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DOE OSTI2023

Impact of Radiolysis on Iodine Speciation in a Variety of Matrices

This report details the work conducted in fiscal year 2023 with the aim of providing a better understanding of the speciation of iodine in support of domestic 99 Mo production Shine who uses a sulfate matrix in their process. Laboratory experiments to study the behavior of iodine speciation were conducted using spectroscopic methods including UV-vis and Raman as well as electrochemical studies. Radiolysis occurring in solution presents a unique challenge that is largely unaddressed with thermodynamic modeling, therefore a kinetic model was developed to address these issues. The focus of the studies conducted this year used surrogate radiolytic conditions induced by hydrogen peroxide for comparison to radiolytic conditions from direct gamma irradiation and neat acid solutions. For each case a model incorporating the kinetics was used to predict iodine speciation in the defined conditions where the output was directly compared to the experimental conditions. Further modeling was conducted incorporating the process conditions that are specific to Shine with incorporation of radiolysis and gas sparging.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Speciation Correlations between Uranium and Iodine

The National Nuclear Security Administration’s Office of Conversion has been supporting development of molybdenum-99 production methods utilizing non-highly enriched uranium targets. One of the novel methods being considered is the use of an aqueous homogeneous subcritical system. Concerns about controlling potentially volatile radioiodine have led to efforts to characterize the iodine species in this system. This report examined the potential to characterize the chemical species of the iodine by correlating it with analyses of uranium, which is more ubiquitous in the target material and suitable for on-line monitoring. Direct measurement of the iodine species would be infeasible due to the difficulty in performing analyses. The use of chemometric modeling utilizing spectral analyses is viable for the uranium measurements, however, a better understanding of the radiolytic effects of neutrons and the corresponding relationship between uranium and iodine is required to correlate any measurements to iodine speciation.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Predicting Radiation-Induced Plutonium Redox Chemistry using Multiscale Modeling Methods

Over the the last 70 years plutonium (Pu) has been integral in the development of several technologies that have changed the world, yet our fundamental understanding of its chemistry is still far from complete. This is a testament to this element?s unique and complex properties, such as its ability to coexist as multiple oxidation states in aqueous solution. Careful manipulation of plutonium oxidation states is essential in the study and utilization of its rich chemistry. To achieve this level of control, a comprehensive mechanistic understanding of radiation-induced plutonium redox chemistry is critical due to the unavoidable exposure of plutonium to ionizing radiation fields, both inherent and from in-process applications. For this reason, we have developed an experimentally evaluated multi-scale computer model for the prediction of gamma radiation-induced Pu(IV) redox chemistry in concentrated nitric acid solutions (1.0, 3.0, and 6.0 M).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

193 mIr/ 192 Ir ratio from irradiation with various NCERC critical assemblies

Eight irradiations of K 2 IrCl 6 pellets were carried out with three critical assemblies: Flattop-Oy, Flattop-Pu and Godiva IV. One of the irradiations was performed with three pellets at different locations of the critical assembly, while the other irradiations with pellets at the center, but operating the critical assembly at different power and, respectively, different neutron flux. Counting samples were prepared from the irradiated pellets and X-ray spectra measured with silicone drift detectors. The 193 mIr/ 192 Ir ratio is reported for each irradiation.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Chemical Process Safety at TRISO-Based, Metal-Based, and Salt-Based Fuel Fabrication Facilities: Technical Assessment and Guidance Assessment

As part of efforts to prepare for potential and ongoing safety reviews for licensing of advanced non-light-water reactor fuel cycles, the U.S. Nuclear Regulatory Commission (NRC) tasked Pacific Northwest National Laboratory to prepare an assessment on the state of knowledge of potential chemical processes at fuel cycle facilities supporting the front end of these fuel cycles, and to assess the associated regulatory guidance. This report provides a technical assessment of chemical process safety considerations to support NRC licensing reviews of fabrication processes for tri-structural isotropic (TRISO) based, metallic-based, and salt-based fuels. The assessments involved collecting publicly available information on the fuel fabrication processes to (i) identify the operational process steps, characteristics and chemicals involved, (ii) identify the physical safety considerations and health safety considerations during licensing reviews of the various process steps, and (iii) collect information to support assessments of severity of accidents and potential mitigative measures to be implemented. The assessment provides a foundational basis on chemical process safety considerations for advanced fuel fabrication activities, although it is recognized that licensing reviews may necessitate design-specific considerations. The specific conditions under which chemical hazards emerge will require process-specific considerations, highlighting the importance of process-informed interpretation. The assessment also determined that exposure guidelines and limits to assess the consequences of acute exposures are limited for some chemicals, although alternative limits and supplementary information from databases or safety data sheets provide sufficient information to evaluate consequences of acute exposures. In addition, it was identified that metallic and salt fuel fabrication processes may involve beryllium, which is an exposure hazard. The regulatory framework for the licensing of advanced fuel cycle facilities, per 10 CFR Part 70 Domestic Licensing of Special Nuclear Material, is deemed robust and flexible to address the chemical safety considerations in this report. A review was conducted on various regulatory guidance and technical basis documents. This included reviewing NUREG-1520, Revision 2, Standard Review Plan for Fuel Cycle Facilities License Applications – Final Report and the process descriptions in Appendix A of NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook, to address advanced fuel types. As new fuels will involve process-specific chemical uses, process-specific considerations are provided in this report. Additionally, it is noted that the U.S. Department of Energy protective action criteria database includes Temporary Emergency Exposure Limits (TEELs) for process-specific chemicals. This report provides technical information to support chemical safety assessments of new advanced fuel cycle facilities and identifies technical and safety information to support licensing reviews. No regulatory barriers were identified for the licensing of advanced fuel cycle facilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

New Methods for Predicting Non-Born-Oppenheimer Chemistry

Current methods for modeling non-adiabatic molecular dynamics face fundamental limitations when treating geometric phase effects: quantum mechanical phenomena where nuclear wavepackets acquire phase shifts when encircling conical intersections. Existing approaches either neglect these effects entirely or rely on potential energy surfaces arising from the Born-Oppenheimer approximation, which introduce artificial singularities and can overestimate geometric phase contributions. This project developed a new theoretical framework based on exact factorization (XF) methods to overcome these limitations. We derived mathematical formulations for hybrid quantum-classical XF dynamics that selectively treat critical nuclear degrees of freedom quantum mechanically while propagating others classically. This approach addresses the computational intractability that has previously limited exact methods to toy systems. Key innovations include a new approach to systematically identifying nuclear coordinates requiring quantum treatment, as well as novel implementation strategies that interface with existing quantum chemistry codes. The project also developed a proof-of-concept code for treating Jahn-Teller systems and creation of educational materials on non-adiabatic dynamics geared at the graduate level. The theoretical framework developed will enable future systematically improvable calculations of nuclear quantum effects in realistic molecular systems, filling a critical gap in non-adiabatic dynamics methods. This foundation supports future development of predictive tools for designing energy-relevant photochemical processes where quantum coherence effects may be exploited to control reaction outcomes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Nuclear Technology Selection and Preliminary Integration at St. Charles and Carrollton Chemical Plants

Nuclear energy is identified as a valuable candidate for decarbonizing industrial facilities as a non-emitting baseload energy source capable of dispatching energy as needed for an integrated facility. As industrial facilities are unique, analyzing what nuclear technology is the best candidate should be evaluated on a case-by-case basis. For two identified facilities of the Dow Chemical Company, an ideal mix of nuclear energy resources is calculated, and integration strategies are put forward. This research was funded through a Gateway for Accelerated Innovation in Nuclear (GAIN) Nuclear Energy Voucher, specifically seeking near-term engineering solutions for the technical issue of introducing nuclear generated energy into the Dow chemical facilities. The research is split into four main phases: background on reactors and chemical facilities, preliminary engineering analysis determining how reactor technologies may be integrated with specific facilities, site characterization and discrete hazards analysis, and finally an optimization of reactor-facility selection through the use of Framework for Optimization of ResourCes and Economics (FORCE) tools.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Evaluations for neutron-induced activation cross sections for Y and Zr isotopes

Activation cross sections are an important input for evaluating results of radiochemical analyses, both for basic science and programmatic applications. We present an update on cross sections for a select set of Y and Zr target isotopes, including several isomers. This effort will focus on a methodology to obtain uncertainty quantified systematic trends for reaction model parameters and propagate them to cross sections. We discuss our results in the context of previous calculations.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Effects of Borate and Organics on U(VI) Solubility in WIPP Brine

This report provides an update to a previously issued report on the solubility of uranium (VI) at different borate concentrations and in the presence of organics. The solubility of uranium (VI) in the Waste Isolation Pilot Plant (WIPP)-relevant brine was determined to support ongoing WIPP recertification activities (CRA-2026). This research was performed by the Los Alamos National Laboratory-Carlsbad Operations (LANL-CO) Actinide Chemistry and Repository Science Program (ACRSP).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Neptunium (V) Solubility in WIPP Brine: Presence and Absence of WIPP-Relevant Organic and Inorganic Ligands

The oxidation state and solubility of neptunium (Np) in 5 M NaCl and high ionic strength synthetic WIPP (Waste Isolation Pilot Plant) brines were systematically investigated as a function of pCH+ (7-11) in the presence and absence of WIPP-relevant organic ligands, including EDTA (Ethylenediaminetetraacetic acid), oxalate, citrate, and acetate, as well as inorganic ligands such as borate and carbonate at a temperature of 23 ± 2°C. This study was conducted through long-term batch solubility experiments spanning approximately 950 days, using an undersaturation approach. A comprehensive set of experimental and spectroscopic techniques, including UV-VIS-NIR spectroscopy and Np L III -edge X-ray Absorption Spectroscopy (XAS), was employed to identify the solubility-controlling Np solid phases and the predominant aqueous Np species present in the samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Effect of Citrate on the Solubility of Uranium (VI) in WIPP Brine

The Waste Isolation Pilot Plant (WIPP) is the only active deep geological repository in the United States for the disposal of defense-related transuranic (TRU) waste, located in the northern part of the Delaware Basin in southeastern New Mexico, approximately 26 miles east of Carlsbad. The repository is situated 2,150 feet (about 610 meters) underground within the Salado Formation, a thick layer of stable salt deposits. This unique geological formation provides a secure and long-term environment for isolating TRU waste, ensuring its safe containment for thousands of years.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

The Oak Ridge National Laboratory Glenn T. Seaborg Institute: 2025 highlights

The Glenn T. Seaborg Institute (GTSI) at Oak Ridge National Laboratory (ORNL) serves as a dedicated hub for advancing actinide R&D, aimed at strengthening the United States’ capabilities in actinide science. With a focus on medical, industrial, and national security applications, the ORNL GTSI seeks to foster the next generation of scientists and engineers. Collaborating closely with corresponding Seaborg Institutes at Lawrence Livermore National Laboratory (LLNL), Lawrence Berkeley National Laboratory (LBNL), Los Alamos National Laboratory (LANL), and Idaho National Laboratory, the ORNL GTSI leverages its unique strengths. These include unparalleled expertise in isotope production, handling of rare isotopes, and leadership in neutron and computational sciences. Notably, the ORNL GTSI is positioned as the nation’s cornerstone for actinide R&D, capitalizing on the exceptional capabilities of the High Flux Isotope Reactor—the premier reactor for isotope production—and the Hazard Category 2 Radiochemical Engineering Development Center, which supports advanced production, processing, and applications research. Furthermore, the GTSI is committed to serving as a national center of excellence for training across all levels of actinide science education. This report provides an overview of the activities, milestones, and accomplishments of the ORNL GTSI over the past year.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Morphological Characterization of UO 2 F 2 and UF 4 Hydrolysis Products

Bulk samples of UO 2 F 2 and UF 4 were aged for 3 weeks under various relative humidity (RH) conditions to examine morphological changes associated with the hydrolysis of the materials. Aged samples were analyzed using scanning electron microscopy (SEM) to examine morphological changes, and Raman spectroscopy was used to correlate these changes with chemical composition. Little to no effect was evident for either material under lower RH conditions (i.e., 16% and 33%–38% RH). Under higher RH conditions (83% and 99%), multiple morphological changes were evident within the UO 2 F 2 samples because of the formation of hydrolysis products and possible interactions at the solid–vapor interface. Only one major morphological change was observed over the aging period—within the UF 4 system at 99% RH. This change can be attributed to the initial formation of anhydrous UO 2 F 2 and, subsequently, its hydrolysis products. After 37 weeks, UO 2 F 2 persisted as a hydrolysis product of UF4; however, no further hydrolysis products were observed. These results suggest that further research into a possible divergence of UF 4 and UO 2 F 2 degradation pathways is warranted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Atomistic Simulations for Thermophysical Properties of Uranium-Containing Halide Molten Salts

Characterizing the thermophysical properties in both fuel and coolant salts are critical in modeling, developing, process optimizing and utilizing molten salt reactors (MSRs), as these properties directly relate to operation metrics and can inform on the selection of candidate salts. The demand for consistent, accurate and publicly available thermophysical property data has become more apparent in recent years as interests have increased from molten salt reactor developers. There are a number of challenges in experimentally measuring properties such as thermal conductivity, viscosity, density and heat capacity , which have led to sparse and often times conflicting data points or molten salts in general. Additionally, there are a number of hazards to consider when synthesizing, storing, using, treating and disposing of molten salts. With the advances in computational capabilities over the last 10 years, the use of atomistic simulations can be implemented to support these efforts. The primary objective of this work is characterize the thermophysical transport properties in a number of molten chloride salts, and in particular NaCl-UCl 3 using ab-initio molecular dynamic (AIMD) simulations. In this binary salt the UCl 3 acts as the primary fissile material and NaCl acts as a carrier salt due with its’ high solubility for actinides A number of studies on the thermophysical properties of NaCl-UCl 3 have been published but there is not a vast amount of viscosity data for this system. In 1975, Desyatnik, et al published a study reporting dynamic viscosities that were calculated from kinematic viscosity measurements, and using the coefficients provided the viscosity in a 70:30 NaCl:UCl 3 mixture is 2.29 cP and 2.88 for a 60:40 mixture. Termini et al. recently reported viscosities in the range of 2.75 – 3 cP for the 63:37 NaCl-UCl 3 mixture in the same temperature range using rolling ball viscosity measurements. Computational viscosity of a similar mixture (64:36) can be obtained from the work Andersson et al. using the reported diffusion coefficients, and the hydrodynamic radius from the pair-radial distribution functions (RDFs). Using Eq (1) (vida infra), the viscosity would be 2.50 cP at 1100K. This is not to say that these values are incorrect due to the varying reported values, but aims to highlight the necessity of this work. The data reported in this ongoing work are computations on a 64:36 mixture of NaCl-UCl 3 at 987K. This work is likely to be expanded into varying concentrations of this mixture along with the inclusion of other salt candidate mixtures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

High Gradient Magnetic Simulation Final Report 2025

Over the course of nearly 2 years, the goal of this project has been to identify methods for the separation of uranium oxide particles from other geologic materials. The primary motivation is to assess methods for enhancing the extraction and throughput of actinide particles from environmental samples, enabling the evaluation of these trace actinides through more traditional methods such as thermal ionization mass spectrometry or secondary ion mass spectrometry.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

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.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Setup for Optical Absorption Measurements of Molten Salt Mixtures

In this work we describe the setup of a spectrophotometer to take absorbance measurements of a heated sample in an inert atmosphere. By using fiber optics to bring light into and out of a glovebox we are able to use a pitch-catch optical setup to observe samples in a cuvette heated up to 400℃. The functionality of the setup is demonstrated by observing the signal produced by Li 2 S in a eutectic mixture of LiCl/LiBr/KBr at elevated temperature overtime.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗