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

Criticality Benchmark Results Using Various MCNP Data Libraries

A suite of 86 criticality benchmarks has been recently implemented in MCNP{trademark} as part of the nuclear data validation effort. These benchmarks have been run using two sets of MCNP continuous-energy neutron data: ENDF/B-VI based data through Release 2 (ENDF60) and the ENDF/B-V based data. New evaluations were completed for ENDF/B-VI for a number of the important nuclides such as the isotopes of H, Be, C, N, O, Fe, Ni, {sup 235,238}U, {sup 237}Np, and {sup 239,240}Pu. When examining the results of these calculations for the five manor categories of {sup 233}U, intermediate-enriched {sup 235}U (IEU), highly enriched {sup 235}U (HEU), {sup 239}Pu, and mixed metal assembles, we find the following: (1) The new evaluations for {sup 9}Be, {sup 12}C, and {sup 14}N show no net effect on k{sub eff}; (2) There is a consistent decrease in k{sub eff} for all of the solution assemblies for ENDF/B-VI due to {sup 1}H and {sup 16}O, moving k{sub eff} further from the benchmark value for uranium solutions and closer to the benchmark value for plutonium solutions; (3) k{sub eff} decreased for the ENDF/B-VI Fe isotopic data, moving the calculated k{sub eff} further from the benchmark value; (4) k{sub eff} decreased for the ENDF/B-VI Ni isotopic data, moving the calculated k{sub eff} closer to the benchmark value; (5) The W data remained unchanged and tended to calculate slightly higher than the benchmark values; (6) For metal uranium systems, the ENDF/B-VI data for {sup 235}U tends to decrease k{sub eff} while the {sup 238}U data tends to increase k{sub eff}. The net result depends on the energy spectrum and material specifications for the particular assembly; (7) For more intermediate-energy systems, the changes in the {sup 235,238}U evaluations tend to increase k{sub eff}. For the mixed graphite and normal uranium-reflected assembly, a large increase in k{sub eff} due to changes in the {sup 238}U evaluation moved the calculated k{sub eff} much closer to the benchmark value. (8) There is little change in k{sub eff} for the uranium solutions due to the new {sup 235,238}U evaluations; and (9) There is little change in k{sub eff} for the {sup 239}Pu metal assemblies, but a decrease in k{sub eff} for the solution assemblies, moving them closer to the benchmark value

A Suite of Criticality Benchmarks for Validating Nuclear Data

The continuous-energy neutron data library ENDF60 for use with MCNP{trademark} was released in the fall of 1994, and was based on ENDF/B-Vl evaluations through Release 2. As part of the data validation process for this library, a number of criticality benchmark calculations were performed. The original suite of nine criticality benchmarks used to test ENDF60 has now been expanded to 86 benchmarks. This report documents the specifications for the suite of 86 criticality benchmarks that have been developed for validating nuclear data

Spectral measurements in critical assemblies: MCNP specifications and calculated results

Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented

Improving the LLNL pulsed sphere experiments database and MCNP models.

During the last 33 years, numerous high energy pulsed-sphere experiments have been performed in which small, medium, and large spheres of 32 different materials were pulsed with a burst of high-energy neutrons. Measured time-dependent detector responses at distant locations provide a benchmark by which various neutron transport codes and cross-section libraries may be judged

Neutron-capture gamma-ray data for obtaining elemental abundances from planetary spectra.

Determination of elemental abundances is a top scientific priority of most planetary missions. Gamma-ray spectroscopy is an excellent method to determine elemental abundances using gamma rays made by nuclear reactions induced by cosmic-ray particles and by the decay of radioactive nuclides [Re73,Re78]. Many important planetary gamma rays are made by neutron-capture reactions. However, much of the data for the energies and intensities of neutron-capture gamma rays in the existing literature [e.g. Lo81] are poor [RF99,RF00]. With gamma-ray spectrometers having recently returned data from Lunar Prospector and NEAR and soon to be launch to Mars, there is a need for good data for neutron-capture gamma rays

Improvements to the photon-production data for radiative capture in ENDF

High-quality photon-production data from thermal-neutron-capture reactions are important for many applications, including oil-well logging, planetary gamma-ray spectroscopy, and environmental techniques. Radiation transport codes usually access photon-production data from the national U.S. nuclear database, ENDF/B. To improve the photon-production data for thermal-neutron capture, we have compiled and evaluated the energies and yields of gamma rays for most naturally occurring isotopes with Z\u3c31 and for a few heavier isotopes. The impact of improved photon-production cross sections for thermal-neutron capture in Cl is illustrated by comparing MCNP simulations and experimental measurements for a pulsed neutron oil-well logging instrument. The new compilation of photon-production data is being incorporated into the latest ENDF/B evaluations for each isotope and will be provided for future releases of the ENDF/B database. Copyright (C) 2000 Elsevier Science Ltd