Search NASA⌕ Search

SEARCH · Search NASA

Results for “nuclear data evaluation”

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 181 records · Page 10

Our Future Nuclear Data Needs

A well-established knowledge of nuclear phenomena including fission, reaction cross sections, and structure/decay properties is critical for applications ranging from the design of new reactors to nonproliferation to the production of radioisotopes for the diagnosis and treatment of illness. However, the lack of a well-quantified, predictive theoretical capability means that most nuclear observables must be measured directly and used to calibrate empirical models, which in turn provide the data needed for these applications. In many cases, either there is a lack of data needed to guide the models or the results of the different measurements are discrepant, leading to the development of evaluation methodologies to provide recommended values and uncertainties. In this review, we describe the nuclear data evaluation process and the international community that carries it out. We then discuss new measurements and improved theory and/or modeling needed to address future challenges in applied nuclear science.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Information Files of 238 U(n,f)/ 235 U(n,f) NIFFTE fissionTPC Cross-sections for the Neutron Data Standards Project

This report documents input files, output files and underlying assumptions for an uncertainty estimate for nuclear data evaluations purposes of the normalized 238 U/ 235 U neutron-induced fission cross-section ratios, 238 U(n,f)/ 235 U(n,f), measured by the NIFFTE fission Time Projection Chamber (fissionTPC). This uncertainty estimate and the resulting files were based on information provided by fissionTPC experimentalists, R.J.Casperson, N.S. Bowden, L. Snyder and K.T. Schmitt, and an EXFOR file compiled by N. Otuka. These files were generated for the Neutron Data Standards project coordinated by the IAEA in order to simplify the incorporation of 238 U(n,f)/ 235 U(n,f) fissionTPC data into the standards database.

238 U(n,f)/ 235 U(n,f)↗

Validation of Recent Changes in Actinide Evaluations U-235, U-233, Pu-239 [Slides]

This presentation discusses U-235 updates, including fluctuations in the fission cross section in the URR range that were refined to represent measured data following the Paradela et al. evaluation. There were small changes to the low-energy resonances to improve the fit to measured capture data from RPI (capture reduced by 5% from 0.06-7.8 eV, by 7.7% from 7.8-11 eV) and there was a very small change to the thermal nu-bar within uncertainties. Discrete level data are stored in MF=6 (format requirement, no impact on calculations) and issues leading to negative eigenvalues in cross-covariances in the RR have been sorted out. In conclusion, many improvements were made to evaluated nuclear data of actinides in the resonance region and above, such as minor changes to U-235 including assessment of changes that depend on the finalization of the O-16 evaluation. U-233 improvements show significant progress in performance and some further improvement might be needed at intermediate energies. Pu-239 ORNL/IAEA evaluation is promising; however, more work is needed.

07 ISOTOPE AND RADIATION SOURCES↗

Quasi-differential neutron induced neutron emission reaction measurements at WNR

The quasi-differential neutron induced neutron emission reaction measurement (QD-NINE) method provides an additional benchmark which constrains cross-section evaluations and provides a measure of verification for existing evaluations. Experiments are preformed using an array of EJ-309 liquid scintillators at a variety of angles measuring the neutron scattering response from a sample.. An MCNP simulation is then compared to this experimental data and is run using various sets of evaluated nuclear data. Since the isotopes of interest for this measurement, {sup 235}U and {sup 239}Pu, are difficult to obtain, the WNR facility at LANL was chosen to perform the measurements utilizing the chi-nu detector array. A measurement of carbon has been performed and the high energy cutoff has been extended from 20 to 60 MeV compared with previous measurements using this method. An early stage simulation was compared to the experimental data and the results show good agreement. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

WPEC SG44 Subgroup Report: Investigation of Covariance Data in General Purpose Nuclear Data Libraries [Slides]

Presentations were as follows: Progress on the Development of Thermal Scattering Covariance Formats, New Paradigm for Nuclear Data Evaluation, The Role of Fission Yield Correlations to Obtain Realistic Uncertainty Values in the Summation Method, ENDF/B-VIII.0 Augmented Covariance Data: The First Iteration, Status and Requirements of Nuclear Data Variance-Covariance Matrices for the Neutronic Assessment of Fast Reactor Cores, The Identification and Treatment of Unrecognized Uncertainties and the Impact on Evaluated Uncertainties– SG44, Improved Calibration of Nuclear Resonance Parameters, Report on the CSEWG Covariance and Measurement Session Initiative on Creating Templates of Expected Measurement Uncertainties, The Nuclear Data Belief Network, Investigation of Covariance Data in General Purpose Nuclear Data Libraries, and Computational Benchmark Proposal.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

WPEC SG50: Developing an Automatically Readable, Comprehensive and Curated Experimental Nuclear Reaction Database

The Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) Working Party on International Nuclear Data Evaluation Co-operation (WPEC) subgroup (SG) 50 was formed in 2020 to develop an automatically readable, comprehensive and curated experimental nuclear reaction database. This database is called MEDUSAL (Machine-readable Experimental Data User Application & Library), and will draw from EXFOR. The EXFOR database preserves experimental nuclear reaction data true to its original documentation and information from the authors of the data. MEDUSAL will deviate from EXFOR by storing additional information from users of the data for their fields of work (evaluation, model development, validation, etc.). This includes expert judgment on the data sets, identification of data points as outliers, renormalization of the data to the newest monitor reactions, and estimations of missing uncertainty sources. The format for MEDUSAL is being developed to enable easy automatic parsing of large amounts of data. Here, we will summarize the use cases, high-level requirements and first steps towards developing the database MEDUSAL and the API to access it.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

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↗

Light Elements $R$-matrix Analyses with the SAMMY code towards the Foundation of Charged-particle Nuclear Data Libraries [Slides]

This presentation covers newly developed SAMMY module for inverse channel transformation. Additionally covered is the R-matrix analysis of 7 Be compound nucleus and the R-­matrix analysis of 17 O compound nucleus. Further touched on is the evaluated Nuclear Data File generation and processing with the AMPX code. The presentation concludes with talks on future evaluation work and tests on light nuclei.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Evaluation of Sandia NCS Benchmark Suite Updates

The Sandia Nuclear Criticality Safety (NCS) program’s benchmark suite was recently updated. This suite is used to ensure that NCS calculations using computer-based neutron transportation codes have an established baseline comparison of calculated versus known experimental results. The Evaluated Nuclear Data File (ENDF) version used for the MCNP models in the suite was changed from ENDF/B-VII.1 to ENDF/B-VIII.0. Additionally, relevant thermal scattering law data libraries (TSLs) were updated. The sensitivity of the calculational bias of each benchmark model to these changes is discussed. Implementation of the ENDF/B-VIII.0 library and updated TSLs results in improvements to bias distribution in the intermediate enriched uranium, plutonium, and mixed uranium–plutonium (IEU, PU, and MIX) fissionable material benchmark categories, but a small bias increase in low- and high-enriched uranium categories (LEU and HEU, respectively). The results also highlight the sensitivity of the benchmarks, with average lethargy of neutrons causing fission energies (EALF) in the intermediate energy range to ENDF/B library changes. The most numerous bias changes were observed in the thermal energy region when transitioning from ENDF/B-VII.1 to ENDF/B-VIII.0. In conclusion, most of the unique bias changes observed in MCNP 6.3.0 between the two nuclear data libraries were in the LEU-COMP-THERM evaluation subset.

ICSBEP↗

Software Quality Assurance and Verification for the MPACT Library Generation Process

This report fulfills the requirements for the Consortium for the Advanced Simulation of Light-Water Reactors (CASL) milestone L2:RTM.P14.02, “SQA and Verification for MPACT Library Generation,” by documenting the current status of the software quality, verification, and acceptance testing of nuclear data libraries for MPACT. It provides a brief overview of the library generation process, from general-purpose evaluated nuclear data files (ENDF/B) to a problem-dependent cross section library for modeling of light-water reactors (LWRs). The software quality assurance (SQA) programs associated with each of the software used to generate the nuclear data libraries are discussed; specific tests within the SCALE/AMPX and VERA/XSTools repositories are described. The methods and associated tests to verify the quality of the library during the generation process are described in detail. The library generation process has been automated to a degree to (1) ensure that it can be run without user intervention and (2) to ensure that the library can be reproduced. Finally, the acceptance testing process that will be performed by representatives from the Radiation Transport Methods (RTM) Focus Area prior to the production library’s release is described in detail.

97 MATHEMATICS AND COMPUTING↗

Nuclear Data: From Measurement, Evaluation, Processing to Practical Applications [Slides]

Any analysis, design of a nuclear system will be as good as the knowledge of the nuclear data used whether differential and integral. ORNL is engaged in generating high level nuclear data for reactor, shielding, and safety applications. Nuclear data remains a major source of data uncertainties. Uncertainties on nuclear data rely upon measurements and evaluations. Tight connections with experimental facilities around the world have helped ORNL to address issues on ND and uncertainty. It is extremely important for ORNL to retain expertise on nuclear data in support of the research and expertise activities.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Status of the n+ 234 U evaluation in the resolved resonance region (*), (**)

In natural uranium, the 234 U isotope represents only 0.0055%, however, this minor isotope can affect highly enriched uranium metal benchmark calculations. In fact, enriched uranium contains more 234 U than natural uranium as the result of the uranium enrichment process. Therefore, the n+ 234 U nuclear data evaluation is one of the milestones of the APPENDIX B within the Nuclear Criticality Safety Program (NCSP).

AMPX↗

Characterization of a mixed high-energy spallation neutron–proton field using monoisotopic activation detectors

In this paper, we used monoisotopic Co, Au, and Bi threshold activation detectors to characterize the high-energy ( > 5 MeV) portion of a mixed spallation neutron–proton (n/p) field. The field was produced by irradiating the natural uranium target–blanket subcritical assembly Quinta with 660-MeV protons from Phasotron accelerator at the Joint Institute for Nuclear Research (JINR) in Dubna. Experimental yields of the (n,x) and (p,x) reactions were compared to the calculated reaction rates, which represent the convolution of particle flux and reaction cross section. The flux was simulated using the radiation transport code MCNPX 2.7.0, and the cross sections were extracted from the following three sources: the intranuclear cascade model INCL4.2, the phenomenological spherical optical model from TALYS-1.8, and the evaluated nuclear data file TENDL-2015. We show that, in most cases, the evaluated data provide a closer agreement with the experimental data in comparison with INCL4.2 model, which is an integral part of MCNPX. Finally, our results bring a contribution to those fields of physics research that utilize computer prediction of neutron generation in the energy range of tens to hundreds MeV, such as the medical radioisotope production, radiation shielding of spallation neutron sources, and nuclear transmutation in accelerator-driven systems.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

The experimental nuclear reaction data (EXFOR): Extended computer database and Web retrieval system

The EXchange FORmat (EXFOR) experimental nuclear reaction database and the associated Web interface provide access to the wealth of low- and intermediate-energy nuclear reaction physics data. This resource is based on numerical data sets and bibliographical information of ~22,000 experiments since the beginning of nuclear science. The principles of the computer database organization, its extended contents and Web applications development are described. New capabilities for the data sets uploads, renormalization, covariance matrix, and inverse reaction calculations are presented in this paper. The EXFOR database, updated monthly, provides an essential support for nuclear data evaluation, application development, and research activities. Finally, it is publicly available at the websites of the International Atomic Energy Agency Nuclear Data Section, http://www-nds.iaea.org/exfor, the U.S. National Nuclear Data Center, http://www.nndc.bnl.gov/exfor, and the mirror sites in China, India and Russian Federation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Integral nuclear data validation using experimental spent nuclear fuel compositions

Measurements of the isotopic contents of spent nuclear fuel provide experimental data that are a prerequisite for validating computer codes and nuclear data for many spent fuel applications. Under the auspices of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) and guidance of the Expert Group on Assay Data of Spent Nuclear Fuel of the NEA Working Party on Nuclear Criticality Safety, a new database of expanded spent fuel isotopic compositions has been compiled. The database, Spent Fuel Compositions (SFCOMPO) 2.0, includes measured data for more than 750 fuel samples acquired from 44 different reactors and representing eight different reactor technologies. Measurements for more than 90 isotopes are included. This new database provides data essential for establishing the reliability of code systems for inventory predictions, but it also has broader potential application to nuclear data evaluation. Furthermore, the database, together with adjoint based sensitivity and uncertainty tools for transmutation systems developed to quantify the importance of nuclear data on nuclide concentrations, are described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Analysis Capability and Data Needs Identified During the Evaluation of the SILENE CAAS Benchmarks

In October 2010, a series of benchmark experiments was conducted at the Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the US Department of Energy (DOE) and the French CEA to create three benchmarks for verification and validation of radiation transport codes, as well as evaluated nuclear data, used to analyze criticality accident alarm systems (CAASs). Many preliminary details of these experiments and their evaluation have been discussed in previous publications. The first benchmark evaluation was published in 2015, and the second and third evaluations will be published at the end of 2016.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

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↗

Neutron Leakage Spectra Sensitivities for ICSBEP Benchmarks

Neutron leakage spectra have been measured, simulated, and investigated by many groups. These spectra have many uses, including determining shielding requirements and calculating dose for radiation protection purposes, validating nuclear data, and determining material composition. It has even been proposed to use measurements of neutron leakage spectra to determine the soil composition of Mars. As part of the Los Alamos National Laboratory (LANL) Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) Laboratory Directed Research Development (LDRD) project, the authors are working on developing methods to calculate sensitivities to neutron leakage spectra. This may allow nuclear data evaluators to better use neutron leakage spectra data to constrain or adjust nuclear data. This may also be particularly useful as many neutron leakage spectra measurements do not require fissile material and can be performed with well characterized neutron sources. Neutron leakage spectra measurements are usually performed by placing a detector at the outside of a nuclear system. This may be a reactor, a neutron generator, or a neutron source. Certain detection systems can use pulse height data to infer neutron energies, other detector systems rely on other ways of determining neutron energy (for example, a Bonner sphere with multiple moderator thicknesses can be used to measure neutron spectrum). These measurements typically rely on some unfolding of the measured results. For pulsed systems like the "Livermore Pulsed Spheres," time-of-flight information can also help to constrain the neutron energy spectra data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗