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At least 289 records · Page 16

Assembly of Uranyl Peroxides from Ball Milled Solids

Mechanochemistry enables transformations of highly insoluble materials such as uranium dioxide or the mineral studtite [(UO 2 )(O 2 )(H 2 O) 2 ]·(H 2 O) 2 into uranyl triperoxide compounds that can subsequently assemble into hydroxide-bridged uranyl peroxide dimers in the presence of lithium hydroxide. Dissolution of these solids in water yields uranyl peroxide nanoclusters including U 24 , Li 24 [(UO 2 )(O 2 )(OH)] 24 . Insoluble uranium solids can transform into highly soluble uranyl peroxide phases in the solid state with miniscule quantities of water. Furthermore, such reactions are potentially applicable to uranium processing in the front and back end of the nuclear fuel cycle.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

3D printed features on nuclear fuel cladding for optimized heat transfer

A nuclear fuel cladding with improved thermomechanical properties is provided. The nuclear fuel cladding includes a double-walled construction having inner and outer hexagonal sidewalls. The inner sidewall and the outer sidewall are spaced apart from each other to form a cooling channel therebetween, and the inner sidewall surrounds a nuclear fuel and is spaced apart from the nuclear fuel by a small gap. Helical fins extend into the cooling channel to interconnect the inner sidewall and the outer sidewall. Resilient fingers extend toward the nuclear fuel through the small gap to comply with variations in the size of the nuclear fuel due to fabrication tolerances as well as thermal expansion and swelling of the nuclear fuel, for example UO 2 , when undergoing fission. The nuclear fuel cladding is formed according to an additive manufacturing process, for example laser powder bed fusion printing.

Petrie, Christian M.↗

Disruptive thermal-spectrum molten salt breeder reactor fuel cycle technologies

Multiple, disruptive fuel salt processing technologies for thermal-spectrum molten salt breeder reactors (TS-MSBRs) have been conceived of and brought to laboratory levels of maturity since the termination of the historic U.S. government program. This paper describes these technologies and the role that they could serve in decreasing the remaining technical risks for TS-MSBR deployment. In conclusion, the focus of the paper is on describing how the technologies could integrate into a dramatically simpler, proliferation resistant fuel salt processing system as well as identifying remaining development hurdles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Advanced nuclear reactor integration opportunities for the pulp and paper industry in the U.S. context: Technical perspectives, gap analysis, and preliminary technoeconomic assessment

Pulp and paper (P&P) manufacturing requires a large amount of low-pressure (LP) steam to digest, wash wood fibers and dry pulp into paper. Most of the LP steam is extracted from backpressure turbines that produce power from high-pressure (HP) steam. This HP steam is generated from burning wood waste material; bark is burned in hog boilers, and lignin is boiled in a black liquor recovery boiler. In a typical integrated P&P mill, 50–100% of the steam is produced from these sources, while additional steam is produced in natural gas (NG), fuel oil, or coal boilers. The other energy-intensive process in the plant is the chemical-recovery section (e.g., lime kiln), which requires high-temperature processing from NG combustion to retrieve and recirculate spent chemicals. This paper assesses the energy and heat demand and material balances of a typical generic kraft pulp mill, along with the nuclear heat, steam, and power integration opportunities to replace conventional combustion systems. The paper also addresses steam and electricity generation through a comprehensive technical and engineering gap analysis of five different nuclear-integration opportunities and their process economics, thus enabling the lignin and bark to be further processed into biobased chemicals or fuels, as well as the potential to reduce overall emissions from kraft pulping. Preliminary findings have shown that the P&P industry could achieve technological benefits by integrating their current manufacturing process with small modular nuclear reactors (SMNRs) on a national level. This research aims to set the path forward for a cleaner and more resilient P&P industry.

Advanced nuclear reactor↗

Developing a Process for Collaboration-Based Siting of a Federal Consolidated Interim Storage Facility in the United States: Learning from Communities Living with Legacy Waste

In June of 2023, the U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) announced the selection of its Collaboration-Based Siting (CBS) Consortia, a group of 12 awardees, including the Consortium for Risk Evaluation with Stakeholder Participation (CRESP) (led by Vanderbilt University) to assist DOE-NE with the development of its process for siting a federal consolidated interim storage facility (FCISF) for spent nuclear fuel (SNF) storage. At this time, DOE NE is not soliciting interested host communities, rather the CBS Consortia are tasked with in-depth engagement, mutual learning, and capacity building to provide DOE NE with feedback on the CBS process. CRESP’s objective is to engage communities in two regions (the Pacific Northwest and the Southeast) with sites currently storing defense- and research-related SNF to foster learning concerning the best and worst practices in community participation in risk-informed decision making. This includes learning from existing structures for public input in radioactive waste management decision making [e.g. citizen advisory boards (CABs)] and other local parties about how to build and sustain trust among the parties. CRESP has been working to engage stakeholders and Tribes surrounding two DOE sites historically differing in receptiveness to engaging with DOE and trusting in DOE to accomplish its missions. CRESP’s approach to date has consisted of (1) engaging voluntarily members and former members of the DOE Office of Environmental Management’s CABs to develop a mutual understanding of their perspectives, values, and experiences related to risk and participatory decision making; (2) engaging members of those communities that would likely be part of any radioactive waste management discussions and who may provide valuable feedback for the development of the CBS process; and in the longer term, (3) engaging communities to foster knowledge sharing on understanding and definitions of risk and how risk is factored into community decision making. Our emphasis is to identify opportunities for improving risk communication frameworks, strategies, and decision making. The selection of a future FCISF site is likely to result in the creation of a structure similar to a CAB composed of members of the local community. We anticipate that these insights can help DOE NE learn from existing participatory risk communication structures in place at sites storing defenseor research-related SNF to understand best practices for fostering enduring and participatory relationships with future CABs. Within this paper, we (1) describe CRESP’s overall approach to assisting DOE NE with maturing the CBS process; (2) present a summary of preliminary phase 1 project results, including the development of a body of knowledge (describing available resources to support community engagement, risk communica tion, and participatory decision making) and community ecosystem information (leveraging demographic, social, economic, and environmental attribute mapping and advanced sentiment analysis of social media data); and (3) based on these results, provide observations for future research opportunities to support the development of CBS processes for radioactive waste management facilities, generally.

collaboration-based siting↗

Update on Parallel Process Execution in the Next Generation System Analysis Model

As of the end of 2021, 88,880 metric tons of heavy metal (MTHM) (44,741 MTHM in dry storage; 44,139 MTHM in wet storage) of spent nuclear fuel (SNF) were stored at various reactor sites across the United States [1]. The Office of Storage and Transportation in the Department of Energy is planning for the transportation, storage, and eventual disposal of SNF and high-level radioactive waste (HLW). To aid in this effort and inform decision-makers about the backend of the spent fuel cycle, systems analysis tools capable of analyzing the various options with respect to SNF and HLW management are being used as well as continuously improved to meet the evolving needs of the program. System analysts typically use these tools to vary underlying assumptions (shipping rates, allocation priority, available facilities, start dates, etc.) and study the implications of these changes on site clearance schedules, campaign costs, transportation infrastructure acquisition, etc.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Results of Re-evaluation of FEPs Related to Higher Fissile Content in HLW Glass at SRS

One of the objectives of the United States Department of Energy Office of Nuclear Energy’s Office of Spent Fuel and High-Level Waste Disposition is to better understand the technical bases, risks, and uncertainties associated with the safe and secure disposition of spent nuclear fuel and high-level radioactive waste. Domestic defense and research activities have generated a few thousand metric tons of spent nuclear fuel and high-level radioactive waste, much of which has been or will be processed and vitrified into high-level waste glass. The Nuclear Waste Policy Act 1982 makes the Department of Energy responsible for disposal of these materials.

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Assessment of Impurities in Neptunium Metal Using Double-Focusing SF-ICP-MS

In the transmutation process, minor actinides such as neptunium, americium and curium obtained from spent nuclear fuel are incorporated into new fuel that can be placed into a fast reactor, where these elements fission into products with shorter half-lives which reduces the health and safety hazards. In order to better understand the all the involved processes, fundamental physical properties of actinides should be investigated, and for that reason, high-purity materials need to be produced. One of the main challenges in characterizing these materials is the lack of pure metal available for study, and lack of analytical methods to determine the purity. Trans-uranic elements are challenging to study because of the need to be handled in specialized facilities, safeguards concerns and they are not easily purchased and/or produced. High-purity Np metal is being produced at the Materials and Fuels Complex (MFC) of Idaho National Laboratory (INL), in a process of direct oxide reduction developed by Squires and Lessing, with a target purity of 99.999%. The goal of this research is to develop methods to determine the impurities in the Np metal, with detection limits low enough to achieve this purity level. In this work, we have used several ICP-MS and ICP OES techniques, but the focus was the method development using a recently acquired Double-Focusing Sector Field Inductively Coupled Plasma Mass Spectrometer (SF-ICP-MS). To produce the Np metal, approximately 20 g of the neptunium oxide powder was calcined at 800 oC for 6 h in a magnesium oxide crucible, which is placed inside a stainless steel crucible, that goes into the Hot Uniaxial Press (HUP) furnace insert. After the calcining step, the Np oxide was mixed with approximately 9 g of calcium metal and 100 g of ultra-dry calcium chloride. The HUP furnace was heated in a ramp rate of 50 oC min -1 up to 1000 oC. When the reaction mixture reached 850 oC, a tantalum stirrer was activated at 100 rpm. After the tantalum stirrer was removed and the mixture cooled overnight, the MgO crucible was opened with a hammer and chisel to retrieve the Np metal button at the bottom of the reaction mixture. Prior to the dissolution process, the surface contaminants of the sample were removed. The Np button was washed in 10 mL of ultra-pure water for 1 min, with a faint reaction observed on the surface, indicating the presence of Ca metal. The water-washed sample was placed in a tube with 2 M HCl for only 3 s, as a vigorous reaction in the button surface was observed. The leachates were saved for further analysis. The sample was then rinsed with ultra-pure water and air-dried prior to dissolution. The cleaned sample was placed in a PTFE beaker and 25 mL of 6 M HCl and 150 µL of 24 M HF were added. After 10 minutes of reaction, the beaker was placed on a hot plate for 1 h and 30 min, followed by the addition of 1 mL of 16 M HNO3 and allowed to react for a further 1 h and 20 min. The sample was removed from heat and 10 mL of HNO3 16 M was added to react with the HCl. The beaker was covered and left for overnight reaction, producing a solution with approximately 20 mL of final volume. A Quadrupole Inductively Coupled Plasma Mass Spectrometer (Q-ICP-MS, Perkin Elmer ELAN DRC II Q-ICP-MS) was used to obtain qualitative mass spectra for the dissolution blank, each of the leachates and the dissolved sample. In the water leachate spectrum mainly Ca and Mg were observed, while the acid wash had a similar peak pattern to the one observed for the Np metal sample, indicating that the process not only removed surface contamination, but also reacted with the sample itself. In the Np material spectrum, Ta and actinides were identified as candidates for major contaminants, as the most intense peaks corresponded to the mass-to-charge ratios of these elements. Minor contaminants included Al, Au, Ca, Cr, Fe, I, Na, Nb, Pb, Sn, Ti, Tl, V, Y. The presence of impurities in the material could have originated from several sources, such as the materials used for the casting process, the use of impure NpO2 or contamination of reagents and materials used for the analysis. Some of the isotopes listed as potential contaminants are susceptible to spectral interferences, so Double-Focusing SF-ICP-MS (Nu Instruments Attom ES HR-ICP-MS) was used for the quantitative analysis. For each isotope determined, several items were evaluated: raw signal of a blank solution, sensitivity and determination coefficient of calibration curves, possibility of interferences and comparison of results obtained for at least two resolving powers (RP): 300 and 4000. The evaluation of the signal in a blank solution can indicate if a spectral interference from the plasma gas and/or the solvent was successfully resolved (commonly observed for 52Cr+/40Ar12C+, for example). It is important to highlight that a signal reduction is expected when a higher RP is used, as the ion beam transmission is decreased by the use of narrower slits

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Sister Rod Destructive Examinations (FY23) Appendix F: Cyclic Integrated Reversible-Bending Fatigue Tests

Spent nuclear fuel (SNF) assemblies must be shipped to other sites for processing and disposal. During shipment, the fuel is typically oriented horizontally, and the fuel rods are subject to periodic alternating loads related to the movement of the vehicle that results in alternating bending of the SNF fuel rods. The number of bending cycles is related to the length of the shipping route, with longer routes producing more cycles. Because it is well-known that cyclic loads can produce failures, even when the stress and strain imposed are below the material’s yield point, investigation of the SNF’s fatigue behavior is prudent. This report discusses the results of fatigue testing conducted at Oak Ridge National Laboratory (ORNL) using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) created by the US Department of Energy Office of Nuclear Energy for the High Burnup Spent Fuel Data Project and its sister rods. The results of the tests are compared with results obtained over the last decade using the same CIRFT for the US Nuclear Regulatory Commission (NRC).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Ground Test Technology Demonstration for Nuclear Thermal Propulsion Engines

Nuclear Thermal Propulsion (NTP) engines have been deemed a key technology to enable human missions to Mars due to their high efficiency, also known as specific impulse (Isp). Ground testing the NTP engine is critical in maturing the technology and increasing the design’s Technology Readiness Level (TRL), thus mitigating risk from NTP engine performance/operations. NTP engine development began with open-air ground testing through the Rover/NERVA program back in the 1960s when regulatory requirements were not as stringent as they are today. Due to the formation of regulatory bodies, increase in oversight and environmental requirements, the exhaust gas released from NTP engines must be captured or processed to gain approval in the case that the exhaust gas contains fission products from the nuclear fuel elements within the reactor. Ground test campaigns following Rover/NERVA focused on the certification of a full scale NTP engine, which led to exhaust processing systems such as the Rocket Exhaust Capture System (RECS) and Real Time (RT) exhaust processing. These systems were deemed favorable for regulatory compliance but have a high initial investment cost. Reassurance on the feasibility of ground testing and the NTP engine technology likely need to be achieved before NASA invests in these systems. Recently the objectives for NTP engine ground testing have shifted from certification testing of a full scale engine to demonstration testing of a subscale engine. The shift to demonstration testing allows for a shorter testing duration and a lower operational thrust (for demo testing purposes only) to demonstrate the NTP engine (5k-12.5k lbf). Due to these factors, the infrastructure, consumables, total footprint, and exhaust system complexity are able to be drastically reduced, thus reducing cost significantly. The High-pressure Exhaust Capture System (HECS) concept was designed for a NTP engine ground test demonstration. The HECS concept greatly reduces cost compared to previous concepts and suggests favorable regulatory acceptance due to its ability to capture all of the exhaust gas from the NTP engine.

Nuclear Thermal Propulsion↗

Nuclear Thermal Propulsion Engine Technology Demonstration Testing

Nuclear Thermal Propulsion (NTP) engines have been deemed a key technology to enable human missions to Mars due to their high efficiency, also known as specific impulse (Isp). Ground testing the NTP engine is critical in maturing the technology and increasing the design’s Technology Readiness Level (TRL), thus mitigating risk from NTP engine performance/operations. NTP engine development began with open-air ground testing through the Rover/NERVA program back in the 1960s when regulatory requirements were not as stringent as they are today. Due to the formation of regulatory bodies, increase in oversight and environmental requirements, the exhaust gas released from NTP engines must be captured or processed to gain approval in the case that the exhaust gas contains fission products from the nuclear fuel elements within the reactor. Ground test campaigns following Rover/NERVA focused on the certification of a full scale NTP engine, which led to exhaust processing systems such as the Rocket Exhaust Capture System (RECS) and Real Time (RT) exhaust processing. These systems were deemed favorable for regulatory compliance but have a high initial investment cost. Reassurance on the feasibility of ground testing and the NTP engine technology likely need to be achieved before NASA invests in these systems. Recently the objectives for NTP engine ground testing have shifted from certification testing of a full scale engine to demonstration testing of a subscale engine. The shift to demonstration testing allows for a shorter testing duration and a lower operational thrust (for demo testing purposes only) to demonstrate the NTP engine (5k-12.5k lbf). Due to these factors, the infrastructure, consumables, total footprint, and exhaust system complexity are able to be drastically reduced, thus reducing cost significantly. The High-pressure Exhaust Capture System (HECS) concept was designed for a NTP engine ground test demonstration. The HECS concept greatly reduces cost compared to previous concepts and suggests favorable regulatory acceptance due to its ability to capture all of the exhaust gas from the NTP engine.

Nuclear Thermal Propulsion↗

Characterization of Urbix Resources Natural Flake Graphite Material for Nuclear Applications (CRADA NFE-18-07340 Final Report)

Urbix Resources is a company specializing in graphite powder processing and purification. Urbix is looking to become the first domestic supplier of nuclear-grade graphite in the United States. This has led to a Cooperative Research and Development Agreement (CRADA) being established between the Oak Ridge National Laboratory (ORNL) and Urbix Resources, with the goal of characterizing a purified graphite feedstock supplied by Urbix Resources. The resulting data will then be used to determine the viability of a large-scale nuclear graphite production and qualification campaign. These characterization methods have been performed on the as-received graphite feedstock, a heat-treated batch of the feedstock, a milled batch of the feedstock, and a nuclear-grade graphite powder historically used in fuel production. The data from the heat-treated and milled feedstock will determine how different processing methods affect the powder’s physical properties, whereas the data from the nuclear grade graphite will provide information about the characteristics of a graphite previously accepted for nuclear applications. All of these powders have been characterized individually and as part of a resin-graphite blend (herein referred to as matrix) that has been used in past fuel experiments. The results of these experiments show that the Urbix powder and nuclear-grade graphite powder display notably different properties from one another in many cases, but this discrepancy may have been the result of the typical particle size of each powder. This hypothesis is supported by tests conducted on the milled powder, which showed a closer resemblance in properties to the nuclear-grade graphite than the as-received powder. The method for pressing the graphite matrix also had a noticeable effect on the measured properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Possible impacts of molten salt reactors on the International Monitoring System

Molten salt reactors (MSRs) are gaining support as many countries look for ways to increase power generation and replace aging nuclear energy production facilities. MSRs have inherently safe designs, are scalable in size, can burn transuranic wastes from traditional solid fuel nuclear reactors, can store excess heat in thermal reservoirs for water desalination, and can be used to produce medical isotopes as part of the real-time liquid-fuel recycling process. The ability to remove 135 Xe in real time from the fuel improves the power production in an MSR because 135 Xe is the most significant neutron-absorbing isotope generated by nuclear fission. Xenon-135, and other radioactive gases, are removed by sparging the fuel with an inert gas while the liquid fuel is recirculated from the reactor inner core through the heat exchangers. Without effective abatement technologies, large amounts of radioactive gas could be released during the sparging process. This work examines the potential impact of radioxenon releases on samplers used by the International Monitoring System (IMS) to detect nuclear explosions. Here, atmospheric transport simulations from seven hypothetical MSRs on different continents were used to evaluate the holdup time needed before release of radioxenon so IMS samplers would register few detections. Abatement technologies that retain radioxenon isotopes for at least 120 d before their release will be needed to mitigate the impacts from a molten salt breeder reactor used to replace a nuclear power plant. A holdup time of about 150 d is needed to reduce emissions to the average level of current nuclear power plants.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Structural Materials and Fuels for Space Power Plants

A fission reactor combined with Stirling convertor power generation is one promising candidate in on-going Fission Surface Power (FSP) studies for future lunar and Martian bases. There are many challenges for designing and qualifying space-rated nuclear power plants. In order to have an affordable and sustainable program, NASA and DOE designers want to build upon the extensive foundation in nuclear fuels and structural materials. This talk will outline the current Fission Surface Power program and outline baseline design options for a lunar power plant with an emphasis on materials challenges. NASA first organized an Affordable Fission Surface Power System Study Team to establish a reference design that could be scrutinized for technical and fiscal feasibility. Previous papers and presentations have discussed this study process in detail. Considerations for the reference design included that no significant nuclear technology, fuels, or material development were required for near term use. The desire was to build upon terrestrial-derived reactor technology including conventional fuels and materials. Here we will present an overview of the reference design, Figure 1, and examine the materials choices. The system definition included analysis and recommendations for power level and life, plant configuration, shielding approach, reactor type, and power conversion type. It is important to note that this is just one concept undergoing refinement. The design team, however, understands that materials selection and improvement must be an integral part of the system development.

Bowman, Cheryl↗

Uncovering Uranium Isotopic Heterogeneity of Fuel Pellets from the Fifth Collaborative Materials Exercise of the Nuclear Forensics International Technical Working Group

In 2017, the Nuclear Forensics International Technical Working Group (ITWG) organized their fifth 37 Collaborative Materials Exercise (CMX-5). The exercise samples were two uranium dioxide fuel pellets 38 manufactured from the same starting materials by different processes to have similar bulk isotopic 39 composition, but different spatial uranium isotopic distributions. Sets of identical materials were sent to 40 all participating laboratories, who then utilized their existing nuclear forensic capabilities to 41 independently analyse fuel pellets and identify similarities and differences of the materials’ 42 characteristics. The analytical methods used to probe the fuel pellets included ex situ, such as sectioning 43 or breaking up the pellets and analyzing dissolved pieces using inductively coupled plasma mass 44 spectrometry (ICP-MS), analyzing particles collected from intact or fragmented pellets by secondary ion 45 mass spectrometry (SIMS), as well as in situ methods, such as laser ablation coupled with ICP-MS, 46 autoradiography and nanoSIMS. In this paper we present the results of these independent analyses and 47 compare the capabilities of those nuclear forensic analytical methods to uncover details of the isotopic 48 heterogeneity of uranium fuel pellets.

Nuclear Forensic Analysis of Uranium Fuel Pellets,↗

An Alternative Conceptual Model for the Spent Nuclear Fuel–Water Interaction in Deep Geologic Disposal Conditions

For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO 2 matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H 2 may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (10 -7 year -1 ) is over estimated because it assumes that there is net UO 2 oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel–water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.

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Sentiment analysis of the United States public support of nuclear power on social media using large language models

This study utilized large language models (LLMs) to analyze public sentiment in the United States (US) regarding nuclear power on social media, focusing on X/Twitter, considering climate change challenges and advancements in nuclear power technology. Approximately, 1.26 million nuclear tweets from 2008–2023 were examined to fine-tune LLMs for sentiment classification. We found the crucial role of accurate data labeling for model performance, with potential implications for a 15% improvement, achieved through high-confidence labels. LLMs demonstrated better performance compared to traditional machine learning classifiers, with reduced susceptibility to overfitting and up to 96% classification accuracy. LLMs are used to segment the US public tweets into policy and energy-related categories, revealing that 68% are politically themed. Policy tweets tended to convey negative sentiment, often reflecting opposing political perspectives and focusing on nuclear deals and international relations. Energy-related tweets covered diverse topics with predominantly neutral to positive sentiment, indicating broad support for nuclear power in 48 out of 50 US states. The US public positive sentiments toward nuclear power stemmed from its high power density, reliability regardless of weather conditions, environmental benefits, application versatility, and recent innovations and advancements in both fission and fusion technologies. Negative sentiments primarily focused on waste management, high capital costs, and safety concerns. The neutral campaign highlighted global nuclear facts and advancements, with varying tones leaning towards positivity or negativity. An interesting neutral theme was the advocacy for the combined use of renewable and nuclear energy to attain net-zero goals.

Energy & Fuels↗

Exploring the influence of transition metals on f-element bonding

The United States aims to triple its nuclear energy production by 2050, which will result in increased uranium usage and spent nuclear fuel generation. This highlights the need for a comprehensive understanding of actinide coordination chemistry, which is crucial for the extraction, processing, purification, and fabrication of uranium-based fuels. Moreover, it is vital for the reprocessing or safe disposal of nuclear waste and effective remediation efforts. Achieving this successfully requires an in-depth understanding of f-electron behavior, as “the role of 5f electrons in bond formation remains a fundamental topic in actinide chemistry”. Introducing a second metal into the system can increase structural dimensionality and diversify structural architecture. Heterometallic systems can also alter material properties, such as magnetic and spectroscopic characteristics, luminescence, and actinide mobility. Additionally, secondary transition metals, even when present only in the second coordination sphere and not directly coordinated, can influence the electron density at the actinyl metal center. In this study, uranium heterometallic single crystals were synthesized by incorporating transition metals such as iron(III), iron(II), nickel(II), manganese(II), copper(I), and cobalt(II). The crystals were formed using 2,6-pyridine dicarboxylic acid and other structurally similar ligands with varying functional groups. The synthesized crystals were analyzed using an extensive array of analytical and computational characterization techniques, including single crystal X-ray diffraction, Raman and infrared spectroscopy, and density functional theory calculations.

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