21 Matching Results

This system will be undergoing maintenance January 24th 9:00-11:00AM CST.

Search Results

Advanced search parameters have been applied.

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

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


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

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

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

AFCI-2.0 Neutron Cross Section Covariance Library

Description: 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 materials and fission products, and 20 actinides. ...
Date: March 1, 2011
Creator: Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S. et al.
Partner: UNT Libraries Government Documents Department

Evidence of water ice near the lunar poles

Description: Lunar Prospector epithermal neutron data were studied to evaluate the probable chemical state of enhanced hydrogen, [H], reported previously to be near both lunar poles [1,2]. Improved versions of thermal and epithermal neutron data were developed for this purpose. Most important is the improved spatial resolution obtained by using shortened integration times. A new data set was created, Epi* = [Epithermal - 0.057 x Thermal], to reduce effects of composition variations other than those due to hydrogen. The Epi* counting rates are generally low near both lunar poles and high over terrane near recent impact events such as Tycho and Jackson. However, other lunar features are also associated with high Epi* rates, which represent a wide range of terrane types that seem to have little in common. If we postulate that one property all bright Epi* features do have in common is low [H], then measured Epi* counting rates appear to be quantitatively self consistent. If we assume that [H]=O above the top 98th percentile of Epi* counting rates at 2{sup o} x 2{sup o} spatial resolution, then [H]{sub ave} = 55 ppm for latitudes equatorward of [75{sup o}]. This value is close to the average found in returned lunar soil samples, [H]{sub ave} {approx} 50 ppm [3]. Using the foregoing physical interpretation of Epi* counting rates, we find that the Epi* counts within most of the large craters poleward of {+-}70{sup o} are higher, and therefore [H] is lower, than that in neighboring inter-crater plains, as shown in Figure 1. Fourteen of these craters that have areas larger than the LP epithermal spatial resolution (55 km diameter at 30 km altitude), were singled out for study. [H] is generally found to increase with decreasing distance from the poles (hence decreasing temperature). However, quantitative estimates of the diffusivity of hydrogen ...
Date: January 1, 2001
Creator: Feldman, W. C. (William C.); Maurice, S. (Sylvestre); Lawrence, David J. (David Jeffery),; Little, R. C. (Robert C.); Lawrence, S. L. (Stefanie L.); Gasnault, O. M. (Olivier M.) et al.
Partner: UNT Libraries Government Documents Department

ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

Description: We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes. The new evaluations are based on both experimental data and nuclear reaction theory predictions. The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, {sup 6}Li, {sup 10}B, Au and for {sup 235,238}U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron- and proton-induced reactions up to an energy of 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; and (10) New methods developed to provide uncertainties and covariances, together with covariance evaluations for some sample cases. The paper provides an overview of this library, consisting of 14 sublibraries in the same, ENDF-6 format, as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched U thermal assemblies is removed; (b) The {sup 238}U, {sup 208}Pb, and {sup 9}Be reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good ...
Date: October 2, 2006
Creator: Chadwick, M B; Oblozinsky, P; Herman, M; Greene, N M; McKnight, R D; Smith, D L et al.
Partner: UNT Libraries Government Documents Department