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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Effect of initial γ -irradiation on infrared laser ablation of poly(vinyl alcohol) studied by infrared spectroscopy

We report the effect of initial γ-irradiation of poly(vinyl alcohol) (PVA) on the kinetics of polymer ablation under CO 2 laser irradiation was studied using infra-red spectroscopy. The rate of laser ablation rapidly decreases with increasing time of laser ablation up to 10 s. Initial γ-irradiation of a dose up to 100 kGy reduces the rate of the laser ablation, just like the effect of laser exposure. The number of OH and CH groups decrease in PVA both with laser irradiation and with γ-irradiation. The number of OH groups decreases more under laser irradiation. Unsaturated bonds are formed in γ–irradiated PVA on the surface of the crater formed by laser ablation of the polymer that has undergone radiolysis. Increasing the dose of initial γ-irradiation from 100 to 2300 kGy leads to an increase in the intensity of the IR bands of the unsaturated bonds. The number of IR bands in the spectra of the laser ablation crater of PVA does not change with the dose of initial γ-irradiation. A dose of radiolysis of 3500 kGy significantly changes the IR spectra of both the ablation crater and the coating formed from the ablation products. Many bands characteristic for PVA are no longer present and the intensities of bands corresponding to unsaturated bonds are increased. A computational model for loss of water and ketone formation was developed for PVA and PVA including a diol using G3(MP2). The highest activation energy barrier step in the decomposition reaction was formation of the enol species and water was shown to catalyze all decomposition steps.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Effects of He Ion Irradiation in Simple Borosilicate Glasses for HLW Immobilisation - 20324

High-level radioactive waste (HLW) consists of fission products and minor actinides. Borosilicate glass is the host material of choice in most countries to safely immobilize HLW, followed by eventual geological disposal. HLW glasses will experience both self-heating and self-irradiation effects due to the decay of fission products and minor actinides present therein. During the first 500 years of geological disposal, beta-gamma decay will be the main source of radiation due to short half-lives of fission products, as compared to very long-lived actinides which will be responsible for irradiation over much longer timescales. High-energy alpha particles and alpha recoil nuclei can potentially have significant impact (stored energy, volume change, accumulated dose, chemical durability) on the performance of the wasteform over longer timescales. Thus, to ensure safe and economic disposal over geological periods (10{sup 4} - 10{sup 6} years), the assessment of performance and properties in terms of radiation effects becomes indispensable. This work aims to simulate the effects of alpha decay in radioactive waste glasses. Two international glasses, namely UK 'MW' lithium-sodium borosilicate and Indian sodium-barium borosilicate, which are being used as HLW immobilization matrices by the UK and India respectively, were prepared by a melt-quench method and annealed below their glass transition temperatures. Both glasses were irradiated with 650 KeV He{sup 2+} ions with doses ranging from 10{sup 15} to 10{sup 17} ions/cm{sup 2}. Structural analysis was carried out using electron paramagnetic resonance (EPR) and Raman spectroscopies. First derivative continuous wave EPR spectra were recorded for all pristine and irradiated glass monoliths at room temperature using an X-band (frequency 9.6 GHz) EPR Bruker spectrometer. Results from EPR analyses showed the formation of irradiation induced paramagnetic centers at g ∼ 2.00 in all irradiated glasses. Raman spectra of pristine and irradiated samples were acquired, and showed notable changes in glass structure post-irradiation, with the emergence of a broad Raman band at 1560 cm{sup -1} upon irradiation with high fluences. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

First-of-a-Kind Fuel-bearing Molten Chloride Irradiation Experiment

Molten salt reactors (MSRs) have garnered increasing attention recently with several demonstration efforts on the way. A key challenge to the licensing basis for these reactors is the lack of experimental data on fueled salts. This is expected to be crucial to the safety evaluation and licensing basis of reactors of this type deployed in the future. While capability for irradiating molten salts has been reestablished in the recent decade, no enriched fuel irradiation capability has been developed and tested as of yet. A new experiment vehicle under development at Idaho National Laboratory (INL) is presented here. The Molten-salt Research Temperature-controlled Irradiation (MRTI) experiment was developed to host enriched uranium bearing salt samples to be irradiated at a test reactor within the lab complex. One of the key scientific objectives is to provide irradiated salt samples for post-irradiation examination (PIE) to study the impact of fission product generation and neutron/gamma radioactivity on the salt solution and salt-facing wall material. This paper provides a detailed overview of the mechanical design of the experiment, followed by an overview of the fabrication and assembly of an initial prototype vehicle (with non-fuel-bearing salt). A summary of the key analyses conducted as a part of the performance and safety evaluation is then provided. Lastly, an overview of the test conducted in prototypic out-of-pile (non-neutron) environment are shown. These evaluations provide the foundation for a planned irradiation of an enriched uranium-bearing chloride salt sample in the near term. The upcoming irradiation will contain 13 cm3 of UCl3-NaCl salt (93% enrichment) generating around 20 W/cm3 of fission energy during irradiation and a temperature range that can be contained between bounds of 525-900°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Atom Probe Tomography Characterization of High-Dose Ion Irradiated MA957

This paper presents a comprehensive characterization study of the microstructural responses to ion irradiation of the oxide dispersion strengthened (ODS) ferritic alloy MA957 using atom probe tomography (APT). The MA957 specimens were irradiated by 1.8 MeV Cr3+ ions to doses and temperatures relevant to advanced fast reactors (100 and 500 dpa at 400-500 °C). The average sizes of the YTiO particles decreased slightly after irradiation at all temperatures, with the most pronounced change at 400 °C. The particle number densities for 400 and 500 °C were slightly below number densities for the unirradiated condition, while the values were slightly higher at 420 and 450 °C. This behavior is consistent with the Nelson-Hudson-Mazey rate theory model that considers the competing effects between recombination and dissolution. An increase in the Y/Ti ratio of the YTiO particles was observed at the two lower temperatures. The formation of Cr-rich precipitates (a') was observed in specimens irradiated at temperatures up to 450 °C. Irradiation-induced segregation on grain boundaries was also investigated. Grain boundary element concentration profiles and ion maps demonstrate differences among the irradiation temperatures, with most datasets differing from the unirradiated condition. A key focus of this ion irradiation study is the high-dose response, and this quantitative assessment revealed relatively small but distinct effects of irradiation on microstructure evolution. Overall, these results showing minor second phase evolution suggest that MA957 should retain mechanical performance generally well up to 500 dpa for the irradiation temperatures studied.

Wang, Jing↗
DOE OSTI2024

The microstructure effects on irradiation response of ferritic – martensitic steels

Microstructural optimization to achieve greater mechanical strength has been one of the focuses in ferritic–martensitic steels development. However, these optimized microstructures’ effects on the radiation response are not well known. In this work, two ferritic–martensitic steels (9Cr-NbMo and 9Cr-Ta) underwent neutron irradiation in the High Flux Isotope Reactor, and their room-temperature post-irradiation tensile properties and microstructure evolutions were investigated and compared. These two steels exhibit similar pre-irradiation tensile behavior, and their yield strengths are higher than that of other ferritic–martensitic steels by about 200–250 MPa. Microstructural characterization on pre-irradiated materials reveals a smaller grain size in 9Cr-Ta (2.8 ± 0.3 μm in 9Cr-Ta versus 4.3 ± 0.5 μm in 9Cr-NbMo) but higher dislocation density and precipitate density in 9Cr-NbMo. As is common for ferritic–martensitic steels at low irradiation temperatures (less than about 0.45T m ), irradiation-induced hardening at 400 °C was observed for both alloys. Irradiation at 490 °C causes the two alloys to exhibit different tensile behavior: 9Cr-Ta softens by 208 MPa in yield stress, whereas 9Cr-NbMo maintains strength. Additionally, microstructural characterizations were performed, including precipitate growth, dislocation, and defect formation. Using the barrier hardening model for microstructure–property correlation, the softening in irradiated 9Cr-Ta is primarily attributed to the significant dislocation recovery, while the strength lost from the slight dislocation recovery in 9Cr-NbMo was compensated by the additional strength from the irradiation-induced cavities. The microstructure effect (primarily precipitate, dislocation and boundary) on the radiation response is discussed herein.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Neutron irradiation induced defects in oxides and their impact on the oxide properties

Understanding the irradiation-induced defects in oxides is of interest for a wide range of applications. ZnO is an interesting oxide with mixed ionic and covalent bonding that contains a variety of point defect structures—making it an excellent model for studying irradiation-induced defects and their impact on properties. Here, we investigate the effects of neutron irradiation on the formation of defects and on the structural, optical, and electrical properties of ZnO single crystals. We observe the formation of vacancies and voids via positron annihilation spectroscopy. Neutron irradiation led to a significant deterioration of the ZnO structure and formed a high concentration of point defects, vacancy clusters, and voids with large disparities in their structure across variable irradiation times. It also led to significant changes in the optical properties and sample color. Irradiation for 444 h induced a high concentration of Cu acceptors as well as a high concentration of Ga donors. Additionally, temperature-dependent Hall effect measurements revealed the competing production of donors and acceptors and showed an increase in the slope of the carrier freeze-out curve with increasing irradiation dose. This work demonstrates the combined effects of neutron irradiation in producing a wide range of structural defects, impurities, and dopants in oxides and their enormous impact on modifying the oxide structure and both the optical and electronic properties. It particularly emphasizes the importance of considering the production of new impurities and dopants during the neutron irradiation of oxides.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2020

Comparative RNA-Seq transcriptome analyses reveal dynamic time-dependent effects of 56 Fe, 16 O, and 28 Si irradiation on the induction of murine hepatocellular carcinoma

One of the health risks posed to astronauts during deep space flights is exposure to high charge, high-energy (HZE) ions (Z > 13), which can lead to the induction of hepatocellular carcinoma (HCC). However, little is known on the molecular mechanisms of HZE irradiation-induced HCC. We performed comparative RNA-Seq transcriptomic analyses to assess the carcinogenic effects of 600 MeV/n 56 Fe (0.2 Gy), 1 GeV/n 16 O (0.2 Gy), and 350 MeV/n 28 Si (0.2 Gy) ions in a mouse model for irradiation-induced HCC. C3H/HeNCrl mice were subjected to total body irradiation to simulate space environment HZE-irradiation, and liver tissues were extracted at five different time points post-irradiation to investigate the time-dependent carcinogenic response at the transcriptomic level. Our data demonstrated a clear difference in the biological effects of these HZE ions, particularly immunological, such as Acute Phase Response Signaling, B Cell Receptor Signaling, IL-8 Signaling, and ROS Production in Macrophages. Also seen in this study were novel unannotated transcripts that were significantly affected by HZE. To investigate the biological functions of these novel transcripts, we used a machine learning technique known as self-organizing maps (SOMs) to characterize the transcriptome expression profiles of 60 samples (45 HZE-irradiated, 15 non-irradiated control) from liver tissues. A handful of localized modules in the maps emerged as groups of co-regulated and co-expressed transcripts. The functional context of these modules was discovered using overrepresentation analysis. We found that these spots typically contained enriched populations of transcripts related to specific immunological molecular processes (e.g., Acute Phase Response Signaling, B Cell Receptor Signaling, IL-3 Signaling), and RNA Transcription/Expression. A large number of transcripts were found differentially expressed post-HZE irradiation. These results provide valuable information for uncovering the differences in molecular mechanisms underlying HZE specific induced HCC carcinogenesis. Additionally, a handful of novel differentially expressed unannotated transcripts were discovered for each HZE ion. Taken together, these findings may provide a better understanding of biological mechanisms underlying risks for HCC after HZE irradiation and may also have important implications for the discovery of potential countermeasures against and identification of biomarkers for HZE-induced HCC.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

Mechanical and Thermophysical Properties of 3D-Printed SiC before and after Neutron Irradiation – FY21

This report presents the mechanical and thermophysical properties of 3D-printed SiC before and after neutron irradiation that have been evaluated to assess the fuel matrix material for the Transformational Challenge Reactor (TCR). The TCR fuel form consists of an additively-manufactured silicon carbide (SiC) matrix and uranium nitride tristructural isotropic (UN TRISO) fuel particles, which is manufactured through a newly developed processing route combining binderjet 3D printing, TRISO fuel particle loading, and chemical vapor infiltration (CVI). Because the fuel matrix is a primary component of the TCR core and its response to mechanical and thermal loads during operation is one of the most influential factors on the integrity of TCR core, testing and evaluation have focused on producing mechanical and thermophysical properties data for the binderjet/CVI SiC. Baseline mechanical and thermophysical properties were measured from the disk specimens printed for different and sizes orientations, which included equibiaxial flexural failure strength, elastic constants, thermal diffusivity and conductivity, density, and the coefficient of thermal expansion. Flexural failure strength datasets showed similar Weibull distributions regardless of sample variants including different orientations. The mean failure strengths of the 3D-printed SiC variants were in the range of 280–310 MPa, which are slightly lower than that of the chemical vapor deposition (CVD) SiC. Thermophysical test results showed that specific heat and thermal expansion are not sensitive to the build direction of SiC samples, while thermal conductivity is highly dependent on the build direction and can be correlated to the anisotropic character of the 3D-printed SiC. Neutron irradiation tests were carried out on the 3D-printed 6-mm diameter SiC disk specimens. Irradiation was carried to 2.3 dpa over a temperature range of 360–880°C. No significant degradation in strength was observed in SiC after irradiations in various conditions and with different orientations. Anisotropy that had been observed in the thermal conductivity of 3D-printed SiC prior to irradiation vanished after irradiation as the irradiation defect thermal resistivity accumulated in the material. Electron microscopy of the microstructure after neutron irradiation showed distinct defect morphologies in the heterogenous material, but no evidence for irradiation-induced cracking or degradation in the microstructure was observed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Irradiation of Sintered Mo Disks with Presence of Organic Impurities in He Gas Flow

Argonne National Laboratory (Argonne) is assisting NorthStar Medical Technologies in the development of a domestic supply of 99 Mo. Specifically, the present study focuses on the production of 99 Mo-feed solution used by the RadioGenix™ 99m Tc generator. During the target-irradiation phase of production, impurities can potentially be introduced into the feed, and can lead to disturbance of ligand- 99m Tc complexation chemistry and contamination of the final radiopharmaceutical that directly interacts with the patient. To address this issue, Argonne performed irradiations and chemical processing to identify whether the potential contamination of He flow with hydrocarbon oil during irradiation affects the radiochemical purity of the final K 2 MoO 4 (K 2 TcO 4 ) in 5M KOH solution. To mimic the conditions of real irradiation at NorthStar, Argonne used its electron linear accelerator and Van de Graaff facilities, heated the target and oil source to >800°C, and controlled oxygen in the presence of He during irradiation. Following irradiation, scanning electron microscopy (SEM) and carbon analysis (CA) were used to detect carbon contamination on the solid targets. The radiochemical purity of the dissolved targets was studied via thin-layer paper chromatography (TLC). As a result of these experiments, small regions of the surface of some irradiated disks were found to be high in carbon, but the total carbon content was still negligible in comparison to the reference sample, which had experienced no irradiation or contact with oil. The summarized results from the SEM, CA, and TLC tests lead to the conclusion that even an excess of oil and heating during irradiation do not affect the radiochemical purity of the final product.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Experimental Evaluation of Deformation and Fracture Mechanisms in Highly Irradiated Austenitic Steels

The present report documents recent experimental results of analysis using scanning electron microscopy/electron backscatter diffraction (SEM-EBSD) of plastic deformation mechanisms and strain localization phenomena in austenitic steels irradiated by neutrons. Experiments were performed with specimens irradiated to 125 dpa and, additionally, with specimens that experienced radiation-induced swelling up to 3%. Section 1 briefly analyzes the deformation localization in irradiated steels and its consequences on the material performance. The section describes the advantages and importance of the SEM-EBSD approach combined with in situ mechanical testing capability. Section 2 briefly introduces the experimental tools and methods (i.e., SEM/EBSD in situ tensile frame, electric discharge machine to manufacture irradiated specimens) and describes the investigated materials (i.e., element composition, irradiation conditions, and general microstructure). Section 3 describes the key experimental results and provides a brief analysis and comparison with the datasets obtained earlier within the same task (i.e., low-dose specimens). The discussion focuses on EBSD microstructure maps with strain localization features, misorientation evolution as a function of strain, and observed deformation mechanisms. Section 4 evaluates data collected in recent years on highly irradiated steel and estimates the possible misorientation evolution under irradiation. The section introduces and discusses the concept of in-service-induced damage as an irradiation-assisted stress-corrosion cracking precursor. Section 5 summarizes the work performed. As expected, the present work results are beneficial for exploring and understanding degradation mechanisms in highly irradiated in-core materials found in light water reactors after long-term in-service life.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

The Recovery of Medical Isotope 188 W from Irradiated W Metal Target - A New Approach

Tungsten-188 is in widespread use in 188 W(t 1/2 = 69 d )/ 188 Re(t 1/2 = 16.9 h ) biomedical generators. Oak Ridge National Laboratory has been providing this product to the world since 1999. At ORNL, 188 W is produced via irradiation in ORNL’s High Flux Isotope Reactor (HFIR). Enriched 186 W targets in the form of sintered metallic pellets or rings achieve a compact loading in the irradiation vessel, providing a high yield per unit target. The enrichment of the target is >90% 186 W, and this isotope undergoes double neutron capture to produce the desired 188 W product. While 188 W is produced by neutron bombardment, 191 Os(t 1/2 = 15.4 d ) is simultaneously produced as a by-product and expected to be separated from 188 W by postirradiation treatment.In the current processing pathway, the irradiated W metal rings are first converted into an oxide form of WO 3 by heating the irradiated W metal target at 750°C in a quartz reaction vessel inside a vertical furnace under a constant flow of air. During heating, W metal reacts with oxygen in the air to produce WO 3 , which is soluble in 6 M NaOH for preparation of 188 W product. This oxidation process also converts 188 Os (the decay daughter of 188 W) and 191 Os (15.4 d , the irradiation produced byproduct) into OsO 4 , a highly volatile and toxic gas. The gaseous effluents driven from the quartz reaction vessel are passed through a scrubbing array to remove OsO 4 before the air is discharged from the process. This heterogeneous oxidation method simultaneously achieves goals of (1) converting metal target to a soluble oxide form and (2) removing volatile OsO 4 away from the solid WO 3 product by air flow and absorbing the harmful Os species by the scrubbing array. But this method has two potential problems as well: (1) O 2 reacts with only W metal at high temperatures, not with W alloyed with other elements. The O 2 –W reaction will be retarded when formation of WRe or WC occurs, or even when a layer of non-W materials on the surface of the irradiated W rings.; (2) 100% absorption of OsO 4 of high yield (>90%) from the reaction of Os + O 2 is a strict requirement to the OsO 4 scrubbing system--so NaOH scrubbers of a redundant size (2x 1.5 L) are in use for safety reasons.To resolve above two potential problems, direct dissolution of the irradiated W metal target by a selected reagent is a preferred pathway to avoid heating step with generation of tremendous amount of volatile OsO 4 . Hydrogen peroxide (H 2 O 2 ) is such a candidate to dissolve W in forms of either metal or alloys, although literature lacks information of solubilities of Re or Os in H 2 O 2 . With experimental results of dissolving non-radioactive W, Re and Os in H 2 O 2 under various conditions, this report illustrates a method of H 2 O 2 dissolution for irradiated W target, with a complete dissolution of W and Re, but ≤10% dissolution of Os (converted into gaseous OsO 4 and carried out into a scrubbing for absorption) during processing irradiated W target. The portion of undissolved Os can be separated from W solution by a follow up filtration step. Solubilities of W, Re and Os in H 2 O 2 at a temperature range from 14° to 50°C are presented. And a dissolution rate of W metal per surface area of W metal in H 2 O 2 is calculated based on results of dissolving a W metal cylinder of known surface area in H 2 O 2 at room temperature without stirring.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

In-situ and ex-situ characterization of irradiated AM materials

This study investigates the irradiation performance of laser powder bed fusion (LPBF) 316L and 316H stainless steels (SS), and their wrought counterparts. While LPBF 316L has been more studied, LPBF 316H is relatively new, with little prior data on its irradiation behavior. The research involved fabricating, preparing, irradiating, and characterizing six materials: LPBF 316L (two variants), LPBF 316H (two variants), and wrought 316L and 316H. LPBF materials were subjected to various heat treatments, including solution annealing and stress relief, and were irradiated using in-situ and ex-situ ion techniques at temperatures of 300°C and 600°C, with doses ranging from 0.2 dpa to 25 dpa. Under ex-situ irradiation with 4 MeV Ni 2+ ions at 600°C, the dislocation cell structures in LPBF316L-1 and LPBF316H-1 gradually evolved into a uniform dislocation network. At a low irradiation dose (0.2 dpa), the cell structures were still partially visible. At 2 dpa, a uniform dislocation network formed, though remnants of the original cell structure were still discernible as contrast domains. At 5 dpa and 10 dpa, even the contrast domains vanished. For irradiation at 300°C, the cell structure was still clearly visible at 0.2 dpa for LPBF316L-1. For LPBF316H-1, the dislocation cell structure was less apparent at the same dose but was still recognizable. With further increase in dose at 300°C, the dislocation cell walls were completely replaced by irradiation-induced defects and no longer visible. Void formation was observed for irradiation at 600°C. At 2 dpa, no evident voids were observed in either LPBF316L-1 or LPBF316H-1.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Flux effects in precipitation under irradiation simulation of Fe-Cr alloys

Radiation-enhanced precipitation of Cr-rich α’ in irradiated Fe-Cr alloys, which results in hardening and embrittlement, depends on the irradiating particle and the displacement per atom (dpa) rate. Here, we utilize a Cahn-Hilliard phase-field based approach, that includes simple models for nucleation, irradiating particle and rate dependent radiation-enhanced diffusion and cascade mixing to simulate α’ evolution under neutrons, heavy ions, and electron irradiations. Different irradiating particles manifest very different cascade mixing efficiencies. The model was calibrated using neutron data. For cascade inducing neutron/heavy-ion dpa rates at 300 °C between 10-8 and 10-6 dpa/s the model predicts approximately constant number density, decreasing radius, decreasing α’ Cr composition, and lower α’ volume fraction. The model then predicts a dramatic transition to no α’ formation above approximately 10-5 dpa/s, while electron irradiation, with weak mixing, had little effect at dpa rates up to 10-3 dpa/s. These model predictions are consistent with experiments. We explain the results in terms of the flux dependence of the radiation enhanced diffusion, cascade mixing, and their ratio, which all vary significantly in relevant flux ranges for neutron and cascade inducing ion irradiations. These results show that both cascade mixing and radiation enhanced diffusion must be accounted for when attempting to emulate neutron-irradiation effects using accelerated ion irradiations.

Ke, Jia-Hong↗
DOE OSTI2025

Raman spectroscopy of neutron irradiated silicon carbide

The effects of neutron irradiation on microstructural evolution and the resultant changes in physical and mechanical properties are of critical importance for the development of silicon carbide (SiC) materials for nuclear applications. This study neutron-irradiated βSiC under a wide range of conditions at temperatures between 235 and 750°C and neutron doses of 0.01–11.8 displacements per atom, and then evaluated the effects on the SiC structure using Raman spectroscopy. The SiC optical phonon lines were shifted to lower wavenumbers by irradiation. Correlations were found among the wavenumber of the longitudinal optical phonon line, irradiation-induced swelling, and irradiation temperature. The peak shift also correlated indirectly with decreasing thermal conductivity of irradiated SiC. The irradiation-induced peak shift is explained by combinations of lattice strain, reduction of the elastic modulus, and other factors including decreasing coherent domain size. These findings bridge irradiation-induced microstructural changes and property changes and illustrate how Raman spectroscopy is a useful tool for nondestructively assessing irradiated SiC materials for nuclear applications.

Koyanagi, T.↗
NASA NTRS1976

Health protection and food preservation by gamma irradiation

Results of several major studies on food systems for space missions beginning with Apollo 12 through Apollo-Soyuz and investigations of the application of irradiation to food for manned space flight are reported. The study of flight food systems involved the application of radurization (pasteurizing levels) doses of gamma irradiation to flour and bread supplied by Pepperidge Farms in advance of the missions. All flights from Apollo 12 through 17 carried irradiated fresh bread. On Apollo 17, cooperation with Natick Laboratories permitted the introduction of a ham sandwich using irradiated bread and irradiated sterile ham. Investigations centered on irradiated bread were conducted during the course of these missions. Studies were applied to the concept of improving fresh bread from the point of view of mold inhibition. The studies considered how irradiation could best be applied at what levels and on a variety of bread types. Throughout the studies of the application of gamma irradiation the emphasis was placed upon using low levels of irradiation in the pasteurizing or radurizing doses--under a Megarad. The primary goal was to determine if a public health benefit could be demonstrated using radurization along with food preservation and food quality improvements. The public health benefit would be parallel to that of pasteurization of milk as a concept. Publications are included providing the details of these observations, one dealing with the flour characteristics and the other dealing with the influence on fresh bread types. These demonstrate the major findings noted during the period of the studies examining bread.

Source record↗
NASA NTRS2007

Fluence Uniformity Measurements in an Electron Accelerator Used for Irradiation of Extended Area Solar Cells and Electronic Circuits for Space Applications

It is common to have liquid crystal displays and electronic circuit boards with area sizes of the order of 20x20 sq cm on board of satellites and space vehicles. Usually irradiating them at different fluence values assesses the radiation damage in these types of devices. As a result, there is a need for a radiation source with large spatial fluence uniformity for the study of the damage by radiation from space in those devices. Kent State University s Program on Electron Beam Technology has access to an electron accelerator used for both research and industrial applications. The electron accelerator produces electrons with energies in the interval from 1 to 5 MeV and a maximum beam power of 150 kW. At such high power levels, the electron beam is continuously scanned back and forth in one dimension in order to provide uniform irradiation and to prevent damage to the sample. This allows for the uniform irradiation of samples with an area of up to 1.32 sq m. This accelerator has been used in the past for the study of radiation damage in solar cells (1). However in order to irradiate extended area solar cells there was a need to measure the uniformity of the irradiation zone in terms of fluence. In this paper the methodology to measure the fluence uniformity on a sample handling system (linear motion system), used for the irradiation of research samples, along the irradiation zone of the above-mentioned facility is described and the results presented. We also illustrate the use of the electron accelerator for the irradiation of large area solar cells (of the order of 156 sq cm) and include in this paper the electrical characterization of these types of solar cells irradiated with 5 MeV electrons to a total fluence of 2.6 x 10(exp 15) e/sq cm.

Uribe, Roberto M.↗
NASA NTRS2015

Fe and O EELS Studies of Ion Irradiated Murchison CM2 Carbonaceous Chondrite Matrix

Introduction: The physical and chemical response of hydrated carbonaceous chondrite materials to space weathering processes is poorly understood. Improving this understanding is a key part of establishing how regoliths on primitive carbonaceous asteroids respond to space weathering processes, knowledge that supports future sample return missions (Hayabusa 2 and OSIRISREx) that are targeting objects of this type. We previously reported on He+ irradiation of Murchison matrix and showed that the irradiation resulted in amorphization of the matrix phyllosilicates, loss of OH, and surface vesiculation. Here, we report electron energy-loss spectroscopy (EELS) measurements of the irradiated material with emphasis on the Fe and O speciation. Sample and Methods: A polished thin section of the Murchison CM2 carbonaceous chondrite was irradiated with 4 kilovolts He(+) (normal incidence) to a total dose of 1 x 10(exp 18) He(+) per square centimeter. We extracted thin sections from both irradiated and unirradiated regions in matrix using focused ion beam (FIB) techniques with electron beam deposition for the protective carbon strap to minimize surface damage artifacts from the FIB milling. The FIB sections were analyzed using a JEOL 2500SE scanning and transmission electron microscope (STEM) equipped with a Gatan Tridiem imaging filter. EELS spectra were collected from 50 nanometer diameter regions with an energy resolution of 0.7 electronvolts FWHM at the zero loss. EELS spectra were collected at low electron doses to minimize possible artifacts from electron-beam irradiation damage. Results and Discussion: Fe L (sub 2,3) EELS spectra from matrix phyllosilicates in CM chondrites show mixed Fe(2+)/Fe(3+) oxidation states with Fe(3+)/Sigma Fe approximately 0.5. Fe L(sub 2,3) spectra from the irradiated/ amorphized matrix phyllosilicates show higher Fe(2+)/Fe(3+) ratios compared to spectra obtained from pristine material at depths beyond the implantation/amorphization layer. We also obtained O Ka spectra from phyllosilicates in both regions of the sample. The O Ka spectra show a pre-edge feature at approximately 530.5 electronvolts that is related to O 2p states hybridized with Fe 3d states. The intensity ratio of the O Ka pre-edge peak relative to the main part of the O Ka edge (that results from transitions of O 1s to 2p states) is lower in the irradiated layer compared to the pristine material and may reflect the loss of O (as OH) as was observed by IR spectroscopy. Conclusions: In addition to amorphization and OH loss, EELS spectra of He(+) irradiated matrix phyllosilicates in Murchison show that some of the Fe(3+) is reduced to Fe(2+). Spectral deconvolution is underway to extract quantitative ratios from the EELS spectra.

Keller, L. P.↗
DOE OSTI2021

Destructive Examination of a FeCrAl-UO 2 Irradiation Test

Several destructive postirradiation examinations were performed on an irradiation specimen that coupled an early iron-chromium-aluminum (FeCrAl) candidate alloy cladding with UO2 pellets. This irradiation test was designed to investigate the early life performance and compatibility of the FeCrAl-UO 2 cladding-fuel system under prototypic light water reactor neutronic conditions. Additionally, these tests were expected to provide neutron irradiated samples for severe accident testing of this fuel system. The irradiation studied in this work was part of the ATF-1 series of drop-in style irradiations performed in the Idaho National Laboratory Advanced Test Reactor. The rodlet studied in this work is one of three similar rodlets irradiated in ATF-1 that had approximately 7.6 cm of cladding machined from a wrought FeCrAl alloy and fueled with a 6.1 cm stack of UO 2 pellets. The outer diameter of the cladding was ~0.94 cm and the inner diameter was ~0.83 cm. After irradiation and non-destructive examination in Idaho this rodlet was shipped to the Oak Ridge National Laboratory hot-cells for further examination. The irradiated rodlet was sectioned into several samples for microstructural, micromechanical, and severe accident testing. During sectioning, it was noted that the fuel was not firmly bonded to the cladding and could be readily removed from small cladding slices. Microstructural characterization of fuel cross sections also revealed no significant interaction between the fuel and the cladding. Samples were also prepared for microhardness testing. To prepare for high temperature oxidation testing, the fuel was dissolved from segments of the cladding. High temperature oxidation testing of cladding segments was performed at 1200°C and 1300°C in a steam environment. Comparisons between the oxidation of this FeCrAl alloy in its neutron irradiated state, as-fabricated state, and the oxidation of Zircaloy-2 are made. The oxidation testing will be followed by ring compression testing to evaluate ductility. The microstructure of the samples after oxidation and ring compression testing will also be analyzed.

Harp, Jason↗