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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.

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At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Measurement of the 28 Si ⁢(𝑛,𝑛′⁢𝛾) cross section with 𝑛, 𝛾, and correlated 𝑛−𝛾 angular distributions

Silicon has become an unavoidable element in the circuitry central to everyday life. In turn, the interactions of silicon isotopes with neutrons for nuclear physics applications, among other motivations, have become increasingly important to understand. The dominant isotope of silicon, 28 Si, is thus of primary interest for enhanced understanding for neutron transport calculations and related investigations. Unfortunately, the existing measurement database for neutron scattering reactions on 28 Si is minimal, and nuclear data evaluations on this topic have not been updated for decades. This article details new measurements of the 28 Si ⁢(𝑛,𝑛′⁢𝛾) reaction utilizing multiple analysis methods available within the correlated gamma neutron array for scattering (CoGNAC). Specifically, high-precision near-threshold results and high-incident-energy results were obtained using the 𝛾-only and correlated 𝑛−𝛾 techniques. First-ever measurements of the correlated 𝑛−𝛾 angular distribution for particles emitted following population of the first excited state in 28 Si were obtained as well, which provide unique insight into theoretical descriptions of the inelastic neutron scattering reaction mechanism itself and detailed guidance for nuclear reaction models. The results agree well with literature data where they exist, and substantially expand on the current database for neutron reactions on 28 Si .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

SPARTA: A flux adjustment methodology to interpret complex experiments

For the accurate determination of reactivity from a detector count rate, correction of spatial effects is of prime importance. This spatial correction is often provided using simulation methodologies, but this may introduce a bias if the result of the experiment is also used as input data for the simulation. Here, this work presents a flux adjustment methodology able to infer experimental reactivity and correction of spatial effects without the need for a simulation. It can process the signal from a complex experiment such as a heat balance measurement in the TREAT reactor, where control rods are continuously adjusted to maintain a constant power. In the present work, this methodology successfully computed the reactivity and the local spatial variation of the flux of a generated signal. It also proved to be robust against noise and errors on kinetic parameters and provides a credible interpretation of a heat balance experiment in TREAT. Efficiency of flux adjustment methods for complex experiment enable a better experiment interpretation less reliant on nuclear data evaluation.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Alpha Reaction Sublibrary

The alphas sublibrary aims to describe nuclear reactions between incident alpha particles and different nuclei. For ENDF/B-VIII.1, the LLNL ECPL evaluation for 6Li, and the JENDL-based NNL evaluations of 9 Be and 16,17 O were adopted. Also, there were minor fixes done to 4 He.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Decay Reaction Sublibrary

The decay sublibrary aims to describe all known spontaneous decays and branching ratios for all nuclei across the nuclide chart. ENDF/B-VIII.1 remains unchanged from ENDF/B-VIII.0 for the decay sublibrary.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Deuteron Reaction Sublibrary

The deuterons sublibrary aims to describe nuclear reactions between incident deuteron particles and different nuclei. For ENDF/B-VIII.1, we adopted additions and corrections to the ENDF/B-VIII.0 d+t file, as well as new LANL evaluation updates for 3 He and 6 Li.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Helium-3 Reaction Sublibrary

The Helium-3 (Helions) sublibrary aims to describe nuclear reactions between incident 3 He particles and different nuclei. For ENDF/B-VIII.1, we adopted a new 4He evaluation from the INDEN collaboration and the LLNL ECPL evaluation for 7 Li.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Neutron Reaction Sublibrary

The neutrons sublibrary aims to describe nuclear reactions between incident neutron particles and different nuclei. For ENDF/B-VIII.1, many neutron files were re-evaluated or received major changes. A new 239Pu evaluation was jointly-produced by IAEA, LANL, LLNL and ORNL bringing important updates to fission neutron multiplicity, Prompt Fission Neutron Spectrum, resonance and fast regions. Around one third of the new neutron evaluations were performed as part of the INDEN collaboration, including 16,18 O, 19 F, 28,29,30 Si, 63,65 Cu, 50,51,52,53,54 Cr, 55 Mn, 54,56,57 Fe, 139 La, 233,235,238 U, 240,241 Pu. Important non-INDEN evaluations include 234,236 U, 206,207,208 Pb, 181 Ta, 88 Sr, 140,142 Ce, Pt and Dy isotopes, and many others. Also, dosimetry reactions from IRDFF-II were adopted for many materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Neutron Induced Fission Product Yields Sublibrary

The neutron-induced fission yields (NFY) sublibray aims to describe fission yields in neutron-induced reactions. For ENDF/B-VIII.1, there were no new fission yields complete evaluations. However, an important fix the 241 Pu NFY file was implemented, solving a long-standing anomalous discontinuity issue accidentally introduced in ENDF/B-VI.2, decades ago. Additionally, many erroneously large uncertainties, present in NFY from ENDF/B-VIII.0, were corrected.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Proton Reaction Sublibrary

The protons sublibrary aims to describe nuclear reactions between incident proton particles and different nuclei. For ENDF/B-VIII.1, compared to ENDF/B-VIII.0, there were only evaluation updates for 4 He.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Spontaneous Fission Product Yields Reaction Sublibrary

The spontaneous fission yields (SFY) sublibray aims to describe fission yields that happen spontaneous, without any incident particle. For ENDF/B-VIII.1, there were no new fission yields complete evaluations. However, many erroneously large uncertainties, present in both SFY from ENDF/B-VIII.0, were corrected.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Neutron Standards Sublibrary

The neutron standards sublibrary describes specific reaction cross sections, in a limited range, that are so well-known that they are used as ratios or references in other measurements. ENDF/B-VIII.1 remains unchanged from ENDF/B-VIII.0 for the neutron standards sublibrary

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Thermal Neutron Scattering Sublibrary

The thermal neutron scattering law (TSL) sublibrary aims to describe the interaction of incident neutrons at thermal or sub-thermal energies with different compound materials such fuels, moderators and special-purpose materials. In ENDF/B-VIII.1 there was a large number of new and updated TSL evaluations, including traditional moderators (light water, Beryllium metal, Beryllium Oxide, Calcium Hydride, plastics (Polystyrene and Lucite), graphite (reactor-grade and crystalline), anhydrous Hydrogen Fluoride, and heavy paraffinic oil); exotic moderators (Beryllium Carbide, Zirconium Hydride, Yttrium Hydride, Lithium-7 Hydride and Deuteride), FLiBe molten salt, structural materials and cladding (Silicon Carbide, Silicon Dioxide, Zirconium Carbide), fuels (Plutonium Dioxide, Uranium Carbide, Uranium metal, Uranium Nitride, Uranium Dioxide, Uranium Hydride), and special purpose materials. In ENDF/B-VIII.1 we also distribute alongside the evaluated files, a comma-separated file (CSV), named TSL_MAT_numbers.csv, which lists all evaluated files in the current release and their corresponding unique MAT number.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Triton Reaction Sublibrary

The tritons sublibrary aims to describe nuclear reactions between incident triton particles and different nuclei. For ENDF/B-VIII.1, compared to ENDF/B-VIII.0, there were only evaluation updates for 4 He.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Adjusted Graphite Thermal Scattering Data Applied to Fluoride-Salt-Cooled Reactors

Thermal scattering data is a key component of fission reactor simulation. While most materials can be reasonably modeled with a free gas approximation, the moderators in thermal systems require a more rigorous treatment to sufficiently capture their specific neutron interaction physics. Graphite does not typically feature as a principle moderator in power reactors in service today, but remains an attractive prospect for a number of new reactor concepts currently in development. In comparison to light water, a material nearly ubiquitous in thermal reactors, graphite offers a better neutron economy by way of a small absorption cross section, high thermodynamic stability and can be incorporated into the structure of the core itself. Graphite in its purest form is a crystalline material composed of stacked layers of graphene, a hexagonal planar lattice of carbon atoms. In-plane, each carbon atom is covalently bonded to 3 neighbouring carbon atoms, creating a strong bond, while between planes only a weak van der Waals force exists. Pure graphite has a density of 2.26 g/cm{sup 3}. However, reactor-grade contains a significant amount of carbon binder, leading to a porous bulk medium with randomly oriented micro crystals. Densities of 1.5 - 1.9 g/cm{sup 3} are typical for reactor-grade graphite, corresponding to porosities on the order of 10-30%. This deviation from pure crystalline graphite should be accounted for in the thermal scattering data. When an incident neutron is in or near the thermal region, its de Broglie wavelength is the same order as the molecular bond lengths, leading to elastic scattering. At similar energies, the dynamics of the moderator lattice (in the case of solid moderators) can cause inelastic scattering. For these moderators, special cross sections should be calculated and used in reactor simulations. To generate this thermal scattering data, parameters pertaining to the properties and structure of the material in question are fed into a nuclear data processing code, such as NJOY. Within NJOY is a module called LEAPR which is used to generate the thermal scattering law S(α,β), which is in turn is fed into another module THERMR which is used to generate thermal cross sections that factor in the molecular-scale scattering phenomena. The core of the thermal scattering law is the phonon spectrum ρ(β) which describes the modes of excitation of the material lattice. An investigation into the phonon frequency spectrum was conducted at North Carolina State University (NCSU) for reactor-grade graphite which accounts for porosity. This phonon spectrum was calculated using the ab initio density functional theory (DFT) code VASP. The inelastic cross section calculated using this spectrum was found to have good agreement with relevant experimental data. The resulting spectra differed from the benchmark spectrum produced by Young and Koppel, the same data used in current nuclear data evaluations. The latter uses the lattice dynamics of perfect crystalline graphite with four force constants fitted to the model from thermodynamics data from reactor-grade (porous) graphite, which yields a reasonable approximation to the phonon spectrum of reactor-grade graphite.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Resolving the Discontinuity Between Thermal Scattering Data and Fast Data

The task of resolving discontinuities between thermal cross sections and fast cross sections in AMPX processing is part of a broader project to provide guidance on the potential effect of unknown thermal neutron scattering law (TNSL) data, primarily from the standpoint of criticality safety. The primary goal of this work is to deliver code fixes in AMPX that increase confidence that SCALE libraries provide physically correct cross sections across the full range of neutron energies encountered in transport applications. This report discusses the resolution of two sources of discontinuity. The first issue occurs when the Bragg edges in an evaluated nuclear data file do not extend to 5 eV. The second issue concerns parameters chosen during the processing of the nuclear data library—specifically, the parameter that defines the energy at which the thermal data ends and the fast nuclide data begins. This report also discusses a third issue, which is related to the calculation of probability distributions from a densely gridded double-differential cross section.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Uncertainty Analysis for a Criticality Benchmark [Slides]

This presentation discusses how KRUSTY was analyzed with LANL’s MCNP neutron transport code, Monte Carlo N-Particle. In the Monte Carlo method, neutrons are born in fission events and fly around the geometry having events (e.g., scatter, capture, fission, escape) according to the natural probabilities. “Natural probabilities” are given as neutron cross sections , which are measured and evaluated nuclear data. One of the goals of the benchmark project is to improve the nuclear cross sections (and the codes that use them). In this work, the response of interest is called k eff , which describes the neutron multiplication in a near critical system. Like most quantities, k eff is not measured directly; it is only inferred. In summation, MCNP results for k eff matched the measurements extremely well. The evaluated uncertainties are asymmetric and smaller than 0.00100 (e.g., + 0.00080 / – 0.00062). The largest uncertainties are due to alignment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Uncertainty Analysis for a Criticality Benchmark [Slides]

This presentation discusses how KRUSTY was analyzed with LANL’s MCNP neutron transport code, Monte Carlo N-Particle. In the Monte Carlo method, neutrons are born in fission events and fly around the geometry having events (e.g., scatter, capture, fission, escape) according to the natural probabilities. “Natural probabilities” are given as neutron cross sections , which are measured and evaluated nuclear data. One of the goals of the benchmark project is to improve the nuclear cross sections (and the codes that use them). In this work, the response of interest is called k eff , which describes the neutron multiplication in a near critical system. Like most quantities, k eff is not measured directly; it is only inferred. In summation, MCNP results for k eff matched the measurements extremely well. The evaluated uncertainties are asymmetric and smaller than 0.00100 (e.g., + 0.00080 / – 0.00062). The largest uncertainties are due to alignment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗