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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.

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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Cooling plate assembly for plasma windows positioned in a beam accelerator system

A beam accelerator system operable to produce a medical isotope, including an ion accelerator that generates an ion beam; a low-pressure chamber; an anode adjacent and fluidly connected to the low-pressure chamber; a plasma window adjacent and fluidly connected to the anode; and a cathode housing adjacent and fluidly connected to the plasma window. The plasma window has a plurality of plates, each plate having an aperture that is aligned with an aperture in one or more adjacent plates to form a plasma channel. One or more plates in the plurality of plates includes a unitary plate having an aperture therein, and one or more cooling channels entering the unitary plate at a first side of the unitary plate and exiting the unitary plate at a second side of the unitary plate. The one or more cooling channels run through a thickness of the unitary plate.

Gribb, Tye↗
DOE OSTI2020

Recycling Isotopically Enriched Mo from the RadioGenix Waste Stream

NorthStar’s RadioGenix generator utilizes an aqueous biphasic extraction chromatography column to separate Tc from low-specific-activity 99 Mo. During this separation process, a fraction of the source Mo material can be found in the waste. This is not of concern when natural Mo targets are used. However, the introduction of isotopically enriched 98 Mo for neutron capture or 100 Mo for accelerator-based production requires recovery of this valuable enriched Mo material. The generator waste is absorbed and solidified in a superabsorbent hydrogel polymer. In this contribution, we report results from our small-scale experimental study on development of a separation process for recovery of molybdenum from the polymeric gel waste using common chemicals and simple chemical techniques. Our investigation confirms that the acidic deswelling of the alkaline hydrogel with concentrated hydrochloric acid converts the occluded molybdate to a neutral molybdenyl dichloride species that is freely released from the gel. Mo can then be concentrated on a commercial tri-n-butyl phosphate resin, eluted with ammonium hydroxide, and crystallized as ammonium heptamolybdate for conversion back to Mo metal. Typical Mo recoveries are 90-95%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fracture Toughness Characterization of Generation II FeCrAl Alloys after ~18 dpa Irradiation

FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding applications due to their excellent corrosion resistance to the elevated temperature steam environment. Currently, the handbook on FeCrAl material properties contains only limited data regarding the fracture toughness properties of any FeCrAl alloy. This includes alloys currently under investigation within the Advanced Fuels Campaign (AFC) at Oak Ridge National Laboratory (ORNL). In this project, a series of irradiation capsules have been irradiated in the High Flux Isotope Reactor (HFIR) at ORNL with two Generation II FeCrAl candidate alloys, i.e., C06M and C36M, to assess the fracture response of these alloys after neutron irradiation. These alloys represent the “book-end” compositions for C26M, the alloy currently being developed as the leading candidate for LWR cladding. A total of six irradiation capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. These damage doses represent the expected middle and end of life damage levels for typical LWR cladding while the irradiation temperature regimes will provide insight into the role of varying microstructural features on the fracture toughness properties of neutron irradiated FeCrAl alloys. To date, irradiation of all capsules has been completed in HFIR. This report summarizes the latest results of microhardness and fracture toughness PIE for the 16 dpa capsules (FCAB2, FCAB4, and FCAB6), for which the measured irradiation conditions were: 204°C/17.6dpa, 343°C/18.3dpa, and 507°C/18.6dpa. The main conclusions of this study can be summarized as follows: 1) After the 204°C/17.6dpa irradiation, both C06M and C36M exhibited significant irradiation hardening and embrittlement 2) After the 343°C/18.3dpa irradiation, both C06M and C36M exhibited small irradiation hardening without irradiation embrittlement 3) After the 507°C/18.6dpa irradiation, both C06M and C36M exhibited irradiation softening without irradiation embrittlement 4) Comparing the microhardness and Master Curve reference temperature T 0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M. This should be mainly due to a flat response of the Master Curve reference temperature T 0q to the irradiations at 166-204°C and 315-343°C ranges 5) C06M showed a lower T 0q , meaning better toughness, than C36M at the unirradiated condition and such trend was kept even after neutron irradiation except for the 166-204°C irradiation where both materials had similar T 0q . 6) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Particle-in-cell techniques for the study of space charge effects in the Advanced Cryogenic Gas Stopper

The Facility for Rare Isotope Beams, FRIB, will significantly increase the rates and varieties of rare isotopes compared to those presently available at existing rare-isotope facilities. The production of stopped beams via the current generation of linear gas stopping devices available is not well matched to the beam intensities that will be available at FRIB. Many experiments that make use of reaccelerated beams, from the ReA facility, greatly benefit from having the highest-intensity, rare-isotope beams possible. Some experiments would also benefit greatly from stopped beams with improved purity. In this paper, the intensity bottlenecks of FRIB’s newest linear gas stopper are studied via experiment and simulation in support of a future detailed conceptual design of a next-generation linear gas stopper that can provide rare-isotope beams with rates of 10^8 pps, a factor of 50 higher than the current state-of-the-art system, with improved beam purity.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

A potential emerging issue concerning repair welding of out-of-core PWR components involving tritium exposure and 3 He retention

This paper identifies a potential but previously unrecognized risk of helium-induced embrittlement and cracking during repair welding of out-of-core pressurized water reactor (PWR) components exposed to tritium-contaminated coolant. While previous weldability concerns centered on 4 He accumulation in neutron-irradiated alloys located within the in-core or near-core regions, new measurements show that 3 He generated by tritium decay can accumulate in out-of-core components. Because hydrogen isotopes readily diffuse to grain boundaries and become trapped there, significant 3 He generation at grain boundaries may lead to cracking during weld repairs. Initial data from far-below-core PWR flux thimble tubes confirm the presence of 3 He levels above known cracking thresholds for repair welds. In conclusion, these findings indicate that out-of-core regions should be considered when defining safe weld repair windows in reactors operating for 60–100 years.

Extended PWR lifetimes↗
DOE OSTI2022

Production Review of Accelerator-Based Medical Isotopes

The production of reactor-based medical isotopes is fragile, which has meant supply shortages from time to time. This paper reviews alternative production methods in the form of cyclotrons, linear accelerators and neutron generators. Finally, the status of the production of medical isotopes in China is described.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2025

Preliminary Insights Into the Feasibility of Determining the Purification Date of Enriched Uranium by Direct Measurement of the 230 Th/ 234 U Ratio Using an All-Faraday Detector Configuration on the Neoma MC-ICP-MS

Rationale: Mass spectrometric measurement of the 230 Th/ 234 U ratio to calculate the purification age of enriched uranium is typically conducted via a combination of ion counters and faraday detectors, thus requiring an inter-detector calibration scheme. Here, our aim is to understand whether the pursuit of a simplified measurement scheme involving only faraday detectors is feasible. Methods: We investigate the possibility of determining U-Th model ages for two enriched uranium standards (NBL U630 and U850) by direct measurement of the 230 Th/ 234 U ratio (without chromatographic separation or isotope dilution) on a ThermoFisher Scientific Neoma MC-ICP-MS utilizing both solution and laser ablation (LA)-based sampling techniques and an all-faraday detector configuration. Results: For the solution mode analyses conducted on aliquots containing sub μg/mL total U, we produce composite average 230 Th/ 234 U model dates of May 19, 1988 (± 351 days), and March 26, 1961 (± 2.5 years) using the directly measured 230 Th/ 234 U ratios for the NBL U630 and U850 uranium standards, which have certified purification dates of June 6, 1988 (± 190 days), and December 31, 1957 (± 36.5 days), respectively. The ages produced by LA-based sampling of dried residues of the same standards deposited onto cotton TexWipes are less accurate and of poorer precision (June 23, 2004 ± 8.7 years for U630 and December 21, 1965 ± 7.9 years for U850) but still yield meaningful information in regards to the purification date. Conclusions: We believe that further refinement of the all faraday detector measurement approach to include development of a more robust Th/U relative sensitivity factor determination, signal cutoff selection, and data processing protocols will allow for this approach to be confidently applied to enriched uranium materials with unknown purification histories. Potential advantages of the method include the reduced sample handling and infrastructure requirements as well as the ability to simultaneously generate a broad picture of the uranium isotopic composition in tandem with the U-Th age determination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Laser-induced fluorescence of filament-produced plasmas

Self-guided ultrafast laser filaments are a promising method for laser beam delivery and plasma generation for standoff and remote detection of elements and isotopes via filament-induced breakdown spectroscopy (FIBS). Yet, there are several challenges associated with the practical application of FIBS, including delivery of sufficient laser energy at the target for generating plasma with a copious amount of emission signals for obtaining a high signal-to-noise ratio. Here, in this study, we use laser-induced fluorescence (LIF) to boost the emission signal and reduce self-reversal in the spectral profiles. Ultrafast laser filaments were used to produce plasmas from an Al 6061 alloy target at various standoff distances from 1 to 10 m. For LIF emission enhancement, a narrow linewidth continuous-wave laser was used in resonance with a 394.40 nm Al I resonant transition, and the emission signal was monitored from the directly coupled transition at 396.15 nm. Emission signal features of Al I are significantly enhanced by resonant excitation. In addition, LIF of filament ablation plumes reduces the self-reversal features seen in the thermally excited spectral profiles. Time-resolved two-dimensional fluorescence spectroscopy was performed for evaluating the optical saturation effects, which are found to be non-negligible due to high Al atomic densities in the filament-produced plasmas.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Interplay between Catalyst Corrosion and Homogeneous Reactive Oxygen Species in Electrochemical Ozone Production

In this study, electrochemical ozone production (EOP), a six-electron water oxidation reaction, offers promising avenues for creating value-added oxidants and disinfectants. However, progress in this field is slowed by a dearth of understanding of fundamental reaction mechanisms. In this work, we combine experimental electrochemistry, spectroscopic detection of reactive oxygen species (ROS), oxygen-anion chemical ionization mass spectrometry, and computational quantum chemistry calculations to determine a plausible reaction mechanism on nickel- and antimony-doped tin oxide (Ni/Sb–SnO 2 , NATO), one of the most selective EOP catalysts. Antimony doping is shown to increase the conductivity of the catalyst, leading to improved electrochemical performance. Spectroscopic analysis and electrochemical experiments combined with quantum chemistry predictions reveal that hydrogen peroxide (H 2 O 2 ) is a critical reaction intermediate. We propose that leached Ni 4+ cations catalyze hydrogen peroxide into solution phase hydroperoxyl radicals ( • OOH); these radicals are subsequently oxidized to ozone. Isotopic product analysis shows that ozone is generated catalytically from water and corrosively from the catalyst oxide lattice without regeneration of lattice oxygens. Further quantum chemistry calculations and thermodynamic analysis suggest that the electrochemical corrosion of tin oxide itself might generate hydrogen peroxide, which is then catalyzed to ozone. The proposed pathways explain both the roles of dopants in NATO and its lack of stability. Our study interrogates the possibility that instability and electrochemical activity are intrinsically linked through the formation of ROS. In doing so, we provide the first mechanism for EOP that is consistent with computational and experimental results and highlight the central challenge of instability as a target for future research efforts.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2025

Amplification of 10 μm picosecond pulses in a compact optically pumped multiatmosphere CO 2 laser

Picosecond pulses, generated by a commercial mid-IR optical parametric source (∼0.5 ps, 1-5 μJ), have been regeneratively amplified to an energy of ≤2 mJ in a palm-size high-pressure CO 2 amplifier optically pumped by a Q-switched 2.8 μm Cr:Er:YSGG laser. The characteristics of the amplified pulses have been studied as a function of seed wavelength centered around 10.3 or 10.6 μm, and the pressure in the gain medium. The measured amplified pulse length is ∼5 ps. Numerical modeling suggests that pulses as short as ∼0.5 ps with multi-GW peak power can be generated in a mixture of CO 2 isotopes. Robustness of the output of a compact CO 2 regenerative amplifier to the seed and pump fluctuations measured at 3 Hz and a perfect Gaussian beam profile indicates the potential of such laser technology for producing high-power ultrashort long-wave infrared pulses at high repetition rates.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Radial Deconsolidation and Leach-Burn-Leach of Eight As-Irradiated AGR-3/4 TRISO Fuel Compacts

Eight as-irradiated AGR-3/4 fuel compacts were subjected to destructive post-irradiation examination via radial-deconsolidation-leach-burn-leach (RDLBL) at INL. The RDLBL process deconsolidated the compacts in multiple, radial steps, followed by a final, single-step axial deconsolidation. The samples generated at each step were analyzed for isotopes of key fission products and actinides. After each deconsolidation step, the compact volume was assessed, and this was used to normalize the measured quantity of nuclides of interest to give a volumetric concentration as a function of radial position within the compact. The total inventories of measured fission products and actinides and the radial concentration profiles were compared among the eight compacts deconsolidated at INL and four other as-irradiated compacts examined at ORNL. The results were analyzed for the effects of irradiation temperature. These results will be used for comparisons with fission product transport models and as input from which fission product diffusivities can be calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Elucidating the Radiation-Induced Redox Chemistry of Plutonium Under Used Nuclear Fuel Reprocessing Conditions

Plutonium plays a critical role in the development of sustainable nuclear fuel cycles, and yet, our fundamental understanding of this element’s inherent radiation-induced redox chemistry and associated impacts on nuclear fuel cycle technologies is limited. Unanticipated changes in oxidation state distribution can influence the speciation and transport of plutonium in a given process. Control of these parameters is especially important for used nuclear fuel reprocessing technologies, wherein the separation and recovery of plutonium is typically achieved by the selective formation, maintenance, and complexation of specific oxidation states. Furthermore, plutonium’s inherent radiation-induced redox chemistry has the capacity to influence the radiolytic behavior of its complexes, the longevity of which are critical in the design of efficient and cost-effective advanced reprocessing technologies. These radiation-induced processes are unavoidable under fuel cycle conditions owing to the inherency of ionizing radiation fields to the decay of plutonium’s isotopes and to the various other radioisotopes generated by nuclear fission and neutron-capture process and the subsequent radioactive decay of their products. As such, mechanistically understanding the response of plutonium’s multiple oxidation states to multi-component ionizing radiation fields is essential for predicting the behavior of this critical element under used nuclear fuel reprocessing conditions. Here, through a combination of time-resolved (electron pulse) and steady-state (alpha and gamma) irradiation experiments complemented by quantitative, multiscale modeling calculations, we present advances in our understanding of radiation-induced plutonium redox chemistry!

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Radial Deconsolidation Leach-Burn-Leach of Five AGR-3/4 TRISO Fuel Compacts after Post-Irradiation Heating Tests

Five AGR-3/4 fuel compacts were previously heated in the Fuel Accident Condition Simulator (FACS) furnace. Three of these compacts were also reirradiated in the Neutron Radiography (NRAD) reactor before the FACS tests. One of these was heated to 1200°C, two were heated to 1400°C, one was heated to 1600°C, and one was held for a period of time at 1600 and 1700°C. After these FACS tests, the compacts were subjected to destructive post-irradiation examination (PIE) via radial-deconsolidation-leach-burn-leach (RDLBL) at INL. The RDLBL process deconsolidated the compacts in multiple radial steps, followed by a final, single-step axial deconsolidation. The samples generated at each step were analyzed for isotopes of key fission products and actinides. After each deconsolidation step, the compact volume was assessed, and this was used to normalize the measured quantities of nuclides of interest to give volumetric concentrations as a function of radial position within the compact. The total inventories of measured fission products and actinides and the radial concentration profiles were compared with their sibling compacts within the same capsule (similar irradiation conditions) that went through RDLBL in the as-irradiated state. These data will be used to refine the fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Technetium-99M generator for enriched molybdenum

A method for separating a parent isotope from a daughter isotope is provided, the method comprising supplying irradiated target; dissolving the irradiated target; treating the dissolved irradiated target to a precipitation step to form a first solid phase of the parent isotope and a first liquid phase of the daughter isotope; filtering the first liquid phase of daughter isotope to create a first purified fraction of the daughter isotope to create a second solid phase of the parent isotope and a second liquid phase of the daughter isotope; and filtering the second liquid phase to create a second liquid fraction of the daughter isotope. The process can be repeated until the 99 Mo is exhausted and the enriched target material 98 Mo or 100 Mo can be easily re-used.

Brown, Michael A.↗
DOE OSTI2022

Rates and Reversibilities in Interconnected Reaction Networks

Mathematical relations prescribing unidirectional forward and reverse rates originally derived based on single-path reaction sequences do not apply to interconnected reaction networks. The presence of branches in reaction networks, leading to alternative stable products, decreases unidirectional rates in reference to those calculated by single-path functional forms, as shown by simulated isotopic exchange rates, but impacts the unidirectional forward and reverse rates equally such that the functional form of effective reversibility remains unchanged. Regardless of stoichiometric numbers and network connectivity, the application of the pseudo-steady-state hypothesis on reactive intermediates in conjunction with consideration of unidirectional rates toward the product of interest and all alternative stable products results in mathematical expressions that accurately reflect simulated isotopic exchange rates. Further analyses based on kinetic resistance, a property akin to electrical resistance, illustrate the manifestation of nodal resistances in addition to the single-path kinetic resistance for interconnected reaction networks. The generalized formalism for assessing rates and reversibilities in interconnected networks derived herein enables us to demonstrate that unidirectional rates cannot be assessed solely from effective reversibilities and net rates of generation of stable species in such networks. Furthermore, isotopic exchange rates, under the condition that each elementary step in the overall reaction sequence forms a unique reactive intermediate that is consumed solely by the subsequent step, can be utilized to determine unidirectional rates and can serve to validate postulated reaction pathways in highly interconnected reaction networks (e.g., CO x hydrogenation).

42 ENGINEERING↗
DOE OSTI2025

Development and Demonstration of a Prototype Molten Salt Sampling System

Molten salt reactors (MSRs) offer potential operability and safety advantages when compared to commercial light water reactors (LWRs). However, operating experience with MSRs is sparse in comparison to what exists for LWRs. Further, the chemical and isotopic composition of the fuel and/or coolant salt is dynamic and difficult to characterize continuously, posing potential safety, operability, and safeguards unknowns that need to be addressed. A molten salt sampling system (MSSS) is regarded as a necessary subsystem within first generation MSRs used to obtain samples of salt for chemical and isotopic analysis in support of the need to monitor and control salt composition during operation. The MSSS is being developed using the Safety-in-Design (SiD) methodology, which incorporates incremental integration of safety analysis into the design process. The MSSS conceptual design emerging from the application of the early stages of the SiD methodology consists of a sample collection system and its housing, a freeze port, and inert gas control and delivery systems. This article describes the prototypes developed to test the functions of these MSSS subsystems, presents the results of testing in both dry and molten salt environments (including reliability data collection performed in accordance with the principles of SiD and the development of a semiquantitative fault tree model), and summarizes the opportunities for future design and testing enhancements based on the results of prototype testing.

molten salt reactor↗
DOE OSTI2021

Cross Section Generation Capability in Griffin

The Griffin code is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory. The code includes a variety of steady-state solvers for fixed-source, k-eigenvalue, adjoint, and subcritical multiplication, as well as transient solvers for point-kinetics, improved quasi-static, and spatial dynamics. The code reads multigroup cross sections in the ISOXML format generated from external deterministic or Monte Carlo cross section generation codes. The implementation of the cross section generation capability in Griffin was initiated last year by plugging in the cross section application programming interface (CSAPI) and reviewing the methodologies for treating particulate fuels. The focus this year was on improving the CSAPI integration and implementing advanced self-shielding methods for applications to advanced reactor problems with TRISO fuels. First, the process for cross section library generation was updated to accurately and rigorously produce isotopic cross section data. Second, the equivalent Dancoff factor cell method performing slowing down calculations on the fly for the resonance treatment was implemented in CSAPI to improve the accuracy of effective multigroup cross sections in the resonance energy range. Third, the iterative local spatial self-shielding method was implemented under the calculation framework of the equivalent Dancoff factor cell method to accurately deal with the double heterogeneity effect of particulate fuel. The updated CSAPI with the advanced self-shielding methods, together with the cross section libraries generated based on the improved process, were tested for pin-cell, unit-cell, and fuel assembly problems with various resonance self-shielding conditions based on very high temperature reactor, high temperature test reactor, and Empire benchmark cores, indicating that the updated CSAPI in Griffin is able to produce multigroup cross sections accurately and efficiently. We also showed that the methodology worked well for pebble bed fuel from HTR-10, but the capability still needs to be fully integrated into CSAPI. In the future, further benchmark tests will be performed for various thermal reactor core problems, including particulate fuel-based pebble bed reactors.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗