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in creator/contributor: "Little, R C"

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Nuclear data for MCNP

Sources of neutron and photon transport data are described as well as the processing of the evaluated data sets into continuous-energy and multigroup cross-section sets. The procedures for checking and validating the processed data are discussed. The question of why so many data sets are available is addressed by indicating the differences between data sets as well as their relative strengths and weaknesses. Suggestions are made to help the MCNP user in selecting appropriate cross-section sets.… more

Date: January 1, 1985

Creator: Little, R. C. & Seamon, R. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Neutron-induced photon production in MCNP

An improved method of neutron-induced photon production has been incorporated into the Monte Carlo transport code MCNP. The new method makes use of all partial photon-production reaction data provided by ENDF/B evaluators including photon-production cross sections as well as energy and angular distributions of secondary photons. This faithful utilization of sophisticated ENDF/B evaluations allows more precise MCNP calculations for several classes of coupled neutron-photon problems.

Date: January 1, 1983

Creator: Little, R. C. & Seamon, R. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Direct sulfur recovery during sorbent regeneration. Final report

The objective of this research project was to improve the direct elemental sulfur yields that occur during the regeneration of SO{sub 2}-saturated MgO-vermiculite sorbents (MagSorbents) by examining three approaches or strategies. The three approaches were regeneration-gas recycle, high-pressure regeneration, and catalytic reduction of the SO{sub 2} gas using a new catalyst developed by Research Triangle Institute (RTI). Prior to the project, Sorbent Technologies Corporation (Sorbtech) had deve… more

Date: August 1, 1993

Creator: Nelson, S. G. & Little, R. C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above approx.14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies … more

Date: January 1, 1987

Creator: Brenner, D. J.; Prael, R. E. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Determination of the effective delayed neutron fraction using MCNP4B

The capability to calculate effective delayed neutron fractions has now been implemented into MCNP4B and is in the testing phase. This option should prove to be most useful for multiplying systems which are not easily modeled using deterministic codes.

Date: February 2, 1999

Creator: Werner, C. J. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Impact of MCNP Unresolved Resonance Probability-Table Treatment on Uranium and Plutonium Benchmarks

A probability-table treatment recently has been incorporated into an intermediate version of the MCNP Monte Carlo code named MCNP4XS. This paper presents MCNP4XS results for a variety of uranium and plutonium criticality benchmarks, calculated with and without the probability-table treatment. It is shown that the probability-table treatment can produce small but significant reactivity changes for plutonium and {sup 233}U systems with intermediate spectra. More importantly, it can produce substa… more

Date: September 20, 1999

Creator: Mosteller, R. D. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Modern nuclear data for Monte Carlo codes

No Description Available.

Date: October 1, 2000

Creator: Little, R. C. & Frankle, S. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

TEMPERATURE DEPENDENCE OF THERMAL NEUTRONS FROM THE MOON

Planetary thermal neutron fluxes provide a sensitive proxy for mafic and feldspathic terranes, and are also necessary for translating measured gamma-ray line strengths to elemental abundances. Both functions require a model for near surface temperatures and a knowledge of the dependence of thermal neutron flux on temperature. We have explored this dependence for a representative sample of lunar soil compositions and surface temperatures using MCNP{trademark}. For all soil samples, the neutron d… more

Date: October 1, 2000

Creator: LITTLE, R.C.; FELDMAN, W. & AL, ET

Item Type: Article

Partner: UNT Libraries Government Documents Department

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New calculations for critical assemblies using MCNP4B

A suite of 41 criticality benchmarks has been modeled using MCNP{trademark} (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes {sup 233}U, {sup 235… more

Date: July 1997

Creator: Adams, A. A.; Frankle, S. C. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Improved photon production data for MCNP{trademark}

Computer simulations with MCNP are often used to obtain information from measurements of neutron induced gamma-ray spectra. For such simulations to be useful, the complicated spectra produced by a wide variety of nuclides must be reproduced, requiring high quality nuclear data. A previous assessment of the neutron induced photon production data in the MCNP data libraries indicated a need for improvement. The photon production data were often based on outdated experiments and binned in such wide… more

Date: April 1, 1998

Creator: Adams, A. A.; Frankle, S. C. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

ENDF/B-V cell comparisons and MCNP-3A

This study compares calculated results from the CELL-2 spectrum code to calculated results from the MCNP Monte Carlo code for typical pressurized water reactor (PWR) fuel lattices at elevated temperatures. The MCNP calculations represent the first analysis of commerical fuel lattices by continuous energy Monte Carlo at realistic temperaturs with ENDF/B-V data. Eigenvalue results demonstrate that the two codes agree to within the statistical uncertainty of MCNP. No systematic variations or biase… more

Date: January 1, 1988

Creator: Eich, W. J.; Eisenhart, L. D.; Little, R. C.; Mosteller, R. D. & Chao, J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Static and Dynamic Criticality: Are They Different?

Let us start by stating that this paper does not contain anything new. It only contains material that has been known for decades, but which is periodically forgotten. As such this paper is intended merely to refresh people's memories. We will also mention that this paper is an example of the occasional discrepancy between textbook methodologies and real world applications, in the sense that the conclusions reached here contradict what it says in most textbooks, i.e., most textbooks incorrectly … more

Date: December 12, 2003

Creator: Cullen, D E; Clouse, C J; Procassini, R & Little, R C

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Impact of MCNP unresolved resonance probability-table treatment on uranium and plutonium benchmarks

Versions of MCNP up through and including 4B have not accurately modeled neutron self-shielding effects in the unresolved resonance energy region. Recently, a probability-table treatment has been incorporated into a developmental version of MCNP. This paper presents MCNP results for a variety of uranium and plutonium critical benchmarks, calculated with and without the probability-table treatment.

Date: December 31, 1998

Creator: Mosteller, R. D. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Nuclear data libraries for incident neutrons and protons to 150 MeV in ENDF-6 format

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 analys… more

Date: February 1, 1998

Creator: Chadwick, M. B.; Frankle, S. C.; Little, R. C. & Young, P. G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Ion transport calculations using MCNP

The MCNP Monte Carlo code (version 4B) has been adapted to perform multi-dimensional ion transport calculations in amorphous media for microelectronics materials applications. In this application, focused ion beams are used to implant donor ions through a mask into an underlying semiconductor substrate, achieving tailored implantation profiles as a function of penetration depth with a minimum of radial spread past the mask edge. However, as the device feature size shrinks below submicron scale,… more

Date: November 1997

Creator: Keen, N. D.; Prinja, A. K.; Little, R. C. & Adams, K. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Photoneutron production in electron beam stop for dual-axis radiographic hydrotest facility (DARHT)

A beam stop design for an electron linear accelerator was analyzed from the perspective of photoneutron production and subsequent dose. Sophisticated nuclear data modeling codes were used to generate the photoneutron production cross sections and spectra that were then used in MCNP transport calculations. The resulting neutron dose exceeded limits for workers present in the experimental area while the accelerators are producing electron beam pulses. Therefore, the beam stop was redesigned to li… more

Date: March 1, 1998

Creator: Chadwick, M. B.; Brown, T. H. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Utilization of new 150-MeV neutron and proton evaluations in MCNP

MCNP{trademark} and LAHET{trademark} are two of the codes included in the LARAMIE (Los Alamos Radiation Modeling Interactive Environment) code system. Both MCNP and LAHET are three-dimensional continuous-energy Monte Carlo radiation transport codes. The capabilities of MCNP and LAHET are currently being merged into one code for the Accelerator Production of Tritium (APT) program at Los Alamos National Laboratory. Concurrently, a significant effort is underway to improve the accuracy of the phys… more

Date: October 1, 1997

Creator: Little, R. C.; Frankle, S. C.; Hughes III, H. G. & Prael, R. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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MCNP (Monte Carlo Neutron Photon) capabilities for nuclear well logging calculations

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. The general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo Neutron Photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another comp… more

Date: January 1, 1989

Creator: Forster, R. A.; Little, R. C. & Briesmeister, J. F.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Verification of MCNP and DANT/sys With the Analytic Benchmark Test Set

The recently published analytic benchmark test set has been used to verify the multigroup option of MCNP and also the deterministic DANT/sys series of codes for criticality calculations. All seventy-five problems of the test set give values for K{sub eff} accurate to at least five significant digits. Flux ratios and flux shapes are also available for many of the problems. All seventy-five problems have been run by both the MCNP and DANT/sys codes and comparisons to K{sub eff} and flux shapes ha… more

Date: September 20, 1999

Creator: Parsons, D. K.; Sood, A.; Forster, R. A. & Little, R. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Criticality benchmark calculations using PARTISN: Comparisons using MENDF5 and MENDF6 nuclear data libraries.

A project was undertaken to assess the MENDF5 and MENDF6 nuclear data libraries through the analysis of 86 critical assembly benchmarks using the LANL discrete ordinates transport code PARTISN. As an initial analysis of the effects of some limitations in the MENDF libraries, this current work assesses differences in k,,a calculations between the PARTISN cases (with MENDF5 and MENDF6 nuclear data libraries) and MCNP cases, and compares these results to the experimental data.

Date: January 1, 2003

Creator: Ellis, Ronald J.; Yugo, James J.; Frankle, S. C. (Stephanie C.) & Little, R. C. (Robert C.)

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Prototype demonstration of radiation therapy planning code system

This is the final report of a one-year, Laboratory-Directed Research and Development project at the Los Alamos National Laboratory (LANL). Radiation therapy planning is the process by which a radiation oncologist plans a treatment protocol for a patient preparing to undergo radiation therapy. The objective is to develop a protocol that delivers sufficient radiation dose to the entire tumor volume, while minimizing dose to healthy tissue. Radiation therapy planning, as currently practiced in the… more

Date: September 1, 1996

Creator: Little, R. C.; Adams, K. J.; Estes, G. P.; Hughes III, L. S. & Waters, L. S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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New probability table treatment in MCNP for unresolved resonances

An upgrade for MCNP has been implemented to sample the neutron cross sections in the unresolved resonance range using probability tables. These probability tables are generated with the cross section processor code NJOY, by using the evaluated statistical information about the resonances to calculate cumulative probability distribution functions for the microscopic total cross section. The elastic, fission, and radiative capture cross sections are also tabulated as the average values of each of… more

Date: April 1, 1998

Creator: Carter, L. L.; Little, R. C.; Hendricks, J. S. & MacFarlane, R. E.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Development and benchmarking of higher energy neutron transport data libraries

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 … more

Date: January 1, 1988

Creator: Arthur, E. D.; Young, P. G.; Perry, R. T.; Madland, D. G.; MacFarlane, R. E.; Little, R. C. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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{sup 7}Li(p,n) Nuclear Data Library for Incident Proton Energies to 150 MeV

Researchers at Los Alamos National Laboratory are considering the possibility of using the Low Energy Demonstration Accelerator (LEDA), constructed at LANSCE for the Accelerator Production of Tritium program (APT), as a neutron source. Evaluated nuclear data are needed for the p+{sup 7}Li reaction, to predict neutron production from thin and thick lithium targets. In this report we describe evaluation methods that make use of experimental data, and nuclear model calculations, to develop an ENDF… more

Date: November 1, 2000

Creator: Mashnik, S.; Chadwick, M. B.; Hughes, H. G.; Little, R. C.; MacFarlane, R. E.; Waters, L. S. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department