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Presentation discussing the research and development of deep neural network models for modeling microscopic neutron cross-section data in the Griffin reactor physics application for pebble-bed reactors. This work details advancements made between the last review meeting in July 2024 until July 2025.
A rudimentary code system has been developed for propagating neutron cross-section uncertainties to response uncertainties using the PARTISN multigroup discrete-ordinates neutron transport code. The code system uses the first-order sandwich rule with sensitivities computed using SENSMG and covariances computed using NJOY. Databases of reaction cross sections and covariances using ENDF/B-VIII.0 data in the LANL standard 30-group structure have been precalculated. The final uncertainty in a k eff example problem compares well with the result from TOFFEE, a similar code that propagates neutron cross-section uncertainties to response uncertainties using the MCNP6 Monte Carlo neutron transport code. Presently, neither code system propagates uncertainties in $\overline{v}$, $\overline{μ}$, or χ, and neither code system propagates nuclide-nuclide covariances
This work details the generation of enrichment-dependent thermal neutron scattering cross sections for several crucial uranium fuel compounds. The evaluations of the thermal scattering law (TSL) and associated cross sections for uranium dioxide (UO 2 ), uranium carbide (UC), and uranium nitride (UN) were performed using standard ab initio lattice dynamics (AILD) methods. The data for uranium metal was produced using a novel hybrid approach of molecular dynamics combined with lattice dynamics methods. 235 U enrichments of 5%, 10% (LEU+), 19.75% (HALEU), 93% (HEU), and 100% were considered, in addition to natural uranium. The enrichment-dependent masses and free atom cross sections were used in the generation of elastic and inelastic thermal neutron scattering cross sections, while the calculation of the phonon density of states (DOS) and resulting TSL considered only the natural isotopic composition of uranium. The use of an identical DOS for all enrichments is expected to have minimal impact on the final data, as the small change in uranium mass should not significantly affect lattice vibrations. The cross sections are shown to exhibit significant dependence on 235 U enrichment. The submission of this data to the National Nuclear Data Center (NNDC) for release in the ENDF/B-VIII.1 database should support the design of advanced reactor concepts.
The understanding of astrophysics processes and the performance of nuclear reactors and other nuclear systems depend on a precise description of the neutron interaction cross sections for materials and nuclei present in these environments. At low neutron energies, these cross sections exhibit resonance structure represented by sharp enhancements when the neutron energy is sufficiently close to excited levels in a compound nucleus. Such resonances can be characterized by their quantum numbers relative to angular momenta, which are often deduced in an ad hoc and irreproducible manner from the shape of the cross sections. The correct assignment of the quantum numbers of neutron resonances is therefore of paramount importance. To address this we have developed a machine-learning method to automate the identification and correction of these spin assignments. The algorithm is trained from simulated data, generated from statistical properties of resonance data for a given nucleus, to mimic the errors found in real data. In this project we describe five independent approaches to further develop and expand the applicability of the machine-learning spin classifier: i) Feature impact; ii) Integration with the Atlas; iii) Training optimization; iv) Spacings systematics; and v) Validation with polarized data. The premises, methods, results, and future perspectives are discussed.
We report the neutron capture cross section, the probability per unit area a neutron is absorbed by a nucleus followed by the emission of gamma ray(s), is an important quantity in several fields such as astrophysics, nuclear criticality safety, radiochemical diagnostics, nuclear medicine, nuclear forensics and nuclear security. Accurate knowledge of how neutrons interact with matter is a crucial component to understand physical processes in depth. A great example is the understanding of how our universe was created, a quest that requires accurate modeling in which neutron capture cross sections are a key ingredient
The Mars Science Laboratory (MSL) Curiosity rover recently traversed over plateaus of mafic aeolian sandstones (the 'Stimson' formation) that overlie mudstones (the 'Murray' formation). Within the Stimson formation we observed many lighter-toned, halo-forming features, that are potentially indicative of fluid alteration (see Fig. 1). These halo features extend for tens of meters laterally and are approx.1 meter wide. The halo features were characterized by Curiosity's geochemical instruments: Alpha Proton X-Ray Spectrometer (APXS), Chemin, Chemcam and Sample Analysis at Mars (SAM). With respect to the host (unaltered) Stimson rocks, fracture halos were significantly enriched in silicon and low in iron [1]. Changes in hydrogen abundance (due to its large neutron scattering cross section) greatly influence the magnitude of the thermal neutron response from the Dynamic Albedo of Neutrons (DAN) instrument [2]. There are also some elemental species, e.g. chlorine, iron, and nickel, that have significant microscopic neutron absorption cross sections. These elements can be abundant and variable results provide a useful estimate of the lower bound for bulk hydrogen content (assuming a homogeneous distribution).
Commercial and laboratory neutron detection systems use indirect neutron response of materials like pressurized helium-3 (via nuclear reaction 3He(n, p)3H) to measure and count neutrons emanating from a source. Recently a host of semiconductors, especially a ternary semiconductor of lithium indium diselenide (6LiInSe2) and a quaternary alloy of enriched lithium-6, indium, phosphorous, and selenium (6LiInP2Se6), have shown promising neutron counting possibilities by directly converting neutrons into charge-carrying elements within the body of the semiconductor. These semiconductors have high thermal neutron capture cross sections, suitable energy bandgaps (~2.0 electron volts) for room-temperature operations, and a favorable electronic band structure for efficient electron charge transport. The article examines the semiconductor properties of these compounds in terms of their neutron counting capabilities and possible ways to extract neutron energy information from them. Lithium-6 and boron-10 (with thermal neutron absorption cross sections of 938 ± 6 and 3855 ± 26 barns, respectively) produce charged particles to be measured via indirect neutron interactions. The efficiency of indirect conversion neutron detectors is limited because of the inefficiencies in conversion mechanism. In case of direct conversion, the neutrons create charged particles in a single material for neutron capture and charge collection, increasing detection efficiency. Unlike 3He proportional counters, which provide no neutron energy information, the semiconductors can be used as neutron energy spectrometer. Fully resolved neutron energy by 6LiInP2Se6 from a plutonium-beryllium source has been reported in the literature. We will discuss the influence of these multilayered semiconductors’ crystallographic structures and growth techniques on neutron energy determination.
Improved neutron inelastic scattering cross sections have repeatedly been identified as a top priority nuclear data need, important for basic science and a range of applications in nuclear energy, stockpile stewardship, and proliferation detection. For the C 12 ( n , n ′ γ ) reaction in particular, recent measurements have unveiled some structural discrepancies, demonstrating incongruities among themselves and in relation to the ENDF/B-VIII.0 nuclear data evaluation. To help resolve these disagreements, a measurement was performed at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory using a broad-spectrum neutron beam and a 99.8% pure natural carbon target. The Gamma Energy Neutron Energy Spectrometer for Inelastic Scattering (GENESIS) was employed to measure energy-differential γ -ray emission spectra as a function of incident neutron energy in the energy range of 5.5 to 16.7 MeV. The C 12 partial γ -ray cross sections were extracted at 63 ∘ , 122 . 5 ∘ , and 150 ∘ with respect to the incoming neutron beam and integrated using angular distribution data available in the literature. The data show agreement with a recent literature measurement and evaluation from 11 to 15 MeV, but indicate a larger cross section for incident neutron energies between 5.5 and 8.5 MeV. The measured relative angular distributions are also reported and were found to agree with evaluation. Published by the American Physical Society 2025
Uranium-233 plays an important role in the Th-U fuel cycle, with substantial production in the cycle. This cycle has been proposed as an alternative to the U-Pu fuel cycle due to its reduced production of transuranic elements. An accurate measurement of the 233 U($\eta$,$\gamma$) cross section is required by the National Criticality Safety Program (NCSP) to complete the neutron-induced cross section data, where experimental capture cross section data are scarce and were measured decades ago. The most recent capture cross section data available in the literature were measured in 2007 at the n_TOF facility (CERN); in the 60s measurements were performed at Rensselaer Polytechnic Institute (RPI) and at LANL. Finally, as reported by ORNL, a new evaluation with a revised (renormalized) fission cross section is needed on 233 U. The challenge for this measurement lies in the difficulty of measuring the capture cross section data in the competing fission background, as the fission cross section is around one order of magnitude larger than the capture cross section for 233 U. The accuracy of a capture cross section measurement depends on discrimination between $\gamma$’s produced in capture and fission reactions, for which an experimental setup combining capture and fission detectors is needed. For the ($\eta$,$\gamma$) cross section measurement at LANSCE, this discrimination is achieved by combining the Detector for Advanced Neutron Capture Experiments (DANCE), to measure $\gamma$’s from capture reactions, with a Parallel Plate Avalanche Counter (PPAC) to tag the $\gamma$’s produced by fission. This method was successfully used to measure 235 U and 239 Pu capture cross sections. In these measurements, the neutron capture cross section was determined in a large fission background well above 100 keV. As part of the NCSP nuclear data effort, we have looked at past DANCE measurements on 233 U($\eta$,$\gamma$) and evaluated whether existing data is adequate to apply this technique.
The interactions of galactic cosmic rays (GCR) with extraterrestrial bodies produce small amounts of radionuclides and stable isotopes. The production rates of many relatively short-lived radionuclides, including 2.6-year Na-22 and 312-day Mn-54, have been measured in several meteorites collected very soon after they fell. Theoretical models used to calculate production rates for comparison with the measured values rely on input data containing good cross section measurements for all relevant reactions. Most GCR particles are protons, but secondary neutrons make most cosmogenic nuclides. Calculated production rates using only cross sections for proton-induced reactions do not agree well with measurements. One possible explanation is that the contribution to the production rate from reactions initiated by secondary neutrons produced in primary GCR interactions should be included explicitly. This, however, is difficult to do because so few of the relevant cross sections for neutron-induced reactions have been measured.
Nuclear data is the linchpin underwriting several fundamental capabilities and mission needs at LANL. New techniques such as machine learning can be brought to bear to solve old problems such as multigroup cross-section accuracy. In neutron transport, generating multigroup cross sections is a complex and arcane task, but a crucial one, as accurate solutions require appropriate cross sections. There are two key challenges when generating multigroup cross sections: (1) choosing an accurate weight function, and (2) choosing appropriate energy boundaries. Often, energy boundaries are chosen using “expert judgment” that is not documented and is difficult to replicate. The long-standing Los Alamos 30-group structure has been in use since at least 1969 and is still in use today. Simplistic attempts over the years since to improve on the 30-group structure have been met with limited success. Machine learning algorithms would enable the selection of appropriate, problem-dependent group boundaries without an inordinate investment of scientist time. We will develop workflows and tools to enable these improved group boundary choices, which will reduce uncertainty and increase predictive capability of neutron-transport applications at LANL.
We have developed a set of modeled nuclear reaction cross sections for use in radiochemical detector diagnostics. Systematics for the input parameters required by the Hauser-Feshbach statistical model developed in the TALYS code system are used to calculate neutron induced nuclear reaction cross sections for targets ranging from Thallium (Z = 81) to Bismuth (Z = 83).
Inelastic neutron and capture gamma ray cross section libraries for COHORT Monte Carlo procedure
The Intentional Forensics (IF) project seeks to design a suite of tagging materials that can be used to unobtrusively “barcode” nuclear fuel. The idea being that, if the fuel leaves institutional controls and is later recovered, the “barcode” can be used to determine the provenance of the fuel. The IF project is investigating several schemes for this “barcode” and taggants with tailored isotopic compositions are once leading concept. Recently 18 O was suggested as a potential taggant [IF2023]. This report assesses the quality of 16,17,18 O evaluations in the ENDF/B-VIII.0 library [ENDF8.0]. In a neutron-rich environment such as a nuclear reactor, we are interested in two main processes – neutron scattering and neutron absorption (or capture). For the typical neutron energies encountered in a reactor, absorption is determined exclusively by the neutron radiative capture cross section (in which a neutron is absorbed, and the residual nucleus emits one or more gamma rays). The capture cross sections are described by and can be completely determined from R-matrix parameters in the ENDF evaluations. Therefore, it is sufficient to consider the quality of the neutron resonance part of these evaluations. Neutron scattering is determined mainly by the elastic scattering cross section which is very similar for all oxygen isotopes and is essentially constant until around 100 keV. The cross section varies from 3-4 barns, depending on the isotope, and 16 O has the largest elastic cross section of all isotopes. Therefore, we expect that significant amounts of 17 O and 18 O could diminish the moderating role of oxygen in fuel. We note that, in traditional power reactor, the hydrogen in the coolant water provides the vast majority of the moderation in the reactor, so the impact of 17 O and 18 O in any taggant would be minimal. At higher energies, neutron resonances start to play a role and the R-matrix parameters determine the cross sections. Fig. 1 shows the mass region near stable oxygen isotopes. As nuclei capture neutrons, they transmute into a different isotope, one unit to the right. In particular, 16 O transmutes to 17 O, 17 O transmutes to 18 O and 18 O transmutes to 19 O. 19 O, being unstable, beta decays in 27 s to 19 F. So, even if the 18 O neutron capture data is good, the fact that it transmutes into 19 F may be problematic and needs investigation. Fig. 2 from [NuDat] shows the thermal neutron capture cross sections for all nuclei in the ENDF/B-VII.1 library [ENDF7.1]. Clearly oxygen thermal capture values are quite small compared to other nuclei. Note, there is no 18 O evaluation in the ENDF/B-VII.1 library. In any event, given that neutron capture events lower the reactor reactivity and potentially induce chemical changes in the fuel, we will focus our attention on the capture part of the neutron resonances.
Cross sections for neutron scattering from hydrogen and deuterium taking into account spin correlations, rotations and vibrations
A diffusion-based and a discrete ordinates method are used to simulate a neutron noise experiment in the research reactor AKR-2 at the Technical University in Dresden, Germany. The AKR-2 reactor provides an interesting case for the comparison between the two methods because it is characterized by large heterogeneities and regions with low macroscopic neutron cross-sections. For the calculations, the same spatial discretization and the same set of two-energy macroscopic neutron cross-sections with isotropic scattering are used. Significant discrepancies between the diffusion-based and discrete ordinates methods are found in regions of the systems where the diffusion approximation is expected to be inaccurate in reproducing characteristics of the static neutron flux and neutron noise. (authors)
The Rare-RI Ring (R3) is a recently commissioned cyclotronlike storage ring mass spectrometer dedicated to mass measurements of exotic nuclei far from stability at Radioactive Isotope Beam Factory (RIBF) in RIKEN. The first application of mass measurement using the R3 mass spectrometer at RIBF is reported. Rare isotopes produced at RIBF— 127 Sn, 126 In, 125 Cd, 124 Ag, 123 Pd—were injected in R3. Masses of 126 In, 125 Cd, and 123 Pd were measured whereby the mass uncertainty of 123 Pd was improved. This is the first reported measurement with a new storage ring mass spectrometry technique realized at a heavy-ion cyclotron and employing individual injection of the preidentified rare nuclei. The latter is essential for the future mass measurements of the rarest isotopes produced at RIBF. The impact of the new 123 Pd result on the solar r-process abundances in a neutron star merger event is investigated by performing reaction network calculations of 20 trajectories with varying electron fraction Y e . It is found that the neutron capture cross section on 123 Pd increases by a factor of 2.2 and β-delayed neutron emission probability, P 1n , of 123 Rh increases by 14%. The neutron capture cross section on 122 Pd decreases by a factor of 2.6 leading to pileup of material at A = 122, thus reproducing the trend of the solar r-process abundances. The trend of the two-neutron separation energies (S 2n ) was investigated for the Pd isotopic chain. The new mass measurement with improved uncertainty excludes large changes of the S 2n value at N = 77. Finally, such large increase of the S 2n values before N = 82 was proposed as an alternative to the quenching of the N = 82 shell gap to reproduce r-process abundances in the mass region of A = 112 – 124.
Engineers often stack printed circuit boards (PCBs), populated with semiconductor devices, one behind the other in neutron beams to improve the error statistics in radiation-effects tests. As the neutron beam traverses through the boards, the neutron flux is both attenuated and the energy spectrum changes, which may have a significant impact on the results of such tests. In this work, we have developed a technique to measure the energy-dependent neutron attenuation due to the PCBs and to determine the energy-dependent effective cross section of neutron interactions with a single PCB. With this effective cross section, engineers can calculate the change in the neutron energy spectrum and the total neutron attenuation after any number of identical PCBs in the stack. This information will allow test engineers to partially correct inaccuracies in soft error tests.