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Nuclear data for MCNP

Description: 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. 31 refs.
Date: January 1, 1985
Creator: Little, R.C. & Seamon, R.E.
Partner: UNT Libraries Government Documents Department

Neutron-induced photon production in MCNP

Description: 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.
Partner: UNT Libraries Government Documents Department

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

Description: 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 substantial reactivity increases for systems with large amounts of {sup 238}U and intermediate spectra.
Date: September 20, 1999
Creator: Mosteller, R.D. & Little, R.C.
Partner: UNT Libraries Government Documents Department

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

Description: 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 from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs.
Date: January 1, 1987
Creator: Brenner, D.J.; Prael, R.E. & Little, R.C.
Partner: UNT Libraries Government Documents Department

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

Description: 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 biases appear as a function of enrichment or temperatute. This study constitutes the first verification of the reaction rates of CELL-2 and its processed ENDF/B-V data library via accurate numerical benchmarks.
Date: January 1, 1988
Creator: Eich, W.J.; Eisenhart, L.D.; Little, R.C.; Mosteller, R.D. & Chao, J.
Partner: UNT Libraries Government Documents Department

Direct sulfur recovery during sorbent regeneration. Final report

Description: 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 developed a sorbent-regeneration process that yielded directly a pure elemental sulfur product. In the process, typically about 25 to 35 percent of the liberated S0{sub 2} was converted directly to elemental sulfur. The goal of this project was to achieve a conversion rate of over 90 percent. Good success was attained in the project. About 90 percent or more conversion was achieved with two of the approaches that were examined, regeneration-gas recycle and use of the RTI catalyst. Of these approaches, regeneration-gas recycle gave the best results (essentially 100 percent conversion in some cases). In the regeneration-gas recycle approach, saturated sorbent is simply heated to about 750{degree}C in a reducing gas (methane) atmosphere. During heating, a gas containing elemental sulfur, water vapor, H{sub 2}S, S0{sub 2}, and C0{sub 2} is evolved. The elemental sulfur and water vapor in the gas stream are condensed and removed, and the remaining gas is recycled back through the sorbent bed. After several recycles, the S0{sub 2} and H{sub 2}S completely disappear from the gas stream, and the stream contains only elemental sulfur, water vapor and C0{sub 2}.
Date: August 1, 1993
Creator: Nelson, S.G. & Little, R.C.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Partner: UNT Libraries Government Documents Department

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

Description: 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 {sup 2}H, {sup 12}C, {sup 14}N, {sup 16}O, {sup 27}Al, {sup 28,29,30}Si, {sup 40}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.
Date: February 1, 1998
Creator: Chadwick, M.B.; Frankle, S.C.; Little, R.C. & Young, P.G.
Partner: UNT Libraries Government Documents Department

New calculations for critical assemblies using MCNP4B

Description: 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}U, {sup 238}U, and {sup 239}Pu. The values of k{sub eff} for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, {sup 235}U, {sup 238}U, and {sup 239}Pu can have a significant impact on the values of k{sub eff}. In addition to the integral quantity k{sub eff}, several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k{sub eff}. Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data.
Date: July 1997
Creator: Adams, A. A.; Frankle, S. C. & Little, R. C.
Partner: UNT Libraries Government Documents Department

TEMPERATURE DEPENDENCE OF THERMAL NEUTRONS FROM THE MOON

Description: 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 density is found to be independent of temperature, in accord with neutron moderation theory. The thermal neutron flux, however, does vary with temperature in a way that depends on {Delta}, the ratio of macroscopic absorption to energy-loss cross sections of soil compositions. The weakest dependence is for the largest {Delta} (which corresponds to the Apollo 17 high Ti basalt in our soil selection), and the largest dependence is for the lowest {Delta} (which corresponds to ferroan anorthosite, [FAN] in our selection). For the lunar model simulated, the depth at which the thermal neutron population is most sensitive to temperature is {approx}30 g/cm{sup 2}.
Date: October 1, 2000
Creator: LITTLE, R.C.; FELDMAN, W. & AL, ET
Partner: UNT Libraries Government Documents Department

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

Description: 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 limit doses to acceptable values, consistent with the ALARA philosophy.
Date: March 1, 1998
Creator: Chadwick, M.B.; Brown, T.H. & Little, R.C.
Partner: UNT Libraries Government Documents Department

Ion transport calculations using MCNP

Description: 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, this radial migration of ions becomes important. Our goal is to simulate ion implantation in materials containing two and three dimensional heterogeneities for a variety of implant conditions (e.g., incident ion type, energy, incident angle, and problem geometry).
Date: November 1997
Creator: Keen, N. D.; Prinja, A. K.; Little, R. C. & Adams, K. J.
Partner: UNT Libraries Government Documents Department

Improved photon production data for MCNP{trademark}

Description: 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 energy groups as to be of limited value for some applications. This paper describes the work that is underway at Los Alamos National Laboratory to improve the photon production data for thermal neutron capture reactions. To date, high quality photon production data for each stable isotope of chlorine, chromium, iron, copper, and nickel have been obtained. The improved spectra have been incorporated into ENDF formatted evaluations and processed into corresponding MCNP data files. Similar improvements for aluminum, manganese, silicon, calcium, and vanadium are also planned. The methodology used to produce the spectra is discussed, and sample results for chlorine are presented.
Date: April 1, 1998
Creator: Adams, A. A.; Frankle, S. C. & Little, R. C.
Partner: UNT Libraries Government Documents Department

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

Description: 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.)
Partner: UNT Libraries Government Documents Department

Prototype demonstration of radiation therapy planning code system

Description: 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 field, suffers from inaccuracies made in modeling patient anatomy and radiation transport. This project investigated the ability to automatically model patient-specific, three-dimensional (3-D) geometries in advanced Los Alamos radiation transport codes (such as MCNP), and to efficiently generate accurate radiation dose profiles in these geometries via sophisticated physics modeling. Modem scientific visualization techniques were utilized. The long-term goal is that such a system could be used by a non-expert in a distributed computing environment to help plan the treatment protocol for any candidate radiation source. The improved accuracy offered by such a system promises increased efficacy and reduced costs for this important aspect of health care.
Date: September 1, 1996
Creator: Little, R.C.; Adams, K.J.; Estes, G.P.; Hughes, L.S. III & Waters, L.S.
Partner: UNT Libraries Government Documents Department

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

Description: 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 physics in the merged code. In particular, full nuclear-data evaluations (in ENDF6 format) for many materials of importance to APT are being produced for incident neutrons and protons up to an energy of 150-MeV. After processing, cross-section tables based on these new evaluations will be available for use fin the merged code. In order to utilize these new cross-section tables, significant enhancements are required for the merged code. Neutron cross-section tables for MCNP currently specify emission data for neutrons and photons only; the new evaluations also include complete neutron-induced data for protons, deuterons, tritons, and alphas. In addition, no provision in either MCNP or LAHET currently exists for the use of incident charged-particle tables other than for electrons. To accommodate the new neutron-induced data, it was first necessary to expand the format definition of an MCNP neutron cross-section table. The authors have prepared a 150-MeV neutron cross-section library in this expanded format for 15 nuclides. Modifications to MCNP have been implemented so that this expanded neutron library can be utilized.
Date: October 1, 1997
Creator: Little, R.C.; Frankle, S.C.; Hughes, H.G. III & Prael, R.E.
Partner: UNT Libraries Government Documents Department

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

Description: 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 have been made. Results from this verification exercise are given below.
Date: September 20, 1999
Creator: Parsons, D.K.; Sood, A.; Forster, R.A. & Little, R.C.
Partner: UNT Libraries Government Documents Department

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

Description: 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 component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data. A rich collections of variance reduction features can greatly increase the efficiency of a calculation. MCNP is written in FORTRAN 77 and has been run on variety of computer systems from scientific workstations to supercomputers. The next production version of MCNP will include features such as continuous-energy electron transport and a multitasking option. Areas of ongoing research of interest to the well logging community include angle biasing, adaptive Monte Carlo, improved discrete ordinates capabilities, and discrete ordinates/Monte Carlo hybrid development. Los Alamos has requested approval by the Department of Energy to create a Radiation Transport Computational Facility under their User Facility Program to increase external interactions with industry, universities, and other government organizations. 21 refs.
Date: January 1, 1989
Creator: Forster, R.A.; Little, R.C. & Briesmeister, J.F.
Partner: UNT Libraries Government Documents Department

New probability table treatment in MCNP for unresolved resonances

Description: 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 these partials conditional upon the value of the total. This paper summarizes how the probability tables are utilized in this MCNP upgrade and compares this treatment with the approximate smooth treatment for some example problems.
Date: April 1, 1998
Creator: Carter, L.L.; Little, R.C.; Hendricks, J.S. & MacFarlane, R.E.
Partner: UNT Libraries Government Documents Department

Status of the MCNP{trademark}/LCS{trademark} merger project

Description: The MCNPX code is now in limited release in a beta-test version. We provide a brief status report on the physics modules now in the code and of the enhanced capabilities to use new evaluated neutron data. We also present new benchmark calculations in which LAHET and MCNPX are compared with experimental results from the Japan Atomic Energy Research Institute.
Date: April 19, 1998
Creator: Hughes, H.G.; Adams, K.J.; Chadwick, M.B.; Comly, J.C.; Frankle, S.C.; Hendricks, J.S. et al.
Partner: UNT Libraries Government Documents Department

Coupled proton/neutron transport calculations using the S sub N and Monte Carlo methods

Description: Coupled charged/neutral article transport calculations are most often carried out using the Monte Carol technique. For example, the ITS, EGS, and MCNP (Version 4) codes are used extensively for electron/photon transport calculations while HETC models the transport of protons, neutrons and heavy ions. In recent years there has been considerable progress in deterministic models of electron transport, and many of these models are applicable to protons. However, even with these new models (and the well established models for neutron transport) deterministic coupled neutron/proton transport calculations have not been feasible for most problems of interest, due to a lack of coupled multigroup neutron/proton cross section sets. Such cross sections sets are now being developed at Los Alamos. Using these cross sections we have carried out coupled proton/neutron transport calculations using both the S{sub N} and Monte Carlo methods. The S{sub N} calculations used a code called SMARTEPANTS (simulating many accumulative Rutherford trajectories, electron, proton and neutral transport slover) while the Monte Carlo calculations are done with the multigroup option of the MCNP code. Both SMARTEPANTS and MCNP require standard multigroup cross section libraries. HETC on the other hand, avoids the need for precalculated nuclear cross sections by modeling individual nucleon collisions as the transporting neutrons and protons interact with nuclei. 21 refs., 1 fig.
Date: January 1, 1991
Creator: Filippone, W.L. (Arizona Univ., Tucson, AZ (USA). Dept. of Nuclear and Energy Engineering); Little, R.C.; Morel, J.E.; MacFarlane, R.E. & Young, P.G. (Los Alamos National Lab., NM (USA))
Partner: UNT Libraries Government Documents Department

AFCI-2.0 Library of Neutron Cross Section Covariances

Description: Neutron cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The primary purpose of the library is to provide covariances for the Advanced Fuel Cycle Initiative (AFCI) data adjustment project, which is focusing on the needs of fast advanced burner reactors. The covariances refer to central values given in the 2006 release of the U.S. neutron evaluated library ENDF/B-VII. The preliminary version (AFCI-2.0beta) has been completed in October 2010 and made available to the users for comments. In the final 2.0 release, covariances for a few materials were updated, in particular new LANL evaluations for {sup 238,240}Pu and {sup 241}Am were adopted. BNL was responsible for covariances for structural materials and fission products, management of the library and coordination of the work, while LANL was in charge of covariances for light nuclei and for actinides.
Date: June 26, 2011
Creator: Herman, M.; Herman,M.; Oblozinsky,P.; Mattoon,C.; Pigni,M.; Hoblit,S. et al.
Partner: UNT Libraries Government Documents Department