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Oxidative Cross-Esterification and Related Pathways of Co-Adsorbed Oxygen and Ethanol on Pd–Au
Ethanol is one of the most important industrial chemicals considering its wide range of uses. From ultrahigh vacuum (UHV) experiments, we were able to effectively produce acetaldehyde and ethyl acetate on a Pd–Au model catalyst. The presence of oxygen at the Pd–Au interface sites was found to be highly reactive for ethanol dehydrogenation, significantly increasing acetaldehyde production and promoting cross-esterification to produce ethyl acetate. Density functional theory calculations show that compared to the bare PdAu(111) surface, the presence of preadsorbed oxygen and the subsequently formed hydroxyl group leads to highly active and selective initial ethanol dehydrogenation at the O–H and α–C-H position, which was corroborated by isotope-labeling experiments. Specifically, using temperature-programmed desorption with the isotope CD 3 CH2OH, we identified C–C bond breakage through methane production. Furthermore, the simultaneous desorption of D2 with ethyl acetate from the oxidation of CD 3 CD 2 OD suggests that the abstraction of hydrogen from acetaldehyde and/or ethanol is a relevant step in ethyl acetate production, which is supported by our theoretical predictions. These results shed light on the mechanistic pathways of having oxygen and ethanol coadsorbed on Pd–Au surfaces, which has significant implications for industrial applications such as fuel cells where catalytic design plays a major role in the efficiency of power production.
Paths to superheavy nuclei
This document summarizes the discussions and outcomes of the Facility for Rare Isotope Beams (FRIB) Theory Alliance topical program ‘The path to Superheavy Isotopes’ held in June 2024 at FRIB. Its content is non-exhaustive, reflecting topics chosen and discussed by the participants. The program aimed to assess the current status of theory in superheavy nuclei (SHN) research and identify necessary theoretical developments to guide experimental programs and determine fruitful production mechanisms. This report details the intersection of SHN research with other fields, provides an overview of production mechanisms and theoretical models, discusses future needs in theory and experiment, explores other potential avenues for SHN synthesis, and highlights the importance of building a strong theory community in this area.
Determination of Collision Cross Sections Using a Fourier Transform Electrostatic Linear Ion Trap Mass Spectrometer
Collision cross sections (CCSs) were determined from the frequency-domain linewidths in a Fourier transform electrostatic linear ion trap. With use of an ultrahigh-vacuum precision leak valve and nitrogen gas, transients were recorded as the background pressure in the mass analyzer chamber was varied between 4× 10{sup -8} and 7 × 10{sup -7} Torr. The energetic hard-sphere ion–neutral collision model, described by Xu and coworkers, was used to relate the recorded image charge to the CCS of the molecule. In lieu of our monoisotopically isolating the mass of interest, the known relative isotopic abundances were programmed into the Lorentzian fitting algorithm such that the linewidth was extracted from a sum of Lorentzians. Although this works only if the isotopic distribution is known a priori, it prevents ion loss, preserves the high signal-to-noise ratio, and minimizes the experimental error on our homebuilt instrument. Six tetraalkylammonium cations were used to correlate the CCS measured in the electrostatic linear ion trap with that measured by drift-tube ion mobility spectrometry, for which there was an excellent correlation (R{sup 2} ≈ 0.9999). Although the absolute CCSs derived with our method differ from those reported, the extracted linear correlation can be used to correct the raw CCS. With use of [angiotensin II]{sup 2+} and reserpine, the corrected CCSs (334.9 ± 2.1 and 250.1 ± 0.5, respectively) were in good agreement with the reported ion mobility spectrometry CCSs (335 and 254.3, respectively). With sufficient signal-to-noise ratio, the CCSs determined are reproducible to within a fraction of a percent, comparable to the uncertainties reported on dedicated ion mobility instruments. .
Strategies for Managing Unique Waste Streams Resulting from Heavy Isotope Production and Advanced Fuel Cycle Development Programs at ORNL
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Significant Roles of Surface Hydrides in Enhancing the Performance of Cu/BaTiO 2.8 H 0.2 Catalyst for CO 2 Hydrogenation to Methanol
Abstract Tuning the anionic site of catalyst supports can impact reaction pathways by creating active sites on the support or influencing metal‐support interactions when using supported metal nanoparticles. This study focuses on CO 2 hydrogenation over supported Cu nanoparticles, revealing a 3‐fold increase in methanol yield when replacing oxygen anions with hydrides in the perovskite support (Cu/BaTiO 2.8 H 0.2 yields ~146 mg/h/gCu vs. Cu/BaTiO 3 yields ~50 mg/h/gCu). The contrast suggests that significant roles are played by the support hydrides in the reaction. Temperature programmed reaction and isotopic labelling studies indicate that BaTiO 2.8 H 0.2 surface hydride species follow a Mars van Krevelen mechanism in CO 2 hydrogenation, promoting methanol production. High‐pressure steady‐state isotopic transient kinetic analysis (SSITKA) studies suggest that Cu/BaTiO 2.8 H 0.2 possesses both a higher density and more active and selective sites for methanol production compared to Cu/BaTiO 3 . An operando high‐pressure diffuse reflectance infrared spectroscopy (DRIFTS)‐SSITKA study shows that formate species are the major surface intermediates over both catalysts, and the subsequent hydrogenation steps of formate are likely rate‐limiting. However, the catalytic reactivity of Cu/BaTiO 2.8 H 0.2 towards the formate species is much higher than Cu/BaTiO 3 , likely due to the altered electronic structure of interface Cu sites by the hydrides in the support as validated by density functional theory (DFT) calculations.
HFIRCON Version 1.0.5 User Guide
The High Flux Isotope Reactor (HFIR) Controller (HFIRCON) code is a collection of python routines and C plugins that automate the workflow for fuel and single- or multicycle target depletion analyses for HFIR at Oak Ridge National Laboratory (ORNL). This code calls the LAVAMINT (LAVA Model Interrogator) parallel (MCNP) Monte Carlo N-Particle model interrogator to stochastically calculate cell volumes and bounding boxes, the ADVANTG (Automatic Variance Reduction Generation) code package for all variance reduction and source biasing calculations, the ORNL-Transformative Neutronics/MCNP5 transport solver for all transport solutions, and the MSX_DEPLETE module to perform all depletion calculations via the ORIGEN (Oak Ridge Isotope Generation) application programming interface. It also performs a robust set of postprocessing functions to automatically provide summaries of several key metrics that are common to a wide variety of typical HFIR design and safety-basis analyses.
Management of Fissionable Equivalent Mass Materials at ORNL - 20146
The US DOE manages an inventory of materials that contains a range of long-lived radioactive isotopes that were produced from the 1960's through the 1980's by irradiating targets in production reactors to produce special heavy isotopes for US DOE programmatic use, scientific research, and industrial and medical applications. Since the production reactors and enrichment facilities that produced many of these materials have been shut down, they are considered unique materials that are not likely to be produced again. ORNL uses these materials in US DoE's center for production, storage, and distribution of TRU isotopes (plutonium through californium) and other related nuclear research programs. ORNL also operates the High Flux Isotope Reactor, which provides a high neutron source for production of isotopes for medical, industrial, and nuclear research programs. As a result, ORNL has an inventory of radioisotopes that are being managed for ongoing research programs and being held for reuse because they have potential intrinsic value to US DOE. An initiative is underway at ORNL to better manage these materials, particularly focusing on those with high fissionable equivalent mass that could impact the ability to perform ongoing or new research projects. This paper describes the actions ORNL is taking to manage these inventories, many of which consist of TRU materials, through reuse and disposal. (authors)
Mathematical Modeling of Nonstationary Separation Processes in a Cascade of Gas Centrifuges for Separation of Tungsten Isotopes
We have developed and program-realized a mathematical model of nonstationary separation processes proceeding in cascades of gas centrifuges for separation of multicomponent isotope mixtures. With the help of this model we have calculated the parameters of nonstationary separation processes proceeding in the cascades of gas centrifuges for tungsten isotope separation. It has been shown that the model adequately describes the nonstationary processes in the cascade and is suitable for calculating their parameters during separation of multicomponent isotope mixtures.
Zirconium Nuclear Data Campaign: Measurement of 90 Zr ( n, γ ) Cross Section
The isotopes of Zr with A = [90, 91, 92, 94] make up more than 97% of naturally occurring Zr and are important to many nuclear applications such as nuclear reactors. One of the attractive qualities of naturally occurring Zr isotopes is that they have a low σ γ /σ t ratio at most neutron energies. Thus, they improve the neutron economy in reactors by preferentially scattering neutrons rather than absorbing them. This same quality also presents a challenge to measuring the capture cross section, σ γ , of Zr isotopes. The ENDF/B VIII.0 library has a relative uncertainty of approximately 10–20% for incident neutron energies < 0.1 MeV and an uncertainty greater than 20% for energies > 0.1 MeV for the majority of natural Zr isotopes. This motivated the Nuclear Criticality Safety Program to embark on a campaign to accurately measure and evaluate these Zr isotopes. In this work, we demonstrate energy-dependent neutron capture cross section measurements for the first enriched sample to be measured: 90 Zr.
Advanced Fuels Campaign Execution Plan
The Advanced Fuels Campaign (AFC) Execution Plan outlines the strategy, mission, scope, near-term and long-term goals, structure, and organization associated with nuclear fuels and materials research, development, and demonstration activities within the Department of Energy’s (DOE) Nuclear Fuel Cycle and Supply Chain (NFCSC) program. NFCSC has been given responsibility to identify and mature advanced fuel technologies for the DOE using a science-based approach, focused on developing a fundamental understanding of nuclear fuels and materials to drive development of integrated nuclear fuel and materials technology. This science-based approach combines theory, experiments, and multiscale modeling and simulation to achieve a predictive understanding of relevant behaviors ranging from fuel fabrication processes (and their resulting fuel microstructures) through fuel/cladding performance under irradiation (in contrast to more empirical, observation-based approaches frequently used in fuel performance modeling and fuel qualification). The traditional scope of AFC includes the evaluation and development of multiple fuel forms to support two fuel cycle options: once-through and full recycle. The word “fuel” is used generically to include conventional fuels, transmutation targets, and any associated cladding or duct materials. The once-through fuel cycle addresses advanced light water reactor fuels with enhanced performance, extended burnup, and reduced waste generation. In fiscal year (FY) 2012, AFC’s scope expanded to include research, development, and demonstration (RD&D) for light water reactor (LWR) fuels with enhanced accident tolerance. Fuel fabrication activities include the development of innovative methods to enhance process efficiencies, reduce waste, and improve control over as-fabricated fuel microstructural properties to achieve desired in-reactor performance. Using modern modeling and simulation approaches, the objective is to predict fresh fuel properties given the feedstock characteristics and fabrication process parameters. The performance-related activities include small-scale, in-reactor, and out-of-reactor phenomenological testing (distinct from, but synergistic with, integral prototypic testing) and extensive, quantitative characterization (focusing on characterization of fuel and cladding materials at the scale of microstructure) both before and after testing. Larger-scale, prototypic experiments are conducted in concert with phenomenological testing to drive a Fuel Development and Qualification program, incorporating a fundamental understanding of fuel behavior performance characteristics. Then, using the tools developed under the productive science-based approach, fuels will be optimized to meet specific performance requirements, thereby minimizing the need to repeatedly perform large-scale, integral experiments over a wide parametric range as a means of experimental exploration. Two significant initiatives are underway within AFC. First, a gap analysis completed in early FY 2019 identified critical irradiation testing needs that are lacking within the national light water reactor (LWR) fuels testbed since the shutdown of the Halden Reactor in 2018. The identified gaps are for instrumented, prototypic testing of LWR fuels, especially under boiling water reactor conditions, ramp conditions, and conditions leading to fuel failure; these needs exist for supporting current LWR fuels and their possible extension to higher burnups, but are especially urgent relative to near-term development and qualification of accident-tolerant fuels. Recommendations that resulted from the Halden Gap Analysis focused on enhancements at Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) to fill gaps in testing capabilities relative to these needs. Second, a concerted effort to develop and demonstrate a systematic approach to accelerating the development, testing, and qualification of new fuel systems has been initiated. This is highlighted by a test strategy that combines the considerable advances in multiscale, mechanistic fuel modeling of recent years with a MiniFuel separate effects test program in the High Flux Isotope Reactor (HFIR) and a Fission Accelerated Steady-state Testing (FAST) semi-integral accelerated test program in ATR. This approach is being tested/demonstrated using the metallic fuel system, but if successful it is expected to be extensible to multiple fuel types and diverse applications. This document includes an overview of the NFCSC program, a definition of science-based development of nuclear fuels, near-term goals for Advanced LWR fuels (ALFs), and longer-term goals for Advanced Reactor Fuels (ARFs) RD&D. This includes the activities that will be conducted to achieve success toward the grand challenge, as well as the goals and milestones to be achieved over the next few decades of research and development. Long-term goals are based on the DOE Office of Nuclear Energ
Advanced Fuels Campaign Execution Plan
The Advanced Fuels Campaign (AFC) Execution Plan outlines the strategy, mission, scope, near-term and long-term goals, structure, and organization associated with nuclear fuels and materials research, development, and demonstration activities within the Department of Energy’s (DOE) Nuclear Fuel Cycle and Supply Chain (NFCSC) program. NFCSC has been given responsibility to identify and mature advanced fuel technologies for the DOE using a science-based approach, focused on developing a fundamental understanding of nuclear fuels and materials to drive development of integrated nuclear fuel and materials technology. This science-based approach combines theory, experiments, and multiscale modeling and simulation to achieve a predictive understanding of relevant behaviors ranging from fuel fabrication processes (and their resulting fuel microstructures) through fuel/cladding performance under irradiation (in contrast to more empirical, observation-based approaches frequently used in fuel performance modeling and fuel qualification). The traditional scope of AFC includes the evaluation and development of multiple fuel forms to support two fuel cycle options: once-through and full recycle. The word “fuel” is used generically to include conventional fuels, transmutation targets, and any associated cladding or duct materials. The once-through fuel cycle addresses advanced light water reactor fuels with enhanced performance, extended burnup, and reduced waste generation. In fiscal year (FY) 2012, AFC’s scope expanded to include research, development, and demonstration (RD&D) for light water reactor (LWR) fuels with enhanced accident tolerance. Fuel fabrication activities include the development of innovative methods to enhance process efficiencies, reduce waste, and improve control over as-fabricated fuel microstructural properties to achieve desired in-reactor performance. Using modern modeling and simulation approaches, the objective is to predict fresh fuel properties given the feedstock characteristics and fabrication process parameters. The performance-related activities include small-scale, in-reactor, and out-of-reactor phenomenological testing (distinct from, but synergistic with, integral prototypic testing) and extensive, quantitative characterization (focusing on characterization of fuel and cladding materials at the scale of microstructure) both before and after testing. Larger-scale, prototypic experiments are conducted in concert with phenomenological testing to drive a Fuel Development and Qualification program, incorporating a fundamental understanding of fuel behavior performance characteristics. Then, using the tools developed under the productive science-based approach, fuels will be optimized to meet specific performance requirements, thereby minimizing the need to repeatedly perform large-scale, integral experiments over a wide parametric range as a means of experimental exploration. Two significant initiatives are underway within AFC. First, a gap analysis completed in early FY 2019 identified critical irradiation testing needs that are lacking within the national light water reactor (LWR) fuels testbed since the shutdown of the Halden Reactor in 2018. The identified gaps are for instrumented, prototypic testing of LWR fuels, especially under boiling water reactor conditions, ramp conditions, and conditions leading to fuel failure; these needs exist for supporting current LWR fuels and their possible extension to higher burnups, but are especially urgent relative to near-term development and qualification of accident-tolerant fuels. Recommendations that resulted from the Halden Gap Analysis focused on enhancements at Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) to fill gaps in testing capabilities relative to these needs. Second, a concerted effort to develop and demonstrate a systematic approach to accelerating the development, testing, and qualification of new fuel systems has been initiated. This is highlighted by a test strategy that combines the considerable advances in multiscale, mechanistic fuel modeling of recent years with a MiniFuel separate effects test program in the High Flux Isotope Reactor (HFIR) and a Fission Accelerated Steady-state Testing (FAST) semi-integral accelerated test program in ATR. This approach is being tested/demonstrated using the metallic fuel system, but if successful it is expected to be extensible to multiple fuel types and diverse applications. This document includes an overview of the NFCSC program, a definition of science-based development of nuclear fuels, near-term goals for Advanced LWR fuels (ALFs), and longer-term goals for Advanced Reactor Fuels (ARFs) RD&D. This includes the activities that will be conducted to achieve success toward the grand challenge, as well as the goals and milestones to be achieved over the next few decades of research and development. Long-term goals are based on the DOE Office of Nuclear Energ
Using Monte Carlo methods to estimate occupational dosage during 225 Ac production
Since alpha emitting isotopes such as 225 Ac, ideal for radiation treatment, are in short supply, the Medical Isotope Research and Production program at Brookhaven National Laboratory (BNL) is devel- oping new pathways for 225 Ac production. One proposed pathway is to resurrect the EBCO cyclotron at BNL to irradiate a 226 Ra target for 40 hours in a 16 MeV proton beam. Using MCNPX and CINDER software to simulate this process, we found that 1.26 Ci of 225 Ac could be produced from 300 mg of 226 Ra. To ensure workplace safety, we modeled the detailed geometry of lab 2-66C in Bldg. 801 and performed Monte Carlo simulations to estimate the radiation outside the lab as the irradiated sample is processed in a hot cell. We found the maximum dose outside the lab to be about 9 x 10 -5 rem/h. To reduce occupational dose, we recommend 200 thick steel shielding covering the sample insertion passage. These 225 Ac production runs could help cover the EBCO cyclotron's operational costs.
A membrane-based approach to medical isotope purification (Final Report)
This is the final technical report from the grant "A membrane-based approach to medical isotope purification" which was awarded by the Isotope Research and Development Program through the early career research program to Christine Duval at Case Western Reserve University. The project developed adsoprtive membranes with covalently bound diglycolamide-derived ligands for the purification of Ac-225 and benchmarked them against the current state of the art adsorbent materials.
Removal and Decommissioning of Disused Gamma Irradiator in Minsk, Belarus
Under the scope of The Department of Energy’s Office of Radiological Security (ORS) Program the Belarusian company, “Isotope Technologies” (IT), in concert with assistance of “Radii” (a company based in Moscow, Russia) dismantled a disused gamma irradiator RKhM-γ-20. This project required innovation, flexibility, and prior experience. The team successfully removed the radioactive sources from the site and moved them to the local radioactive waste facility Ekores. The gamma unit had been installed at the Minsk, Belarus Research Institute of Physical Organic Chemistry in 1971. This research center is a part of the National Academy of Sciences of Belarus. There were 48 highly radioactive Cobalt 60 sources inside the unit. During the dismantling process it was discovered that the source retaining drum was broken. Therefore, the removal proved to be much more complicated than originally envisioned. IT and Radii were unable to remove a number of sources from the drum by means of standard removal equipment and finally needed to remove the entire unit itself along with the remaining sources. The problem was further complicated by a very short timeframe allowed for work implementation. This was because the unit was located near an elementary school requiring the team to complete the project prior to the end of summer school holidays. Moreover, the unit itself was not easily accessible. The project team developed a technical solution and accomplished the removal.
Development of a Computational Framework for Multiphysics Multiphase Species Tracking using NEAMS Tools
This report implements a high-fidelity multiphysics modeling framework using the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program tools to track isotopic species in Molten Salt Reactors (MSRs), with a specific focus on the 91-depletion chain within the Molten Salt Reactor Experiment (MSRE). The model integrates neutronics, thermal-hydraulics, depletion, and thermochemistry to simulate the production, transport, and phase transitions of isotopes under steady-state and transient conditions. The main findings reveal that isotopes such as bromine-91 largely remain in the liquid phase, while others, including krypton-91and yttrium-91, transition to the gas phase, significantly influencing the reactor’s radiological source term. The study also shows that during transients, like a reactivity insertion transient, rapid void formation and the expansion of the liquid-gas interface led to substantial transfers of dissolved isotopes into the gas phase, altering isotope distribution and largely increasing the source term in the off-gas system. Additionally, the research highlights that short-lived isotopes dominate the initial off-gas response during transients, while longer-lived isotopes determine the equilibrium state, underscoring the necessity of dynamic simulations for accurate species tracking and reactor safety analysis. The developed methodology will be applied in the future to the tracking of a larger number of species and introduce other species tracking mechanisms, such as deposition and plating.
Cell design for laser-induced breakdown spectroscopy measurements on reactive gas samples
Uranium hexafluoride (UF 6 ) is the primary material used for the enrichment of uranium in the production of light water nuclear reactor fuels worldwide. Accurate, rapid quantification of uranium isotopic composition in nuclear materials is required for safeguards programs and nonproliferation purposes. One potential technique for isotopic measurements in uranium species in the field is laser induced breakdown spectroscopy (LIBS). Safe and effective application of LIBS to UF 6 for enrichment measurements is uniquely challenging due to the chemical and physical properties of UF 6 , which necessitate specific handling procedures. The objective of this work is to design a cell for isotopic analysis of UF 6 that is (1) compatible with chemical and physical properties of UF 6 , (2) compatible with LIBS laser, and (3) portable-sized for nuclear safeguards applications. Along with cell design, initial testing of the cell for basic performance and chemical compatibility is performed. As designed and constructed, the portable gas cell was gas-tight, chemically compatible with UF 6 , and withstood long-duration laser exposure. In this work, the cell has proven capability for handling reactive gasses, such as UF 6 , with specification application to isotopic analysis.
SRF Resonators for a bi-directional Energy Ramping Upgrade of the Isotope Production Facility Beamline at LANL
The LANL Isotope Production Facility (IPF) is one of the two high-current proton linear accelerators operated under the auspices of the DOE Office of Nuclear Physics Isotope Development and Production for Research Applications Program. The IPF beamline operates with a fixed 100-MeV energy proton beam, diverted from the LANSCE accelerator and guided towards the isotope production target station. The fixed beam energy limits the number of isotopes that can be produced at IPF, and adversely affects the purity of isotopes produced at non-optimal energies. In particular, the IPF program is receiving increasing numbers of requests for high purity isotopes such as Np-236 (national security applications) and Re-186 (medical therapy). These are typically produced via (p,n) and (p,2n) nuclear reactions, requiring very tight control of the incident energies on the production targets to ensure an end product with a high radionuclidic purity. It is well known that this level of energy control cannot be achieved when a 100 MeV beam of mono-energetic protons is degraded to the required energies, typically below 30 MeV, as too much energy spread is acquired. Operating the IPF beamline at lower, mono-energetic beam energies such as at 40 MeV will allow the required tight energy control in target assemblies to ensure a high purity product. In addition, NP has requested a LANL/BNL/ORNL collaboration focus on establishing a national large-scale production capability for the therapy isotope Ac-225. Recent nuclear cross section measurements done at LANL show that due to a steep rise in production cross section towards higher proton energies (See Figure 2), proton beams at energies substantially greater than 100 MeV will be highly beneficial for Ac-225 production [AC255]. Such higher-energy beams cannot be delivered to IPF at present. Lower-energy beams (70 MeV and 40 MeV) can be delivered but only in sole-use mode; all other LANSCE users must be offline during 70 MeV or 40 MeV operation.