Albert C. Kahler

ENDF70: A Continuous-Energy MCNP Neutron Data Library Based on ENDF/B-VII.0

Abstract Following the release of ENDF/B-VII.0 evaluations, an ACE-formatted continuous-energy neutron data library called ENDF70 for MCNP has been produced at Los Alamos National Laboratory. This new library contains data for 387 isotopes and three elements at five temperatures: 293.6, 600, 900, 1200, and 2500 K. It can be obtained as part of the MCNP5 1.50 release. The new library was created using ENDF/B-VII.0 neutron evaluations and primarily version 248 of NJOY99. A processing script was created that set up the input files for NJOY and employed checking codes to test the content of the processed data. A sample MCNP run was performed for each isotope and temperature, and cross sections for each isotope were plotted to make sure there were no major problems. The processed ACE libraries did not always pass all quality assurance tests. For example, energy-balance problems were identified for several evaluations having negative heating numbers or inconsistencies between total and partial heating. Similarly, some problems were found with unresolved resonance probability tables, resulting in probability tables being excluded from the final library for several materials. Certain evaluations were modified and reprocessed as a result of the quality assurance tests, and some data points in the final ACE files were changed because they were too small or had other problems. The new ENDF70 library provides MCNP users with the latest ENDF/B data available. This collection of data includes a larger range of isotopes and temperatures than previously released, which will be beneficial in numerous applications. The upgrades included as part of ENDF/B-VII.0 and, hence, ENDF70 should improve calculations.

Whisper: Sensitivity/Uncertainty-Based Computational Methods and Software for Determining Baseline Upper Subcritical Limits

Abstract Nuclear criticality safety analysis using computational methods such as a Monte Carlo method must establish, for a defined area of applicability, an upper subcritical limit (USL), which is a calculated multiplication factor k that can be treated as actually subcritical and is derived from a calculational margin (combination of bias and bias uncertainty) and a margin of subcriticality. Whisper, a nonparametric, extreme-value method based on sensitivity/uncertainty techniques and the associated software are presented. Whisper uses benchmark critical experiments, nuclear data sensitivities from the continuous-energy Monte Carlo transport software MCNP, and nuclear covariance data to set a baseline USL. Comparisons with a traditional parametric approach for validation, which requires benchmark data to be normally distributed, show that Whisper typically obtains similar or more conservative calculational margins; comparisons with a rank-order nonparametric approach show that Whisper obtains less stringent calculational margins.

Applications of Integral Benchmark Data

Abstract The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) provide evaluated integral benchmark data that may be used for validation of reactor physics/nuclear criticality safety analytical methods and data, nuclear data testing, advanced modeling and simulation, and safety analysis licensing activities. The handbooks produced by these programs are used in over 30 countries. Five example applications are presented in this paper: (a) use of IRPhEP data in uncertainty analyses and cross-section adjustment, (b) uncertainty evaluation methods for reactor core design at Japan Atomic Energy Agency using reactor physics experimental data, (c) application of benchmarking data to a broad range of criticality safety problems, (d) cross-section data testing with ICSBEP benchmarks, and (e) use of the International Handbook of Evaluated Reactor Physics Benchmark Experiments to support the power industry.

Verification and Validation of the ENDF/B-VII.1–Based Continuous-Energy Data Tables for MCNP6

Abstract In December 2011, the National Nuclear Data Center released ENDF/B-VII.1, the “latest recommended evaluated nuclear data file for use in nuclear science and technology applications.” The data were released in the standard Evaluated Nuclear Data Format (ENDF). This release represents the advances made in nuclear data during the 5 years since the release of ENDF/B-VII.0. The Nuclear Data Team at Los Alamos National Laboratory has processed the ENDF/B-VII.1 library and made available a library of ACE data tables at several temperatures for each of the ENDF/B files. The ACE data library is called ENDF71x and is available through the Radiation Safety Information Computational Center with MCNP6. The files can also be used with MCNP5 or other Monte Carlo codes.

Validation of MCNP6.1 for Criticality Safety of Pu-Metal, -Solution, and -Oxide Systems

Guidance is offered to the Los Alamos National Laboratory Nuclear Criticality Safety division towards developing an Upper Subcritical Limit (USL) for MCNP6.1 calculations with ENDF/B-VII.1 nuclear data for three classes of problems: Pu-metal, -solution, and -oxide systems. A benchmark suite containing 1,086 benchmarks is prepared, and a sensitivity/uncertainty (S/U) method with a generalized linear least squares (GLLS) data adjustment is used to reject outliers, bringing the total to 959 usable benchmarks. For each class of problem, S/U methods are used to select relevant experimental benchmarks, and the calculational margin is computed using extreme value theory. A portion of the margin of sub criticality is defined considering both a detection limit for errors in codes and data and uncertainty/variability in the nuclear data library. The latter employs S/U methods with a GLLS data adjustment to find representative nuclear data covariances constrained by integral experiments, which are then used to compute uncertainties in keff from nuclear data. The USLs for the classes of problems are as follows: Pu metal, 0.980; Pu solutions, 0.973; dry Pu oxides, 0.978; dilute Pu oxide-water mixes, 0.970; and intermediate-spectrum Pu oxide-water mixes, 0.953.