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At least 271 records · Page 15

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DOE OSTI2025

Imaging the Progression of Radiolytic Damage in Molecular Crystals with Scanning Nanobeam Electron Diffraction

Almost every electron microscopy experiment is fundamentally limited by radiation damage. Nevertheless, little is known about the onset and progression of radiolysis in beam-sensitive materials. Here we apply ambient-temperature scanning nanobeam electron diffraction to record simultaneous dual-space movies of organic and organometallic nanocrystals at sequential stages of beam-induced radiolytic decay. We show that the underlying mosaic of coherently diffracting domains undergoes internal rearrangement as a function of accumulating electron fluence, causing the intensities of some associated Bragg reflections to fade nonmonotonically. Furthermore, we demonstrate that repeated irradiation at a single probe position leads to the isotropic propagation of delocalized radiolytic damage well beyond the direct footprint of the incident beam. We refer to these expanding tides of amorphization as “impact craters.”

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2021

Metalloprotein catalysis: structural and mechanistic insights into oxidoreductases from neutron protein crystallography

Metalloproteins catalyze a range of reactions, with enhanced chemical functionality due to their metal cofactor. The reaction mechanisms of metalloproteins have been experimentally characterized by spectroscopy, macromolecular crystallography and cryo-electron microscopy. An important caveat in structural studies of metalloproteins remains the artefacts that can be introduced by radiation damage. Photoreduction, radiolysis and ionization deriving from the electromagnetic beam used to probe the structure complicate structural and mechanistic interpretation. Neutron protein diffraction remains the only structural probe that leaves protein samples devoid of radiation damage, even when data are collected at room temperature. Additionally, neutron protein crystallography provides information on the positions of light atoms such as hydrogen and deuterium, allowing the characterization of protonation states and hydrogen-bonding networks. Neutron protein crystallography has further been used in conjunction with experimental and computational techniques to gain insight into the structures and reaction mechanisms of several transition-state metal oxidoreductases with iron, copper and manganese cofactors. Here, the contribution of neutron protein crystallography towards elucidating the reaction mechanism of metalloproteins is reviewed.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2021

Imaging the short-lived hydroxyl-hydronium pair in ionized liquid water

The radiolysis of water is ubiquitous in nature and plays a critical role in numerous biochemical and technological applications. Although the elementary reaction pathways for the ionized water have been studied, the short-lived intermediate complex and structural dynamic response after the proton transfer reaction remain poorly understood. Here, using liquid-phase ultrafast electron diffraction technique to measure the intermolecular O··O and O··H bonds, we captured the short-lived radical-cation complex OH(H 3 O + ) that was formed within 140 femtoseconds through a direct oxygen-oxygen bond contraction and proton transfer, followed by the radical-cation pair dissociation and the subsequent structural relaxation of water within 250 femtoseconds. These measurements provide direct evidence of capturing this metastable radical-cation complex before separation, thereby improving our fundamental understanding of elementary reaction dynamics in ionized liquid water.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Chemical and physical transformations of carbon-based nanomaterials observed by liquid phase transmission electron microscopy

This article addresses recent advances in liquid phase transmission electron microscopy (LPTEM) for studying nanoscale synthetic processes of carbon-based materials that are independent of the electron beam—those driven by nonradiolytic chemical or thermal reactions. In particular, we focus on chemical/physical formations and the assembly of nanostructures composed of organic monomers/polymers, peptides/DNA, and biominerals. The synthesis of carbon-based nanomaterials generally only occurs at specific conditions, which cannot be mimicked by aqueous solution radiolysis. Carbon-based structures themselves are also acutely sensitive to the damaging effects of the irradiating beam, which make studying their synthesis using LPTEM a unique challenge that is possible when beam effects can be quantifi ed and mitigated. With new direct sensing, high frame rate cameras, and advances in liquid-cell holder designs, combined with a growing understanding of irradiation effects and proper experimental controls, microscopists have been able to make strides in observing traditionally problematic carbon-based materials under conditions where synthesis can be controlled, and imaged free from beam effects, or with beam effects quantified and accounted for. These materials systems and LPTEM experimental techniques are discussed, focusing on nonradiolytic chemical and physical transformations relevant to materials synthesis.

Parent, Lucas R.↗

DOE OSTI2020

Electron-beam-driven chemical processes during liquid phase transmission electron microscopy

Liquid phase (or liquid cell) transmission electron microscopy (LP-TEM) has been established as a powerful tool for observing dynamic processes in liquids at nanometer to atomic length scales. However, the simple act of observation using electrons irreversibly alters the nature of the sample. A clear understanding of electron beam-driven processes during LP-TEM is required to interpret in situ observations and utilize the electron beam as a stimulus to drive nanoscale dynamic processes. In this review, we discuss recent advances toward understanding, quantifying, mitigating, and harnessing electron beam-driven chemical processes occurring during LP-TEM. We will highlight progress in several research areas, including modeling electron beam-induced radiolysis near interfaces, electron beam-induced nanocrystal formation, and radiation damage of soft materials and biomolecules.

Woehl, Taylor J.↗

DOE OSTI2020

Chemical and bonding analysis of liquids using liquid cell electron microscopy

Liquid cell transmission electron microscopy (TEM) has become an essential tool for studying the structure and properties of both hard and soft condensed-matter samples, as well as liquids themselves. Liquid cell sample holders, often consisting of two thin window layers separating the liquid sample from the high vacuum of the microscope column, have been designed to control in situ conditions, including temperature, voltage/current, or flow through the window region. While high-resolution and time-resolved TEM imaging probes the structure, shape, and dynamics of liquid cell samples, information about the chemical composition and spatially resolved bonding is often difficult to obtain due to the liquid thickness, the window layers, the holder configuration, or beam-induced radiolysis. Here, we review different approaches to quantitative liquid cell electron microscopy, including recent developments to perform energy-dispersive x-ray and electron energy-loss spectroscopy experiments on samples in a liquid environment or the liquid itself. We also cover graphene liquid cells and other ultrathin window layer holders.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Cellular Response to Proton Irradiation: A Simulation Study with TOPAS-nBio

The cellular response to ionizing radiation continues to be of significant research interest in cancer radiotherapy, and DNA is recognized as the critical target for most of the biologic effects of radiation. Incident particles can cause initial DNA damages through physical and chemical interactions within a short time scale. Initial DNA damages can undergo repair via different pathways available at different stages of the cell cycle. The misrepair of DNA damage results in genomic rearrangement and causes mutations and chromosome aberrations, which are drivers of cell death. This work presents an integrated study of simulating cell response after proton irradiation with energies of 0.5–500 MeV (LET of 60–0.2 keV/µm). A model of a whole nucleus with fractal DNA geometry was implemented in TOPAS-nBio for initial DNA damage simulations. The default physics and chemistry models in TOPAS-nBio were used to describe interactions of primary particles, secondary particles, and radiolysis products within the nucleus. The initial DNA double-strand break (DSB) yield was found to increase from 6.5 DSB/Gy/Gbp at low-linear energy transfer (LET) of 0.2 keV/µm to 21.2 DSB/Gy/Gbp at high LET of 60 keV/µm. A mechanistic repair model was applied to predict the characteristics of DNA damage repair and dose response of chromosome aberrations. It was found that more than 95% of the DSBs are repaired within the first 24 h and the misrepaired DSB fraction increases rapidly with LET and reaches 15.8% at 60 keV/µm with an estimated chromosome aberration detection threshold of 3 Mbp. The dicentric and acentric fragment yields and the dose response of micronuclei formation after proton irradiation were calculated and compared with experimental results.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2020

Hydrogen Transport in a Model 9979 Shipping Package with Inner Convenience Cans

Radiolytic hydrogen production and accumulation inside containment packages is a concern at any facility responsible for their packaging, storage, transportation, and/or disposal. When hydrogen gas accumulates to concentrations above the Lower Flammability Limit (LFL) which is 4% or 40,000 ppm in air, the possibility of a deflagration or explosion increases. This concern persists over the course of the package lifetime which is unlimited when disposed of by burial or in permanent repositories. Here, we report on a numerical model used to predict the concentration of hydrogen within each layer of a Model 9979 package containing a convenience can assembly. Simulations show the hydrogen concentration to always be highest in the inner convenience can containing the radioactive source. When the radioactive source is within the Los Alamos National Laboratory (LANL) Packaging Limits, the hydrogen concentration is shown to remain well below the LFL at all times including packaging, storage, transportation, and disposal. A hydrogen transport model is presented for a Model 9979 package system containing a nested arrangement of convenience cans, which are tin oxide coated steel cans of various sizes with a slip-lid assembly. The inner convenience can contains the radioactive source material along with an unknown quantity of incidental water acquired from humid air or processing. While visible organic materials such as paper and plastics were purposely excluded from the inner can, it is not possible to claim the wastes are entirely organic free. The inner convenience can is tape sealed and placed into a plastic bag which is horsetail closed (i.e., twisted and taped). The bagged can is placed into an outer convenience can that is also tape sealed. The can assembly is then placed into the 30 gallon drum and subsequently placed inside the 55 gallon drum in the 9979 package. Here we assume hydrogen gas is produced in the inner convenience can from alpha radiolysis of water at a rate dependent on the quantity of uranium isotopes and water present. The hydrogen transport model was used to calculate hydrogen accumulations within the package’s five layers at different times and conditions. These simulations serve two purposes; (i) to build confidence in the model by comparing predicted values to measured values, and (ii) to check the steady state hydrogen concentrations that are approached at long times in the package’s lifetime. Model simulations were compared to gas samples taken from the 30 gallon drum after storage at LANL’s Chemistry and Metallurgy Research (CMR) building for around 500 days. Hydrogen concentration calculations over much longer periods (i.e., more than 270 years) included extreme storage durations, transportation at extreme cold temperatures, and disposal of packages assuming different average temperatures.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Evaluation of Hydrogen Generation in High Burnup Demonstration Dry Storage Cask

This report provides a best-estimate evaluation of residual water content (post-dry out) in the High Burnup (HBU) LWR Spent Fuel Demonstration project TN-32 cask, and evaluates the radiolysis of the residual free water, and the physisorbed and chemisorbed waters on the surfaces of the fuel and cask internal contents. The evaluation of radiolytic breakdown of those waters with gamma radiation causing the generation of hydrogen gas (H 2 ) is made using available literature data and models. This evaluation is part of the overall materials performance evaluation of the SNF-in-canister system, and is part of the technical bases for their continued safe dry storage. The TN-32 cask contents included 32 HBU LWR spent fuel assemblies each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless steel structural components. The residual free and surface (physisorbed/chemisorbed) waters are ascribed to water vapor in the free volume and to components’ surfaces, respectively. The total potential radiolytic hydrogen inventory from the water vapor and from waters ascribed to surfaces has been calculated assuming all the water produced molecular H 2 . The residual water that is chemically incorporated into the bulk of a hydrated oxide, i.e., chemisorbed water, and its total potential hydrogen inventory has been calculated. These calculations are at the physical limit of material available and are used for a bounding assessment purpose only.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Surrogate Model Development of Spent Fuel Degradation for Repository Performance Assessment

In model simulations of deep geologic repositories, UO 2 fuel matrix degradation typically begins as soon as the waste package breaches and groundwater contacts the fuel surface. The initial degradation rate depends on the timing of these events, burnup of the fuel, temperature, and concentrations of dissolved reactants. Estimating the initial rate of degradation is fairly straightforward, but as UO 2 corrosion products precipitate on the fuel surface and the movement of dissolved species between the fuel surface and environment is impeded by the precipitated solids, the rate is more difficult to quantify. At that point, calculating the degradation rate becomes a reactive-transport problem in which a large number of equations must be solved by iteration for a large number of grid cells at each time step. The consequence is that repository simulations, which are already expensive, become much more expensive, especially when hundreds or thousands of waste packages breach. The Fuel Matrix Degradation (FMD) model is the process model of the Spent Fuel and Waste Science and Technology (SFWST) campaign of the US Department of Energy (DOE). It calculates spent fuel degradation rates as a function of radiolysis, redox reactions, electrochemical reactions, alteration layer growth, and diffusion of reactants through the alteration layer. Like other similar fuel degradation process models, it is a complicated model requiring a large number of calculations and iterations at each time step.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Modifications to the Bubble experiment and preparation for additional irradiations

SHINE Medical Technologies is planning to use neutron-induced fission in a subcritical low-enriched-uranium uranyl sulfate target solution for production of 99 Mo. During this operation, the solution will undergo self-heating due to fissioning of the uranium, radiolytic decomposition of the water, and circulation due to thermal gradients generated in the solution, and will be cooled by cooling tubes running through the annulus and from cooling outside the annulus. Because the formation of the radiolysis-induced bubbles (H 2 and O 2 ) and their size and dynamics will impact the operational parameters of the liquid target, an understanding of bubble behavior is critical for the ability to predict the behavior of the target solution during this operation. It is also important to be able to predict the thermal gradients and the circulation in the vessel. Researchers at Argonne National Laboratory have designed an experimental setup to study radiolytic gas formation in uranyl sulfate under direct electron beam irradiation and have conducted initial experiments. Results of those experiments provided invaluable information on thermal hydraulic behavior of the solution and some information on bubble formation and behavior, but those initial experiments fell short in the measurements of the gas generation rates and bubble behavior. To address the shortcomings of the original experiment, the irradiation setup was modified to improve our abilities to measure gas generation rates and measure the temperature distribution in the solution with better precision. Modifications to the experimental setup and preparation for the irradiations are described below.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Crack Stability in Breached Fuel

This report describes the fracture mechanics formalism to evaluate the stability of cracks in a fuel cladding. Recognized consensus-body linear elastic fracture mechanics (LEFM) was applied to identify the crack instability length, or the length at which unstable mechanical crack extension would occur, as a function of pellet swelling loading (radial strain and fraction conversion from UO 2 to U 3 O 8 ) at the local cracked cladding region for two postulated fracture toughness (K IC ) values of the cladding. The crack opening displacement (COD) and the crack opening area (COA) were also identified. This analysis informs evaluations for crack extension and the potential for pellet debris loss from the fuel rod for cases of pellet oxidation in dry storage canisters where inadvertent residual water may undergo radiolysis causing oxidizing conditions to pellets exposed to the canister environment through breached cladding.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Milestone 2.8: Preliminary Radiolytic Gas Generation Measurements from Helium-Backfilled Samples

The Department of Energy (DOE) is currently managing nearly 13 metric tons of aluminum-clad spent nuclear fuel (ASNF) with the intention of extended (> 50 years) dry storage in helium-backfilled canisters. Due to in-reactor and cooling pond conditions, oxyhydroxide corrosion layers have formed on the surface of the ASNF elements. These corrosion layers are susceptible to radiolysis and the formation of molecular hydrogen gas (H 2 ) due to the fuel’s inherent radiation field. Consequently, a rigorous evaluation of the effect of helium gas on radiolytic H 2 production is necessary to support the Technical Considerations and Challenges for Extended (> 50 yrs) Dry Storage of ASNF program, especially as previous Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution work demonstrated a significant effect of gas composition on the radiolytic yield (G-value) of H 2 . Here we report preliminary G-values for the radiolytic formation of H 2 from the gamma irradiation of pre-corroded aluminum alloy 1100 coupons flame sealed in helium environments. Irradiations yielded G(H 2 ) values of (5.1 ± 0.5) x 10 –4 and (9.4 ± 0.9) x 10 –4 µmol J –1 for pristine coupons, and (10.1 ± 0.4) x 10 –4 and (15.1 ± 1.2) x 10 –4 µmol J –1 for pre-corroded coupons for 0% and 50% relative humidity, respectively. These helium environment G(H 2 ) values are between 28% and 58% higher than previously reported values for argon environments. This enhancement is attributed to the significant difference in first ionization energy between helium (24.59 eV) and argon (15.76) facilitating additional processes, e.g., Penning ionization. These new preliminary helium environment G(H 2 ) values will be employed by Task 3 - Sealed and Vented System Episodic Breathing and Gas Generation Prediction to model the effect of radiolytic H 2 accumulation in helium environments to evaluate the practicality of the extended storage standard canister design.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2021

Crack Stability in Breached Fuel

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Corrosion of 347 stainless steel in the presence of uranyl sulfate solution and radiation

The U.S. molybdenum 99 (Mo-99) industry is pursuing production of fission-made Mo-99 using a uranium solution such as uranyl sulfate. In this process, uranyl sulfate solution containing low-enriched uranium will be bombarded by neutrons creating Mo-99 and other fission products. During the production, the uranyl sulfate solution will be irradiated until an acceptable activity level of Mo-99 is produced. The uranyl sulfate solution containing Mo-99 and other fission products will then undergo a series of separation steps. First, uranyl sulfate can be separated from Mo-99 using a primary titania column to recover Mo-99, with the uranyl sulfate solution to be used for another irradiation cycle. Then, raffinate from a primary titania column containing Mo-99 can be concentrated and purified using a LEU modified Cintichem process developed by Argonne National Laboratory. During irradiation, the temperature of the uranyl sulfate solution can reach near boiling (up to ~80° C assumed), causing radiolysis of water and the resultant formation of hydrogen peroxide. Because high-radiation fields will be present during each irradiation cycle, it is important to determine the corrosion rates of SS-347 under such conditions to estimate the life cycle of the target solution vessel. The buildup of corrosion products from the SS components in the uranyl sulfate solution also needs to be well understood because potential accumulation of iron, nickel, and other corrosion products may affect the Mo-99 recovery and purification process. To study the corrosion rates of SS-347 material under conditions relevant to future Mo-99 production facility, SS-347 coupons in uranyl sulfate solution at ~80° C were irradiated using Argonne’s Van de Graaff generator, which can generate high-radiation fields without fissioning of uranium or production of activation products.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Integrated Fuel Cycle Materials and Chemistry Program (FY21 Plutonium Science Task Report)

Task 2 of the IFCMCP was designed to utilize vibrational spectroscopy, microscopy, X-ray diffraction, and calorimetry to explore the spectroscopic, structural, and chemical properties of plutonium compounds relevant to the nuclear fuel cycle. An important precursor to this work is production of fresh, high-quality plutonium samples. Given the health hazards and material accountability associated with plutonium, sample production is not trivial and production experiments require extensive planning and coordination by research and support staff. Further, because alpha-emissions from plutonium produce time- and dose dependent radiolysis resulting in structural defects, particularly in the presence of water, freshly prepared plutonium samples must be carefully packaged and analyzed with urgency following their production. In FY21, SRNL and UND engaged in numerous research studies involving production and subsequent characterization of plutonium dioxide, plutonium fluoride, plutonium oxalate, and plutonium nitrate. Specific efforts involving each of these compounds are described in detail in the body of this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Literature Review on Next Generation Solvent Isopar ® L Vapor Pressure Curve and the Partitioning of its Modifier and Extractant

The Next Generation Solvent (NGS) is set to replace the Original Caustic Side Solvent Extractant (CSSX) at the Salt Waste Processing Facility (SWPF). The Savannah River National Laboratory (SRNL) was requested by Savannah River Mission Completion (SRMC), formerly Savannah River Remediation (SRR), to perform a literature review on the following topics to address flammability concerns with the current solvent: Isopar ® L vapor pressure curve for NGS, partitioning ratio for the extractant MaxCalix and the modifier Cs-7SB, and high cesium concentration impacts on NGS radiolysis and potential solvent degradation rates in high cesium concentrations. The following conclusions and recommendations are made based on previous experimental work and literature: (1) Current SWPF flammable gas generation calculations use an Isopar ® L vapor pressure curve based on experimental testing with the Original CSSX solvent. No such testing to date has been performed with NGS. It is suggested that the decrease in Cs-7SB concentration for NGS compared to the Original CSSX solvent would lead to a slightly higher vapor pressure at all temperatures in SWPF vessels. A bounding NGS vapor pressure curve has been provided; it is recommended to see if these values would challenge current flammability controls and to perform testing if needed.(2) The partitioning ratio for Cs-7SB is known in the Original CSSX solvent with dilute nitric acid and caustic solutions. No tests could be found for the partitioning of Cs-7SB to dilute boric acid solutions; however, a similar partitioning ratio is expected. Due to the lipophilic alkyl chains on MaxCalix, it is expected to be even less soluble than BOBCalixC6 in the aqueous phase and should not be considered a significant contributor to the f organic term. Additionally, the reaction rate of N,N’,N’’-Tris(3,7-dimethyloctyl)guanidine (TiDG) or its degradation products with a hydrogen radical should be estimated/determined if they are found to be significant contributors to the f organic term. (3) NGS is expected to see much higher Cs concentrations at SWPF in comparison to its use at the Modular CSSX Unit (MCU). These higher Cs concentrations could influence radiolytic degradation rates of the solvent. NGS appears to be fairly stable to radiolytic degradation based on previous testing and its use at MCU. However, there has not been radiolytic flammable gas generation testing with NGS to date. There is a risk that the continued use of G-values obtained for flammable gases produced from the irradiation of the Original CSSX solvent is not bounding for NGS, but this is considered a very low risk due to the similarities in the composition of the solvents, as well as the conservatisms in the experimental design of the Original CSSX testing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Gamma Irradiation of NaCl-UCl 3 Salt for the Molten Chloride Fast Reactor

This report documents the results of the gamma ray irradiation of NaCl-UCl 3 eutectic salt in the Advanced Test Reactor (ATR) spent fuel pool gamma tube and subsequent analysis of capsules utilizing the Gas Assay Sample and Recharge (GASR) system. NaCl-UCl 3 salt capsules at four different temperatures, 75C, 150C, 300C and 600C, were irradiated shortly after the core internal changeout (CIC) of the ATR, beginning April 28 th of 2021, and ending on August 18 th of 2021. Re-positioning of the fuel elements in the spent fuel pool to move more fresh fuel elements around the gamma tube occurred on May 11 th of 2021, resulting in an approximate 20% increase in the dose rate and consequently the total absorbed dose. In total, the salt-filled capsules underwent 2638 hours of gamma irradiation from the adjacent freshly discharged ATR spent fuel. Capsule internal pressure measurements were taken via the Gas Assay Sample and Recharge (GASR) system in the INL Hot Fuels Examination Facility (HFEF) following completion of the irradiation. Based on the GASR analysis results, the primary conclusion from this experiment to-date is that radiolytic generation of chlorine gas is insignificant for the dose rates and total absorbed dose that these samples experienced. To elucidate more detailed effects of radiolysis on NaCl-UCl 3 eutectic salts and reach a more definitive conclusion, further advanced analyses such as electron microscopy, simultaneous thermal analysis, and electron paramagnetic resonance are recommended to analyze the capsule wall and salt samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗