Search NASA⌕ Search

SEARCH · Search NASA

Results for “ENDF”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

NJOY2016

NJOY2016 is used to convert evaluations in the Evaluated Nuclear Data Files (ENDF) format into forms useful for practical applications such as fission and fusion reactor analysis, criticality safety, radiation shielding, nuclear waste management, nuclear medicine procedures, and more. It begins with the generation of pointwise libraries, including reaction and resonance reconstruction, Doppler broadening, radiation heating and damage, thermal scattering data, unresolved resonance data, and gas production. It reviews the production of libraries for the continuous-energy Monte Carlo code MCNP, multigroup neutron, photon, and particle cross sections and matrices, and photon interaction data. Uncertainty information for ENDF data is processed, including the capability for calculating covariances of resonance data, angular distributions, energy distributions, and radioactive nuclide production. NJOY can prepare and process thermal scattering data evaluations into application libraries. It can also plot cross sections, angular distributions and secondary emission spectra.

Kahler III, A. C. (Skip)↗
DOE OSTI2016

Improving Capture-gamma Libraries for Nonproliferation Applications

This report describes the measurement, evaluation and incorporation of new -ray spectroscopic data into the Evaluated Nuclear Data File (ENDF) for nonproliferation applications. Analysis and processing techniques are described along with key deliverables that have been met over the course of this project. A total of nine new ENDF libraries have been submitted to the National Nuclear Data Center at the Brookhaven National Laboratory and are now available in the ENDF/B-VIII.beta2 release. Furthermore, this project has led to more than ten peer-reviewed publications and provided theses for ve graduate students. This project is a component of the NA-22 venture collaboration on \Correlated Nuclear Data in Fission Events" (LA14-V-CorrData-PD2Jb).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

The NJOY Nuclear Data Processing System, Version 2016

The NJOY Nuclear Data Processing System, version 2016, is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from evaluated nuclear data in the ENDF-4 through ENDF-6 legacy card-image formats. NJOY works with evaluated files for incident neutrons, photons, and charged particles, producing libraries for a wide variety of particle transport and reactor analysis codes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

A tale of two tools: mcres.py, a stochastic resonance generator, and grokres.py, a resonance quality assurance tool

We detail two software tools, now integrated into the fudge code system. The first tool, mcres.py, can be used to generate stochastic ensembles of resonances which are both consistent with the expectations of the Gaussian Orthogonal Ensemble of Random Matrix Theory and with the level densities and widths encoded in ENDF formatted files. The second tool, grokres.py, can be used to assess global and local features of sequences of resonances found in ENDF files and make comparisons to known results from Random Matrix Theory. We apply these tools to 54 Fe and other nuclei.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

WPEC Subgroup-38 Final Report part II: Specifications for a new database structure

This is the revision of the specifications for the Generalised Nuclear Database Structure (GNDS), corresponding to GNDS version 1.9. GNDS is designed to serve as a replacement for the Evaluated Nuclear Data Format (ENDF). Although GNDS supersedes the ENDF6 format, GNDS should be viewed as a natural evolution of the ENDF-6 format, using modern computational tools to make nuclear data more accessible.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Comparison of the Baseline USL Calculation Methods for Loosely-Coupled and Novel Neutronic Systems [Slides]

Current work includes reconstructing 187-group ENDF B/VII.1 covariance matrix for comparison study to 44-group ENDF B/VII.1 matrix to determine how covariance matrix structures affect the USL calculations and investigating the exact domination mechanism of region-wise sensitivities in a loosely-coupled system. Future work will involve seeking to understand how bias distributions change with reactivity and how to calculate propagate distribution error into USL calculations, as well as how cross section perturbation studies can be extended to other types of calculations, such as shielding calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Release of ENDF/B-VIII.1β1-Based ACE Data Files

On March 1 st , 2023, the National Nuclear Data Center (NNDC) released the ENDF/B-VIII.1β1 nuclear data library, which contained neutron, spontaneous and neutron-induced fission yield, and alpha sublibraries. The library was released in the standard Evaluated Nuclear Data File (ENDF) format. The files can be accessed on Brookhaven National Laboratory’s GitLab. The thermal scattering law (TSL) sublibrary will be released at a later date. The files provided in the neutron sublibrary were processed into A Compact ENDF (ACE)-formatted files, verified, and validated by the XCP-5 Nuclear Data Team. The resulting collection of ACE files will be referred to as the “e81b1” library. This report details the processing of these files and the quality assurance approach taken. This is not intended to be a full validation effort; rather, this library is intended to simply reproduce the released files for further validation testing by the community. The e81b1 ACE library will be released to the NNDC for access through GitLab and hosted locally on LANL computing resources. This library is only intended for analysis to support the future ENDF/B-VIII.1 release and should not be used for production results.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Validation of ENDF/B-VIII.1 Nuclear Data Files [Slides]

ENDF-6 formatted files were processed into A Compact ENDF (ACE) files using NJOY2016. Several validation tests were performed: (1) LANL Legacy Benchmark Suite (2) “Modern” Benchmark Suite (3) HEU Benchmark Suite (4) LEU Benchmark Suite (5) Mixed (U+Pu) Benchmark Suite (6) Pu Benchmark Suite (7) 233 U Benchmark Suite.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Release of ENDF81SaB: ENDF/B-VIII.1-Based ACE Data Files for Thermal Scattering

On August 30, 2024, the National Nuclear Data Center (NNDC) released the ENDF/B-VIII.1 nuclear data library. The library was released in the standard Evaluated Nuclear Data File (ENDF) format. These files can be accessed on the NNDC's website (www.nndc.bnl.gov). The files provided in the thermal neutron scattering sublibrary were processed into A Compact ENDF (ACE)-formatted files, verified, and validated by the XCP-5 Nuclear Data Team, resulting in the ENDF81SaB application library. This report details the processing of these files and the quality assurance approach taken. This is not intended to be a full validation effort; rather, this library is intended to simply reproduce the released files for further validation testing by the community. The validation basis and details of the evaluations are documented in the forthcoming ``Big Paper''.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ORNL FY2024 Nuclear Data Evaluation Contributions: Cu, La, N, and Ta [Slides]

63,65 Cu angular distributions resolved for ENDF/B-VIII.1, consistent with both differential data and integral benchmark performance. 139 La RRR and URR evaluations will be merged with the LANL high energy evaluation, and to be submitted to the future ENDF-B release (post-VIII.1). 14 N RRR evaluation is planned to produce n+ 14 N, p+ 14 C, and a+ 11 B contributions to be submitted to the future ENDF-B release (post-VIII.1). 181 Ta covariances repaired and reported as intended in an errata to ENDF/B-VIII.1.

139-La↗
DOE OSTI2026

Resonance Self-Shielding: Why it is so Important

This paper is one of a series that I am writing to document my 58 years of experience with ENDF and Neutron Transport calculations, beginning when I worked at the National Nuclear Data Center (NNDC), Brookhaven National Laboratory (BNL), from 1967 to 1972. During those years I was the head of the computer unit of NNDC, assigned to develop computer codes to pre-process, view and test ENDF/B data. Since then, I have continued to support the ENDF effort without any official position or monetary compensation, because I realized how important accurate nuclear data is for use in use in our Engineering applications. It is so important to realize that regardless of how accurate or even perfect our application codes may be to transport particles, without accurate nuclear data we are in a “Garbage In = Garbage Out” situation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Evaluation of δ-Phase ZrH1.4 to ZrH1.7 Thermal Neutron Scattering Laws Using Ab Initio Molecular Dynamics Simulations

Zirconium hydride is commonly used for next-generation reactor designs due to its excellent hydrogen retention capacity at temperatures below 1000 K. These types of reactors operate at thermal neutron energies and require accurate representation of thermal scattering laws (TSLs) to optimize moderator performance and evaluate the safety indicators for reactor design. In this work, we present an atomic-scale representation of sub-stoichiometric ZrH2−x(0.3≤x≤0.6), which relies on ab initio molecular dynamics (AIMD) in tandem with velocity auto-correlation (VAC) analysis to generate phonon density of states (DOS) for TSL development. The novel NJOY+NCrystal tool, developed by the European Spallation Source community, was utilized to generate the TSL formulations in the A Compact ENDF (ACE) format for its utility in neutron transport software. First, stoichiometric zirconium hydride cross sections were benchmarked with experiments. Then sub-stoichiometric zirconium hydride TSLs were developed. Significant deviations were observed between the new δ-phase ZrH2−x TSLs and the TSLs in the current ENDF release. It was also observed that varying the hydrogen vacancy defect concentration and sites did not cause as significant a change in the TSLs (e.g., ZrH1.4 vs. ZrH1.7) as was caused by the lattice transformation from ϵ- to δ-phase.

42 ENGINEERING↗
DOE OSTI2016

NJOY2016

NJOY2016 is used to convert evaluations in the Evaluated Nuclear Data Files (ENDF) format into forms useful for practical applications such as fission and fusion reactor analysis, criticality safety, radiation shielding, nuclear waste management, nuclear medicine procedures, and more. It begins with the generation of pointwise libraries, including reaction and resonance reconstruction, Doppler broadening, radiation heating and damage, thermal scattering data, unresolved resonance data, and gas production. It reviews the production of libraries for the continuous-energy Monte Carlo code MCNP, multigroup neutron, photon, and particle cross sections and matrices, and photon interaction data. Uncertainty information for ENDF data is processed, including the capability for calculating covariances of resonance data, angular distributions, energy distributions, and radioactive nuclide production. NJOY can prepare and process thermal scattering data evaluations into application libraries. It can also plot cross sections, angular distributions and secondary emission spectra.

Kahler III, A. C. (Skip)↗
DOE OSTI2022

Modernization efforts for the R -Matrix code SAMMY [Abstract]

The R-Matrix code SAMMY is a widely used nuclear data evaluation code focused on the resolved range, which includes corrections for experimental effects. The code is still mostly written in Fortran 77, and uses a memory management system suitable for the time of its initial writing (1984). A modernization effort is under way to bring the code in-line with modern software development practices. A continuous-integration testing framework was added, automating the large existing set of test cases. It is run on every commit. The memory management was updated to current standard practices suitable for modern software analysis tools. The code can be obtained from https://code.ornl.gov/RNSD/SAMMY. The resonance parameters and covariance information are now stored in C++ objects shared by SAMMY and AMPX, the processing code that generates nuclear data libraries for SCALE. This allows for easier maintenance and access to the resonance parameters inside and outside of SAMMY. This feature is already used by accessing and changing parameters in memory in the Bayesian Monte Carlo Evaluation Framework for Cross Sections Nuclear Data and Integral Benchmark Experiments project, Further plans include the switch to the ENDF reading and writing routines in AMPX, as these routines are more robust, easier to maintain, and support more features. Of note here is support for the new GNDS format. Previously it wasn’t easy to share the full covariance matrix for evaluations containing more than one isotope due to limitations on the ENDF format; this is now supported in GNDS. The data are currently available in a binary SAMMY format and can be exported to GNDS to make them more widely available and sharable. The next step will be to use the same resonance processing code at 0K in AMPX and SAMMY as one of the available Reich-Moore R-Matrix formalism. The first step toward this goal is to isolate the reconstruction into a module that takes resonance parameters as its input and does not depend on SAMMY global parameters. This goal has been achieved and it should now be possible to more easily change the resonance formalism and add enhancements as the Phenomenological R-Matrix parameterization of direct, doorway, and compound nuclear reactions discussed elsewhere on this conference. This concerted modernization and enhancement effort provides multiple advantages to the nuclear data community. It will allow parameter optimization using enhanced formalisms, including experimental effects, that better match complex experimental data. Then those evaluated parameters can immediately be passed off to AMPX to be reconstructed with the exact same cross section model and be put into a data library for subsequent testing using SCALE and the Valid Benchmark suite or other suitable benchmark suites.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

ENDF/B-VIII.0 testing with AMPX and SCALE

As the new ENDF/B-VIII.0 data has continued development from its initial beta release into the final release version, it has subsequently undergone processing and testing at Oak Ridge National Laboratory with AMPX and SCALE. Each of the beta releases has been processed with AMPX to create both multigroup (MG) and continuous-energy (CE) libraries. The MG libraries were all processed using the SCALE 252-group structure. These libraries were then used with the VALID suite for benchmarking the new data library's performance against the currently distributed ENDF/B-VII.1 (VII.1)-based SCALE libraries. Here, we compare the MG and CE VALID results for the two current ENDF releases, versions VII.1 and the most recent beta release of ENDF/B-VIII.0 as of this writing (VIII.0.β4). Generally, both the MG and CE VALID aggregate results for VIII.0.β4 compare reasonably well with their VII.1-derived counterparts. As can be expected of any new ENDF release, some of the individual cases have substantially different computed neutron multiplicity (k{sub eff}) values between the two releases, although typically the differences are less than 100 percent-milli (pcm). However, changes to the plutonium cross sections have induced much larger changes in the ratio of the computed k{sub eff} to the experimental value (C/E) for thermal spectrum plutonium solution critical systems. This paper details the comparative performance of the VII.1 and VIII.0.β4 data for the broad range of test cases covered by the VALID suite. A companion paper at this conference provides details of the baseline validation of SCALE 6.2, including the first release of the VII.1 data. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Improvements in Electron-Photon-Relaxation Data for MCNP6

The Los Alamos Monte Carlo transport code MCNP6 has recently benefitted from significant work to improve the photon-electron transport and atomic-relaxation capabilities of the code. The improvements include extension of existing photon data to lower energies than previously supported; addition of new data such as subshell-specific photoelectric cross sections; completion of the incoherent and coherent form factor data to cover all scattering angles at all supported energies; new data for electron interactions including Bremsstrahlung, elastic scattering, subshell-specific electroionization, and atomic excitation; a much more detailed representation for atomic relaxation; and an entirely new (to MCNP) approach to electron transport: the single-event method. These improvements were made available with the initial production release of MCNP6 along with an extended electron, photon, and relaxation data file eprdata12 to support the new capabilities. More recently new data have become available and have now been incorporated into MCNP6, correcting a few issues with the previous work, and adding new sets of cross sections that were not included in eprdata12. The purpose of this paper is to introduce these newer changes and enhancements, which will be included with the next public release of the MCNP6 code system. The database that supports these developments relies on an evolving set of libraries originally released by D. E. Cullen and collaborators as separate data for photo-atomic interactions (EPDL), atomic relaxation (EADL), and electro-atomic interactions (EEDL). An updated version of this release was officially adopted in the ENDF/B VI.8 release of the Evaluated Nuclear Data File and it was this source that was used to produce the ACE (A Compact ENDF) format library eprdata12 for use in the initial production release of MCNP6.1 (and the subsequent interim release MCNP6.1.1). The latest improved version of this database was released by Cullen in 2014. This is the EPICS2014 collection, which includes both an ENDF-formatted file and separate EPDL, EADL, and EEDL files. In the area of photon-electron transport and atomic-relaxation, the newest enhancements to the current version of MCNP6 (scheduled for release as MCNP6.2) are primarily based on improvements and additions drawn from the EPICS2014 data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2018

EXSAN: An Interactive Version of the Nuclear Cross Section Data Handbook - 25687

Neutron cross section measurements and data files have advanced significantly since 1989 when Los Alamos published 'A Nuclear Cross Section Data Handbook', a detailed compendium of point-wise cross section plots and pie charts of relative contributions of individual reactions to the total cross section for 129 isotopes. Other compendiums with similar data and analysis tools have also been published. The intent of this work is to develop a modern, open-source application for interactive exploration and analysis of Evaluated Nuclear Data File (ENDF) libraries that also allows users to create their own 'Nuclear Cross Section Data Handbooks'. EXSAN is an easy-to-use graphical user interface application for exploring ENDF photon and neutron cross sections developed in Python. In addition to interactive plotting, EXSAN includes methods for computing rank-order lists of cross section for any energy and calculating composite cross sections for isotopic and elemental compositions and additional batch automation features. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗