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DOE OSTI2020

Leveraging Machine Learning Capabilities for the Characterization of Irradiated Uranium: A Case Study of Analysis Methods for Nuclear Safeguards and Nuclear Forensics

Nondestructively determining the initial enrichment of irradiated uranium is a complex and laborious multivariable problem due to the presence of fission products. This work demonstrates the capabilities of machine learning to analyze gamma-ray spectral data to determine initial enrichment without knowledge of the decay time of the sample. The approach developed is agnostic to the particular scenario and is applicable to a wide variety of applications in nuclear forensics and nuclear safeguards. We irradiated 5 mg uranium standard reference materials at discrete enrichment values ranging from 0.02% to 97% 235 U (weight percent) in UT Austin’s Nuclear Engineering Teaching Laboratory TRIGA Mark II 1.1 MW research reactor, allowed each to decay for 8 hours, and then measured each sample via gamma-ray spectrometry for 50 hours post-irradiation yielding 1,400 individual gamma-ray spectra discretized into 8,192 energy bins. We then trained decision trees models to analyze individual gamma-ray spectra and estimate the associated initial enrichment without knowledge of the time since end of irradiation. We evaluated the performance of the models with a reserved test set not used for training or calibrating the model. A decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 2.3% (weight percent 235 U). Next, we implemented a principal component analysis pre-processing routine of the gamma-ray spectrometry data to reduce the dimensionality of the dataset from 8,192 channels in the spectrum to 10 principal components while retaining over 99% of the inherent variance in the data. Decision tree models constructed with these data demonstrated decreased mean absolute error in enrichment determination, reduced computational time, and decreased complexity. A single decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 0.05% (weight percent 235 U). Furthermore, we analyzed these models with learning curves to ensure that overfitting did not occur. The capabilities provided by these models can be naturally extended to other application-focused measurements in the fields of nuclear safeguards, nuclear forensics, and nuclear non-proliferation.

Drescher, Adam↗

DOE OSTI2025

Solute Redistribution During Aging in a Laser-Based Powder Bed Fusion-Produced Fe-18Ni-9Co-5Mo-Ti Maraging Steel

Solute distribution during the laser-based powder bed fusion of metals process is highly complex because of solidification. This study builds on a previous investigation of how solute heterogeneities across different scales influence phase distribution and aging physical phenomena in a Fe-18Ni-9Co-5Mo-Ti maraging steel (Santana et al. in Addit Manuf 104494, 2024). Previous work using energy-dispersive X-ray spectroscopy revealed heterogeneous Ni and Mo enrichment at intercellular regions. Using X-ray diffraction, 8 ± 3% retained austenite was quantified in as-built samples. Additionally, electron backscatter diffraction analysis exposed the location of retained austenite along cell boundaries. In this work, scanning transmission electron microscopy results revealed the presence of non-enriched retained austenite in the as-built condition. With increased aging temperature, Ni and Mo enrichment in austenite increased, especially at the austenite/martensite boundary. Solute diffusion was simulated in DICTRA Thermo-Calc® software to assess solute redistribution between the matrix and second phases and to comprehend whether elemental diffusion processes are interconnected. Three different scenarios were considered: (1) solute diffusion in the matrix; (2) diffusion with enriched austenite in the intercellular region; (3) diffusion with non-enriched austenite in the intercellular region. Results showed that Ni partitioning controls austenite growth and that boundary regions exhibit higher element enrichment than the core.

42 ENGINEERING↗

DOE OSTI2025

Understanding the effect of minor alloying elements on helium bubble formation in ferritic-martensitic steels

Ferritic-martensitic steels are promising structural materials for advanced nuclear reactors. To minimize long-term radioactivity, reduced-activation ferritic-martensitic steels have been developed by substituting high-activation elements like Ni and Mo with low-activation elements such as W. However, the impact of these alloying modifications on helium bubble formation, which plays a key role in material swelling, remains unclear. Here, in this study, we compared helium bubble formation in ferritic-martensitic steel T91 and reduced-activation ferritic-martensitic steel F82H. Both materials were irradiated with sequential 100 keV, 150 keV, and 200 keV helium ions to a dose of 0.5 dpa and a helium concentration of 9,000 appm at 500°C. The helium bubbles in F82H exhibited a larger average size and a lower density than those in T91, suggesting differences in minor alloying elements may influence the bubble growth. Here, to investigate the effects of these alloying elements, we characterized radiation-induced segregation near bubbles and grain boundaries. Prominent Ni-Mn-Si enriched clusters were found near bubbles in T91, while only Mn-Si enriched clusters were found near bubbles in F82H. In addition, the obvious Cr enrichment near grain boundaries was absent around bubbles in both steels. The different segregation trends among elements revealed the variations in element diffusion mechanisms and the different sink biases between bubbles and grain boundaries. Cr enrichment near grain boundaries is mostly driven by interstitial-mediated diffusion. However, since bubble growth relies on net vacancy flux, vacancy-mediated diffusion plays a dominant role in controlling element segregation near bubbles. Therefore, Cr enrichment was not found near bubbles. Because of preferential vacancy-drag diffusion for Ni, Si and Mn, these elements were enriched near bubbles. Due to the strong binding energies of vacancies with these solute atoms, the vacancy diffusivity can be reduced near these solutes. Therefore, the more prominent Ni-Si-Mn clustered near helium bubbles in T91 lead to stronger suppression of helium bubble growth compared to F82H.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Effects of experimental nitrogen deposition on soil organic carbon storage in Southern California drylands

Atmospheric nitrogen (N) deposition is enriching soils with N across biomes. Soil N enrichment can increase plant productivity and affect microbial activity, thereby increasing soil organic carbon (SOC), but such responses vary across biomes. Drylands cover ~45% of Earth's land area and store ~33% of global SOC contained in the top 1 m of soil. Nitrogen fertilization could, therefore, disproportionately impact carbon (C) cycling, yet whether dryland SOC storage increases with N remains unclear. To understand how N enrichment may change SOC storage, we separated SOC into plant–derived, particulate organic C (POC), and largely microbially derived, mineral–associated organic C (MAOC) at four N deposition experimental sites in Southern California. Theory suggests that N enrichment increases the efficiency by which microbes build MAOC (C stabilization efficiency) if soil pH stays constant. But if soils acidify, a common response to N enrichment, then microbial biomass and enzymatic organic matter decay may decrease, increasing POC but not MAOC. We found that N enrichment had no effect on C fractions except for a decrease in MAOC at one site. Specifically, despite reported increases in plant biomass in three sites and decreases in microbial biomass and extracellular enzyme activities in two sites that acidified, POC did not increase. Furthermore, microbial C use and stabilization efficiency increased in a non–acidified site, but without increasing MAOC. Instead, MAOC decreased by 16% at one of the sites that acidified, likely because it lost 47% of the exchangeable calcium (Ca) relative to controls. Indeed, MAOC was strongly and positively affected by Ca, which directly and, through its positive effect on microbial biomass, explained 58% of variation in MAOC. Finally, long–term effects of N fertilization on dryland SOC storage appear abiotic in nature, such that drylands where Ca–stabilization of SOC is prevalent and soils acidify, are most at risk for significant C loss.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2022

Impacts of LEU+ and ATF on Fresh Fuel Storage Criticality Safety

The use of increased fuel enrichment, which is still in the realm of low-enriched uranium (LEU) fuel, has been of interest to commercial light water reactor operators as part of the next iteration in fuel cycle technological advances and research and development. Using increased enrichment fuel, or high-assay LEU (HALEU), in power plants has clear benefits for being able to load cores with additional power-producing fuel. Although HALEU enrichments can range up to 20%, the more guarded approach of investigating enrichments above current fuels within 10% enrichment is referred to as LEU plus (LEU+) to reflect the less drastic change in operating conditions and requirements and similarity to current fuel cycles. Of additional interest and increasing maturity is the incorporation of accident-tolerant fuel (ATF) concepts, which are also applicable to the current fleet. This class of technologies involves changes such as cladding (e.g., chromium coating or FeCrAl) and fuel composition (e.g., chromia dopant) alterations to demonstrate improved fuel performance under accident scenarios. The ability to properly store fuel before and after residence time in the reactor is crucial to plant operation. Typically, this is done in either a new fuel vault (NFV) or spent fuel pool (SFP). Storing, loading, and unloading dozens of fuel assemblies within the same general area provides opportunities for obvious criticality concerns. These concerns are addressed with regulations to the subcriticality margin that the NFV and SFP must maintain in certain conditions. Adopting LEU+ fuel results in inherent reactivity increases, which are extremely relevant for safe fuel storage. Therefore, a clear understanding of the effects of LEU+ fuel and ATF on criticality safety margins to regulatory limits is required, as well as an understanding of the degree of absorber crediting under normal and accident conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Culturing of “Unculturable” Subsurface Microbes: Natural Organic Carbon Source Fuels the Growth of Diverse and Distinct Bacteria From Groundwater

Recovery and cultivation of diverse environmentally-relevant microorganisms from the terrestrial subsurface remain a challenge despite recent advances in modern molecular technology. Here, we applied complex carbon (C) sources, i.e., sediment dissolved organic matter (DOM) and bacterial cell lysate, to enrich groundwater microbial communities for 30 days. As comparisons, we also included enrichments amended with simple C sources including glucose, acetate, benzoate, oleic acid, cellulose, and mixed vitamins. Our results demonstrate that complex C is far more effective in enriching diverse and distinct microorganisms from groundwater than simple C. Simple C enrichments yield significantly lower biodiversity, and are dominated by few phyla (e.g., Proteobacteria and Bacteroidetes ), while microcosms enriched with complex C demonstrate significantly higher biodiversity including phyla that are poorly represented in published culture collections (e.g., Verrucomicrobia , Planctomycetes , and Armatimonadetes ). Subsequent isolation from complex C enrichments yielded 228 bacterial isolates representing five phyla, 17 orders, and 56 distinct species, including candidate novel, rarely cultivated, and undescribed organisms. Results from this study will substantially advance cultivation and isolation strategies for recovering diverse and novel subsurface microorganisms. Obtaining axenic representatives of “once-unculturable” microorganisms will enhance our understanding of microbial physiology and function in different biogeochemical niches of terrestrial subsurface ecosystems.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2020

Drought Drives Spatial Variation in the Millet Root Microbiome

Efforts to boost crop yield and meet global food demands while striving to reach sustainability goals are hindered by the increasingly severe impacts of abiotic stress, such as drought. One strategy for alleviating drought stress in crops is to utilize rootassociated bacteria, yet knowledge concerning the relationship between plant hosts and their microbiomes during drought remain under-studied. One broad pattern that has recently been reported in a variety of monocot and dicot species from both native and agricultural environments, is the enrichment of Actinobacteria within the drought-stressed root microbiome. In order to better understand the causes of this phenomenon, we performed a series of experiments in millet plants to explore the roles of drought severity, drought localization, and root development in provoking Actinobacteria enrichment within the root endosphere. Through 16S rRNA ampliconbased sequencing, we demonstrate that the degree of drought is correlated with levels of Actinobacterial enrichment in four species of millet. Additionally, we demonstrate that the observed drought-induced enrichment of Actinobacteria occurs along the length of the root, but the response is localized to portions of the root experiencing drought. Finally, we demonstrate that Actinobacteria are depleted in the dead root tissue of Japanese millet, suggesting saprophytic activity is not the main cause of observed shifts in drought-treated root microbiome structure. Collectively, these results help narrow the list of potential causes of drought-induced Actinobacterial enrichment in plant roots by showing that enrichment is dependent upon localized drought responses but not root developmental stage or root death.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2022

Concentration-response gene expression analysis in zebrafish reveals phenotypically-anchored transcriptional responses to retene

Polycyclic aromatic hydrocarbons (PAHs) are ubiquitous environmental contaminants and are associated with human disease. Canonically, many PAHs induce toxicity via activation of the aryl hydrocarbon receptor (AHR) pathway. While the interaction between PAHs and the AHR is well-established, understanding which AHR-regulated transcriptional effects directly result in observable phenotypes and which are adaptive or benign is important to better understand PAH toxicity. Retene is a frequently detected PAH in environmental sampling and has been associated with AHR2-dependent developmental toxicity in zebrafish, though its mechanism of toxicity has not been fully elucidated. To interrogate transcriptional changes causally associated with retene toxicity, we conducted whole-animal RNA sequencing at 48 hours post-fertilization after exposure to eight retene concentrations. The concentrations were selected to produce effects ranging from no phenotype to mortality and malformations in 100% of animals at 5 days post-fertilization. We identified a concentration-response relationship between retene teratogenicity and differential gene expression in both number of DEGs and magnitude of expression change. Elevated expression of cyp1a at retene concentrations below the threshold for teratogenicity suggested that while cyp1a expression is a sensitive biomarker of AHR activation, it may be too sensitive to serve as a biomarker of AHR-dependent teratogenicity. Genes differentially expressed at only non-teratogenic concentrations were enriched for transforming growth factor-ß (TGF-ß) signaling pathway disruption while DEGs identified at only teratogenic concentrations were significantly enriched for response to xenobiotic stimulus and reduction-oxidation reaction activity. DEGs which spanned both non-teratogenic and teratogenic concentrations showed similar disrupted biological processes to those unique to teratogenic concentrations, indicating these processes were disrupted at low exposure concentrations. Gene co-expression network analysis identified several gene modules, including those associated with PAHs and AHR2 activation. One, Module 7, was strongly enriched for AHR2-associated genes and contained the strongest responses to retene. Benchmark concentration (BMC) of Module 7 genes identified a median BMC of 7.5 µM, nearly the highest retene concentration with no associated teratogenicity, supporting the hypothesis that Module 7 genes are largely responsible for retene toxicity.

Toxin, Zebrafish, Retene, Transcriptomics, network↗

DOE OSTI2023

Production of Radioisotopes of Scandium for Imaging and Therapy (Final Report)

The theranostic approach is a powerful technique that enables scientists and physicians to diagnose and treat diseases using the same targeting moiety labeled with radionuclides for imaging and therapy. A challenge behind synthesizing and evaluating these theranostic radiopharmaceuticals lies in the production of suitable radionuclides. This study aimed to investigate the production of high radionuclidic purity 43 Sc from enriched [ 46 Ti]TiO 2 for Positron Emission Tomography (PET) and 47 Sc from enriched [ 50 Ti]TiO 2 for targeted radiotherapy, as well as establish a target recycling technique for continued production of these radionuclides. Methods: Targets were prepared with 100 mg of either enriched [ 46 Ti]TiO 2 or [ 50 Ti]TiO 2 , pressed into pellets, and placed into Ta target holders with a 0.127 mm natTi cover foil. The 46 Ti targets were bombarded with 18 MeV protons at 20 µA for 1.5 h, while 50Ti targets were bombarded with 24 MeV protons at 20 µA for 4 or 8 h. The radioscandium nuclides were purified using branched N, N, N’, N’ -tetra-2-ethylhexyldiglycolamide (DGA) resin. The titanium targets were recycled by precipitating the dissolved Ti collected from the column effluent and wash streams, pH adjustment, vacuum filtration, and then drying the solid at 250 °C prior to reforming the target for the subsequent bombardment. The apparent molar activity of the isolated radioscandiums was analyzed by titration with 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA). The relative imaging performance of 18 F, 43 Sc, and 68 Ga were analyzed with PET images acquired with 30-min static scans. Results: The average radioscandium recovery for 43Sc was 91.7 ± 7.4 %, with a radionuclidic purity of 98.8 %. The average recovery of 47Sc was 89.9 ± 3.9 %, with a radionuclidic purity of 90.8%. The average recovery of enriched TiO2 was 96 ± 4.0 %. The highest apparent molar activity for [43Sc]Sc-DOTA was 23.2 GBq/µmol and 3.39 GBq/µmol for [ 47 Sc]Sc-DOTA. The order of PET resolution from highest to lowest was 18 F > 43 Sc > 68 Ga. Conclusions: This work demonstrates the feasibility of using enriched recycled [ 46 Ti]TiO 2 and [ 50 Ti]TiO 2 targets for the production of the 43 Sc and 47 Sc theranostic isotope pair in high purity.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2021

Neutron Irradiation Induced Changes in Isotopic Abundance of 6Li and 3D Nanoscale Distribution of Tritium in LiAlO2 Pellets Analyzed by Atom Probe Tomography

Tritium, a radioactive isotope of hydrogen is produced by neutron irradiation of 6Li enriched LiAlO2 pellets in Tritium producing burnable absorber rods (TPBARs). Three-dimensional nanoscale mapping of Tritium in irradiated pellets is critical to understand its spatial variation and its association with various microstructural features in the pellets. Spatially resolved analysis of 6Li isotopic enrichment in such ceramic pellets before and after irradiation can provide insights to heterogeneities in Li isotopic distribution. However, Li being a light isotope, its spatially resolved compositional or isotopic analysis is a challenging task to most analytical methods. Here we used atom probe tomography, to evaluate the Li isotopic enrichment in both the LiAlO¬2 matrix and the Li deficient secondary phase LiAl5O8 before irradiation. These results were then compared with 6Li isotopic ratio measured by APT analysis of neutron irradiated pellets to clearly demonstrate the unique capability of APT to quantitatively analyze the isotopic enrichment of 6Li as well as for other light elements present in these materials. Evidence for heterogenous nanoscale distribution of 3H and O3H within irradiated pellet microstructure is also provided, which is attributed to 3H trapping in irradiation induced vacancy clusters or voids within the LiAlO2 matrix. The characterization results given here, now prove the feasibility for using APT for analyzing nanoscale spatially resolved isotopic enrichment of Li as well as other light elements such as Tritium in TPBARs. This work also paves way to adopt APT for analyzing light elements such as Li and tritium in materials used for nuclear fusion reactor applications, geochemistry, astrophysics as well as for energy storage applications.

Devaraj, Arun↗

DOE OSTI2021

The effects of low- Z shielding on uranium isotope discrimination using the time-emission profiles of long-lived delayed neutrons

Characteristic radiation signals can often be obscured or distorted by relatively small amounts of shielding, as is the case for highly-enriched uranium (HEU). Methods that can detect and identify special nuclear material (SNM) even in the presence of shielding are therefore of high value in many nuclear security and nonproliferation applications. It has been previously shown that the buildup and decay time profiles of long-lived delayed neutron emission from induced fission are unique and can be used to successfully discriminate between uranium isotopes and infer enrichment without the need for a calibration standard. Because the long-lived delayed neutron precursors exhibit decay times on the order of tens of seconds, their time profiles are not sensitive to the delays associated with scattering or diffusion in shielding materials, which occur on much shorter time scales. The use of capture-based neutron detectors can also mitigate the effects of shielding, because unlike recoil-based detectors, they do not have a low-energy detection threshold, so that neutrons that lose their energy in the shielding are still detectable. In this work, the effects of low-Z shielding on the discrimination of uranium isotopes using delayed neutron time profile measurements are investigated for the first time. Furthermore, the delayed neutron buildup and decay time profiles of polyethylene-shielded HEU and depleted uranium objects are measured using two different capture-gated detector designs. The recorded profiles show good agreement with the shapes predicted from six-group delayed neutron models even with the shielding in place, and the enrichment of the HEU and DU objects was estimated accurately to within 5%. While this method shows potential for discriminating and characterizing extreme cases of enrichment when a moderate amount of shielding is present, simulation results suggest that significant limitations exist for intermediate enrichment levels due to distortions to the delayed neutron time profile caused by neutron reflection in the shielding.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Radiation hot spot formation on UF 6 30B cylinders

The accurate material accountancy of a country's enriched uranium hexafluoride (UF 6 ) stockpiles is imperative to the International Atomic Energy Agency (IAEA) for determining appropriate material safeguards. Uranium hexafluoride is typically stored in large cylinders with enrichment verification typically performed with a NaI gamma detector using the enrichment meter method. This approach for calculating enrichment assumes that the UF 6 is homogeneous inside the cylinder. Here, the measurements and tests presented in this manuscript show that the UF 6 in cylinders left outside in the elements can experience fractionation, leading to inhomogeneity, and subsequent development of radiation hot spots. These hot spot locations can produce errant enrichment measurements. In dark-colored cylinders that experience significant surface heating from the sun, UF 6 sublimates off the cylinder walls, leaving only the uranium daughter products in a horizontal patch (stripe) along the cylinder's midline. This daughter patch is hypothesized to be the source of the radiation hot spots. The patch forms during the summer and tends to decay away during the winter, when solar intensity and temperatures are reduced.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

CAPOS: The bulge Cluster APOgee Survey: III. Spectroscopic tomography of Tonantzintla 2

Here, we performed the first detailed spectral analysis of red giant members of the relatively high-metallicity globular cluster (GC) Tononzintla 2 (Ton 2) using high-resolution near-infrared spectra collected with the Apache Point Observatory Galactic Evolution Experiment II survey (APOGEE-2), obtained as part of the bulge Cluster APOgee Survey. We investigated chemical abundances for a variety of species including the light, odd-Z, α-, Fe-peak, and neutron-capture elements from high S/N spectra of seven giant members. The derived mean cluster metallicity is [Fe/H] = –0.70 ± 0.05, with no evidence for an intrinsic metallicity spread. Ton 2 exhibits a typical α-enrichment that follows the trend for high-metallicity Galactic GCs, similar to that seen in 47 Tucanae and NGC 6380. We find a significant nitrogen spread (> 0.87 dex), and a large fraction of nitrogen-enriched stars that populate the cluster. Given the relatively high-metallicity of Ton 2, these nitrogen-enriched stars are well above the typical Galactic levels, indicating the prevalence of the multiple-population phenomenon in this cluster that also contains several stars with typical low first-generation N abundances. We also identify the presence of [Ce/Fe] abundance spread in Ton 2, which is correlated with the nitrogen enhancement, indicating that the s-process enrichment in this cluster has likely been produced by relatively low-mass asymptotic giant branch stars. Furthermore, we find a mean radial velocity of the cluster, –178.6 ± 0.86 km s –1 , with a small velocity dispersion, 2.99 ± 0.61 km s –1 , which is typical of GCs. We also find a prograde bulge-like orbit for Ton 2 that appears to be radial and highly eccentric. Finally, the considerably nitrogen-enhanced population observed in Ton 2, combined with its dynamical properties, makes this object a potential progenitor for the nitrogen-enriched field stars identified so far toward the bulge region at similar metallicity.

79 ASTRONOMY AND ASTROPHYSICS↗

DOE OSTI2023

Melanization slows the rapid movement of fungal necromass carbon and nitrogen into both bacterial and fungal decomposer communities and soils

Microbial necromass contributes significantly to both soil carbon (C) persistence and ecosystem nitrogen (N) availability, but quantitative estimates of C and N movement from necromass into soils and decomposer communities are lacking. Additionally, while melanin is known to slow fungal necromass decomposition, how it influences microbial C and N acquisition as well as elemental release into soils remains unclear. Here, we tracked decomposition of isotopically labeled low and high melanin fungal necromass and measured 13 C and 15 N accumulation in surrounding soils and microbial communities over 77 d in a temperate forest in Minnesota, USA. Mass loss was significantly higher from low melanin necromass, corresponding with greater 13 C and 15 N soil inputs. A taxonomically and functionally diverse array of bacteria and fungi was enriched in 13 C and/or 15 N at all sampling points, with enrichment being consistently higher on low melanin necromass and earlier in decomposition. Similar patterns of preferential C and N enrichment of many bacterial and fungal genera early in decomposition suggest that both microbial groups co-contribute to the rapid assimilation of resource-rich soil organic matter inputs. While overall richness of taxa enriched in C was higher than in N for both bacteria and fungi, there was a significant positive relationship between C and N in co-enriched taxa. Collectively, our results demonstrate that melanization acts as a key ecological trait mediating not only fungal necromass decomposition rate but also necromass C and N release and that both elements are rapidly co-utilized by diverse bacterial and fungal decomposers in natural settings.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2021

Performance of HALEU and HEU-Fueled Nuclear Thermal Propulsion Reactors

Interplanetary spaceflight may be significantly improved by rockets that utilize nuclear thermal propulsion (NTP) instead of the chemically propelled rockets in use today. To be efficient, NTP reactors must be light, which presents unique design opportunities and challenges to reactor designers. One such opportunity is the enrichment of the nuclear fuel that powers the NTP reactor. The use of high-enriched uranium (HEU) instead of high-assay low-enriched uranium (HALEU) will result in a lower mass of fuel required but may present additional challenges beside the mass considerations, such as proliferation concerns, availability, public acceptance and authorization basis. Whether or not a NTP reactor fueled with HEU is significantly lighter than its HALEU counterpart is unclear and an assessment has been performed here as part of the Mars Transportation Assessment Study. This report assesses the impact of reducing the fuel enrichment form HEU (93 wt.% 235U) to HALEU fuel (19.5 wt.% 235U) on reactor design outcomes for a conceptual NTP reactor. Two reactor designs, one fueled with HEU and the other with HALEU, were prepared based on a multi-stage design process. The neutronics and thermal-hydraulics characteristics of these reactors are compared and contrasted to illustrate the impacts of fuel enrichment. Despite differences in fuel and moderator mass, the overall mass difference between the HALEU and HEU cores was only 9%. The required core size and assembly dimensions are nearly equivalent. This finding is based on a rapid design process that targeted core volume minimization. Each features some neutronics advantages - depletion is much less significant in the HEU core, but the Doppler reactivity feedback is stronger in the HALEU core. Additional discussions are made about the possible benefits of other design changes, including the possibility of utilizing a fast-spectrum reactor. Given the scope of this assessment the findings discussed in this report are not exhaustive and are only providing an initial comparison of the most obvious design and performance parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Delayed Critical and Subcritical Experiments with Polyethylene Moderated Unreflected Thin 15 in. Diameter HEU Metal Plates

The thin ~15 in. diameter highly enriched uranium (HEU) metal plates were assembled to delayed criticality at the Oak Ridge Critical Experiments Facility (ORCEF) in 1969 with various thicknesses of polyethylene (varying from 1/16 to 2$\frac{3}{8}$ inches) between uranium metal plates. The average 235 U enrichment was 93.27 wt. %. These unreflected critical configurations contained 4$\frac{2}{3}$ to 20$\frac{5}{6}$ thin 15 in. diameter HEU metal plates (on loan from Los Alamos National Laboratory [LANL] and shipped to Oak Ridge National Laboratory [ORNL] on June 3, 1969). Depending on the thickness of polyethylene, the enriched uranium masses varying from 28,053 to 135,148 grams. Fractional plate sections consisted of the appropriate number of 60° pie sections. In addition to the measurement at delayed criticality, subcritical measurements were also performed by the inverse kinetic rod drop method. Prompt neutron decay constant measurements were also performed by the Rossi alpha and randomly pulsed neutron method using a time-tagged spontaneous fission californium neutron source; these are briefly reported here. At the time of these measurements in 1969, the thin HEU metal plates were in near-pristine condition with extremely little oxidation, allowing better descriptions of the uranium plates than the use of these plates in a heavily oxidized and deteriorated condition in recent reflected benchmark experiments at the LANL facility at the Nevada Test Site with these same thin highly enriched uranium metal plates. This report documents the experimental information for the measurements performed so that later researchers can perform the required uncertainty and calculational analyses and documentation to use these data for an International Nuclear Criticality Safety Benchmark Evaluation Program (ICSBEP) or a Nuclear Energy Agency (NEA) benchmark. Data from the experiments described should be acceptable for use as criticality safety benchmark experiments for the ICSBEP and the NEA nuclear criticality safety benchmark program once the uncertainty analysis on the measured neutron multiplication factors is completed. Additional data—such as the dimensional inspection reports, uranium isotopic information, and other relevant particulars—should be retrieved from the Y-12 Plant or LANL and incorporated in the final ICSBEP benchmark. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, the uncertainties in $k_{eff}$ could be as low as ± 0.0002 for some configurations. Other experiments with smaller-diameter than 15 in. diameter HEU metal plates have been benchmarked in HEU-METFAST-001. The prompt neutron time decay measurements could be the basis for an International Reactor Physics Benchmark Program. Preparation of the present report is part of an effort at ORNL to document more than 15 undocumented critical and subcritical experiments enumerated in ORNL/TM-2019/18 and performed by ORNL at ORCEF and other US Department of Energy critical experiments facilities using more than 500 operational days of critical facility time. This work for this report publication was supported by the Nuclear Criticality, Radiation Transport and Safety NCSP Program at ORNL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Technical Evaluation of Accelerated Basin De-Inventory Material Addition to Sludge Batch 11 (Rev. 1)

The Accelerated Basin De-inventory (ABD) program involves discarding spent nuclear fuel that is currently stored in L-Basin to the Defense Waste Processing Facility (DWPF) for vitrification. The first ABD discards will occur during the preparation of Sludge Batch (SB) 11. Savannah River Mission Completion has requested that the Savannah River National Laboratory assess the technical gaps related to the increased gadolinium poisoning requirement and the impacts of performing the Low Temperature Aluminum Dissolution (LTAD) process in Tank 51 with H-Canyon discards present. The following summarizes the evaluation of the impacts of increasing the quantity of gadolinium (and related topics) from what was previously evaluated in the SRNL studies of gadolinium-poisoned ABD material solubility, the overall ABD flowsheet review, and increasing the fissile mass loading in glass: 1) Based on literature surveys, there is no indication that organic interactions with gadolinium will be significant at the high pH (typically >13) conditions of the Concentration, Storage, and Transfer Facilities. Any interactions of gadolinium with organics in DWPF are not expected to adversely impact DWPF or downstream facilities. Thus, there is little-to-no residual risk from organic interactions with gadolinium [Gap closed]; 2) Adding depleted uranium to ABD material, targeting 235 U enrichment of 4.90% within each transfer window, will mitigate potential impacts from an increase in soluble 235 U enrichment during sludge washing and LTAD. The plan to take advantage of previous transfers and allow 235 U enrichment of >5% during the final transfer window carries a risk that Tank 51 supernate will have a 235 U enrichment of >5%, which should be evaluated for acceptance; 3) Increasing the gadolinium mass ratio to 3.0:1 Gd: 235 U(eq SLU ) should lead to the same or higher partitioning of gadolinium into the solid phase within the DWPF Chemical Process Cell, resulting in both liquid and solid phases with expected partitioning of Gd consistent with the prior solubility study [Gap closed for SB11]; 4) There are no expected impacts on DWPF melt temperature and melter operations due to the minimal ~0.2 weight percent (wt%) increase in Gd concentration relative to previous sludge batches [Gap closed for SB11]; 5) As observed previously, Gd is expected to enter the off-gas system via physical entrainment, but at a slightly higher concentration than what was observed for SB9 melter off-gas pluggage deposits (0.07 wt%) [Gap closed for SB11] ; 6) There are no expected impacts on DWPF recycle or the Recycle Collection Tank glycolate destruction process. [Gap closed for SB11]; 7) Gd is projected to be a trace component in the SB11 glass (<0.5 wt%) and can be ignored for process control. Trace components do not significantly impact glass durability, thus the conclusions of the previous Product Consistency Test evaluation at a fissile mass loading of 2,500 g fissile/m3 glass still applies to SB11. The ~0.1 wt% increase in Gd2O3 concentration relative to the previous study will not impact the predictability of SB11 glass with the DWPF Product Composition Control System (PCCS) models for durability or the acceptability of glass according to the Waste Acceptance Product Specifications (WAPS) criterion for product consistency [Gap closed for SB11]; 8) No additional Toxicity Characteristic Leaching Procedure testing is necessary for SB11 and the hazardous waste specification of the SB11 DWPF waste form is unchanged after the addition of the ABD stream [Gap closed for SB11]. The following summarizes the evaluation of the impacts of adding two-thirds of the ABD material to Tank 51 prior to LTAD: 1) The addition of two-thirds of the ABD increases overall aluminum mass from 1.39×10 4 kg to 1.64×10 4 kg (15.5% ABD Al). The form of the insoluble portion of the Al resulting from ABD addition should be the more readily dissolved Al(OH) 3 and amorphous forms. The portion of the ABD aluminum that is processed by LTAD is expected to be completely soluble, thus requiring that less of the boehmite in the sludge be dissolved to reach the same Al target in the SB. [Gap closed for SB11]; The expected LTAD impact on other components, as related primarily to the components in ABD, are discussed. Gd is expected to remain insoluble during LTAD and not impact the solubility of other components. [Gap closed for SB11]; The addition of two-thirds of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 13,100 kg (63% ABD U) and the projected plutonium mass from 86.0 kg to 89.5 kg (3.9% ABD Pu). The addition of all of the ABD increases overall projected SB11 uranium mass from 4,740 kg to 16,100 kg (70% ABD U) and the projected plutonium mass from 86.0 kg to 90.4 kg (5.3% ABD Pu). The 235 U enrichment will be ≤5%. The fissile uranium will be adequately poisoned by Gd and the fissile Pu will be adequately poisoned by Fe from the sludge. [Gap closed for SB11]; There is a low risk that ABD addition will impact the rheology or pumpability of the slurry. There is a low but higher risk of ABD addition prior to LTAD impacting the settling rate; Based on the evaluation of adding two-thirds of the ABD material and all of the ABD material prior to the LTAD process, there is no volume or mass limit that would need to be imposed on ABD additions prior to LTAD. [Gap closed for SB11]. Revision 1 of this report addresses a variation on the ABD additions and LTAD strategy where sodium hydroxide additions for LTAD may be performed intermittently or concurrently with an ABD addition window. The proposed change does not alter the conclusions of this evaluation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Scoping Study for Fast Flux Testing in the Advanced Test Reactor

The value of fast spectrum reactors remains prominent in the nuclear technology portfolio. The performance of these reactors can be maximized with advancements in nuclear fuel technologies, but development of these technologies is currently held back by lack of fast spectrum test reactors available to the United States. Spectral modification of experiment positions in the thermal spectrum Advanced Test Reactor (ATR) has long been used to support fast reactor fuel development, but these methods have not been progressed to their full potential. This study investigated the use of concentric rings of aluminum-clad fuel plates in ATR flux traps and thermal neutron absorbing filters to increase fast neutron flux on test specimens. This concept was termed the Boosted Energy Advanced Spectrum Test (BEAST). This approach will enable irradiation of advanced fuel designs in prototypic-length fuel pins and representative flux environment to support post irradiation exams, enable transient testing, and produce the type of data that will permit lead test assembly irradiations in true Sodium Fast Reactors (SFRs) when they become available. Neutronic predictions were performed to investigate BEAST design options and thermal hydraulic models were produced to ensure feasibility of BEAST. Two versions were considered based on the geometric limitations of ATR’s small and large flux traps. The small version was found to be preferable due to slightly higher fast flux and fast-to-thermal neutron ratio. Perhaps more influentially, the small flux trap option was also preferred to avoid conflict with ongoing very high temperature reactor fuel irradiation programs in ATR’s large northeast flux trap. The small flux trap option provided less than half the test volume of the large version, but still had adequate volume for seven SFR pins in cross section which could be stacked two-high in ATR’s 1.2m long core to accommodate up to 14 EBR-II size pins. The preference for the small flux trap configuration should be revisited if additional collaborative test programs emerge with the need to irradiate a significant volume of additional specimens. Calculations were performed regarding a lithium deuteride ring to convert thermal neutrons into 14 MeV fusion neutrons. At the time this report was written these calculations were partially complete and it remains to be seen whether the concept would be worth including in BEAST. Given the preference for the small flux trap option, which does not afford enough room for the 14 MeV ring, it was concluded to defer future work on the lithium deuteride ring. This decision could be revisited if fusion material research programs emerge for collaborative testing in BEAST. A cadmium-lined specimen holder design was found to be adequate in filtering thermal neutrons and preferred over other neutron absorbers based on past experience with cadmium baskets. It was acknowledged that cadmium-bearing hardware would become depleted and need to be replaced occasionally, which appeared feasible from a mechanical design perspective. Neutronic studies investigated different enrichment levels in the booster fuel using uranium-molybdenum alloy dispersion fuel which has performed well in past ATR irradiations. Both options were able to drive fuel pins to SFR-like fission heating rates. The high enriched booster fuel option outperformed the low enriched option by ~20% on key metrics including fast flux and fast-to-thermal ratio, but the low enriched option was favored in order to broaden options for potential fuel suppliers. The preferred BEAST design options including cadmium filter with low enriched booster fuel in the small flux trap configuration was predicted to achieve 6.2E14 n/cm2sec fast flux (>0.1 MeV) with a fast-to-thermal ratio of 44.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗