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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
NASA NTRS1966

Simulated radioactivity.

Simulated radioactive decay experiment using random number generator or set of random numbers

SIMULATION↗
DOE OSTI2023

Proton nuclear magnetic resonance (1H NMR) of flammable organic chemicals in radioactive high–level supernatant waste at the Savannah River Site (SRS)

The Savannah River Site stores approximately 36 million gallons of radioactive and hazardous waste that contains approximately 245 million curies. The waste is sent through various chemical processes to reduce its volume and to separate various components. The facility plans to replace formic acid (a chemical used to reduce soluble mercury) with glycolic acid. Recycle solution with glycolate may flow back to the tank farm, where the glycolate can generate hydrogen gas by thermal and radiolytic mechanisms. The current analytical method for detecting glycolate (ion chromatography) in supernatant requires a large dilution to reduce interference from the nitrate anions. Hydrogen nuclear magnetic resonance is an analytical method that requires less sample dilution. It takes advantage of the CH 2 group in glycolate. Liquid samples were spiked with four different levels of glycolate to build a calibration line, as it is recommended in the standard addition method. The detection and quantitation limits determined were 1 and 5 ppm, respectively, for 32 scans, which is well below the process limit of 10 ppm. In one test, 800 scans of a supernatant spiked with 1 ppm glycolate resulted in a -CH 2 peak with a signal-to-noise ratio of 36.

1H↗
DOE OSTI2024

Investigation of candidates for reactor produced radioactive materials in support of radiological training exercises

Bromine-82, Potassium-42 and Copper-64 have been successfully adopted as radioactive surrogates for outdoor large area contamination training. The goal of this project was to discover new materials that could supplement potassium bromide (KBr) and copper pellets in radiological dispersal device (RDD) training events to reduce the down time of the training fields and to broaden the toolbox of the RDD surrogate training event program at Idaho National Laboratory. Of the ten different materials investigated, sodium nitrite, gallium metal, and gallium oxide presented the greatest promise as potential materials to replace potassium bromide in RDD training events.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Evaluation of mitigation strategies for radioactive fission gases in fluoride-salt-cooled high-temperature reactors

Here, the Fluoride-salt-cooled High-temperature Reactor (FHR) is a promising Generation IV nuclear reactor due to its improved safety design, near atmospheric working pressure, and high electric power conversion efficiency. However, control of radioactive fission gases, such as tritium, is a critical issue in FHRs due to the significantly larger production rate and potentially larger leakage rate of tritium compared to those in Light Water Reactors (LWRs). Therefore, this study proposes two mitigation strategies for tritium in FHRs: Double-Wall Heat eXchanger with a Tritium Carrier (DWHX-TC) option and Single-Wall Heat eXchanger with a Tritium Barrier (SWHX-TB) option. For the DWHX-TC option, two tube surface configurations, i.e., plain and fluted tubes, and four tritium carriers, i.e., helium, FLiBe, FLiNaK, and KF-ZrF 4 , are initially considered, among which the best tube configuration and promising tritium carrier are then selected and used for design optimization of the DWHX. For the SWHX-TB option, two cases, tritium barriers with and without cracks, are investigated in this study. In addition, the two mitigation strategies are evaluated by an in-house, one-dimensional, coupled HEat and MAss Transfer (HEMAT) code developed to predict the leakage rate of tritium in nuclear systems. Our analysis shows that (1) a double-wall configuration using fluted tubes as both the inner and outer tubes is better than other double-wall configurations investigated; (2) helium, as a tritium carrier, is superior to other candidates, including FLiBe, FLiNaK, and KF-ZrF 4 ; (3) the leakage rate of tritium (into the atmosphere) is reduced from 3400 Ci/day, taking the Advanced High-Temperature Reactor (AHTR) as a reference, to 4.0 Ci/day for an optimum design of the DWHX-TC option using helium as the tritium carrier and 1.7 Ci/day for an optimum design of the SWHX-TB option using 30-µm thick silicon carbide (SiC) as the tritium barrier (no cracks); and (4) potential cracks in tritium barriers may significantly deteriorate mass transfer performance of the SWHX-TB option. For example, a cracking rate of 0.1% leads to an increase of 694.1% in the leakage rate under the conditions investigated.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

On the long-term migration of uranyl in bentonite barrier for high-level radioactive waste repositories: The effect of different host rocks

Repositories for high level radioactive waste hosted in argillite or granite with bentonite backfill are the two disposal concepts that been extensively studied. A general perception is that granite has very low sorption capacity and therefore is disadvantageous in retarding the migration of radionuclides in comparison with argillite. However, when there is bentonite backfill, does the sorption capacity of granite really matter for the migration of radionuclides? How different host rocks would affect the migration of U(VI) through the bentonite backfill of an engineered barrier system (EBS), and which properties of the host rocks cause those differences in migration, given that migration of radionuclides is a critical measure in assessing the performance of a repository? Here we present two coupled thermal, hydrological, and chemical models for the transport of U(VI), each with an identical setup except for the type of host rock (argillite vs granite rock). Comparisons between our models show that the different host rocks exert their influence on migration of U(VI) via regulating the chemical conditions in the bentonite, which affect the concentration of U(VI) at the source and the adsorption of U(VI) in the bentonite. The key chemical conditions are pH, Ca + , and HCO 3 - concentration in pore-water, which are strongly affected by soluble carbonate minerals. Our models also show the occurrence of illitization (dissolution of smectite and precipitation of illite) in the bentonite, which affects the migration of U(VI) through changing pH and the quantity of adsorbents. Although generalization of current model results for granite and argillite should be done carefully, the simulations highlight the importance of pore-water chemistry in host rock, as well as bentonite-host rock interactions, when assessing the performance of a repository. In comparison to properties such as thermal conductivity, permeability, and adsorption capacity, properties that have been studied intensively, the pore-water chemistry of host rocks deserve equal or even more attention because of its strong influence on radionuclide migration.

58 GEOSCIENCES↗
DOE OSTI2021

Neutron diffraction illustrates residual stress behavior of welded alloys used as radioactive confinement boundary

Corrosion-resistant welded alloys are frequently used as a leak-tight boundary in critical applications that require confinement of hazardous and/or radioactive substances, including an increasing population of spent nuclear fuel (SNF) canisters. The behavior of residual stresses generated as a result of irregular elastic–plastic deformation during processes such as welding is one of today's key issues to a full understanding of the aging mechanisms that may compromise the confinement boundary. Whether such processes and any subsequent weld repairs, not subjected to post-weld heat treatment, would negatively affect the initial material by introducing through-thickness tensile stresses remains an open question. Here we report the first residual stress measurements using neutron diffraction on the welded joints of a SNF canister. We found significant tensile residual stresses in the as welded sample, indicating that initiation and through-thickness growth of cracks may be possible. Following repair, we observed a stress redistribution and introduction of beneficial compressive stresses. We anticipate our results will improve understanding of confinement susceptibility to aging and guide improvements in repair techniques.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Production and characterization of thin, self-supporting Si foils for use as targets in radioactive beam experiments

Development of thin, self-supporting silicon foils (both natural and isotopically-enriched) for use as targets in reaction studies with radioactive beams is detailed. Foils with a thickness of ~220 μg/cm 2 were produced using vapor deposition and were floated onto aluminum frames with 10–15 mm diameter holes. During their production, the foil thickness was measured using a quartz crystal monitor. Subsequently, the foil thickness was characterized by particle energy loss measurements and Rutherford backscattering (RBS). These measurements demonstrated that the thickness could be determined to within a 0.5% uncertainty. The elemental purity of the foils was assessed using RBS and X-ray photoelectron spectroscopy. Lastly, this analysis demonstrated that the foils have 87%–90% silicon abundance.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗