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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Materials Data on Pu(Mo3Se4)2 by Materials Project

Pu(Mo3Se4)2 crystallizes in the trigonal R-3 space group. The structure is three-dimensional. Pu3+ is bonded in a body-centered cubic geometry to eight Se2- atoms. There are two shorter (2.88 Å) and six longer (3.17 Å) Pu–Se bond lengths. Mo+2.17+ is bonded to five Se2- atoms to form a mixture of edge and corner-sharing MoSe5 square pyramids. There are a spread of Mo–Se bond distances ranging from 2.54–2.77 Å. There are two inequivalent Se2- sites. In the first Se2- site, Se2- is bonded in a 1-coordinate geometry to one Pu3+ and three equivalent Mo+2.17+ atoms. In the second Se2- site, Se2- is bonded in a 5-coordinate geometry to one Pu3+ and four equivalent Mo+2.17+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Pu(MoO4)2 by Materials Project

Pu(MoO4)2 crystallizes in the orthorhombic Pbca space group. The structure is three-dimensional. Pu4+ is bonded in a 8-coordinate geometry to eight O2- atoms. There are a spread of Pu–O bond distances ranging from 2.28–2.47 Å. There are two inequivalent Mo6+ sites. In the first Mo6+ site, Mo6+ is bonded in a tetrahedral geometry to four O2- atoms. There are a spread of Mo–O bond distances ranging from 1.77–1.83 Å. In the second Mo6+ site, Mo6+ is bonded in a tetrahedral geometry to four O2- atoms. There are a spread of Mo–O bond distances ranging from 1.78–1.80 Å. There are eight inequivalent O2- sites. In the first O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu4+ and one Mo6+ atom. In the second O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu4+ and one Mo6+ atom. In the third O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu4+ and one Mo6+ atom. In the fourth O2- site, O2- is bonded in a 2-coordinate geometry to one Pu4+ and one Mo6+ atom. In the fifth O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu4+ and one Mo6+ atom. In the sixth O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu4+ and one Mo6+ atom. In the seventh O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu4+ and one Mo6+ atom. In the eighth O2- site, O2- is bonded in a distorted bent 150 degrees geometry to one Pu4+ and one Mo6+ atom.

36 MATERIALS SCIENCE↗
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 OSTI2024

Conceptual Spacer Design for the ATR GEN I Target for Pu-238 Production in the Advanced Test Reactor at Idaho National Laboratory

The initial target design used for Pu-238 production at Idaho National Laboratory was designed by Oak Ridge National Laboratory to optimize the production of Pu-238 in the High Flux Isotope Reactor (HFIR) and are referred to as HFIR GEN II targets. To take advantage of the Advanced Test Reactor’s (ATR) taller active core region a redesign of the HFIR GEN II targets was needed. It was proposed to stack two HFIR GEN II targets nose to nose about the core center line; however, this resulted in excessive neutron and photon heating in the pellets located in the center. This peak heating was not desirable so three alternative designs were investigated for the ATR GEN I targets. The python-based code, MCNP to ORIGEN2 in Python (MOPY), was used to calculate the heating rates after 40 days of irradiation to capture the effects of each configuration. The purpose of this paper is to document the details of these conceptual design calculations and comparisons for the ATR GEN I targets.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Conceptual Spacer Design for the ATR GEN I Target for Pu-238 Production in the Advanced Test Reactor at Idaho National Laboratory

The initial target design used for Pu-238 production at Idaho National Laboratory was designed by Oak Ridge National Laboratory to optimize the production of Pu-238 in the High Flux Isotope Reactor (HFIR) and are referred to as HFIR GEN II targets. To take advantage of the Advanced Test Reactor’s (ATR) taller active core region a redesign of the HFIR GEN II targets was needed. It was proposed to stack two HFIR GEN II targets nose to nose about the core center line; however, this resulted in excessive neutron and photon heating in the pellets located in the center. This peak heating was not desirable so three alternative designs were investigated for the ATR GEN I targets. The python-based code, MCNP to ORIGEN2 in Python (MOPY), was used to calculate the heating rates after 40 days of irradiation to capture the effects of each configuration. The purpose of this paper is to document the details of these conceptual design calculations and comparisons for the ATR GEN I targets.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

OVERVIEW OF RECENT PU-238 PRODUCTION ACTIVITIES AT IDAHO NATIONAL LABORATORY

The Plutonium Fuel Services (PFS) program at Idaho National Laboratory (INL) is active in the qualification of irradiation targets containing Np-237 for irradiation in the Advanced Test Reactor (ATR) to produce Pu-238 for future NASA missions. INL qualified and loaded 7 targets in ATR’s South Flux Trap (SFT) for cycle 169A, which occurred in Spring 2021. This program was reinitiated after two baseline production targets in three positions validated significant production of Pu-238 [ref.1]. The validation model was followed by the PFS-1 experimental test in the ATRC (Critical) facility [ref.2]. This paper outlines the progress and status of the PFS program. The qualification effort, safety analysis, hardware status, and future activities for qualification of an updated target design for use in the ATR will be discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Pu-238 Production Calculator and Estimation Tool

A MS Excel tool was created to help track Pu-238 production, Np-237 feedstock use, estimate production scenarios, and coordinate planning for shipments. As-run estimates were used as a basis for beginning and end of life isotopic concentrations. Multiple target types were created to demonstrate a proof of concept while actual results were not yet available The initial goals of this production tool were: • Track unirradiated, currently irradiated, and irradiated targets at the Advanced Test Reactor (ATR) • Estimate production yields from irradiated targets in various positions for each cycle • Provide past, current, and future year quantities of Np-237 and Pu-238 at ATR, amounts shipped to ATR and amounts shipped from ATR • Estimate production from current and future target designs based on data from numerical modeling (values will be provided by Idaho National Laboratory (INL)) • After proof of principle, additional work is planned that will further enhance the tool with: o Incorporate the High Flux Isotope Reactor (HFIR) reactor at Oak Ridge National Laboratory (ORNL) for irradiation o Incorporate irradiations and target shipments from ORNL to INL and back to track Np-237 transfers across the Department of Energy (DOE) Complex

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

A highly-enriched 244 Pu reference material for nuclear safeguards and nuclear forensics measurements

A highly-enriched 244 Pu isotope dilution reference material has been prepared and characterized for metrologically traceable measurements of very small quantities of plutonium. The amount of plutonium in samples associated with nuclear safeguards and nuclear forensic measurements can be significantly less than 1 ng. Accordingly, the ability to quantify the amount and isotopic composition of plutonium from a single mass-spectrometric analysis is particularly desirable. The highly-enriched 244 Pu reference material, described here, will minimize the magnitude of spike corrections necessary to obtain accurate information on plutonium isotopic composition from isotope dilution measurements.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Lattice response to ambient self-irradiation of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures

Self-irradiation in plutonium is a continuous process that introduces radiation damage ranging from point defects to nanoscale radiation damage as well as decay daughter product impurities that require to be accommodated by the host material. These changes are affecting the structure and properties of the material by lattice swelling, mechanical strengthening, and other alterations. X-ray diffraction and dilatometry studies have shown that the room temperature self-irradiation of δ-phase Pu-Ga alloys (2 and 7 at%Ga) results in swelling of the lattice that saturates after 0.1–0.2 displacement per atom (dpa) of accumulated lattice damage. Here, for a better understanding of these self- irradiation processes, defect accumulation and damage evolution, this time-of-flight neutron diffraction study investigates the impact of ambient self-irradiation effects on the structural evolution of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures (<120 K), associated with a varied history of storage time and thermal treatments. Two key behaviors were observed in the lattice evolution after exposure to cryogenic temperatures with respect to room-temperature accumulated self-irradiation damage. Samples with no self- irradiation damage accumulation under ambient conditions showed swelling of the lattice at the end of the measurement, while samples that accumulated significant self-irradiation damage under ambient conditions showed only a partial recovery of the lattice parameter value upon completion of the measurements (cooling followed by heating back to room temperature). Possible causes of the behavior are proposed, and their associated effects are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiation performance of nonfertile (Pu-MA-Zr) fast reactor metal fuels

This work was part of a program begun in 2001 to develop advanced nuclear fuels, originally as carriers for plutonium and minor actinides (neptunium, curium, and americium) taken from spent commercial light-water reactors (LWR) so that the plutonium and minor actinides could be ‘burned’ or transmuted in an accelerator or a fast nuclear reactor. A central part of these experiment programs has been the development of advanced fast reactor fuels, because a fast reactor was considered the most efficient vehicle to transmute the actinide waste products, and metallic fuels is a central focus of these tests. An experiment design was developed in which a thermal test reactor, the Advanced Test Reactor (ATR), was used to test small fuel pin prototypes, by creating areas in the core shielded by cadmium filters to produce a largely epithermal and fast neutron spectrum environment in which the pins could be irradiated. The results of non-fertile metallic fuel (no uranium) tests are presented here. Pu-Am-Np-Zr fuels were irradiated to fission densities up to 33 × 10 20 fission/cm 3 and Pu-239 depletions of up to 39%. The depletions were created by roughly 2/3 by fission and 1/3 by transmutation neutron capture. Up to five fuel ‘rodlets’ were irradiated in three sealed capsules stacked axially in the core, and the peak cladding temperatures ranged from 300°C to 500°C, depending on axial location as those near the core centerline are operating hotter and to higher fission densities. Several post-irradiation examinations (precision gamma scanning and fission gas release) were similar to other historical metal fuel experiments in fast reactors. However, optical metallography indicated that two of the rodlets had breached. The exact reasons are unclear. Due to the design of this irradiation experiment a rodlet breach could have increased the temperature in others in the same capsule by contaminating the thermal gap helium with heavier and less conductive fission product gases. Finally, some of those rodlets showed high amounts of fuel/cladding chemical interaction (FCCI).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Symmetry-correct bonding in density functional theory calculations for delta phase Pu

The long-held conclusion that magnetic order in delta phase Pu makes the structure mechanically unstable in density functional theory calculations is incorrect if the magnetic order is allowed to be non-collinear. The non-collinear 3Q spin structure is intrinsically cubic and makes all structurally equivalent bonds equivalent in their bonding character. Applied to density functional theory calculations on delta phase Pu, the 3Q spin structure results in elastic constants and phonons with the correct symmetry. Finally, the calculated phonon dispersion agrees better with experiment than previous calculations, and the calculated thermal expansion shows the unique behavior seen experimentally: it is negative.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Correlations for the specific heat capacity of ( U x Pu 1 - x ) 1 - y Gd y O 2 - z derived from molecular dynamics

We report UO 2 is the primary conventional fuel used in most nuclear reactors with Gd 2 O 3 commonly added as a burnable absorber to produce a more level power distribution in the reactor core at the beginning of operation. It can also be mixed with other actinide oxides to produce mixed oxide (MOx) fuel. In this study, molecular dynamics simulations were used to predict the specific heat capacity of Gd-doped PuO 2 , UO 2 and (U, Pu)O 2 MOx accommodating Gd 3+ substituted at cation sites via two charge compensation mechanisms - oxygen vacancy formation and the oxidation of U 4+ to U 5+ . The specific heat capacity values for PuO 2 and UO 2 are in good agreement with other studies showing a distinct peak at high temperatures - above 1800 K. As Gd 3+ is added, the peak height reduces for each composition considered. An analytical fit was applied to the data where Gd 3+ was fully charge compensated by either oxygen vacancies or U 5+ . The expression was then validated by predicting the specific heat capacity for three compositions of (Ux Pu 1-x ) 1-y Gd y O 2-z containing both oxygen vacancies and U 5+ , and compared to molecular dynamics data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Computational insights into the structure of anhydrous Pu(III) oxalate

Despite the plutonium oxalate method's wide use in plutonium reprocessing, the method's mechanistic details remain unclear, particularly the identity of the plutonium oxidation state during conversion from an oxalate hydrate to the oxide. Recently, the optical vibrational spectra of Pu(III) oxalate during calcination were measured, providing an experimental reference for computational studies aimed at elucidating the oxalate structure. Here, in this work, we compare the vibrational and electronic properties of two candidate anhydrous Pu(III) oxalate structures calculated using density functional theory with recent experiments. We find that both structures are plausible and may coexist at experimental temperatures, providing insights into the broad features measured in the Raman and infrared spectra.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Transmission Electron Microscopy Characterization of Fuel Cladding Chemical Interaction between Minor Actinides bearing U-Pu-Zr Fuel and AIM1 Cladding

Minor actinides (MA) significantly contribute to the long-term radiotoxicity of spent nuclear fuel (SNF). Separating MA from SNF and incorporating it into metallic fuels for fast reactor transmutation is a potential method to reduce this radiotoxicity. Here, this study focuses on transmission electron microscopy characterization of two samples from the fuel cladding chemical interaction (FCCI) region of an americium (Am) and neptunium (Np)-bearing (MA-bearing) uranium-plutonium-zirconium (U-Pu-Zr) fuel irradiated in the Phenix fast reactor to 9.5 % FIMA burnup at approximately 550 °C cladding temperature. The results show that despite the complex chemical interactions between MA and AIM1 cladding elements, excessive FCCI was not induced, and Am penetration depth in the cladding limited to less than 4 µm. Np remained mostly inside fuel. The Zr-rich compounds layer effectively limited the accumulation of lanthanide on the inner cladding surface. Overall, the FCCI behavior between investigated MA-bearing U-Pu-Zr fuel and AIM1 cladding is benign.

Chemical interaction↗
DOE OSTI2026

Plutonium migration and phase evolution in irradiated U-Pu-Zr metallic fuels: An integrated EPMA-SEM-TEM study

Constituent redistribution is a defining feature of irradiated U-Pu-Zr metallic fuels, yet its mechanisms and effects on fuel performance are not sufficiently resolved to guide model development. Although decades of irradiation testing have established broad trends, a true mechanistic understanding of constituent redistribution has not been achieved. Here, in this study, we use electron probe microanalysis (EPMA), scanning electron microscopy (SEM), and transmission electron microscopy-based (TEM) selective area electron diffraction (SAED) on a EBR-II irradiated U-19 wt.% Pu-6 wt.% Zr fuel pin cross-section to correlate the composition, porosity, and crystallographic phases formed after irradiation. Constituent redistribution is thought to consist of three distinct zones, in which uranium and zirconium migrate while plutonium remains relatively unchanged. Our EPMA results resolve eight distinct compositional regions, and more importantly, show that plutonium redistributes alongside zirconium, contrary to historical assumptions. The distribution of fission products was highly asymmetric with a few large lanthanide precipitates observed at isolated sites on the pin periphery instead of a uniform distribution of smaller precipitates around the periphery. Using thermodynamic data from TAF-ID and the measured EPMA compositions, matrix phase fractions were predicted across the fuel radius. Phase predictions based on composition did not match TEM/SAED results, which revealed a much higher fraction of α−U phase than would be expected if phases were retained from reactor temperatures. These findings highlight the need for expanded SAED phase identification to capture post-irradiation and storage effects, as well as rigorous uncertainty quantification in fuel performance and phase diagram modeling to better constrain predictions from compositional data.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

239 Pu Fission Spectrum Cumulative Fission Product Yield Measurement Using Godiva IV Critical Assembly

Precise integral measurement of fast neutron-induced fission product yields for various actinides is of high interest for applied nuclear science. The goal of this effort is to improve uncertainties in fission product yield values of 239 Pu. Fission was induced in a pure 239 Pu (>99%) target using the Godiva IV critical assembly in burst mode. Further, the irradiated sample was transferred to a high-resolution γ-ray detector within 45 minutes. γ-ray list mode data was collected from 45 minutes to 1 week after the irradiation. γ-ray spectroscopy was performed to analyze the time dependent γ-ray yields using an automated peak search algorithm to identify isotopes by their decay γ-ray energy and half-life. The initial activity for each isotope identified was used to calculate their fission product yield.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Energy dependence of chain fission product yields from neutron-induced fission of 235 U, 238 U, and 239 Pu

The lack of completeness and systematic studies of cumulative fission product yield (FPY) from the three major actinides of 235 U, 238 U, and 239 Pu, over a broad energy range, was the primary motivation for the LLNL-LANL-TUNL collaboration. Given the focus on resolving FPY behavior around fission spectrum neutron energies, our 2016 data provides FPY measurements in fine energy steps in the range of 0.5–5.5 MeV and an additional anchor point at 14.8 MeV. Recently, our collaboration filled the gap between second chance fission and 14.8 MeV, providing a comprehensive study at eleven incident neutron energies from 0.5 to 14.8 MeV. However, during the course of the last measurements, it was found that the reported masses of the actinide deposits in the fission ionization chambers, which were used to measure the number of fissions, were not consistent with the complementary tests. Therefore, some corrections to the old data are needed. The purpose of this paper is to summarize all of the high-yield FPY data obtained for 235 U, 238 U, and 239 Pu isotopes using quasi-monoenergetic neutron beams at 11 incident energies ranging from E n = 0.5 to 14.8 MeV.

Tonchev, A. P. [Lawrence Livermore National Labora↗
DOE OSTI2020

Evaluating 239 Pu(n,f) cross sections via machine learning using experimental data, covariances, and measurement features

In this paper, the neutron-induced 239 Pu fission cross section, 239 Pu(n,f), is evaluated from 1–20 MeV using experimental data and associated covariances while also considering information on the measurement, termed features here. For instance, methods to determine the background, sample backing material, or impurities in the sample, are explicitly taken into account in the evaluation process. To this end, outliers in the experimental data are identified with a modified version of the Hybrid Robust Support Vector Machine. In a second step, two machine learning methods (logistic regression with elastic net regularization and random forest regression with SHAP feature importance metric) are used to highlight measurement features that are common among many of the outlying data points. Based on this analysis, penalty uncertainties are added to the experimental covariances of outlying data points that have outlier measurement features and are put through the generalized-least-squares evaluation. The resulting evaluated mean values and covariances differ distinctly from those data evaluated without the penalty uncertainties. These results highlight that certain measurement features should be more closely examined.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗