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Results for “Neutron Cross Sections”

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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

140,142 Ce Neutron Cross Section Resolved Resonance Region Evaluation

A resolved resonance region evaluation of 140,142 Ce was conducted by Oak Ridge National Laboratory. Requested by the US Nuclear Criticality Safety Program, this evaluation is based on recent high-resolution transmission and capture high-resolution measurements of nat Ce and 142 Ce conducted at JRC-Geel at the Geel Linear Accelerator facility. It is also based on recently measured thermal constants available from the EX FOR database. Starting from the resonance parameters from the ENDF/B-VIII.0 library and following a preliminary R-matrix analysis, an updated set of resonance parameters and corresponding covariance in formation was derived by the fit of these experimental datasets using the Reich–Moore approximation of the R-matrix theory, as implemented in the SAMMY code system. The resolved resonance region upper energy limit for 140 Ce was kept at 200 keV, whereas the 142 Ce resonance region was extended from 13 to 26 keV. This new evaluation was found to be in good agreement not only with several integral quantities of interest to the reactor physics community, but also with the stellar Maxwellian-averaged cross section.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1997

Accurate universal parameterization of absorption cross sections II--neutron absorption cross sections

A recent parameterization (here after referred as paper I, Ref. [4]) of absorption cross sections for any system of charged ions collisions including proton -nucleus collisions, is extended for neutron-nucleus collisions valid from approximately 1 MeV to a few GeV, thus providing a comprehensive picture of absorption cross sections for any system of collision pair (charged and/or uncharged). The parameters are associated with the physics of the problem. At lower energies, the optical potential at the surface is important and the Pauli operator plays an increasingly important role at intermediate energies. The agreement between the calculated and experimental data is better than earlier published results.

NASA Discipline Radiation Health↗
NASA NTRS1988

Nucleon-nucleus interaction data base: Total nuclear and absorption cross sections

Neutron total cross sections are represented for Li to Pu targets at energies above 0.1 MeV and less than 100 MeV using a modified nuclear Ramsauer formalism. The formalism is derived for energies above 100 MeV by fitting theoretical cross sections. Neutron absorption cross sections are represented by analytic expressions of similar form, but shape resonance phenomena of the Ramsauer effect is not present. Elastic differential cross sections are given as a renormalized impulse approximation. These cross section data bases are useful for nucleon transport applications.

Wilson, J. W.↗
DOE OSTI2024

Reduced-Order Modeling of Multigroup Neutron Cross Sections for High-Temperature Gas-cooled Reactors

Abstract – Deterministic neutronics calculations rely on multigroup neutron cross section libraries, which consist of databases of tabulated values, used to calculate the neutron cross sections through multivariate linear interpolation. However, interpolation of the multidimensional cross section data becomes memory inefficient and time consuming as the number of tabulations increases, significantly slowing down the neutronics calculation, especially in the case of microscopic cross section libraries where every isotope (on the order of hundreds) has its own set of specific reactions and cross sections. In order to address this challenge, this work constructs efficient and robust reduced-order models (ROMs) of the multi-group cross sections to support the Griffin simulation of high-temperature gas-cooled reactors (HTGRs). The first part of the study investigates the linearity of the multigroup cross section data across isotopes, reaction types, and energy groups on pre-generated datasets for the purpose of dimensionality reduction. Secondly, a down-selection of ROM techniques is presented on representative classical machine learning (ML) techniques, including variants of linear regression, kernel-based methods, tree-based algorithms, and artificial neural networks. The selection criteria jointly consider the memory efficiency, predictive accuracy, prediction speed, and scalability in comparison to the multidimensional interpolation. Among all the ML techniques, deep neural networks (DNNs) have proven to be the best selection with sufficient accuracy, high robustness, good memory efficiency, great scalability, and superior flexibility. DNNs have been trained for all isotopes in this work and systematic Griffin testing is ongoing to ensure the feasibility of this ROM technique for predicting cross section and reducing memory requirements without a significant sacrifice in computational performance.

42 - ENGINEERING↗
DOE OSTI2024

Reduce-Order Modeling of Multigroup Neutron Cross Sections for High-Temperature Gas-cooled Reactors

Deterministic neutronics calculations rely on multigroup neutron cross section libraries, which usually consists of a database of tabulated values, used to calculate the cross sections through multivariate linear interpolation. However, interpolation of the multidimensional cross section data becomes memory inefficient and time consuming as the number of tabulations increases, significantly slowing down the neutronics calculation, especially in the case of micro cross section libraries where every isotope (on the order of hundreds) has its own set of specific reactions and cross sections. To address this challenge, this work constructs efficient and robust reduced-order models (ROMs) of the multi-group cross sections to support the Griffin simulation of high-temperature gas-cooled reactors (HTGRs). The first part of the study investigates the linearity of the multi-group cross section data across isotopes, reaction types and energy groups on pre-generated datasets for the purpose of dimensionality reduction. Secondly, a down-selection of ROM techniques is presented on representative classical machine learning (ML) techniques, including variants of linear regression, kernel-based methods, tree-based algorithms, and artificial neural networks. The selection criteria jointly consider the memory efficiency, predictive accuracy, prediction speed, and scalability in comparison to the multidimensional interpolation. Among all the ML techniques, deep neural networks (DNNs) have proven to be the best selection with sufficient accuracy, high robustness, good memory efficiency, great scalability, and superior flexibility. DNNs for have been trained for all isotopes in this work and systematic Griffin testing is ongoing at this moment to ensure the feasibility of this ROM technique for cross section predictions.

42 - ENGINEERING↗
DOE OSTI2022

Total thermal neutron cross section measurements of hydrogen dense polymers from 0.0005–20 eV

Hydrogen dense polymers, specifically polyethylene, polystyrene, and Plexiglas, have served as neutron moderator and reflector materials in hundreds of separate critical benchmark experiments because of their low cost and abundance of hydrogen. In order to accurately model and simulate these critical benchmarks, the thermal scattering law (TSL) evaluation that governs how neutrons will thermalize must be well understood and rigorously validated. To support this validation, researchers at Rensselaer Polytechnic Institute performed total neutron cross section measurements for high-density polyethylene & polystyrene over the energy range 0.0005–20 eV and for Plexiglas G & Plexiglas G-UVT over the energy range 0.0005–3 eV. Comparisons were made between the measured cross section and that predicted by the ENDF/B-VIII.0 and Oak Ridge National Laboratory/European Spallation Source/Rensselaer Polytechnic Institute TSL evaluations for polyethylene, Plexiglas, and polystyrene, as available. These experiments represent the first total neutron cross section measurements for polystyrene.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Measurement of the neutron cross section on argon between 95 and 720 MeV

We report an extended measurement of the neutron cross section on argon in the energy range of 95-720 MeV. The measurement was obtained with a 4.3-hour exposure of the Mini-CAPTAIN detector to the WNR/LANSCE beam at LANL. Compared to an earlier analysis of the same data, this extended analysis includes a reassessment of systematic uncertainties, in particular related to unused wires in the upstream part of the detector. Using this information we doubled the fiducial volume in the experiment and increased the statistics by a factor of 2.4. Here we also shifted the analysis from energy bins to time-of-flight bins. This change reduced the overall considered energy range, but improved the understanding of the energy spectrum of incoming neutrons in each bin. Overall, the new measurements are extracted from a fit to the attenuation of the neutron flux in five time-of-flight regions: 140ns-180ns, 120ns-140ns, 112ns-120ns, 104ns-112ns, 96ns-104ns. The final cross sections are given for the flux-averaged energy in each time-of-flight bin with statistical and systematic (syst) uncertainties: σ(146 MeV) = 0.60 $^{+0.14}_{-0.14}$ ±0.08(syst) b, σ(236 MeV) = 0.72 $^{+0.10}_{-0.10}$ ± 0.04(syst) b, σ(319 MeV) = 0.80 $^{+0.13}_{-0.12}$ ±0.040(syst) b, σ(404 MeV) = 0.74 $^{+0.14}_{-0.09}$ ±0.04(syst) b, σ(543 MeV) = 0.74 $^{±0.09}_{-0.09}$ ± 0.04(syst) b.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2025

The total neutron cross section of liquid and solid ammonia

Ammonia is a material of interest for future neutron moderators at high-power sources due to its high hydrogen density, low melting point, and resistance to polymerization in an intense radiation field. Its performance in such applications cannot currently be calculated due to the absence of suitable computer models for the interaction of neutrons with ammonia under relevant conditions. In an effort to develop suitable scattering kernels for computer simulations of moderator performance, we have conducted a series of Density Functional Theory and Molecular Dynamics calculations of the molecular-level thermal properties of ammonia at various temperatures within both the solid and liquid phases. In this paper, we compare computer calculations for the energy-dependent total neutron cross section of ammonia, based on these models, to experimental measurements of those cross sections at temperatures of 221 K, 180 K, and 35 K. The experimental data were collected over an energy range from 0.1 meV to 10 eV using time-of-flight techniques at the Low Energy Neutron Source (LENS) facility at Indiana University. This comparison provides a first validation in the development of thermal scattering libraries for Monte Carlo source design simulations based on liquid and solid ammonia. In conclusion, we also provide some insights into where additional development of tools for creating such models may be needed.

Ammonia↗
DOE OSTI2022

Total thermal neutron cross section measurements of yttrium hydride from 0.0005 - 3 eV

Yttrium hydride serves as a neutron moderator material that enables compact, high temperature nuclear reactors. However, in order to accurately design and simulate a nuclear system relying upon yttrium hydride, the fundamental nuclear data of yttrium hydride must be well understood. Thermal neutron scattering law (TSL) evaluations represent an important aspect of nuclear data as thermal scattering can drastically alter the neutron multiplication factor of a system. Therefore, to support evaluation and validation of thermal neutron scattering for yttrium hydride, researchers at Rensselaer Polytechnic Institute (RPI) performed total thermal neutron cross section measurements for YH 1.68 and YH 1.85 over the energy range of 0.0005 - 3 eV. Further, these measurements represent the first total cross section measurements for yttrium hydride that encompass the entire thermal region. Comparisons were made against the ENDF-B/VIII.0, Zerkle & Holmes and Oak Ridge National Laboratory TSL evaluations, where generally good agreement was found.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Neutron Cross Section Measurement In The Protodune-SP Experiment

Understanding the detector response to neutrons will be critical for performing neutrino oscillation analyses in the next-generation Deep Underground Neutrino Experiment (DUNE). The DUNE physics program is centered around measuring the neutrino flavor composition as a function of their energy both at the near and the far detector. Neutrinos in the DUNE beam will have energies ranging between 100~MeV and 10~GeV, which is significant, because individual neutrino energies will not be known beforehand and will have to be reconstructed. Neutrino interactions in DUNE will produce leptons and hadrons -- including protons, pions, and neutrons. Neutrons can transport energy away from their origin and sometimes go undetected. In addition to the primary neutrons produced by the neutrino, subsequent interactions of the charged hadrons can result in secondary neutrons. Neutrons are a source of missing energy and will bias the neutrino energy measurement. Currently, there is also a 20\% energy scale uncertainty and a 40\% uncertainty on the energy resolution for neutrons in DUNE, which must be addressed. ProtoDUNE Single-Phase (ProtoDUNE-SP) is a 770-ton prototype for the DUNE far detector and was designed to both validate the technology that will be employed in DUNE and to measure cross sections for the charged hadrons (pions, protons, and kaons) at the relevant energies for DUNE. The ProtoDUNE-SP experiment, therefore, is in a unique position to characterize the secondary neutron component for DUNE. This is achieved by searching for candidate neutron interactions in ProtoDUNE-SP events and using these to facilitate a measurement of the neutron inelastic cross section as well as an estimate of the neutron energy and number. The cross section measurement presented here is based on neutrons produced in 1~GeV, $\pi^+$ events captured in 2018 by ProtoDUNE-SP in accordance with the production and cross section models in the GEANT4 simulation toolkit, version 4.1 0.6p1. T he best-fit neutron inelastic cross section, in the kinetic energy range of 114 to 314~MeV, is $1.24_{-0.08}^{+0.10}$~(stat.~$\oplus$~syst.)~barns.

Rivera, David Orlando↗
DOE OSTI2021

Progress on the reevaluation and validation of the n+233U neutron cross sections

The set of 233 U resonance parameters of the ENDF/B-VIII.0 nuclear data library was adopted from the previous ENDF/B-VII.1 evaluation using the external levels to update the thermal values. Adoption of IAEA 2017 thermal standards ( σ f = 533.0 ± 2.2 b, σ c = 44.9 ± 0.9 b, and ν ‾ tot = 2.487 ± 0.011 ) and of the IAEA-recommended thermal-neutron induced prompt fission neutron spectrum (PFNS) with average PFNS energy of 2.030 ± 0.013 MeV requires a re-evaluation of 233 U neutron cross sections in the resolved resonance region. A newly produced evaluation is being tested on benchmarks carefully selected from the Handbook of International Criticality Safety Benchmark Experiments (ICSBEP) which are highly sensitive to 233 U data. An important goal of this work was to eliminate the strong negative gradient of the calculated effective multiplication factors with respect to the epithermal fission fraction observed in the validation of the ENDF/B-VIII.0 library for those assemblies. A significant improvement in integral performance of critical 233 U solutions is observed for the newly proposed evaluation. Further work addressing the fast neutron region is needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

ORNL Neutron Cross Section Measurements of 90 Zr

Nuclear criticality modeling and simulations rely on the quality of the existing evaluated nuclear data libraries such as Evaluated Nuclear Data File (ENDF)/B, the Joint Evaluated Fission and Fusion (JEFF) nuclear data library, or the Japanese Evaluated Nuclear Data Library (JENDL). In some cases, the cross-section evaluations of those libraries were found to be deficient in describing criticality benchmarks accurately. More than two decades ago, the US Nuclear Criticality Safety Program (NCSP) established a Nuclear Data (ND) task which encompassed experiments and evaluations. In response to this, the Oak Ridge National Laboratory (ORNL) formed a Nuclear Criticality and Data group which performed ND experiments, data analysis, and evaluations to produce ENDF files for the ND libraries as identified in the NCSP Five-Year Plan. Before being submitted to the ENDF library, files were processed and tested for performance by running benchmark calculations. This procedure was centralized in the ORNL group and is now often referred to as the ND pipeline. NCSP collaborates with the Joint Research Center (JRC) of the European Commission in Geel, Belgium, to perform high-resolution neutron-induced cross section measurements at the Geel Linear Accelerator (GELINA). The objective is to address emerging ND problems in criticality calculations. Difficulties with ND include insufficient neutron energy range, missing covariances, and previously unrecognized inaccuracies with experiments. New neutron total and capture cross sections of 90 Zr in the neutron energy range from 100 eV to several hundred keV were recently performed. These measured data will be used, together with existing high-resolution transmission data from a metallic 90 Zr sample, to improve representation of the cross sections.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Thermal Neutron Cross Section Measurements at the RPI LINAC [Slides]

Cold polyethylene shows adequate capacity for enhancing sub-thermal neutron flux. Cold moderator system produced high quality thermal cross section measurements. Measurements were performed for polyethylene, polystyrene, Plexiglas, yttrium hydride and beryllium and compared with ENDF/B-VIII.0 TSL evaluations. All data reduction for all measurements was finalized. Future work includes completion of necessary journal publications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Ab Initio Evaluation of Uranium Carbide S(α,β) and Thermal Neutron Cross Sections

Uranium Carbide (UC) is a nuclear fuel material which offers better neutron economy and lower fuel-cycle costs compared to conventional mixed-oxide fuels. UC’s lattice binding and dynamical properties impact thermal neutron scattering and low temperature epithermal resonance absorption. The Thermal Scattering Law (TSL) describes the scattering system available energy and momentum transfer states. There is no TSL evaluation for UC in the ENDF/B-VIII.0 database; herein, ab-initio lattice dynamics (AILD) techniques are invoked to calculate the phonon spectrum for UC using spin-orbit-coupling density functional theory (DFT). The TSLs, inelastic and elastic thermal scattering cross sections for Uranium and Carbon in UC, respectively, are calculated in FLASSH for use in higher fidelity reactor design calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Ab Initio Evaluation of Plutonium Dioxide $S(α,β)$ and Thermal Neutron Cross Sections

Plutonium Dioxide (PuO 2 ) is an advanced fuel-cycle nuclear fuel material for thermal and fast reactors, with a melting point at 3261.8 K and thermal conductivity above 3 Wm -1 K -1 for temperatures below 1500 K. Plutonium from spent UO 2 is reprocessed to form chemically stable mixed oxide (MOX) fuel. In contrast to metallic fuels, the low PuO 2 conductivity results in high thermal gradients between fuel center and surface, resulting in structural annealing and efficient fission product migration over burnup. Crystal binding affects the doppler broadening of epithermal resonances for nuclear fuels and scattering behavior of thermal neutrons. Currently, there is no ENDF/B Thermal Scattering Law, i.e., $S(α,β)$, evaluation for PuO 2 , only free atom cross sections for Plutonium and Oxygen, respectively, which neglect lattice contributions to total cross section. Herein, ab initio lattice dynamics (AILD) techniques are employed to calculate the phonon density of states (DOS) using spin-orbit-coupling density functional theory (DFT) to predict the PuO2 paramagnetic ground-state structure. These results highlight the evaluation of $S(α,β)$ for Plutonium and Oxygen, respectively in PuO 2 and consequential generation of thermal neutron scattering cross sections for high fidelity criticality safety analysis and reactor calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗