Evaluations of the 239 Pu Fission Source Term
239 Pu(n,f) PFNS is being re-evaluated to include new model and high-impact Chi-Nu (LANL-led) and CEA exp. data.
07 ISOTOPE AND RADIATION SOURCES↗
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NuTools: Exploring Practical Roles for Neutrinos in Nuclear Energy and Security
For decades, physicists have used neutrinos from nuclear reactors to advance basic science. These pursuits have inspired many ideas for application of neutrino detectors in nuclear energy and security. While developments in neutrino detectors are now making some of these ideas technically feasible, their value in the context of real needs and constraints has been unclear. This report seeks to help focus the picture of where neutrino technology may find practical roles in nuclear energy and security. This report is the final product of the Nu Tools study, commissioned in 2019 by the DOE National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation Research and Development (DNN R\&D). The study was conducted over two years by a group of neutrino physicists and nuclear engineers. A central theme of the study and this report is that useful application of neutrinos will depend not only on advancing physics and technology but also on understanding the needs and constraints of potential end-users.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Validation testing of 239 Pu IAEA INDEN, LANL and LLNL 239 Pu files from May 2022 [Slides]
239 Pu files validated: ENDF/B-VIII.0, ENDF/B-VIII.0 except for 239 Pu from INDEN (p49) which performs the same in crit testing as (p53), ENDF/B-VIII.0 except for 239 Pu from INDEN (p51), ENDF/B-VIII.0 except for 239 Pu from LANL (4/21), ENDF/B-VIII.0 except for 239 Pu from LLNL (5/16); tweaked nu-bar, ENDF/B-VIII.0 except for 239 Pu from LLNL (5/1)
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
(U) SENSMG: First-Order Sensitivities of Neutron Reaction Rates, Reaction-Rate Ratios, Leakage, k eff , α , and Subcritical Multiplication Using PARTISN
SENSMG is a tool for computing first-order sensitivities of neutron reaction rates, reaction-rate ratios, leakage, k eff , α, and subcritical multiplication using the PARTISN multigroup discrete-ordinates code. SENSMG computes sensitivities to all of the transport cross sections and data (total, fission, a nu, chi, and all scattering moments), two edit cross sections (absorption and capture), and the density for every nuclide and energy group. It also computes sensitivities to the mass density for every material and derivatives with respect to all interface locations and outer boundaries. It computes sensitivities to user specified reactions whose cross sections are available in a user-supplied NJOY output file. The tool can be used for one-dimensional spherical and slab (r) and two-dimensional cylindrical (r-z) geometries. The tool can be used for fixed-source and eigenvalue problems. For most responses, the tool implements Generalized Perturbation Theory (GPT) as discussed by Williams and Stacey. The tool is thus limited to computing sensitivities only for GPT-allowable responses. For subcritical multiplication, the tool implements sensitivities derived by O’Brien and Clark. SENSMG has a similar role as the old SWANLAKE (Ref. 8), FORSS (Ref. 9), and SENSIT (Ref. 10) codes. It has capabilities similar to those of SUSD3D (Refs. 11 and 12), which also uses PARTISN. Section II of this report describes the theory behind adjoint-based sensitivities, gives the equations that SENSMG solves, and defines the sensitivities that are output. Section III describes the user interface, including the input file and command line options. Section IV describes the output. Section V gives some notes about the coding that may be of interest. Section VI presents some sample problems and discusses verification, which is ongoing. Section VII lists needs and ideas for future work. Appendix A lists most of the input files whose results are presented in Sec. VI. Appendix B provides some useful details on one of the cross-section libraries that SENSMG supports.
72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
Measurements for Flattop-HEU Benchmark Reevaluation
In June 2022, high-fidelity measurements of the Flattop critical assembly were taken at the National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site by a team from Los Alamos National Laboratory, Figure 1. Flattop-HEU is composed of a sphere of highly enriched uranium (HEU) surrounded by a thick spherical natural uranium (NU) reflector as shown in Figure 2 and Figure 3. These measurements were taken as part of the reevaluation of the Flattop-HEU benchmark evaluation for the International Criticality Safety Benchmark Evaluation Program (ICSBEP) Handbook. This reevaluation is being completed to update the benchmark to modern standards with significantly improved fidelity and uncertainty analysis. [1] The measurements address the largest identified uncertainties determined during a preliminary reevaluation in 2015. [2]
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Evaluated 238 U(n,f) Average Prompt Fission Neutron Multiplicities Including the CGMF Model
This report documents an evaluation of the average prompt fission neutron multiplicity, $\overline{v}_p$, of 238 U from 800 keV to MeV. This evaluation had to be re-done from “scratch” as the input to previous $\overline{v}_p$ evaluations, specifically ENDF/B-VIII.0, was not found. That means that all available experimental data were re-analyzed and uncertainties were re-estimated. The new evaluated 238 U $\overline{v}_p$ based on only experimental data differs distinctly from ENDF/B-VIII.0 $\overline{v}_p$ from 2 to 4.5 MeV, and from 6 to 7 MeV, and is otherwise similar. The difference from 2 to 4.5 MeV stems from the fact that ENDF/B-VIII.0 was tweaked in this energy range to data of Frehaut, while two other, equally trustworthy, data sets would indicate an evaluated 238 U $\overline{v}_p$ that is up to 2% higher. Also, second chance fission in ENDF/B-VIII.0 was smoothed over from 6–7 MeV. Another major difference to ENDF/B-VIII.0 is that one of the evaluations presented here includes model information from the Hauser-Feshbach fission fragment decay code CGMF, while ENDF/B-VIII.0 is based purely on experimental data. CGMF links several fission quantities with each other; $\overline{v}_p$ is predicted by assumptions made on, e.g., pre-neutron emission yields as a function of mass, the total kinetic energy, or spin and parity of fission fragments. This allows to validate the new 238 U $\overline{v}_p$ by using CGMF parameters obtained from fitting to experimental 238 U $\overline{v}_p$ to predict yields as a function of mass, the average total kinetic energy, or the mean energy of the prompt fission neutron spectrum. These model-predicted values can then be compared to experimental and evaluated data. The model-predicted fission-observable values using evaluated parameters obtained here are reasonably close to experimental data for some observables, but are farther away from experimental data related to TKE observables. In addition to that, the evaluated 238 U(n,f) $\overline{v}_p$ shows similar deviations from ENDF/B-VIII.0 as for the evaluation with only experimental data. This difference is expected to lead to changes in simulated effective neutron multiplication factor, $k_{eff}$ of ICSBEP critical assemblies that are sensitive to 238 U in the fast range (BigTen, Flattop, Flattop-Pu). These changes in $k_{eff}$ need to be counter-balanced. Chi-Nu PFNS experimental data are expected to be released in the next few months that might lead to the needed changes in the PFNS. Until then, we hold off in benchmarking the new 238 U(n,f) $\overline{v}_p$ as well as submitting it to ENDF/B-VIII.1. Also, new high-precision 238 U $\overline{v}_p$ are expected to be measured by the CEA in the next two years that will shed further light on question on 238 U $\overline{v}_p$ from 2–4.5 and 6–7 MeV.
238U↗
Flowsheet Design for the Purification of Mo-99 and Uranium Target Recovery by Solvent Extraction
New production and purification methods are sought to meet the increasing demand of 99 Mo but without the use of highly enriched uranium (HEU). To this end, larger quantities of low-enriched uranium (LEU) or even natural uranium (NU) are being leveraged to maintain high production capacities of 99 Mo. Solvent extraction is widely used to treat bulk quantities of uranium and to separate out trace quantities of fission products. This report describes the calculation of AMUSE-generated flowsheets to process irradiated uranium targets to recover 99 Mo by solvent extraction. Flowsheets are presented and discussed along with a number of advantages and disadvantages.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Enabling Efficient Water Splitting with Advanced Materials Designed for High pH Membrane Interface
This project was focused on developing the durable, high-performance materials and interfaces for advanced water splitting, enabling a clear pathway for achieving <$2/Kg H2 (on scale) with efficiency of 43 kWh/kg H 2 via anion exchange membrane (AEM)-based electrolysis. We aimed to advance these final goals via an improved fundamental understanding of both hydrogen and oxygen evolution reactions (HER/OER) leading to novel platinum group metal (PGM)-free catalyst materials in conjunction with critical improvements in membrane and ionomers and gas evolution electrodes with corresponding characterization and testing. Northeastern University (NU) lead this effort focusing on catalyst development and characterization (both in situ and ex situ) while project partners lead improvements in ionomer and membrane materials and will aid in the development of specialized electrode and membrane electrode assemblies. In addition, close collaboration occured with the HydroGEN Energy Materials Network (EMN) National Laboratory consortium including efforts related to use of advanced ionomers, durability protocols and validation of electrolyzer materials (e.g. NREL), multiscale modeling and computation (e.g. LBNL), and molecular dynamics (MD) simulations of the membrane catalyst interface (e.g. SNL). The interactions with HydroGEN included exchange of data and materials as needed to facilitate project success.
08 HYDROGEN↗
ElectroCat: Platinum-Group-Metal-Free Engineered Framework Nano-Structure Catalyst (Final Technical Report)
Traditional synthesis methodologies that produce durable, highly active PGM-free electrocatalysts typically consist of a top-down approach utilizing a nitrogen-coordinating precursor, a metal precursor, and a support material. Nitrogen-coordinated transition metal complexes typically are absorbed onto a support material, which is then activated via high-temperature pyrolysis. Selection of the support and the nitrogen coordinating precursor has long been the focal point of efforts to maximize the number of accessible, catalytically active sites. In this project, GWE in collaboration with SRNL and NU will develop durable, highly active, low-cost, PGM-free electrocatalysts for polymer electrolyte membrane fuel cells designed and synthesized using a unique, bottom-up, rationally designed approach. Electrocatalysts are produced via polymerization reaction followed by a high-temperature pyrolysis step. Additionally, this project focuses on catalyst site and ORR mechanistic modeling to better understand the structure-property relationship.
08 HYDROGEN↗
Validating ENDF/B-VIII.1beta1 with LLNL Pulsed-Sphere Neutron-Leakage Spectra [Slides]
In this presentation 6 Li Distinct changes are observed. 235 U Discussion shows Improvements (seen in VIII.1beta0) are coming from new PFNS informed by Chi-Nu experimental data. Additionally, 239 Pu large changes coming from inelastic scattering. 9 Be ENDF/B-VIII.0 shows to be better right after elastic peak while 16 O had slight changes. 19 F had small improvements in ENDF/B-VIII.1 and 56 Fe had noticeable changes in thicker spheres. Examples of negligible changes for the sake of completeness are given.
16-Oxygen↗
Plasma-Assisted Catalytic Conversion of CO 2 and Propane to Propylene and CO
Ethylene and propylene are critical pillars of the petrochemical and plastics industry. The current industrial route for producing these olefins, which is via steam cracking process, is extremely endothermic and highly CO 2 -intensive. In this work, Susteon, in partnership with the North Carolina State University (NCSU), New Castle University (NU), and SoCalGas, has investigate catalytic materials and process designs to produce propylene from propane by utilizing CO 2 as a soft oxidant in the presence of low temperature dielectric barrier discharge plasma. This route integrates the effect of catalysis and the presence of a plasma environment to intensify the production of the important three carbon olefin from propane at a lower temperature and near atmospheric pressure. This CO 2 oxidative dehydrogenation (CO 2 -ODH) process coproduces CO, which is a valuable by-product, critical for the petrochemical sector. Through this work, Susteon and the team have made significant progress in developing catalysts and gaining insights on the plasma-assisted CO 2 -ODH process.
01 COAL, LIGNITE, AND PEAT↗
The Monolithic Heat Pipe Microreactor Reference Plant Model
This work introduces a reference plant model for a generic monolithic heat-pipe-cooled microreactor. The model will serve as a springboard to develop future evaluation models in the licensing process of similar microreactor designs at the U.S. Nuclear Regulatory Commission. This model has been developed with the Comprehensive Reactor Analysis Bundle and its specifications are based on open literature publications for the eVinci TM design. BlueCRAB is the U.S. Nu- clear Regulatory Commission non-light-water reactor analysis system based on the Multiphysics Object-Oriented Simulation Environment framework, which can couple the Griffin, BISON, and Sockeye applications to resolve the various physics that are essential for the safety analysis of this type of reactor system. The core specifications includes tristructural isotropic fuel, graphite monolith, graphite reflectors, and drums composed of graphite and B 4 C. No moderator or burnable poison pins are used in the design. The fuel enrichment is reduced to control excess reactivity in the core. This core design is not optimized and only serves for testing purposes, since the primary objective of this work is to exercise the multiphysics coupling for this type of reactor system. A three dimensional (3D) core heterogeneous Griffin discrete ordinates (SN) transport model allows the precise calculation of the flux distribution and pin powers. Griffin transfers the power density distribution and obtains a temperature distribution to and from BISON. The BISON model com- putes the 3D core temperature distribution and is coupled to 876 Sockeye subapplications running a heat pipe model. This 3D conduction model is coupled to the various heat pipes via heat flux boundary conditions. The model includes a small gap between the heat pipe and the monolith. Convective heat transfer boundaries with either ambient temperature or condenser temperature as heat sinks are imposed at the model boundaries. The 2D Sockeye heat pipe model uses a vapor- only methodology, which provides the needed resolution for transient calculations and allows the determination of various heat pipe limits. This approach is superior to the superconductor model traditionally used in steady-state calculations. BlueCRAB computes steady-state power and temperature distributions that serve as the initial condition for a loss-of-heat-sink transient simulation. The steady-state results show significant peaking due to the position of the control drum, but this is a characteristic of the particular design used, which is not optimized at this stage. The transient results show the reactor power slowly stabilizing towards a 3% power level after the partial loss of secondary heat removal. Several recriticalities are observed due to cooling through the secondary system but the reactor is self-stabilizing and behaves as expected.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Measurement of the Prompt Fission Neutron Spectrum from 800 keV to 10 MeV for 240 Pu($sf$) and for the 240 Pu($n,f$) Reaction Induced by Neutrons of Energy from 1-20 MeV
The presence of 240 Pu in nuclear fuels for reactors has resulted in high uncertainties in the results of reactor and nuclear transmutation calculations because of deficiencies in 240 Pu-related nuclear data. Specifically for the prompt fission neutron spectrum (PFNS) of 240 Pu, there is only one neutron-induced, ($n,f$), measurement at 0.85 MeV incident neutron energy and only one complete spontaneous fission, ($sf$), measurement. This limited availability of data does not sufficiently guide nuclear data evaluations of these quantities. Here we report on a measurement of both the 240 Pu($sf$) and the 240 Pu($n,f$) PFNS, both over the emitted neutron energy range of 0.79–10.0 MeV, and from incident neutron energies of 1.0–20.0 MeV for the ($n,f$) reaction. Measurements were made with a hemispherical array of liquid scintillators at the high-energy Los Alamos Neutron Science Center white neutron source at the Weapons Neutron Research facility as part of the joint LANL-LLNL Chi-Nu experimental campaign to measure actinide fission neutron spectra. These measurements are the first of their kind, and provide clear experimental evidence for second-chance fission, third-chance fission, and pre-equilibrium neutron emission processes in neutron-induced fission of 240 Pu, and are the first ever measurements above 1 MeV incident neutron energy.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Formatting and V&V of Consistent 238,240−24 2Pu $\overline{v}_p$ Evaluated Mean Values and Covariances
This report is in answer to the Nuclear Criticality Safety Program FY24 quarter 4 milestone that requires: “Format and V&V nu-bar means and covariances” for 238,240-242 Pu average prompt fission neutron multiplicities, $\overline{ν}$ p , that were obtained by a consistent evaluation leveraging the fission-event generator CGMF and a detailed uncertainty quantification of experimental data. It is described how nuclear data mean values and covariances were formatted using ENDFtk. Implementing the new 238,240-242 Pu $\overline{ν}$ p into the ENDF/B-VIII.1β 4 library leads to only small overall changes in criticality values of the Jezebel, Dirty Jezebel, Jupiter-001, Jupiter-002, EUCLID 3x2 and EUCLID 8x1 critical assemblies. Simulated k eff uncertainties due to $\overline{ν}$ p covariances change only little if cross-isotope covariances are considered or not for those assemblies with low percentage content of minor Pu isotopes. However, for the Dirty Jezebel critical assembly, that has a sizeable 240 Pu and non-negligible 241 Pu content, the simulated k eff uncertainties due to considering or neglecting cross-isotope $\overline{ν}$ p covariances is 443 versus 374 pcm.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
ICARUS at the Short-Baseline Neutrino program: first results
The ICARUS collaboration employed the 760-ton T600 detector in a successful three-year physics run at the underground LNGS laboratory, performing a sensitive search for LSND-like anomalous $\nu_e$ appearance in the CERN Neutrino to Gran Sasso beam, which contributed to the constraints on the allowed neutrino oscillation parameters to a narrow region around 1 $eV^2$. After a significant overhaul at CERN, the T600 detector has been installed at Fermilab. In 2020 the cryogenic commissioning began with detector cool down, liquid argon filling and recirculation. ICARUS then started its operation collecting the first neutrino events from the Booster Neutrino Beam (BNB) and the Neutrinos at the Main Injector (NuMI) beam off-axis, which were used to test the ICARUS event selection, reconstruction and analysis algorithms. ICARUS successfully completed its commissioning phase in June 2022, moving then to data taking for neutrino oscillation physics, aiming at first to either confirm or refute the claim by Neutrino-4 short-baseline reactor experiment. ICARUS will also perform measurement of neutrino cross sections in LAr with the NuMI beam and several Beyond Standard Model studies. After the first year of operations, ICARUS will search for evidence of sterile neutrinos jointly with the Short-Baseline Near Detector, within the Short-Baseline Neutrino program. In this presentation, preliminary results from the ICARUS data with the BNB and NuMI beams are presented both in terms of performance of all ICARUS subsystems and of capability to select and reconstruct neutrino events.
Artero Pons, Maria↗
Multi-Differential Semi-Inclusive Charged-Current Muon Neutrino Cross Sections on Helium-4 in MINERvA and the Helium-to-CH Target Cross Section Ratio
A comprehensive international effort has been underway to elucidate the properties and behaviors of neutrinos. A major source of systematic uncertainties in studying neutrino-induced interactions comes from neutrino-nucleus cross-section models, highlighting the need for more precise statistical measurements. MINERvA, an on-axis neutrino-nucleus scattering experiment located at the Fermi National Accelerator Laboratory, was established to produce neutrino cross-section measurements with many different nuclei. The helium target is the lightest nucleus to be measured by MINERvA. Using the NuMI medium energy muon neutrino data set, we present preliminary results and summarize the extraction of the charged current (CC) muon neutrino - helium-4 semi-inclusive ($\nu_{\mu}$+$^{4}\textrm{He}$$\rightarrow$$\mu^{-}$+N$\pi$+M$\textit{p}$) multi-differential cross-section extraction as a function of transverse ($P_{T}$) and longitudinal ($P_{L}$) muon momentum with respect to the neutrino beamline. We define the final state topology as CC-N$\pi$M$\textit{p}$, with at least two reconstructed tracks: a muon and a combination of N protons and M pions, where $N + M > 0$. To probe the dependence on the size of the nucleus in neutrino-induced interaction phenomena we present a differential cross-section ratio of helium-4 to MINERvA's hydrocarbon target (\textrm{CH}) as a function of transverse muon momentum.
Nguyen, Christian↗
Programmable Catalyst Structures via Adsorbate-Induced Adatom Assembly
The electronic structure and geometric configuration of oxide-supported metal ions are important coordination properties that can be related to catalytic activity and stability. Herein, we interrogate the coordination environment of mononuclear Pd ions supported on ceria using CO adsorption, infrared vibrational spectroscopy, and DFT modeling. We observed the 15 h continuous co-evolution of a palladium- (2167 cm -1 ) and cerium-carbonyl (2177 cm -1 ) complex by monitoring the $\nu$(CO) infrared region. The slow CO adsorption kinetics were caused by the reactive ligand exchange between an oxygen atom of the support and the CO adsorbate to yield an oxygen vacancy and adsorbed CO 2 . We hypothesize that the co-evolved cerium-carbonyl complex was formed upon CO adsorption at or adjacent to this oxygen vacancy. Our hypothesis was experimentally supported by a dramatic attenuation of the cerium carbonyl signal upon pre-adsorption of water through an apparent competitive adsorption mechanism. The attenuation was also accompanied by a 6 cm -1 redshift of the palladium carbonyl band (2161 cm -1 ) attributed to hydrogen bonding between the carbonyl and a nearby hydroxyl. Characteristic n(CO) stretch frequencies catalogued through CO adsorption onto single crystal ceria by Wöll et al.1 led us to index the cerium carbonyl to the {100} nanofacet of the polycrystalline ceria support. It follows from the observed co-evolution of the two carbonyl complexes that Pd was also adsorbed at the {100} nanofacet. Redeployment of a previously developed DFT model by Ivanova-Shor et al.2 featuring square-planar coordination of Pd2+ at the {100} nanofacet (O 4 Pd) of a Ce 21 O 42 nanoparticle model qualitatively reproduced several experimental observations.
36 MATERIALS SCIENCE↗
A Contextually-Aware Sensitivity Analysis to Guide the Design of Randomized Least Squares Solvers in Applications
Our work on the DOE-sponsored project “A Contextually-Aware Sensitivity Analysis to Guide the Design of Randomized Least Squares Solvers in Applications,” was an effort to address critical challenges in nu merical computing and its applications to optimization. The increasing demand for robust and scalable solutions to large-scale linear algebra problems has highlighted the limitations of traditional approaches, particularly in heterogeneous and extreme-scale computing environments. Randomized Numerical Linear Algebra (RandNLA) offers a promising framework to address these challenges, and this proposal builds on this foundation by introducing innovations in sensitivity analysis and computational adaptability.
97 MATHEMATICS AND COMPUTING↗