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DOE OSTI2022

A resonant electro-physical transmutation method for water purification

We present an electro-transmutation (ET) method of water purification using the Quantum Kinetic Fusor™. Electrolysis experiments were performed with residential well water and seawater coupled with two different electrode materials—stainless steel and aluminum—to test the effectiveness of ET in removing contaminants and altering the physiochemistry (e.g., pH, alkalinity) and isotopic composition of water. ET effectively removed contaminants at power inputs as low 0.142 mA cm –2 , i.e., 1.9–4.2 watts, which is substantially lower than other electrochemical techniques. The ET process also reduces, and in some cases eliminates, electrode surface area degradation. This outcome is outside the field of traditional Faraday electrolysis, further distinguishing the ET process as potentially more cost-effective than other water purification methods. Importantly, the experiments reveal ET is an exceptionally low input energy generator. This finding is based on the presence of exotic x-rays (Bremsstrahlung) emissions during operation (1–20 KeV).

42 ENGINEERING↗

DOE OSTI2022

13C-labeling reveals how membrane lipid components contribute to triacylglycerol accumulation in Chlamydomonas

Abstract Lipid metabolism in microalgae has attracted much interest due to potential utilization of lipids as feedstocks for biofuels, nutraceuticals, and other high-value compounds. Chlamydomonas reinhardtii is a model organism for characterizing the synthesis of the neutral lipid triacylglycerol (TAG), from which biodiesel is made. While much of TAG accumulation under N-deprivation is the result of de novo fatty acid (FA) synthesis, recent work has revealed that approximately one-third of FAs, especially polyunsaturated FAs (PUFAs), come from preexisting membrane lipids. Here, we used 13C-isotopic labeling and mass spectrometry to analyze the turnover of glycerol backbones, headgroups, FAs, whole molecules, and molecular fragments of individual lipids. About one-third of the glyceryl backbones in TAG are derived from preexisting membrane lipids, as are approximately one-third of FAs. The different moieties of the major galactolipids turn over synchronously, while the FAs of diacylglyceryltrimethylhomoserine (DGTS), the most abundant extraplastidial lipid, turn over independently of the rest of the molecule. The major plastidic lipid monogalactosyldiacylglycerol (MGDG), whose predominant species is 18:3α/16:4, was previously shown to be a major source of PUFAs for TAG synthesis. This study reveals that MGDG turns over as whole molecules, the 18:3α/16:4 species is present in both DAG and TAG, and the positional distribution of these PUFAs is identical in MGDG, DAG, and TAG. We conclude that headgroup removal with subsequent acylation is the mechanism by which the major MGDG species is converted to TAG during N-deprivation. This has noteworthy implications for engineering the composition of microalgal TAG for food, fuel, and other applications.

Plant Sciences↗

DOE OSTI2024

Krypton-85 chronometry of spent nuclear fuel

We describe the use of the radionuclide 85 Kr, which is produced by nuclear fission and has a half-life of 10.76 years, to determine the age of spent nuclear fuel. The method is based on mass-spectrometric measurement of the relative abundance of fissiogenic Kr isotopes extracted from a fuel sample, and we show that it can be applied to micron-scale particles of spent fuel that are analogous to particles that have been released into the environment from various nuclear facilities in the past. 85 Kr chronometry is potentially valuable for identifying and attributing nuclear materials, grouping samples into collections of common origin that can be used to reconstruct the origin and irradiation history of the material, and verifying declared nuclear activities in the context of international monitoring programs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Structural, Criticality, and Radiation Dose Calculations to support SNF Loading into a DOE Standard Canister

The DOE Standard Canister Demonstration Project includes the development of an internal support structure (ISS) for the 4.6-m long and 45.7-cm diameter canister. This study presents structural evaluations, criticality safety assessments, and dose rate calculations conducted within the scope of the ISS design process to support smooth canister loading operations and to ensure safe storage, transportation, and disposal of Peach Bottom 1 Core II (PB2) and Fort St. Vrain (FSV) SNF, currently stored at the CPP-603 facility of the INL. The ISS includes a 316L stainless steel basket that holds twelve PB2 SNF rods. Further, six individual, 78.7-cm long, 316L stainless steel columns are equally spaced and welded to the inner canister wall at the lower end of the shell. These columns represent the FSV basket and can hold one FSV SNF element. An A92014 T6 aluminum spacer disc is bolted to the bottom plate of the PB2 basket to vertically restrain the FSV SNF element after the PB2 basket is placed inside the canister on top of the FSV basket. The structural evaluations of the ISS followed applicable ASME BPVC.III.3 guidelines and included finite element (FE) analyses of the PB2 basket structure; analyses of welds and bolds; buckling analyses of selected components, and acceptability assessments of the expected basket deformations under loading operations. The criticality safety assessments used the Monte Carlo N-Particle (MCNP) software architecture version 6.2 including ENDF/B-V continuous energy cross-section libraries, considering intact SNF in a single storage overpack or two multi-storage overpack configurations, and intact or failed SNF configured for disposal. The dose rate computations are based on source terms taken from the DOE Spent Fuel Database. The isotopic composition was decay corrected for the year 2022 using the ORIGEN module in the SCALE suite. A 19-group photon spectrum and 27-group neutron-source spectra were generated and used in MCNP to calculate estimated dose-equivalent rates, both on DOE Standard Canister contact and at a radial distance of from the canister surface. The results of this study indicate a structurally sound system that can uphold its criticality safety functions throughout its intended operational phases. Further, they increase confidence that sufficient radiological protection is technically achievable.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Identification of $^{3}$He–$^{3}$H clusters in the $^{6}$Li+$^{89}$Y experiment using particle-$\gamma$ coincidence measurement

The 6 Li+ 89 Y experiment was performed to explore the reaction mechanism induced by a weakly bound nucleus 6 Li and its cluster configuration. Here, the particle-$\gamma$ coincidence method was used to identify the different reaction channels. The $\gamma$-rays coincident with 3 He/ 3 H indicate that the 3 H/ 3 He stripping reaction plays a significant role in the formation of Zr/Nb isotopes. The obtained results support the existence of a 3 He- 3 H cluster in 6 Li. Direct and sequential transfer reactions are adequately discussed, and the FRESCO code is used to perform precise finite-range cyclic redundancy check calculations. In the microscopic calculation, direct cluster transfer is more predominant than sequential transfer in 3 H transfer. However, the direct cluster transfer is of comparable magnitude to the sequential transfer in the 3 He transfer.

CRC calculations↗

DOE OSTI2022

High Flux Isotope Reactor Low Enriched Uranium U-10Mo Fuel Design Parameters

Activities to convert the HFIR from HEU to LEU are ongoing as part of the US Department of Energy (DOE) National Nuclear Security Administration (NNSA) nuclear nonproliferation mission. Design activities to study the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a uranium-molybdenum (U-10Mo) monolithic alloy fuel. This high-density alloy contains 90 wt % uranium and 10 wt % molybdenum and has a uranium density of 15.318gU/cm 3 . The goal of these studies is to generate several candidate HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. Recent advancements in modeling and simulation tools and design methods enabled a thorough analysis of the available design space with U-10Mo fuel. A surrogate model used this analysis as training data to quickly determine the performance of a design given specific design parameters. An optimization module used this surrogate model to quickly search this multidimensional search space given specific desired performance characteristics. This approach was made possible by the large available design space with U-10Mo fuel. Shift, a Monte Carlo tool optimized for high-performance computing (HPC) architectures, was used for faster calculation and better data management for reactor physics simulations. Once most of these design studies were complete, a new suite called the Python HFIR Analysis and Measurement Engine (PHAME) was developed to connect all fuel design analysis steps, making design studies more efficient and reproducible. The post-processing capabilities of these new tools are leveraged for the information provided herein. Leveraging these tools, several candidate fuel designs were selected with varying levels of feature complexity and reactor performance. This report provides design feature details for four selected HFIR LEU U-10Mo fuel designs and their corresponding performance and safety metrics. Nominal best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities, are provided. Simulations show that the high uranium density of U-10Mo fuel provides a large potential design space that enables various LEU designs to meet HEU core performance metrics and safety requirements with a power increase from 85 MW (HEU) to 95 MW or 100 MW (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

The Baghdad Atlas: A relational database of inelastic neutron-scattering (n,n ' γ) data

A relational database has been developed based on the original (n,n'γ) work carried out by A. M. Demidov et al., at the Nuclear Research Institute in Baghdad, Iraq (Demidov et al., 1978) for 105 independent measurements comprising 76 elemental samples of natural composition and 29 isotopically-enriched samples. The information from this Atlas includes: γ-ray energies and relative intensities; nuclide and level data corresponding to the residual nucleus and meta data associated with the target sample that allows for the extraction of the flux-weighted (n,n'γ) cross sections for a given transition relative to a defined value. The optimized angular-distribution-corrected fast-neutron flux-weighted partial γ-ray cross section for the production of the 846.8-keV 21+→0gs+γ-ray transition in 56Fe, determined to be $\langle$σγ$\rangle$=143(29) mb, is used for this purpose. However, different values for the adopted cross section can be readily implemented to accommodate user preference based on revised determinations of this quantity. The Atlas (n,n'γ) data has been compiled into a series of CSV-style ASCII data sets and a suite of Python scripts have been developed to build and install the database locally. The database can then be accessed directly through the SQLite engine, or using alternative methods such as the Jupyter Notebook Python-browser interface. Several examples exploiting different interaction methodologies are distributed with the complete software package.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

New thermal decomposition pathway for TATB

Abstract Understanding the thermal decomposition behavior of TATB (1,3,5-triamino-2,4,6-trinitrobenzene) is a major focus in energetic materials research because of safety issues. Previous research and modelling efforts have suggested benzo-monofurazan condensation producing H 2 O is the initiating decomposition step. However, early evolving CO 2 (m/z 44) along with H 2 O (m/z 18) evolution have been observed by mass spectrometric monitoring of head-space gases in both constant heating rate and isothermal decomposition studies. The source of the CO 2 has not been explained, until now. With the recent successful synthesis of 13 C 6 -TATB ( 13 C incorporated into the benzene ring), the same experiments have been used to show the source of the CO 2 is the early breakdown of the TATB ring, not adventitious C from impurities and/or adsorbed CO 2 . A shift in mass m/z 44 (CO 2 ) to m/z 45 is observed throughout the decomposition process indicating the isotopically labeled 13 C ring breakdown occurs at the onset of thermal decomposition along with furazan formation. Partially labeled (N 18 O 2 ) 3 -TATB confirms at least some of the oxygen comes from the nitro-groups. This finding has a significant bearing on decomposition computational models for prediction of energy release and deflagration to detonation transitions, with respect to conditions which currently do not recognize this oxidation step.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Structural, Criticality, and Radiation Dose Calculations to Support SNF Loading into a DOE Standard Canister

The DOE Standard Canister Demonstration Project includes the development of an internal support structure (ISS) for the 4.6-m-long and 45.7-cm-diameter canister. This study presents structural evaluations, criticality safety assessments, and dose rate calculations conducted within the scope of the ISS design process to support smooth canister loading operations and to ensure safe storage, transportation, and disposal of Peach Bottom 1 Core II (PB2) and Fort St. Vrain (FSV) SNF, currently stored at INL’s CPP-603 facility. The ISS includes a 316L stainless-steel basket that holds 12 PB2 SNF rods. The PB2 basket rests on top of six individual, 78.7-cm-long, 316L stainless-steel columns that are equally spaced and welded to the inner canister wall at the lower end of the shell. These columns represent the FSV basket and can hold one FSV SNF element. An A92014 T6 aluminum spacer disc is bolted to the bottom plate of the PB2 basket to vertically restrain the FSV SNF element after the PB2 basket is placed inside the canister above the FSV basket. The structural evaluations of the ISS followed applicable ASME BPVC.III.3 guidelines and included finite element (FE) analyses of the PB2 basket structure; analyses of welds and bolds; buckling analyses of selected components, and acceptability assessments of the expected basket deformations under loading operations. The criticality safety assessments used the Monte Carlo N-Particle (MCNP) software architecture Version 6.2, including ENDF/B-V continuous-energy cross-section libraries, considering intact SNF in a single storage overpack or two multistorage overpack configurations and intact or failed SNF configured for disposal. The dose rate computations are based on source terms taken from the DOE Spent Fuel Database. The isotopic composition was decay corrected for the year 2022 using the ORIGEN module in the SCALE suite. A 19-group photon spectrum and a 27-group neutron-source spectrum were generated and used in MCNP to calculate estimated dose-equivalent rates, both on DOE Standard Canister contact and at a radial distance of 1 m from the canister surface. The results of this study indicate a structurally sound system that can uphold its criticality safety functions throughout its intended operational phases. Furthermore, this study provides confidence that sufficient radiological protection is technically achievable.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Autotrophic and mixotrophic metabolism of an anammox bacterium revealed by in vivo 13 C and 2 H metabolic network mapping

Anaerobic ammonium-oxidizing (anammox) bacteria mediate a key step in the biogeochemical nitrogen cycle and have been applied worldwide for the energy-efficient removal of nitrogen from wastewater. However, outside their core energy metabolism, little is known about the metabolic networks driving anammox bacterial anabolism and use of different carbon and energy substrates beyond genome-based predictions. Here, we experimentally resolved the central carbon metabolism of the anammox bacterium Candidatus ‘Kuenenia stuttgartiensis’ using time-series 13 C and 2 H isotope tracing, metabolomics, and isotopically nonstationary metabolic flux analysis. Our findings confirm predicted metabolic pathways used for CO 2 fixation, central metabolism, and amino acid biosynthesis in K. stuttgartiensis , and reveal several instances where genomic predictions are not supported by in vivo metabolic fluxes. This includes the use of the oxidative branch of an incomplete tricarboxylic acid cycle for alpha-ketoglutarate biosynthesis, despite the genome not having an annotated citrate synthase. We also demonstrate that K. stuttgartiensis is able to directly assimilate extracellular formate via the Wood–Ljungdahl pathway instead of oxidizing it completely to CO 2 followed by reassimilation. In contrast, our data suggest that K. stuttgartiensis is not capable of using acetate as a carbon or energy source in situ and that acetate oxidation occurred via the metabolic activity of a low-abundance microorganism in the bioreactor’s side population. Together, these findings provide a foundation for understanding the carbon metabolism of anammox bacteria at a systems-level and will inform future studies aimed at elucidating factors governing their function and niche differentiation in natural and engineered ecosystems.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI1968

Nuclear Physics Made Very, Very Easy

The fundamental approach to nuclear physics was prepared to introduce basic reactor principles to various groups of non-nuclear technical personnel associated with NERVA Test Operations. NERVA Test Operations functions as the field test group for the Nuclear Rocket Engine Program. Nuclear Engine for Rocket Vehicle Application (NERVA) program is the combined efforts of Aerojet-General Corporation as prime contractor, and Westinghouse Astronuclear Laboratory as the major subcontractor, for the assembly and testing of nuclear rocket engines. Development of the NERVA Program is under the direction of the Space Nuclear Propulsion Office, a joint agency of the U. S. Atomic Energy Commission and the National Aeronautics and Space Administration. This report is being reprinted for use in the U. S. Atomic Energy Commission and National Aeronautics and Space Administration educational and technology utilization programs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Qualitative and quantitative analysis of neutron irradiation effects in SiC/SiC composites using X-ray computed tomography

Silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composites are candidate materials for cladding of light water reactor (LWR) fuels. Loss of fission product gas retention due to the formation of microcrack networks is considered a potential failure mechanism for SiC/SiC-cladded fuels. In this work, a variety of SiC/SiC composite tubes were irradiated with and without an LWR-relevant radial heat flux in the High Flux Isotope Reactor, followed by detailed characterization with X-ray computed tomography (XCT). This first set of XCT data for neutron-irradiated samples confirmed that the internal stresses arising from a combination of temperature gradients and irradiation-induced swelling act as the primary driver for cracking. Consequently, while the observed cracking patterns varied depending on the tube architectures, the sharp edges of relatively large pores were found to be the common stress concentrator. These findings are useful to help improve the design and manufacturing of SiC/SiC fuel claddings for reduced failure probability.

42 ENGINEERING↗

DOE OSTI2026

Ceramic Composite Experimental Testing Status

Over recent years, ceramic matrix materials such as SiC–SiC and C–C have been gaining interest for use in fusion reactors, light water reactors (LWRs), and high-temperature reactors (HTRs). These materials are good candidates to operate in very high temperature and moderate to high radiation environments. The evaluation of composite materials, in general, is challenging because of variations in precursor materials, variations in the fabrication process across fabricators, and the wide range of potential fiber architectures, to name a few. However, the need to evaluate neutron-irradiated properties adds another layer of complexity, which includes cost, timeline, and specimen size limitations (often associated with irradiation testing). A qualification methodology for the use of ceramic composites is provided in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III-5-HHB. The methodology is supported by ASTM International (ASTM) guides, which provide a pathway to accomplish this effort. Part of the qualification strategy is for the designer to collect material property data on environmental conditions representative of its design envelope. These data include irradiation effects. This report presents an experimental study and test campaign developed to partially address this gap by providing initial mechanical and physical property data required for design. A variety of different materials using different manufacturing techniques are considered as part of this campaign. The test plan suggests performing a screening or partial irradiation study to assist the designer during the material selection process. The designer can then perform a more comprehensive qualification study if the material performance is promising. This work focuses on the status of the specimen preparations (machining of samples), the current test methods and failure analysis as well as the preparation of irradiation vehicles for the irradiation campaign. The irradiation will be performed at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) and at Idaho National Laboratory (INL) in the Advanced Test Reactor (ATR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

High Flux Isotope Reactor Low-Enriched Uranium Low Density Silicide Fuel Design Parameters

High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) to low-enriched uranium (LEU) conversion activities are ongoing as part of the Department of Energy (DOE) National Nuclear Security Administration (NNSA)’s nuclear nonproliferation mission. Design activities studying the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a low density uranium-silicide dispersion (U 3 Si 2 -Al) fuel, which has a uranium density of 4.8 gU/cm 3 . The goal of these studies is to generate several HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. The documented designs will serve as references for fuel fabrication and qualification activities. Recent advancements in modeling and simulation tools enable quick prototyping of fuel designs. Shift, a Monte Carlo neutron transport and depletion tool optimized for high-performance computing (HPC) architectures, is used for efficient fuel cycle and performance metrics calculations. The HFIR Steady State Heat Transfer Code (HSSHTC) is used to vet the thermal safety margin. Also, a new automation tool that connects all fuel design analysis steps, named Python HFIR Analysis and Measurement Engine (PHAME), has been developed to expedite the design study in an efficient and reproducible manner. Leveraging these tools, several candidate fuel designs were selected for varying fabrication complexity. This report provides design feature details for four selected HFIR LEU low density U 3 Si 2 -Al fuel designs and their corresponding performance and safety metrics. Nominal, best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities are provided for candidate fuel designs relevant to framing irradiation experiments to support fuel qualification efforts. Simulations show that the low density U 3 Si 2 -Al, with design features to enhance safety, can meet HEU core performance metrics and safety requirements if the reactor power is increased from 85 MW (HEU) to 95 MW (LEU) and if the active fuel length is increased from 50.80 cm (HEU) to 55.88 cm (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Quantitative near-field water–air spray measurements at elevated pressures by neutron radiography imaging

Extensive experimental research on high-pressure spray has been conducted for decades to deepen our understanding and optimize its use in transportation, aviation, and propulsion applications; however, the near-field and in-nozzle flow characteristics are not fully understood. Dense near-field spray is among the most challenging diagnostic tasks since light is severely scattered and diffused by the liquid droplets and columns. In this work, the near-field spray and in-nozzle flow characteristics of an aeration nozzle at elevated pressures were characterized by neutron radiography imaging at the Oak Ridge National Laboratory High Flux Isotope Reactor. Neutron imaging benefits via strong penetration depths for some metals (i.e., aluminum, lead, and steel) and is sufficiently sensitive to detection of light elements, especially for hydrogen-based molecules, due to the large incoherent scattering cross section of neutrons. Both two-dimensional snapshots of the near-field spray and a three-dimensional tomographic scan of the nozzle geometry and in-nozzle water were obtained. This work provides new quantitative characterization of practical metal nozzle geometry for accurate boundary conditions, internal flow patterns inside the nozzle, and high-pressure spray flows. In conclusion, the findings may be used to improve performance and operating conditions of transportation vehicles and propulsion systems.

42 ENGINEERING↗

DOE OSTI2020

$β$-delayed-neutron studies of 135,136 Sb and 140 I performed with trapped ions

In this work, Beta-delayed-neutron $(βn)$ spectroscopy was performed using the Beta-decay Paul Trap and an array of radiation detectors. The $βn$ branching ratios and energy spectra for 135,136 Sb and 140 I were obtained by measuring the time of flight of recoil ions emerging from the trapped ion cloud. These nuclei are located at the edge of an isotopic region identified as having $βn$ branching ratios that impact the $r$-process abundance pattern around the A ≈ 130 peak. For 135,136 Sb and 140 I, $βn$ branching ratios of 14.6(13)%, 17.6(29)%, and 7.6(28)% were determined, respectively. The $βn$ energy spectra obtained for 135 Sb and 140 I are compared with results from direct neutron measurements, and the $βn$ energy spectrum for 136 Sb has been measured for the first time.

42 ENGINEERING↗

DOE OSTI2024

Specification of FIPD Fission Gas Chemistry Data

The current FIPD library contains two main sets of fission gas chemistry data. The first set is data collected during the Integral Fast Reactor (IFR) program from 1984 to 1994, using a gas mass spectrometry system located in the Analytical Laboratory (AL) at Argonne National Laboratory-West (Argonne-West). Throughout this period, numerous fission gas release and chemistry datasets were gathered from a variety of metallic fuel pins. The fission gas was sampled by the Gas Assay, Sample and Recharge (GASR) System in the Hot Fuel Examination Facility (HFEF) and transferred to the AL to perform gas composition and isotopic abundance analysis. The second set is data collected after the IFR program. The fission gas samples were also collected by the GASR system at the HFEF, but analyzed using a similar gas mass spectrometer located in Pacific Northwest National Laboratory (PNNL). Many fuel pins irradiated in Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF) were measured, including the fuel pins for the MFF series of experiments, designed to qualify metal fuel for use as driver fuel in the FFTF and X496 experiment. For either set of data, fission gas was sampled with the gas sampling line in GASR using sample bottles after the capsule/element volume has been determined and the system is still full of radioactive gas. The sample bottles were then transferred to the sample packaging cylinder or an approved storage location pending transfer to the AL or prepared for shipment to another laboratory (such as PNNL) for analysis of the collected gas as directed on the GASR data form, other approved form. The receiving laboratories (AL or PNNL) required their Analytical Service Request form to be completed prior to sample transfer. Typical sample transfer processes were initiated at HFEF by the principal or process engineer. The laboratories performing the analyses (AL or PNNL) use the sample bottle numbers as well as a sample number produced by the respective laboratory. HFEF and the responsible experimenter tracked the sample using the analysis number, the gas bottle number, and the fuel pin number.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2025

RIMS analysis of isotopically tagged uranium particles with application to Intentional Forensics

Traditional nuclear forensics approaches leverage isotopic measurements for characterizing nuclear forensics signatures in fuel cycle materials. To aid in provenance assessment, isotopically perturbed transition metal taggants can be added to fuel cycle materials. We present a case study for examining natural uranium oxide powder that was tagged with isotopically perturbed Mo before and after irradiation using resonance ionization mass spectrometry. In conclusion, this method accurately and precisely measures taggant and uranium compositions, rapidly enabling the ability to clearly discriminate between tagged particles and other material with application for robust assessment of material provenance.

Forensic Anthropology↗