reaction, relatively near its energy threshold is one of the most promising, due to its high yield and low neutron energy. In this work a thick LiF target irradiated with a proton beam was studied as a neutron source. The 1.88-2.0 MeV proton beam was produced by the tandem accelerator TANDAR at CNEA's facilities in Buenos Aires. A water-filled phantom, containing a boron sample was irradiated with the resulting neutron flux. The 10B(n,αγ)7Li boron neutron capture reaction produces a 0.478 MeV gamma ray in 94% of the cases. The neutron yield was measured through the detection of this gamma ray using a hyperpure germanium detector with an anti-Compton shield. In addition, the thermal neutron flux was evaluated at different depths inside the phantom using bare and Cd-covered gold foils. A maximum neutron thermal flux of 1.4×108 cm-2s-1mA-1 was obtained at 4.2 cm from the phantom surface. In order to optimize the design of the neutron production target and the beam shaping assembly extensive Monte Carlo Neutron and Photon (MCNP) simulations have been performed. Neutron fields from a thick LiF and a Li metal target (with both a D2O-graphite and a Al/AlF3-graphite moderator/reflector assembly) were evaluated along the centerline of a head and a whole body phantom. Simulations were carried out for 1.89, 2.0 and 2.3 MeV proton beams. The results show that it is more advantageous to irradiate the target with 2.3 MeV near-resonance protons, instead of very near threshold, because of the higher neutron yield at this energy. On the other hand, the Al/AlF3-graphite exhibits a more efficient performance than D2O in terms of tumor to maximum healthy tissue dose ratio. Treatment times of less than 15 min and tumor control probabilities larger than 98% are obtained for a 50 mA, 2.3 MeV proton beam. The alternative neutron-producing reaction 13C(d,n) is also briefly reviewed. A

  15. Design of multidirectional neutron beams for boron neutron capture synovectomy

    SciTech Connect

    Gierga, D.P.; Yanch, J.C.; Shefer, R.E.

    1997-12-01

    Boron neutron capture synovectomy (BNCS) is a potential application of the {sup 10}B(n, a) {sup 7}Li reaction for the treatment of rheumatoid arthritis. The target of therapy is the synovial membrane. Rheumatoid synovium is greatly inflamed and is the source of the discomfort and disability associated with the disease. The BNCS proposes to destroy the synovium by first injecting a boron-labeled compound into the joint space and then irradiating the joint with a neutron beam. This study discusses the design of a multidirectional neutron beam for BNCS.

  16. Compact thermal neutron sensors for moderator-based neutron spectrometers.

    PubMed

    Pola, A; Bortot, D; Introini, M V; Bedogni, R; Gentile, A; Esposito, A; Gómez-Ros, J M; Passoth, E; Prokofiev, A

    2014-10-01

    In the framework of the NESCOFI@BTF project of the Italian Institute of Nuclear Physics, different types of active thermal neutron sensors were studied by coupling semiconductor devices with a suitable radiator. The objective was to develop a detector of small dimensions with a proper sensitivity to use at different positions in a novel moderating assembly for neutron spectrometry. This work discusses the experimental activity carried out in the framework of the ERINDA program (PAC 3/9 2012) to characterise the performance of a thermal neutron pulse detector based on (6)Li. PMID:24277874

  17. Neutron producing reactions in PuBe neutron sources

    NASA Astrophysics Data System (ADS)

    Bagi, János; Lakosi, László; Nguyen, Cong Tam

    2016-01-01

    There are a plenty of out-of-use plutonium-beryllium neutron sources in Eastern Europe presenting both nuclear safeguards and security issues. Typically, their actual Pu content is not known. In the last couple of years different non-destructive methods were developed for their characterization. For such methods detailed knowledge of the nuclear reactions taking place within the source is necessary. In this paper we investigate the role of the neutron producing reactions, their contribution to the neutron yield and their dependence on the properties of the source.

  18. Iodine neutron capture therapy

    NASA Astrophysics Data System (ADS)

    Ahmed, Kazi Fariduddin

    A new technique, Iodine Neutron Capture Therapy (INCT) is proposed to treat hyperthyroidism in people. Present thyroid therapies, surgical removal and 131I treatment, result in hypothyroidism and, for 131I, involve protracted treatment times and excessive whole-body radiation doses. The new technique involves using a low energy neutron beam to convert a fraction of the natural iodine stored in the thyroid to radioactive 128I, which has a 24-minute half-life and decays by emitting 2.12-MeV beta particles. The beta particles are absorbed in and damage some thyroid tissue cells and consequently reduce the production and release of thyroid hormones to the blood stream. Treatment times and whole-body radiation doses are thus reduced substantially. This dissertation addresses the first of the several steps needed to obtain medical profession acceptance and regulatory approval to implement this therapy. As with other such programs, initial feasibility is established by performing experiments on suitable small mammals. Laboratory rats were used and their thyroids were exposed to the beta particles coming from small encapsulated amounts of 128I. Masses of 89.0 mg reagent-grade elemental iodine crystals have been activated in the ISU AGN-201 reactor to provide 0.033 mBq of 128I. This activity delivers 0.2 Gy to the thyroid gland of 300-g male rats having fresh thyroid tissue masses of ˜20 mg. Larger iodine masses are used to provide greater doses. The activated iodine is encapsulated to form a thin (0.16 cm 2/mg) patch that is then applied directly to the surgically exposed thyroid of an anesthetized rat. Direct neutron irradiation of a rat's thyroid was not possible due to its small size. Direct in-vivo exposure of the thyroid of the rat to the emitted radiation from 128I is allowed to continue for 2.5 hours (6 half-lives). Pre- and post-exposure blood samples are taken to quantify thyroid hormone levels. The serum T4 concentration is measured by radioimmunoassay at

  19. Research on fusion neutron sources

    NASA Astrophysics Data System (ADS)

    Gryaznevich, M. P.

    2012-06-01

    The use of fusion devices as powerful neutron sources has been discussed for decades. Whereas the successful route to a commercial fusion power reactor demands steady state stable operation combined with the high efficiency required to make electricity production economic, the alternative approach to advancing the use of fusion is free of many of complications connected with the requirements for economic power generation and uses the already achieved knowledge of Fusion physics and developed Fusion technologies. "Fusion for Neutrons" (F4N), has now been re-visited, inspired by recent progress achieved on comparably compact fusion devices, based on the Spherical Tokamak (ST) concept. Freed from the requirement to produce much more electricity than used to drive it, a fusion neutron source could be efficiently used for many commercial applications, and also to support the goal of producing energy by nuclear power. The possibility to use a small or medium size ST as a powerful or intense steady-state fusion neutron source (FNS) is discussed in this paper in comparison with the use of traditional high aspect ratio tokamaks. An overview of various conceptual designs of compact fusion neutron sources based on the ST concept is given and they are compared with a recently proposed Super Compact Fusion Neutron Source (SCFNS), with major radius as low as 0.5 metres but still able to produce several MW of neutrons in a steady-state regime.

  20. Gravitational Waves from Neutron Stars

    NASA Astrophysics Data System (ADS)

    Kokkotas, Konstantinos

    2016-03-01

    Neutron stars are the densest objects in the present Universe, attaining physical conditions of matter that cannot be replicated on Earth. These unique and irreproducible laboratories allow us to study physics in some of its most extreme regimes. More importantly, however, neutron stars allow us to formulate a number of fundamental questions that explore, in an intricate manner, the boundaries of our understanding of physics and of the Universe. The multifaceted nature of neutron stars involves a delicate interplay among astrophysics, gravitational physics, and nuclear physics. The research in the physics and astrophysics of neutron stars is expected to flourish and thrive in the next decade. The imminent direct detection of gravitational waves will turn gravitational physics into an observational science, and will provide us with a unique opportunity to make major breakthroughs in gravitational physics, in particle and high-energy astrophysics. These waves, which represent a basic prediction of Einstein's theory of general relativity but have yet to be detected directly, are produced in copious amounts, for instance, by tight binary neutron star and black hole systems, supernovae explosions, non-axisymmetric or unstable spinning neutron stars. The focus of the talk will be on the neutron star instabilities induced by rotation and the magnetic field. The conditions for the onset of these instabilities and their efficiency in gravitational waves will be presented. Finally, the dependence of the results and their impact on astrophysics and especially nuclear physics will be discussed.

  1. Towards a laser neutron driver.

    PubMed

    Keskilidou, E; Moustaizis, S D; Mikheev, L; Auvray, P; Rouiller, C

    2005-01-01

    During the last few years, important experimental investigations have been made concerning the possibility of induced nuclear fission of high-Z elements by electromagnetic interaction (photofission, electron fission, neutron fission). Fast ions, neutrons and fission fragments from such interactions can be used to pump a laser medium, to produce energy from the (232)Th-(233)U nuclear fission cycle. The main aim of the present work is to study a three-step process, in a relatively new experimental scheme, in order to improve the number of both neutrons and fast ions. In the proposed scheme, high-energy particles and photons are produced by high-intensity laser beam interaction with a solid or gas target, which are utilized later on to trigger the nuclear reactions for the production of (photo) neutrons. These neutrons can give rise to fission of (232)Th that leads through a cascade of decays to (233)U --a highly fissionable material. Such a process will enhance, by an important factor, the final neutron flux and the energetic fission fragments. The use of a high intensity pulsed laser beam will control the turn-on and turn-off of the nuclear reactions and allow one to ensure the security of the whole operation. Finally, the produced neutrons are used to accomplish a major population inversion in an appropriate gas medium for the last stage of amplification of a high-contrast ultra-short laser seed pulse. PMID:15990323

  2. Monoenergetic Neutrons for Stellar Applications

    NASA Astrophysics Data System (ADS)

    Mosconi, M.; Heil, M.; Käppeler, F.; Plag, R.; Mengoni, A.; Nolte, R.

    2009-09-01

    With modern techniques, neutron-capture cross sections can be determined with uncertainties of a few percent. However, Maxwellian averaged cross sections calculated from such data require a correction (because low-lying excited states are thermally populated in the hot stellar photon bath) which has to be determined by theoretical calculations. These calculations can be improved with information from indirect measurements, in particular by the inelastic scattering cross section. For low-lying levels, the inelastically scattered neutrons are difficult to separate from the dominant elastic channel. This problem is best solved by means of pulsed, monoenergetic neutron beams. For this reason, a pulsed beam of 30 keV neutrons with an energy spread of 7 to 9 keV FWHM and a width from 10 to 15 ns has been produced at Forschungszentrum Karlsruhe using the 7Li(p, n)7Be reaction directly at the reaction threshold. With this neutron beam the inelastic scattering cross section of the first excited level at 9.75 keV in 187Os was determined with a relative uncertainty of 6%. The use of monoenergetic neutron beams has been further pursued at the Physikalisch-Technische Bundesanstalt in Braunschweig, including the 3H(p, n)3He reaction for producing neutrons with an energy of 64 keV.

  3. Controlling neutron orbital angular momentum.

    PubMed

    Clark, Charles W; Barankov, Roman; Huber, Michael G; Arif, Muhammad; Cory, David G; Pushin, Dmitry A

    2015-09-24

    The quantized orbital angular momentum (OAM) of photons offers an additional degree of freedom and topological protection from noise. Photonic OAM states have therefore been exploited in various applications ranging from studies of quantum entanglement and quantum information science to imaging. The OAM states of electron beams have been shown to be similarly useful, for example in rotating nanoparticles and determining the chirality of crystals. However, although neutrons--as massive, penetrating and neutral particles--are important in materials characterization, quantum information and studies of the foundations of quantum mechanics, OAM control of neutrons has yet to be achieved. Here, we demonstrate OAM control of neutrons using macroscopic spiral phase plates that apply a 'twist' to an input neutron beam. The twisted neutron beams are analysed with neutron interferometry. Our techniques, applied to spatially incoherent beams, demonstrate both the addition of quantum angular momenta along the direction of propagation, effected by multiple spiral phase plates, and the conservation of topological charge with respect to uniform phase fluctuations. Neutron-based studies of quantum information science, the foundations of quantum mechanics, and scattering and imaging of magnetic, superconducting and chiral materials have until now been limited to three degrees of freedom: spin, path and energy. The optimization of OAM control, leading to well defined values of OAM, would provide an additional quantized degree of freedom for such studies. PMID:26399831

  4. Neutron capture reactions at DANCE

    SciTech Connect

    Bredeweg, T. A.

    2008-05-12

    The Detector for Advanced Neutron Capture Experiments (DANCE) is a 4{pi} BaF{sub 2} array consisting of 160 active detector elements. The primary purpose of the array is to perform neutron capture cross section measurements on small (> or approx.100 {mu}g) and/or radioactive (< or approx. 100 mCi) species. The measurements made possible with this array will be useful in answering outstanding questions in the areas of national security, threat reduction, nuclear astrophysics, advanced reactor design and accelerator transmutation of waste. Since the commissioning of DANCE we have performed neutron capture cross section measurements on a wide array of medium to heavy mass nuclides. Measurements to date include neutron capture cross sections on {sup 241,243}Am, neutron capture and neutron-induced fission cross sections and capture-to-fission ratio ({alpha} = {sigma}{sub {gamma}}/{sigma}{sub f}) for {sup 235}U using a new fission-tagging detector as well as neutron capture cross sections for several astrophysics branch-point nuclei. Results from several of these measurements will be presented along with a discussion of additional physics information that can be extracted from the DANCE data.

  5. Neutron capture reactions at DANCE

    NASA Astrophysics Data System (ADS)

    Bredeweg, T. A.

    2008-05-01

    The Detector for Advanced Neutron Capture Experiments (DANCE) is a 4π BaF2 array consisting of 160 active detector elements. The primary purpose of the array is to perform neutron capture cross section measurements on small (>~100 μg) and/or radioactive (<~100 mCi) species. The measurements made possible with this array will be useful in answering outstanding questions in the areas of national security, threat reduction, nuclear astrophysics, advanced reactor design and accelerator transmutation of waste. Since the commissioning of DANCE we have performed neutron capture cross section measurements on a wide array of medium to heavy mass nuclides. Measurements to date include neutron capture cross sections on 241,243Am, neutron capture and neutron-induced fission cross sections and capture-to-fission ratio (α = σγ/σf) for 235U using a new fission-tagging detector as well as neutron capture cross sections for several astrophysics branch-point nuclei. Results from several of these measurements will be presented along with a discussion of additional physics information that can be extracted from the DANCE data.

  6. Neutron capture therapy with deep tissue penetration using capillary neutron focusing

    DOEpatents

    Peurrung, Anthony J.

    1997-01-01

    An improved method for delivering thermal neutrons to a subsurface cancer or tumor which has been first doped with a dopant having a high cross section for neutron capture. The improvement is the use of a guide tube in cooperation with a capillary neutron focusing apparatus, or neutron focusing lens, for directing neutrons to the tumor, and thereby avoiding damage to surrounding tissue.

  7. THERMAL NEUTRON INTENSITIES IN SOILS IRRADIATED BY FAST NEUTRONS FROM POINT SOURCES. (R825549C054)

    EPA Science Inventory

    Thermal-neutron fluences in soil are reported for selected fast-neutron sources, selected soil types, and selected irradiation geometries. Sources include 14 MeV neutrons from accelerators, neutrons from spontaneously fissioning 252Cf, and neutrons produced from alp...

  8. Boron nitride solid state neutron detector

    DOEpatents

    Doty, F. Patrick

    2004-04-27

    The present invention describes an apparatus useful for detecting neutrons, and particularly for detecting thermal neutrons, while remaining insensitive to gamma radiation. Neutrons are detected by direct measurement of current pulses produced by an interaction of the neutrons with hexagonal pyrolytic boron nitride.

  9. Applications of {sup 3}He neutron detectors

    SciTech Connect

    Testov, D. A.; Briancon, Ch.; Dmitriev, S. N.; Yeremin, A. V.; Penionzhkevich, Yu. E.; Pyatkov, Yu. V.; Sokol, E. A.

    2009-01-15

    Neutron detectors with {sup 3}He-filled proportional counters are described. The use of these detectors in measuring the probability of neutron emission (in particular, multiparticle neutron emission) after the {beta} decay of neutron-rich nuclei and in studying rare events of spontaneous fission of superheavy nuclei is considered.

  10. Neutron detector and fabrication method thereof

    DOEpatents

    Bhandari, Harish B.; Nagarkar, Vivek V.; Ovechkina, Olena E.

    2016-08-16

    A neutron detector and a method for fabricating a neutron detector. The neutron detector includes a photodetector, and a solid-state scintillator operatively coupled to the photodetector. In one aspect, the method for fabricating a neutron detector includes providing a photodetector, and depositing a solid-state scintillator on the photodetector to form a detector structure.

  11. NERO-The Neutron Emission Ratio Observer

    NASA Astrophysics Data System (ADS)

    Lorusso, Giuseppe; Pereira, Jorque; Hosmer, Paul; Kratz, Karl Ludvig; Montes, Fernando; Reeder, Paul; Santi, Peter; Schatz, Hendrik

    2007-10-01

    The Neutron Emission Ratio Observer (NERO), has been constructed for the use at the National Superconducting Cyclotron Laboratory to work in conjunction with the NSCL Beta Counting System in order to detect β-delayed neutrons. The design of the detector provides high and flat efficiency for a wide range of neutron energies, as well as a low neutron background.

  12. HEND Maps of Fast Neutrons

    NASA Technical Reports Server (NTRS)

    2002-01-01

    Observations by NASA's 2001 Mars Odyssey spacecraft show a global view of Mars in high-energy, or fast, neutrons. These maps are based on data acquired by the high-energy neutron detector, one of the instruments in the gamma ray spectrometer suite. Fast neutrons, like epithermal neutrons, are sensitive to the presence of hydrogen. Unlike epithermal neutrons, however, they are not affected by the presence of carbon dioxide, which at the time of these observations covered the north polar area as 'dry ice' frost. The low flux of fast neutrons (blue and purple colors) in the north polar region suggests an abundance of hydrogen in the soil comparable to that determined in the south from the flux of epithermal neutrons. These observations were acquired during the first two months of mapping operations. Contours of topography are superimposed on these maps for geographic reference.

    NASA's Jet Propulsion Laboratory manages the 2001 Mars Odyssey mission for NASA's Office of Space Science, Washington, D.C. Investigators at Arizona State University in Tempe, the University of Arizona in Tucson, and NASA's Johnson Space Center, Houston, operate the science instruments. The gamma-ray spectrometer was provided by the University of Arizona in collaboration with the Russian Aviation and Space Agency, which provided the high-energy neutron detector, and the Los Alamos National Laboratories, New Mexico, which provided the neutron spectrometer. Lockheed Martin Astronautics, Denver, is the prime contractor for the project, and developed and built the orbiter. Mission operations are conducted jointly from Lockheed Martin and from JPL, a division of the California Institute of Technology in Pasadena.

  13. Neutron star structure from QCD

    NASA Astrophysics Data System (ADS)

    Fraga, Eduardo S.; Kurkela, Aleksi; Vuorinen, Aleksi

    2016-03-01

    In this review article, we argue that our current understanding of the thermodynamic properties of cold QCD matter, originating from first principles calculations at high and low densities, can be used to efficiently constrain the macroscopic properties of neutron stars. In particular, we demonstrate that combining state-of-the-art results from Chiral Effective Theory and perturbative QCD with the current bounds on neutron star masses, the Equation of State of neutron star matter can be obtained to an accuracy better than 30% at all densities.

  14. Ground level neutron monitoring instruments

    NASA Astrophysics Data System (ADS)

    Philippov, Maxim; Makhmutov, Vladimir; Stozhkov, Yuri; Maksumov, Osman; Viktorov, Sergey; Kvashnin, Alexander; Kvashnin, Aleksandr

    In the scope of scientific collaboration between Lebedev Physical Institute RAS, University Mackenzie (Brazil) and National Space Institute (INPE, Brazil) we are currenty involved in the developing of neutron detector. This experimental device will help us to study high energy phenomena at the Sun and dynamic processes in the Earth's atmosphere. The device consists of several detecting modules. Each of them includes neutron detectors, pressure and temperature sensors. To receive data from each detecting module uses interface module that connects to computer via serial interface. We present and discuss first experimental results obtained by constructed neutron detector.

  15. Neutron detection via bubble chambers.

    PubMed

    Jordan, D V; Ely, J H; Peurrung, A J; Bond, L J; Collar, J I; Flake, M; Knopf, M A; Pitts, W K; Shaver, M; Sonnenschein, A; Smart, J E; Todd, L C

    2005-01-01

    Research investigating the application of pressure-cycled bubble chambers to fast neutron detection is described. Experiments with a Halon-filled chamber showed clear sensitivity to an AmBe neutron source and insensitivity to a (137)Cs gamma source. Bubble formation was documented using high-speed photography, and a ceramic piezo-electric transducer element registered the acoustic signature of bubble formation. In a second set of experiments, the bubble nucleation response of a Freon-134a chamber to an AmBe neutron source was documented with high-speed photography. PMID:16005238

  16. Amorphous Silicon Based Neutron Detector

    SciTech Connect

    Xu, Liwei

    2004-12-12

    Various large-scale neutron sources already build or to be constructed, are important for materials research and life science research. For all these neutron sources, neutron detectors are very important aspect. However, there is a lack of a high-performance and low-cost neutron beam monitor that provides time and temporal resolution. The objective of this SBIR Phase I research, collaboratively performed by Midwest Optoelectronics, LLC (MWOE), the University of Toledo (UT) and Oak Ridge National Laboratory (ORNL), is to demonstrate the feasibility for amorphous silicon based neutron beam monitors that are pixilated, reliable, durable, fully packaged, and fabricated with high yield using low-cost method. During the Phase I effort, work as been focused in the following areas: 1) Deposition of high quality, low-defect-density, low-stress a-Si films using very high frequency plasma enhanced chemical vapor deposition (VHF PECVD) at high deposition rate and with low device shunting; 2) Fabrication of Si/SiO2/metal/p/i/n/metal/n/i/p/metal/SiO2/ device for the detection of alpha particles which are daughter particles of neutrons through appropriate nuclear reactions; and 3) Testing of various devices fabricated for alpha and neutron detection; As the main results: · High quality, low-defect-density, low-stress a-Si films have been successfully deposited using VHF PECVD on various low-cost substrates; · Various single-junction and double junction detector devices have been fabricated; · The detector devices fabricated have been systematically tested and analyzed. · Some of the fabricated devices are found to successfully detect alpha particles. Further research is required to bring this Phase I work beyond the feasibility demonstration toward the final prototype devices. The success of this project will lead to a high-performance, low-cost, X-Y pixilated neutron beam monitor that could be used in all of the neutron facilities worldwide. In addition, the technologies

  17. QPO Constraints on Neutron Stars

    NASA Technical Reports Server (NTRS)

    Miller, M. Coleman

    2005-01-01

    The kilohertz frequencies of QPOs from accreting neutron star systems imply that they are generated in regions of strong gravity, close to the star. This suggests that observations of the QPOs can be used to constrain the properties of neutron stars themselves, and in particular to inform us about the properties of cold matter beyond nuclear densities. Here we discuss some relatively model-insensitive constraints that emerge from the kilohertz QPOs, as well as recent developments that may hint at phenomena related to unstable circular orbits outside neutron stars.

  18. Neutron Imaging Developments at LANSCE

    SciTech Connect

    Nelson, Ronald Owen; Hunter, James F.; Schirato, Richard C.; Vogel, Sven C.; Swift, Alicia L.; Ickes, Timothy Lee; Ward, William Carl; Losko, Adrian Simon; Tremsin, Anton; Sevanto, Sanna Annika; Espy, Michelle A.; Dickman, Lee Thoresen; Malone, Michael

    2015-10-29

    Thermal, epithermal, and high-energy neutrons are available from two spallation sources at the 800 MeV proton accelerator. Improvements in detectors and computing have enabled new capabilities that use the pulsed beam properties at LANSCE; these include amorphous Si (aSi) detectors, intensified charge-coupled device cameras, and micro-channel plates. Applications include water flow in living specimens, inclusions and fission products in uranium oxide, and high-energy neutron imaging using an aSi flat panel with ZnS(Ag) scintillator screen. images of a metal/plastic cylinder from photons, low-energy and high-energy neutrons are compared.

  19. Crystallization of dense neutron matter

    NASA Technical Reports Server (NTRS)

    Canuto, V.; Chitre, S. M.

    1974-01-01

    The equation of state for cold neutron matter at high density is studied in the t-matrix formulation, and it is shown that energetically it is convenient to have neutrons in a crystalline configuration rather than in a liquid state for values of the density exceeding 1600 Tg/cu cm. The study of the mechanical properties indicates that the system is stable against shearing stresses. A solid core in the deep interior of heavy neutron stars appears to offer the most plausible explanation of speed-ups observed in the Vela pulsar.

  20. Neutron emission prior to fission

    SciTech Connect

    Gavron, A.; Gayer, A.; Boissevain, J.; Britt, H.C.; Nix, J.R.; Sierk, A.J.; Grange, P.; Hassani, S.; Weidenmueller, H.A.; Beene, J.R.

    1986-01-01

    Neutron emission in the /sup 158/Er composite system is studied in order to investigate particle emission with energy spectrum and angular distribution in excess of statistical model predictions. Data are analyzed using a modified statistical model which incorporates effects due to nuclear dissipation, and also calculates neutron emission during the descent from the saddle to the scission point. Calculations consider the Kramers effect and the Transient effect. It is concluded that a detailed interpretation of enhanced neutron emission preceding fission in compound nucleus reactions is possible, and that an upper limit may be set on the reduced nuclear dissipation coefficient. 5 refs., 2 figs. (LEW)