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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Configuration–Interaction Perturbation Theory Calculations of Pu II

Configuration–interaction perturbation theory (CI–PT) is applied to calculations of low-energy states of Pu II. This ion is quite challenging due to a large number of possible determinants arising from seven valence electrons and strong relativistic effects. The CI–PT calculations agree with experiments for the energies and g-factors for many low-energy states that allowed positive identification of the theoretical levels. Isotope shifts were also used to aid in identification, and, in case of the odd states, fitting with three independent parameters was used to match theoretical isotope shifts to the experimental values with good accuracy. The CI–PT approach tested here on the Pu II ion can be generally used to calculate properties of many complex atoms, including U I that can find application in fundamental and applied science.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2025

Evaluate opportunities for monitoring the oxidation states of technetium in both organic and aqueous phases using Raman spectroscopy

An important isotope in the nuclear fuel cycle, 99 Tc is characterized by its long half-life of 2.1 × 10 5 years and its beta emission. It is generated through the fission of 235 U. As the burnup levels of nuclear waste continue to rise, the concentration of technetium in spent fuel increases, leading to a heightened interest in its quantification during the reprocessing of spent nuclear fuel. In HNO3 environments, technetium predominantly exists as TcO 4 − ; however, it tends to interfere with reducing agents, decreasing the separation efficiency for actinides. Therefore, accurate quantitative determination of technetium is critical to controlling its distribution and mitigating its negative effect on the reprocessing of spent fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Assessing high fidelity multi-component models to facilitate safeguards at Gas Centrifuge Enrichment Plants

We report that the International Atomic Energy Agency (IAEA) inspectors routinely carry out environmental sampling (ES) as a verification method. Collection of environmental swipe samples at various locations in Gas Centrifuge Enrichment Facilities (GCEPs) is an important process in detecting misuse of a declared facility and possibly the existence of undeclared nuclear material. These samples are measured for isotopic composition in uranium containing particles by Thermal Ionization Mass Spectrometry (TIMS) or Inductively Coupled Plasma Mass Spectrometry (ICP-MS). Even though ES is highly effective in detecting the absolute value of enrichments and their deviations from the declared values, it cannot explain the cause of those changes. Several potential explanations can serve as possibilities for particles detected above or other than the declared enrichment. These include normal and non-malicious events such as the design of the enrichment cascades, unintentional failures of the machines, or overshoot during startup of a cascade. It can also be the result of deliberate misuse by the facility operators. The primary objective of this work is to understand how these factors affect the enrichments produced by a cascade and quantify anticipated multi-isotopic concentrations for each case. The following methodology is employed to determine signatures at a particular facility. 1) Utilize a new two-dimensional multi-component diffusion code to obtain centrifuge performance data and use that information to design and perform cascade analysis. Compare and contrast the results with previous 1-D radially averaged solutions from the Pancake code. 2) Design a GCEP cascade with the production goal of 19.75% 235 U for each set of machine data above. Investigate two cascade scenarios that include enrichment of natural uranium (NU) feed to 19.75% 235U in a single cascade compared to a two-step process of NU to 5% and then 5% to 19.75%. 3) Simulate the intentional vs. unintentional off-normal scenarios in the cascades to assess the differences in isotopic concentrations. A non-ideal squared-off cascade model developed at the University of Virginia is used to calculate flow rates and isotopic concentrations of the process gas. The analysis is performed using the Rome machine model operated at 600 m/s rotor speed. The upper and lower bounds of normal and abnormal enrichments in a typical facility are used in conjunction with ES results to understand the root causes of such observations.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

Extended Enrichment Accident Tolerant LWR Fuel Isotopic and Lattice Parameter Trends

Commercial light water reactor (LWR) operators and fuel vendors in the United States (US) are pursuing changes to the reactor fuel that include increased enrichment and accident-tolerant fuel (ATF) designs. Enrichments under consideration are between 5 wt % and 10 wt % 235 U, which are a subset of high-assay low-enriched uranium fuels. ATF features are designed to improve fuel system performance under accident conditions. With increased enrichment, fuel cycle economics can be improved if fuel can be licensed for higher burnup (HBU) than typical current limits (e.g., 62 GWd/MTU maximum fuel pin).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Why Lady Godiva Should be Replaced as the Default Validation Experiment for U-235 Nuclear Data

For the past seven decades the Lady Godiva benchmark (HEU-MET-FAST-001) has been the primary experiment used for 235 U nuclear data validation. The papers associated with large nuclear data library releases such as ENDF/B-VII.1 and ENDF/B-VIII.0 refer to it frequently. The reasons why HEU-MET-FAST-001 has been used as the primary validation experiment for 235 U nuclear data validation will be discussed in the following section. However, as discussed among the benchmark community, the standards associated with International Criticality Safety Benchmark Evaluation Project (ICSBEP) evaluations have changed throughout time; this is subject of the OECD/NEA WPEC (Working Party on Nuclear Criticality Safety) Subgroup 8. HEU-MET-FAST-001 is an older benchmark (issued during the inaugural year of ICSBEP in 1996 with only minor revisions occurring since then) and (along with many of the other benchmarks from this era) does not meet the standards for a modern benchmark. This work explores why HEU-MET-FAST-001 is useful for 235 U nuclear data validation and discusses other alternative validation experiments.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Neutron resonance transmission analysis prototype system for thorium fuel cycle safeguards

Emerging thorium-based reactor designs and fuel cycles present challenges to traditional non-destructive assay techniques used in international safeguards. Specifically, assaying the masses of 233 U and 235 U when they are present together in samples with high gamma ray backgrounds is difficult because of similar passive neutron signatures and relatively weak gamma-ray emissions of 233 U. The Pacific Northwest National Laboratory (PNNL) and the Massachusetts Institute of Technology (MIT) are developing a compact neutron resonance transmission analysis (NRTA) system as one potential solution to these challenges. The NRTA technique provides isotopic information for a sample via neutron time-of-flight (TOF) measurements that exploit a sample’s epithermal neutron resonance cross-sections. A recently developed portable NRTA system uses a commercially available, pulsed deuterium-tritium neutron generator with a ~2 m flight path and a GS20 lithium glass scintillator detector. Finally, this paper describes the prototype NRTA system design, a refined radiation transport model of the system, preliminary measurements with thorium and uranium sources, and demonstration of a quantitative isotopic estimation algorithm.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Test Results for Checking Reliability of Selected ENDF/B-VIII.0 Evaluated Uncertainties

This report summarizes error messages obtained from counter-checking whether 1 H, 9 Be, 12 C, 27 Al, 235,238 U and 238–242 Pu relative uncertainties of ENDF/B-VIII.0 are realistic. Large parts of this report were generated automatically by the code CovVal. This code allows to test the reliability of relative uncertainties of any isotope. It gives generic error and warning messages if evaluated relative uncertainties of a particular reaction are likely under- or over-estimated. In addition to these error and warning messages, I added comments whether the flagged uncertainties are truly under- or overestimated. I call out these uncertainties that should be further investigated and corrected by an evaluator for an upcoming release of ENDF/B-VIII.1.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

HIMU geochemical signature originating from the transition zone

Plume volcanism may sample mantle sources deeper than mid-ocean ridge and arc volcanism. Ocean island basalts (OIBs) are commonly related to plume volcanism, and their diverse isotopic and elemental compositions can be described using a limited number of mantle endmembers. However, the origins and depths of these mantle endmembers are highly debated. Here we show that the HIMU (high μ, μ = 238 U/ 204 Pb) endmember may reside in the transition zone. Specifically, we report the geochemical signature of a high-pressure multiphase diamond inclusion, entrapped at 420–440 km depth and 1450 ±50 K, which matches exactly the geochemical patterns of the HIMU-rich OIBs. Since the HIMU component is variably sampled by almost all OIBs, our finding implies that the transition zone causes a major overprint of the geochemical features of mantle plumes. Furthermore, some mantle plumes, like those feeding Bermuda, St Helena, Tubuai and Mangaia, appear to be dominated by this source. Furthermore, our finding highlights the importance of the transition zone in highly incompatible element budget of the mantle.

58 GEOSCIENCES↗
DOE OSTI2023

Molybdenum-99 from Molten Salt Reactor as a Source of Technetium-99m for Nuclear Medicine: Past, Current, and Future of Molybdenum-99

Technitium-99m ( 99m Tc), a widely used radioisotope, is used in tens of millions of medical diagnostic procedures annually. However, it is hard to store and must be immediately used upon production due to its short half-life (i.e., 6 h); thus, it is currently produced from 99 Mo, which itself is a result of 235 U fission. The majority of 99 Mo supplies to U.S. patients are currently provided by foreign producers and produced using highly enriched uranium (HEU). In order to minimize the proliferation risks of HEU-based medical isotope production, the U.S. Department of Energy’s National Nuclear Security Administration has funded a program to accelerate the development of technologies to produce 99 Mo without the use of HEU. Today, the global supply of 99 Mo depends on a limited number of nuclear reactors, and production has been interrupted unexpectedly since 2009 due to the fleet’s advanced age. Herein, alternative options for 99 Mo production are discussed, and one potential option is to obtain 99m Tc from molten salt reactors (MSRs). A MSR is a nuclear fission reactor that can operate at or close to atmospheric pressure with liquid fuel, which allows for producing isotopes in a timely manner. In this paper, the past and current production of 99 Mo via nuclear reactors is described, and the future of 99 Mo production by MSRs is discussed. The behavior and chemical properties of molybdenum in fluoride salts in MSRs and the possible extraction methods are also examined in addition to the limitation of current studies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Development of a Helical Closure for Radioactive Material Shipping Packages

Radioactive Material (RAM) Packagings Used to transport radioactive material • Weapons components • Medical isotopes • Spent nuclear fuel • Etc. • Packages must adhere to 10 CFR 71 which drive to NUREGs, ASME codes, etc. • Transport Index (A2 values) determine if a package is Type A, Type AF, Type B, Type B(U)F

Housley, William M. [Savannah River National Labor↗
DOE OSTI2025

Investigation of a U(IV)/U(III) Thermodynamic Reference Electrode for High-Temperature Molten Fluoride Salts

While thermodynamic reference electrodes with known and stable potentials are common in traditional aqueous systems, the high temperature and corrosive environment of a molten fluoride salt makes achieving long term stability with a thermodynamic reference electrode challenging, especially at temperatures of 600°C or higher. In this work, a thermodynamic reference electrode consisting of U(IV)/U(III) in a boron nitride compartment was evaluated for use in FLiBe at temperatures ≥ 600°C. FLiBe used in the study was purified by AlphaTech's proprietary process and characterized by ICP-MS and square wave voltammetry. The free oxide concentration was <2 ppm. Using the purified FLiBe, the U(IV)/U(III) thermodynamic reference electrode was shown to provide a stable, well-defined, and reproducible potential for more than 600+ hours of use in different tests. Moreover, the thermodynamic reference electrode showed a consistent potential with no signs of failure, even after being cooled between tests and then reheated for reuse. Thus, the U(IV)/U(III) reference electrode is suitable for use in rigorous electrochemical studies in molten fluoride salts. As a result, it may be useful as a common standard, facilitating the advancement of nuclear applications such as isotope separation or online monitoring of reactor systems through improved certainty in the measurement of thermodynamic potentials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Bulk density, grain size, carbon content and isotopes, and nitrogen content of floodplain sediment along the Koyukuk River, Alaska collected in Summer 2018.

The dataset is comprised of grain size, bulk density, carbon and nitrogen content, stable carbon isotopes, and radiocarbon content of soils sampled on the floodplain of the Koyukuk River near Huslia, AK. This dataset was collected to characterize sediment properties and floodplain organic carbon stocks in the discontinuous permafrost floodplain of the Koyukuk River. The data help quantify the flux and storage of sediment and carbon between the Koyukuk River and the surrounding floodplain as the river migrates laterally. The samples were collected in June 27 - July 9, 2018. The soil samples come from a total of 24 locations and were sampled in the upper 490 cm of the river cutbanks, point bars, and floodplain. Samples were analyzed for bulk density, full particle size distribution, nitrogen content and carbon content and isotopes, though not all samples were included in each analysis. An additional 10 samples of floodplain woody debris were measured for radiocarbon content. Data, detailed methods, analyses, and interpretations were previously published at: Douglas, M. M., Lingappa, U. F., Lamb, M. P., Rowland, J. C., West, A. J., Li, G., et al. (2021). Impact of river channel lateral migration on microbial communities across a discontinuous permafrost floodplain. Applied and Environmental Microbiology, 87(20), AEM.01339-21. https://doi.org/10.1128/AEM.01339-21 Douglas, M. M., Li, G. K., Fischer, W. W., Rowland, J. C., Kemeny, P. C., West, A. J., et al. (2022). Organic carbon burial by river meandering partially offsets bank erosion carbon fluxes in a discontinuous permafrost floodplain. Earth Surface Dynamics, 10(3), 421–435. https://doi.org/10.5194/esurf-10-421-2022 The dataset is comprised of a total of three separate data files based on the sample type and analysis, including a file that lists the sample locations, permafrost presence, and geomorphic landform.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Machine learning approach to pattern recognition in nuclear dynamics from the ab initio symmetry-adapted no-core shell model

A novel machine learning approach is used to provide further insight into atomic nuclei and to detect orderly patterns amid a vast data of large-scale calculations. The method utilizes a neural network that is trained on ab initio results from the symmetry-adapted no-core shell model (SA-NCSM) for light nuclei. We show that the SA-NCSM, which expands ab initio applications up to medium-mass nuclei by using dominant symmetries of nuclear dynamics, can reach heavier nuclei when coupled with the machine learning approach. In particular, we find that a neural network trained on probability amplitudes for s- and p-shell nuclear wave functions not only predicts dominant configurations for heavier nuclei but in addition, when tested for the 20 Ne ground state, accurately reproduces the probability distribution. The non-negligible configurations predicted by the network provide an important input to the SA-NCSM for reducing ultralarge model spaces to manageable sizes that can be, in turn, utilized in SA-NCSM calculations to obtain accurate observables. The neural network is capable of describing nuclear deformation and is used to track the shape evolution along the 20-42 Mg isotopic chain, suggesting a shape coexistence that is more pronounced toward the very neutron-rich isotopes. We provide first descriptions of the structure and deformation of 24 Si and 40 Mg of interest to x-ray burst nucleosynthesis, and even of the extremely heavy nuclei such as 166,168 Er and 236 U, that build on first-principles considerations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ROADRUNNER MiniFuel Experiment: Irradiation Target Design and Sample Characterization

High-density uranium nitride (UN) is a fuel candidate for several advanced nuclear reactor designs currently under development. Because there are limited UN performance data relative to fuel fabrication impurity and density variation, an irradiation campaign has been developed as part of a collaborative effort among the University of Texas at San Antonio (UTSA), Westinghouse Electric Company, Oak Ridge National Laboratory (ORNL), and Los Alamos National Laboratory (LANL) under the Nuclear Science User Facilities program. This project, entitled ROADRUNNER, or Research On ADvancing the peRformance of UraNium Nitrides in Extreme enviRonments, aimsto support UN fuel qualification for advanced reactors by investigating the impact of density and impurity variations on UN performance as a function of irradiation temperature and burnup. The MiniFuel experiment vehicle developed by ORNL, which leverages the High Flux Isotope Reactor, was selected to perform this accelerated separate-effects irradiation testing. The experiment test matrix consists of six MiniFuel targets containing miniature UN fuel disks, and targets three distinct burnup levels (37.5, 60, and 75 MWd/kg U) and three distinct temperatures (600, 900, and 1200°C). Neutronics and thermal analyses were performed to determine the experimental parameters needed to meet the desired irradiation conditions and to predict the experiment components temperatures. UN pellets were fabricated at LANL with tightly controlled parameters to produce specimens with three distinct densities and three levels of carbon content. The pellets were then thinned down by UTSA to the experiment-required thickness. The pre-characterization of the specimens includes density measurements, carbon and oxygen contents, microstructure analysis, and x-ray computed tomography. The selected specimens will be assembled into the MiniFuel experiment, and the first ROADRUNNER MiniFuel targets are intended for HFIR insertion during the Fall of 2024. After irradiation, the targets will be shipped to ORNL’s hot cell facility for disassembly. The post-irradiation examination on the fuel specimens includes fission gas release measurements, visual inspection, fuel swelling measurements, gamma spectroscopy, and microstructure analysis. The data collected post-irradiation will be used to develop fuel performance models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

G EANT 4 atomic relaxation data for transfermium nuclei (Z = 101–104)

Advanced theoretical methods can accurately calculate various atomic observables and predict electronic structure. Still, systematic computations of the radiative and non-radiative transition probabilities and energies are missing for the actinides and all the transfermium elements. However, these compilations are needed for comprehensive Monte-Carlo simulations (such as GEANT4) of the radioactive decay of transfermium nuclei. These simulations can forma basis for data analysis of experiments, especially with complex detection setups. Investigation of the transfermium nuclei is crucial for understanding the nature of the nuclear force. In this study, simulations based on data from the Jena Atomic Calculator (JAC) and the data from the Evaluated Atomic Data Library (EADL) present in GEANT4 were found compatible for the three elements Ba(Z = 56), U(Z = 92), and Fm(Z = 100), thus, validating the JAC calculations. For Z> 100, we also found sound agreement between simulations that used data generated with JAC and experimental results involving No(Z = 102) and Rf(Z = 104) isotopes. In conclusion, these results demonstrate that JAC can produce reliable atomic data sets for transfermium elements, which will assist in analyzing nuclear-decay-spectroscopy experiments.

GEANT4↗
DOE OSTI2023

Consistent Nuclear Data Evaluations for Criticality Safety [Slides]

This presentation covers consistent nuclear data evaluations for criticality safety. Topics include the evaluation procedure, n+ 139 La evaluation, the work in progress for the n+ 233 U evaluation (emphasis on capture), and a concluding Summary.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗