Development And Benchmarking Of Higher Energy Neutron Transport Data Libraries

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Development and Benchmarking of Higher Energy Neutron Transport Data Libraries

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ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (727 download)

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Download or read book Development and Benchmarking of Higher Energy Neutron Transport Data Libraries written by and published by . This book was released on 1988 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Neutron cross-section evaluations covering the energy range from 10/sup /minus/11/ to 100 MeV have been prepared for several materials. The principal method used to generate this data base has employed statistical-preequilibrium nuclear models, sophisticated phase shift analyses, and R-matrix techniques. The library takes advantage of formats developed for Version 6 of the Evaluated Nuclear Data File, ENDF. Methods to efficiently utilize the ENDF/B-VI representation of this library in the MCNP Monte Carlo code have been developed. MCNP results using the new library have been compared with calculated results using codes or data based upon intranuclear cascade models. 7 refs., 8 figs.

Transport Data Libraries for Incident Proton and Neutron Energies to 100 MeV.

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ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (727 download)

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Book Synopsis Transport Data Libraries for Incident Proton and Neutron Energies to 100 MeV. by :

Download or read book Transport Data Libraries for Incident Proton and Neutron Energies to 100 MeV. written by and published by . This book was released on 1989 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: A joint effort between the Applied Nuclear Science and Radiation Transport groups at Los Alamos has begun to develop and implement proton, neutron, and photon transport libraries for incident energies up to 100 MeV. The major steps involved in this effort are: (1) development of evaluated (ENDF/B) data formats appropriate for higher energies; (2) extension of low-energy nuclear physics theoretical models for applicability up to 100 MeV; (3) calculation and evaluation of nuclear data in ENDF/B-VI format for appropriate materials up to 100 MeV; (4) development of processing code capabilities to handle the higher energy data; and (5) development of the appropriate interfaces and code patches for use of data in transport codes such as MCNP. In this paper we mainly discuss the development of the basic transport data library, items (2) and (3) above, and summarize the remaining activities. 14 refs.

Improved Neutron Capture Data and Evaluation with Statistical Nuclear Structure Models for Transport Libraries

Author : Bradley William Sleaford
Publisher : ProQuest
ISBN 13 :
Total Pages : 262 pages
Book Rating : 4.:/5 (35 download)

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Book Synopsis Improved Neutron Capture Data and Evaluation with Statistical Nuclear Structure Models for Transport Libraries by : Bradley William Sleaford

Download or read book Improved Neutron Capture Data and Evaluation with Statistical Nuclear Structure Models for Transport Libraries written by Bradley William Sleaford and published by ProQuest. This book was released on 2007 with total page 262 pages. Available in PDF, EPUB and Kindle. Book excerpt:

3-D Neutron Transport Benchmarks

Author : Toshikazu Takeda
Publisher :
ISBN 13 :
Total Pages : 96 pages
Book Rating : 4.:/5 (248 download)

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Book Synopsis 3-D Neutron Transport Benchmarks by : Toshikazu Takeda

Download or read book 3-D Neutron Transport Benchmarks written by Toshikazu Takeda and published by . This book was released on 1991 with total page 96 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Neutron Transport Benchmarks for Binary Stochastic Multiplying Media

Author : Ian Mack Davis
Publisher :
ISBN 13 :
Total Pages : 304 pages
Book Rating : 4.:/5 (614 download)

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Book Synopsis Neutron Transport Benchmarks for Binary Stochastic Multiplying Media by : Ian Mack Davis

Download or read book Neutron Transport Benchmarks for Binary Stochastic Multiplying Media written by Ian Mack Davis and published by . This book was released on 2005 with total page 304 pages. Available in PDF, EPUB and Kindle. Book excerpt: Benchmark calculations are performed for neutron transport in a two material (binary) stochastic multiplying medium. Spatial, angular, and energy dependence are included. The problem considered is based on a fuel assembly of a common pressurized water nuclear reactor. The mean chord length through the assembly is determined and used as the planar geometry system length. According to assumed or calculated material distributions, this system length is populated with alternating fuel and moderator segments of random size. Neutron flux distributions are numerically computed using a discretized form of the Boltzmann transport equation employing diffusion synthetic acceleration. Average quantities (group fluxes and k-eigenvalue) and variances are calculated from an ensemble of realizations of the mixing statistics. The effects of varying two parameters in the fuel, two different boundary conditions, and three different sets of mixing statistics are assessed. A probability distribution function (PDF) of the k-eigenvalue is generated and compared with previous research. Atomic mix solutions are compared with these benchmark ensemble average flux and k-eigenvalue solutions. Mixing statistics with large standard deviations give the most widely varying ensemble solutions of the flux and k-eigenvalue. The shape of the k-eigenvalue PDF qualitatively agrees with previous work. Its overall shape is independent of variations in fuel cross-sections for the problems considered, but its width is impacted by these variations. Statistical distributions with smaller standard deviations alter the shape of this PDF toward a normal distribution. The atomic mix approximation yields large over-predictions of the ensemble average k-eigenvalue and under-predictions of the flux. Qualitatively correct flux shapes are obtained, however. These benchmark calculations indicate that a model which includes higher statistical moments of the mixing statistics is needed for accurate predictions of binary stochastic media k-eigenvalue problems. This is consistent with previous findings.

Neutron Capture Gamma-Ray Libraries for Nuclear Applications

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ISBN 13 :
Total Pages : 9 pages
Book Rating : 4.:/5 (873 download)

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Book Synopsis Neutron Capture Gamma-Ray Libraries for Nuclear Applications by :

Download or read book Neutron Capture Gamma-Ray Libraries for Nuclear Applications written by and published by . This book was released on 2010 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. this can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. They are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

Nuclear Data Libraries for Incident Neutrons and Protons to 150 MeV in ENDF-6 Format

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ISBN 13 :
Total Pages : 10 pages
Book Rating : 4.:/5 (684 download)

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Book Synopsis Nuclear Data Libraries for Incident Neutrons and Protons to 150 MeV in ENDF-6 Format by :

Download or read book Nuclear Data Libraries for Incident Neutrons and Protons to 150 MeV in ENDF-6 Format written by and published by . This book was released on 1998 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: As part of the Accelerator Production of Tritium (APT) program, an effort is underway at Los Alamos National Laboratory to develop nuclear data libraries for incident neutrons and protons to 150 MeV. The libraries will be used in the MCNP Monte Carlo code with appropriate linking to higher energy calculations with the LAHET intranuclear cascade code. The data code system will be used for design of an accelerator-based facility to produce tritium, and will provide information required for analysis of system performance, induced radiation doses, material activation, heating, damage, and shielding analysis. Because of their completeness, the libraries will also be useful for other accelerator-driven applications and for medical, shielding, and space applications at higher energies. The libraries are based primarily on nuclear model calculations with the GNASH reaction theory code, including thorough benchmarking of the model calculations against experimental data. All evaluations are in ENDF-6 format and include specification of production cross sections for light particles, gamma rays, and heavy recoil particles, energy angle correlated spectra for secondary light particles, and energy spectra for gamma rays and heavy recoil nuclei. The neutron evaluations are combined with ENDF/B-VI evaluations below 20 MeV. To date, neutron and proton evaluations have been completed for 2H, 12C, 14N, 16O, 27Al, {sup 28,29,30}Si, 4°Ca, {sup 50,52,53,54}Cr, {sup 54,56,57,58}Fe, {sup 58,60,61,62,64}Ni, {sup 182,183,184,186}W, and {sup 206,207,208}Pb.

One-dimensional Fast-neutron Transport Benchmark Calculations

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ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (96 download)

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Book Synopsis One-dimensional Fast-neutron Transport Benchmark Calculations by :

Download or read book One-dimensional Fast-neutron Transport Benchmark Calculations written by and published by . This book was released on 1975 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A series of state-of-the-art fast-neutron transport calculations were made for several simple shield materials of interest to the Defense Nuclear Agency. The calculations should serve as benchmarks for testing newly evaluated neutron cross-section data and multigroup libraries. Results are presented for one-dimensional spherical systems of iron, air, and concrete (SiO$sub 2$). The results include detailed flux spectra, broad-group fluxes versus distance, and convergence tests for representations of the P/sub l/ scattering kernel and for space, angle, and energy grids. The methods employed in converging the solutions differ from previous benchmark calculations in that sensitivity and point-energy techniques were used to determine the energy grid for the final results. (auth).

Neutron- and Proton-Induced Nuclear Data Libraries to 150 MeV for Accelerator-Driven Applications

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ISBN 13 :
Total Pages : 5 pages
Book Rating : 4.:/5 (684 download)

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Book Synopsis Neutron- and Proton-Induced Nuclear Data Libraries to 150 MeV for Accelerator-Driven Applications by :

Download or read book Neutron- and Proton-Induced Nuclear Data Libraries to 150 MeV for Accelerator-Driven Applications written by and published by . This book was released on 1997 with total page 5 pages. Available in PDF, EPUB and Kindle. Book excerpt: A program is underway at Los Alamos National Laboratory to develop nuclear data libraries for incident neutrons and protons to 150 MeV for accelerator-driven applications. These libraries will be used initially for design of an accelerator-based facility to produce tritium, including analysis of system performance, induced radiation doses, material activation, heating, damage, and shielding requirements. The libraries are based primarily on nuclear model calculations with the GNASH reaction theory code, including thorough benchmark of the model calculations against experimental data. All evaluations include specification of production cross sections for light particles, gamma rays, and heavy recoil particles, energy-angle correlated spectra for secondary light particles, and energy spectra for gamma rays and heavy recoil nuclei. The neutron evaluations are combined with ENDF/B-VI evaluations below 20 MeV. To date, neutron and proton evaluations have been completed or 2H, 12C, 16O, 27Al, 28,29,30Si, 40Ca, 54,56,57,58Fe, 182,183,184,186W, and 206,207,208Pb.

Nuclear Data Development and Shield Design for Neutrons Below 60 MeV

Author : William Bradley Wilson
Publisher :
ISBN 13 :
Total Pages : 114 pages
Book Rating : 4.:/5 (52 download)

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Book Synopsis Nuclear Data Development and Shield Design for Neutrons Below 60 MeV by : William Bradley Wilson

Download or read book Nuclear Data Development and Shield Design for Neutrons Below 60 MeV written by William Bradley Wilson and published by . This book was released on 1978 with total page 114 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Analytical Three-dimensional Neutron Transport Benchmarks for Verification of Nuclear Engineering Codes. Final Report

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ISBN 13 :
Total Pages : 278 pages
Book Rating : 4.:/5 (684 download)

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Book Synopsis Analytical Three-dimensional Neutron Transport Benchmarks for Verification of Nuclear Engineering Codes. Final Report by :

Download or read book Analytical Three-dimensional Neutron Transport Benchmarks for Verification of Nuclear Engineering Codes. Final Report written by and published by . This book was released on 1997 with total page 278 pages. Available in PDF, EPUB and Kindle. Book excerpt: Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade.

Capture Gamma-Ray Libraries for Nuclear Applications

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Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (873 download)

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Book Synopsis Capture Gamma-Ray Libraries for Nuclear Applications by :

Download or read book Capture Gamma-Ray Libraries for Nuclear Applications written by and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF has been used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90percent of all the decay energy an is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We use CASINO, a version of DICEBOX that is modified for this purpose. This can be used to simulate the neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modelling of unknown assemblies.

Experimental Transport Benchmarks for Physical Dosimetry to Support Development of Fast-Neutron Therapy with Neutron Capture Augmentation

Download Experimental Transport Benchmarks for Physical Dosimetry to Support Development of Fast-Neutron Therapy with Neutron Capture Augmentation PDF Online Free

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ISBN 13 :
Total Pages : pages
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Book Synopsis Experimental Transport Benchmarks for Physical Dosimetry to Support Development of Fast-Neutron Therapy with Neutron Capture Augmentation by :

Download or read book Experimental Transport Benchmarks for Physical Dosimetry to Support Development of Fast-Neutron Therapy with Neutron Capture Augmentation written by and published by . This book was released on 2006 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Idaho National Laboratory (INL), the University of Washington (UW) Neutron Therapy Center, the University of Essen (Germany) Neutron Therapy Clinic, and the Northern Illinois University(NIU) Institute for Neutron Therapy at Fermilab have been collaborating in the development of fast-neutron therapy (FNT) with concurrent neutron capture (NCT) augmentation [1,2]. As part of this effort, we have conducted measurements to produce suitable benchmark data as an aid in validation of advanced three-dimensional treatment planning methodologies required for successful administration of FNT/NCT. Free-beam spectral measurements as well as phantom measurements with Lucite{trademark} cylinders using thermal, resonance, and threshold activation foil techniques have now been completed at all three clinical accelerator facilities. The same protocol was used for all measurements to facilitate intercomparison of data. The results will be useful for further detailed characterization of the neutron beams of interest as well as for validation of various charged particle and neutron transport codes and methodologies for FNT/NCT computational dosimetry, such as MCNP [3], LAHET [4], and MINERVA [5].

DEVELOPMENT AND VALIDATION OF THE 7LI(P, N) NUCLEAR DATA LIBRARY AND ITS APPLICATION IN MONITORING OF INTERMEDIATE ENERGY NEUTRONS.

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ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (684 download)

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Book Synopsis DEVELOPMENT AND VALIDATION OF THE 7LI(P, N) NUCLEAR DATA LIBRARY AND ITS APPLICATION IN MONITORING OF INTERMEDIATE ENERGY NEUTRONS. by :

Download or read book DEVELOPMENT AND VALIDATION OF THE 7LI(P, N) NUCLEAR DATA LIBRARY AND ITS APPLICATION IN MONITORING OF INTERMEDIATE ENERGY NEUTRONS. written by and published by . This book was released on 2001 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:

Evaluated Cross-section Libraries and Kerma Factors for Neutrons Up to 100 MeV on 12C.

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Publisher :
ISBN 13 :
Total Pages : 91 pages
Book Rating : 4.:/5 (683 download)

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Book Synopsis Evaluated Cross-section Libraries and Kerma Factors for Neutrons Up to 100 MeV on 12C. by :

Download or read book Evaluated Cross-section Libraries and Kerma Factors for Neutrons Up to 100 MeV on 12C. written by and published by . This book was released on 1995 with total page 91 pages. Available in PDF, EPUB and Kindle. Book excerpt: A program is being carried out at Lawrence Livermore National Laboratory to develop high-energy evaluated nuclear data libraries for use in Monte Carlo simulations of cancer radiation therapy. In this report we describe evaluated cross sections and kerma factors for neutrons with incident energies up to 100 MeV on 12C. The aim of this effort is to incorporate advanced nuclear physics modeling methods, with new experimental measurements, to generate cross section libraries needed for an accurate simulation of dose deposition in fast neutron therapy. The evaluated libraries are based mainly on nuclear model calculations, benchmarked to experimental measurements where they exist. We use the GNASH code system, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms. The libraries tabulate elastic and nonelastic cross sections, angle-energy correlated production spectra for light ejectiles with A(less-than or equal to)and kinetic energies given to light ejectiles and heavy recoil fragments. The major steps involved in this effort are: (1) development and validation of nuclear models for incident energies up to 100 MeV; (2) collation of experimental measurements, including new results from Louvain-la-Nueve and Los Alamos; (3) extension of the Livermore ENDL formats for representing high-energy data; (4) calculation and evaluation of nuclear data; and (5) validation of the libraries. We describe the evaluations in detail, with particular emphasis on our new high-energy modeling developments. Our evaluations agree well with experimental measurements of integrated and differential cross sections. We compare our results with the recent ENDF/B-VI evaluation which extends up to 32 MeV.

Effects of Neutron Data Libraries and Criticality Codes on Iaea Criticality Benchmark Problems

Author : Md. Muslehuddin Sarker
Publisher :
ISBN 13 :
Total Pages : 39 pages
Book Rating : 4.:/5 (355 download)

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Book Synopsis Effects of Neutron Data Libraries and Criticality Codes on Iaea Criticality Benchmark Problems by : Md. Muslehuddin Sarker

Download or read book Effects of Neutron Data Libraries and Criticality Codes on Iaea Criticality Benchmark Problems written by Md. Muslehuddin Sarker and published by . This book was released on 1993 with total page 39 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

Download Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code PDF Online Free

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ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (16 download)

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Book Synopsis Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code by :

Download or read book Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code written by and published by . This book was released on 2000 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V & V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to calculate radiation dose due to the neutron environment around a MEA is shown. An uncertainty of a factor of three in the MEA calculations is shown to be due to uncertainties in the geometry modeling. It is believed that the methodology is sound and that good agreement between simulation and experiment has been demonstrated.