Search NASA⌕ Search

SEARCH · Search NASA

Results for “RADIATION SOURCE”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Effect of Lithium Doping on MgO Hydroxylation and Carbonation

Recovery of magnesium from brines can potentially be used to source MgO (periclase) as a CO 2 sorbent or for Mg-based cements. However, it is not clear how common impurities in brines, such as lithium, affect the resulting MgO reactivity. Here, to test the effect of lithium incorporation on MgO reactivity for hydration and carbonation, we combined computational simulations with experiments. Experimentally altered (Mg,Li)O with a low dopant concentration (0.012 ± 0.002% w/w Li) was characterized using synchrotron-based X-ray scattering and high-resolution electron microscopy to measure reaction layer formation on (Mg,Li)O. Single-crystal X-ray diffraction analysis of (Mg,Li)O demonstrates that the incorporation of lithium leads to the formation of oxygen vacancies. The presence of vacancies is likely causing faster hydroxylation rates as predicted by ab initio molecular dynamics simulations. However, the faster hydroxylation rates likely lead to faster passivation of the surface because we observe thinner reaction layers on (Mg,Li)O samples both over short time periods (30 days) and over long time periods (28 years). After 28 years, the reaction layer on the (Mg,Li)O sample was less than one-third of the thickness of that of the pure MgO sample. In addition, over 30 days, reaction layers on (Mg,Li)O samples primarily formed at steps rather than on terraces, in contrast to our previous observations on MgO. Based on our results, naturally occurring impurities in MgO modify its reactivity even at very low concentrations and need to be considered for accurate reaction rate prediction for application of MgO as a CO 2 sorbent or in cements.

MgO↗
DOE OSTI2026

Titanium alloy response sensitivity to variations in spectral reconstructions of National Ignition Facility xenon line-emission x-ray sources

Thermomechanical shock experiments on the National Ignition Facility (NIF) aim to study high strain rate dynamic material response. In such experiments, the NIF laser is used to generate high fluence x-ray emission sources, which irradiate material samples of interest. Under sufficiently high x-ray energy deposition, thermomechanical impulses are generated in the materials. While it is known that the characteristics of x-ray generated impulses vary as a function of incident x-ray spectra, it remains unclear how spectral assumptions and uncertainties in NIF spectral reconstructions affect our interpretation of impulsive loading. Here, in this paper, we simulate the response of a standard titanium alloy baseline sample to synthetic analytically derived and measured NIF xenon line-emission x-ray sources with a radiation hydrodynamics code. We vary the source spectral characteristics based on different source reconstruction techniques to understand the resulting variation in baseline sample response and compare the simulated response with experimental results. We find that the response is highly sensitive to assumptions made about the spectral contents and that knowledge of spectral uncertainties bounds our understanding of the resulting material response. The results of this effort help to extend our ability to use baseline material samples to extract quantitative properties from x-ray experiments on the NIF.

Alloys↗
DOE OSTI2025

Unfolding of the Villin Headpiece Domain: Revealing Structural Heterogeneity with Time‐Resolved X‐Ray Solution Scattering and Markov State Modeling

Understanding protein folding pathways is crucial to deciphering the principles of protein structure and function. Here, the unfolding dynamics of the 35‐residue villin headpiece (HP35) and a norleucine‐substituted variant (2F4K) using a combination of experimental and computational techniques is investigated. Time‐resolved X‐ray solution scattering coupled with equilibrium molecular dynamics simulations and Markov state modeling reveals distinct unfolding mechanisms between the two variants: HP35 and 2F4K. Specifically, HP35 exhibits a two‐state unfolding process, whereas an intermediate state is identified for the 2F4K mutant. A Markov state model constructed from simulations is used to map atomic‐level transitions to experimental observations, providing insights into the role of sequence variations in modulating folding pathways. The findings underscore the importance of integrating experimental and computational approaches to unravel protein unfolding mechanisms between heterogenous structural ensembles.

Nijhawan, Adam K. [Department of Chemistry Northwe↗
DOE OSTI2024

Californium-252 production at the High Flux Isotope Reactor - II: Comparison between the highly enriched uranium and a proposed low-enriched uranium core

This is the second paper on a 252 Cf production study performed in support of efforts to convert the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The first paper primarily focuses on validating computational tools and nuclear data. This companion paper evaluates another critical aspect: the 252 Cf production capability with a proposed LEU core. HFIR must maintain its world-class performance and missions following conversion and because 252 Cf is a vital, multipurpose neutron-emitting radioisotope, the ability to efficiently produce 252 Cf must be preserved. In this study, the HFIRCON transport and depletion tool, several nuclear data libraries, and Campaign 78 data were used to compute 252 Cf production, sensitivity, and safety metrics. Further, results indicate the 252 Cf production and production rates are slightly higher with a 95MW th LEU core compared with those obtained with the 85MW th HEU core. Additionally, the target peak fission rate densities, discharge cumulative fission densities, and heat deposition rates with the LEU core are within a few percent of those calculated with the HEU core. The findings suggest HFIR’s 252 Cf production capability can be effectively maintained with an LEU core without adversely affecting the safety metrics.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Californium-252 production at the High Flux Isotope Reactor - I: Validation study using campaign data

This paper presents a series of 252 Cf production validation and code-to-code comparison studies performed based on data from the production campaigns at the High Flux Isotope Reactor (HFIR). These studies support efforts to convert HFIR from using highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. HFIR must maintain its world-class performance and missions following this conversion, and because 252 Cf is a vital neutron-emitting radioisotope used for a variety of high-impact applications (e.g., reactor startup, cancer treatment), the ability to efficiently produce 252 Cf must be preserved. In this work, the HFIRCON, Shift, ORIGEN, and TCOMP codes were deployed, and several sets of data libraries were investigated to better understand the calculation codes and the data biases. As-loaded target composition data, as-run irradiation history data, and post-irradiation measurements from recent multi-cycle irradiation campaigns of the HEU core were used to validate and determine methodology biases. Further, the findings demonstrated a good agreement, with results falling within 3 standard deviations of measurements. This paper lays the ground work for the second paper, which evaluates and compares 252 Cf production and safety metrics with the HEU core and a proposed LEU core.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

An approach for spent nuclear fuel containment integrity verification using gas tagging

Verification of containment integrity is required for spent nuclear fuel (SNF) managed by the commercial nuclear industry and U.S. Department of Energy (DOE), especially after extended storage. Certain SNF storage systems, such as the DOE road-ready dry storage system, hold several packaged containments within a welded over-canister. These packaged containments are called Department of Energy Standard Canisters (DOESCs). DOESC leakage identification is challenging because their containment boundary cannot be accessed for testing and their contents (i.e., SNF and fill gas) are often similar. There are concerns that this could result in costly characterization and repackaging operations of DOE road-ready dry storage systems if compromised DOESCs are suspected. Here, to address these concerns, this paper presents an approach for applying a gas tagging process using xenon to uniquely identify compromised inaccessible containments following extended storage. The containments considered for this application are seven DOESCs, each packaged within a single over-canister. Two different SNF loading configurations from the Advanced Test Reactor and Fort Saint Vrain nuclear power plant are considered. These configurations are used to represent research reactor aluminum-clad spent nuclear fuel (ASNF) and TRi-structural ISOtropic (TRISO) SNF types. Results for this application show that for ASNF and TRISO type fuels for which the selected fuels are representative, the volume of taggant required at loading is determined primarily by the lower detection limit and leak rate of taggant from a compromised DOESC, rather than the amount of fission-generated xenon in the loaded fuel. While the application presented is suited for larger leaks, smaller leaks could be detected by modifying certain design parameters. This gas tagging approach can also be applied to other DOE containments and advanced reactor SNF storage systems.

07 - ISOTOPES AND RADIATION SOURCES↗
DOE OSTI2025

Oxygen stable isotopes in the nuclear fuel cycle: Assessment of the potential for determining the fabrication and provenance history of anhydrous and hydrous uranium oxides

Determining the origin and history of interdicted nuclear materials is a central challenge in nuclear forensics. The oxygen stable isotope composition of uranium oxide compounds has emerged as a promising forensic signature, attracting increasing attention since the early 2000s. This review examines analytical techniques for measuring oxygen isotope compositions in uranium oxides and evaluates how the nuclear fuel production cycle introduces or modifies these isotopic signatures. The potential for forensic geolocation is explored through workflows that calibrate the relationship between environmental water oxygen isotopes and those found in uranium oxides. Key strengths and limitations of this approach are assessed, including gaps in knowledge related to isotope fractionation during specific stages of the fuel cycle, and processing facility water inputs. The importance of proper sample handling and storage under inert atmospheres, as well as a deeper understanding of both intra-sample oxygen isotope heterogeneity, and hydrous uranium oxide phase formation, is highlighted for improving the reliability of forensic interpretations. In conclusion, the development of uranium oxide standards with well-characterized δ 18 O values and international collaboration toward consensus on their use are identified as essential steps for advancing the field.

Attribution↗
DOE OSTI2024

Measurement of the 70 Ge( d , n ) 71 As cross section from 2.1 MeV to 10 MeV

This work reports experimental 70 Ge(d,n) 71 As cross sections producing Arsenic-71 (t 1/2 = 65.3 h, 28% β + ), a potentially useful diagnostic radionuclide. Target stacks containing two nat Ge foils, a nat Ni monitor foil, and an Al degrader were irradiated with 5.5–10 MeV deuterons using an FN Tandem Van de Graaff accelerator. 71 As and 61 Cu, 56 Co, and 58 Co (from nat Ni(d,x)) were quantified in the foils by high purity germanium gamma spectrometry. Furthermore, the measured cross sections are larger than previously reported and suitable for producing preclinical quantities of diagnostic radiopharmaceuticals.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Comparison of measurement techniques and sorption of radium-226 in low and high salinity aqueous samples

Human activities have the potential to redistribute radium (Ra) in the marine environment in a manner that may necessitate monitoring or management of subsequent human or environmental exposures. There is therefore a need to identify accurate and accessible techniques for Ra measurement in high salinity samples and to describe the distribution of Ra in estuarine and marine environments, but most efforts in these areas have focused on low salinity matrices. In addition, rapid and reliable measurements are crucial for time-sensitive samples such as short-lived isotopes or emergency situations. The objective of this study is to describe the limits of detection, cost, and relative ease for measurement of Ra in both low and high salinity aqueous samples via three analytical methods: liquid scintillation counting (LSC), high purity germanium (HPGe) gamma spectrometry, and inductively coupled plasma mass spectrometry (ICP-MS). To contextualize these measurements for real-world scenarios, the partitioning of 226 Ra to substrates relevant to the marine environment was also characterized. Although HPGe detection with solid phase extraction had the lowest limit of detection for low salinity samples (0.27 Bq L −1 ), poor 226 Ra recovery for high salinity samples and high materials costs make this method prohibitive for many users. Limits of detection for high salinity samples were lower for LSC (1.28 Bq L −1 ) than for ICP-MS without dilution (11.4 Bq L −1 ), but significant and unexpected degradation of the high salinity LSC standards was observed after six months. Furthermore, our preferred measurement method for high salinity Ra samples is ICP-MS with sample dilution as necessary to reduce matrix effects.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Cyclotron production of a 204 Bi/ 204m Pb generator by 24 MeV proton irradiation of natural Pb foil targets

This work focuses on the formation of a generator of 204 Bi/ 204m Pb for future applications in perturbed angular correlation spectroscopy and biological radiotracing with lead and bismuth isotopes. Generators were formed from the irradiation of natural Pb targets with 24 MeV protons producing 203,204,205,206,207 Bi that were rapidly dissolved in 3 M HNO 3 and 30 % H 2 O 2 and exchanged into a 0.1 M HCl matrix. Use of 100 mg anion exchange resin (AG 1-x8) allowed 203,204m Pb to be eluted in 600 μL of 0.1 M HCl. Furthermore, the longer-lived 205,206 Bi were eluted in 5 mL or 3 mL of either 0.25 M sodium acetate buffer pH 4.4 or 1.0 M H 2 SO 4 . Initial radiolabeling was conducted with eluted 206 Bi using DTPA and DOTA and 203 Pb using TCMC-Azide.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2026

Processing iridium foil targets for the production of radioplatinum isotopes

Dissolving noble metals remains challenging because of their high reduction potentials and tendency to passivate as oxides. This barrier is acute for iridium metal targets used to produce platinum radioisotopes, where post-irradiation processing is costly and hazardous with conventional routes. Here, we report a rapid alkali-fusion method that converts Ir metal directly into chloride media compatible with downstream separations. In the process, an Ir foil is fused with Na 2 O 2 and KCl at temperatures above the KCl melting point; the molten salt enhances heat transfer and facilitates O 2− /Na + transport to the Ir surface, overcoming Ir's oxidation resistance. The fused cake is then cooled and leached in 10 M HCl, yielding Ir in solution without specialized microwave digestion or HF. Pt separation from bulk Ir was evaluated using extraction-chromatography resins. A newly formulated ionic-liquid resin containing Cyphos IL 104 (E104) was compared with two commercial resins (TrisKem TK200 and TK201). All resins achieved >90 % Ir recovery. Pt co-recovery depended on resin chemistry: TK200 recovered 85.9 % Pt, TK201 recovered 54.1 % Pt, and E104 recovered 72.9 % Pt. This study introduces a fast, scalable dissolution route for solid Ir targets and demonstrates resin choices that tune Ir/Pt separation, lowering the barrier to routine production and purification of Pt radioisotopes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Development of a SnO 2 -based 44 Ti/ 44 Sc generator for medical applications

Towards application of 44 Sc for diagnostic nuclear medicine, a 44 Ti/ 44 Sc generator based on an inorganic resin has been evaluated. Unlike other radionuclide generators used for medical applications, the long-term retention of the parent 44 Ti is vital due to its long half life. In this work, tin dioxide (SnO 2 ), a robust inorganic-based resin, has been synthesized and used as the stationary phase for a 44 Ti/ 44 Sc generator. The sorption behavior of 44 Ti/ 44 Sc was tested on SnO 2 with varying acids, concentrations, and times. Preliminary batch study results showed >88 % 44 Ti retention to the resin at lower acid concentrations (0.05 M HNO 3 and 0.05 M HCl). A pilot generator was evaluated for a year, demonstrating 85.3 ± 2.8 % 44 Sc elution yields and 0.71 ± 0.14 % 44 Ti breakthrough in 5 M HNO 3 . Based on capacity studies, a 7.4 MBq (200 µCi) upscaled generator system was constructed for further evaluation of the SnO 2 resin stability and the efficacy of the eluted 44 Sc for radiolabeling. 44 Sc could be regularly eluted from this generator in 5 M HNO 3 with an overall average radiochemical yield 84.7 ± 9.5 %. Post-elution processing of the 44 Sc with DGA-normal resin removed all 44 Ti present and allowed for high 44 Sc-DOTA labeling yields of 94.2 ± 0.5 %. Overall, SnO 2 has been shown to be a viable material for a 44 Ti/ 44 Sc generator.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Identification of uranium oxidation states using oxygen K-edge scanning transmission X-ray microscopy

The field of nuclear forensics is growing in importance, and the increasing capabilities at synchrotron radiation light sources enable non-destructive characterization of oxide particles with better spatial, compositional, and oxidation state speciation resolution than ever before. Here, uranium oxide particles derived from multiple wet chemical processing methods were examined using a scanning transmission X-ray microscope (STXM), and a weakly-supervised method was developed to automatically analyze the collected data. Multiple uranium oxidation states were observed and quantified within and between samples, yielding information about differences between particles produced via the various processing routes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Characterizing secondary neutrons at BLIP for isotope production applications

Fast secondary neutrons created at the Brookhaven Linac Isotope Producer (BLIP) facility following proton irradiation were characterized by the foil activation technique and compared with FLUKA Monte Carlo simulations. The FLUKA-simulated neutron flux was spectrally adjusted following the maximum entropy formalism using the International Reactor Dosimetry and Fusion File (IRDFF-II), with predictions agreeing with experimental measurements to within 9 % following the adjustment procedure. Here, a multitude of degrader configurations were simulated to assess the feasibility of improving the fast (E n > 20 MeV) secondary neutron yield at the proposed neutron target position (“N-slot”). A configuration where the N-slot is closest to the proton degrader produced the highest fast neutron yield, with tungsten degraders achieving the best performance. Assuming the optimized target-degrader configuration proposed in this work, we discuss potential isotope production opportunities with secondary neutrons. In most cases the yields are in the order of several mCi.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Lithium isotopic analysis in depleted lithium salts

Lithium isotope separation processes require characterization of the isotopic compositions of the different output phases of the process. While considerable effort to develop isotope separation methods has been ongoing for decades, the analytical methods associated with these analyses are rarely provided. Here, in this work three types of mass spectrometers (Q-ICP-MS, MC-ICP-MS, TIMS) were tested for the ability to measure depleted lithium salts that are relevant materials for isotopic separations. Results indicate that, once stabilized, the Q-ICP-MS can provide equal precision for Li isotopic samples compared to TIMS or MC-ICP-MS. However, MC-ICP-MS consistently provides the highest precision for salts across both depleted and natural isotopic abundances. These data demonstrate that Q-ICP-MS is an appropriate choice for analysis of depleted lithium salts, especially when rapid analyses are desired for online process monitoring.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Reaction-Induced Fracturing of Porous Carbonate Rocks during Volume-Increasing Replacement by Witherite

Volume-increasing replacement reactions can lead either to fracturing of the parent phase or to the formation of cohesive layers that passivate further reaction, but the factors that drive one outcome or the other are not understood. In this experimental study, we investigated the volume-increasing replacement of carbonate rocks (consisting of CaCO 3 and CaMg(CO 3 ) 2 ) of different porosity by witherite (BaCO 3 ). Samples were characterized using scanning electron microscopy, Raman spectroscopy, small-angle neutron/X-ray scattering, X-ray tomography, and scanning transmission electron microscopy. We observed the formation of a witherite reaction layer and witherite formation within pores and along grain boundaries. Despite this being a volume-increasing replacement reaction, newly formed witherite was porous, potentially allowing further replacement. Filled fractures were observed in the low-porosity carbonates, whereas witherite formed within pores in high-porosity carbonates. We conclude that fracturing of the parent phase versus passivation is contingent on the initial microstructure of the parent with an optimal degree of porosity required for fracturing.

Calcite↗