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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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24 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Scintillation characteristics of the EJ-299-02H scintillator

Here a study of the dead layer thickness and quenching factor of a plastic scintillator for use in ultracold neutron (UCN) experiments is described. Alpha spectroscopy was used to determine the thickness of a thin surface dead layer to be 630 ± 110 nm. The relative light outputs from the decay of 241 Am and Compton scattering of electrons were used to extract Birks’ law coefficient, yielding a kB value of 0.087 ± 0.003 mm/MeV, consistent with some previous reports for other polystyrene-based scintillators. The results from these measurements are incorporated into the simulation to show that an energy threshold of (∼9 keV) can be achieved for the UCNProBe experiment. This low threshold enables high beta particle detection efficiency and the indirect measurement of UCN. The ability to make the scintillator deuterated, accompanied by its relatively thin dead layer, gives rise to unique applications in a wide range of UCN experiments, where it can be used to trap UCN and detect charged particles in situ.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Low Level Calibration of Liquid Scintillation Counting

Liquid Scintillation Counting (LSC) is an analytical laboratory method that quantifies the concentrations of alpha and beta emitting isotopes. The LSC method can be utilized for the detection of tritium, both environmentally and at nuclear facilities. Most LSC samples are aqueous-based and prepared as a radioactive sample dissolved or suspended in a high efficiency scintillation cocktail that converts radioactive energy into light pulses. When placed in counting vials, the liquid scintillation cocktail allows for optical coupling in which energy is released from the samples radioactive decay, and transferred to the scintillator. That energy is then absorbed to produce excited electrons that fall back to their ground-state and emit a pulse of light that is detected by the liquid scintillation analyzers photomultiplier tubes. Quench is a signal attenuation due to physical, chemical, or color interferences. To compensate for quench, quench standards of a known concentration are implemented at increasing increments of quench per vial to create a quench curve. Quench standards can be purchased pre-made or customized from the manufacturer, however, there are several issues that arise from purchasing the pre-made quench standards. The main issues with ordering pre-made quench standards is that they use higher count values, between 30,000 and 250,000 disintegrations per minute (dpm) of tritium per vial and only come in glass vials. Using higher counting standards in glass vials makes it hard to see lower concentrations in plastic vials. Additionally, when using higher energy standards the accuracy of the calibration readings tend to be lower. On the other hand, making a set of quench standards in-house allows for the customization of the dpm concentration to roughly 1,500-3,000 and preparation in plastic vials. This allows for lower activity readings that results in closer precision of the samples' true concentration. Technical Objectives: Create a quench curve and calibrate the LSA for tritium. Verify the calibration of the quench curve prior to cross sample runs. Check calibration against Quality Control (QC) reference value. Conclusion: a tritium quench curve calibration was performed using a scintillation cocktail with ultima gold (AB and AB/F), acetone (only with AB), and water (only with AB/F). By performing the tritium quench curve calibration, the counting efficiency was correlated to various levels of quench. Calibration checks with quality control standards were also performed to verify that the calibration was correct. The observed results provided acceptable values that matched the quality control standard reference values with an acceptable uncertainty. The next step will be fine tuning and cross comparison checks with legacy instrumentation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Thermal Unequilibrium of PdSn Intermetallic Nanocatalysts: From In Situ Tailored Synthesis to Unexpected Hydrogenation Selectivity

Effective control on chemoselectivity in the catalytic hydrogenation of C=O over C=C bonds is uncommon with Pd-based catalysts because of the favored adsorption of C=C bonds on Pd surface. Here we report a unique orthorhombic PdSn intermetallic phase with unprecedented chemoselectivity toward C=O hydrogenation. We observed the formation and metastability of this PdSn phase in situ. During a natural cooling process, the PdSn nanoparticles readily revert to the favored Pd 3 Sn 2 phase. Instead, using a thermal quenching method, we prepared a pure-phase PdSn nanocatalyst. Furthermore, PdSn shows an >96% selectivity toward hydrogenating C=O bonds of various α,β-unsaturated aldehydes, highest in reported Pd-based catalysts. Further study suggests that efficient quenching prevents the reversion from PdSn- to Pd 3 Sn 2 -structured surface, the key to the desired catalytic performance. Density functional theory calculations and analysis of reaction kinetics provide an explanation for the observed high selectivity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Measurement of proton light yield of water-based liquid scintillator

Abstract The proton light yield of liquid scintillators is an important property in the context of their use in large-scale neutrino experiments, with direct implications for neutrino-proton scattering measurements and the discrimination of fast neutrons from inverse $$\beta $$ β -decay coincidence signals. This work presents the first measurement of the proton light yield of a water-based liquid scintillator (WbLS) formulated from 5% linear alkyl benzene (LAB), at energies below 20 MeV, as well as a measurement of the proton light yield of a pure LAB + 2 g/L 2,5-diphenyloxazole (PPO) mixture (LABPPO). The measurements were performed using a double time-of-flight method and a pulsed neutron beam from the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory. The proton light yields were measured relative to that of a 477 keV electron. The relative proton light yield of WbLS was approximately 3.8% lower than that of LABPPO, itself exhibiting a relative proton light yield 15–20% higher than previous measurements of an analogous anoxic sample. The observed quenching is not compatible with the Birks model for either material, but is well described with the addition of Chou’s bimolecular quenching term.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
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 OSTI2021

Tritiated Sulfur Hexafluoride Disposition Strategies

Tritium (T 2 ) is a radioactive isotope of hydrogen that is produced in nuclear fission reactions and is often used in nuclear fusion reactions and accelerator-based applications for medical isotope production. As a hydrogen isotope, tritium can readily bind to hydroxyl radicals (OH), forming tritiated water (HTO or T 2 0), and to carbon atoms. Tritium decays to helium-3 ( 3 He) via beta-decay with max decay energy of 18.6 keV. While it is not an external radiation hazard, it can be an internal radiation hazard if tritium is inhaled, ingested, or absorbed through the skin. In applications where tritium is handled, tritium confinement is performed using different barriers to minimize releases to the environment. For gaseous (elemental), liquid (oxide), and metal (hydride) tritium, process piping and components provide the primary confinement function. Secondary tritium confinement is typically provided by inert (i.e. non-flammable gases such as nitrogen, argon, or helium) gloveboxes which are connected to a tritium stripper system. Primary tritium confinement barriers typically produce low volumes of high tritium concentrations of tritiated methane from carbon in steels or organic materials along with tritiated water/tritium oxide (e.g. HTO) and tritiated ammonia from reactions with oxygen and nitrogen. Tritium escaping primary confinement into secondary confinement atmospheres (e.g. gloveboxes) produce higher volumes of lower activity contamination than found in process piping. Tritium contamination also occurs by leaks or tritium permeation/diffusion through confinement materials. Tritium from inside primary confinement barriers will diffuse or leak out of the primary confinement barrier and usually into the air, if the system is inside an air hood or ventilated hot cell, or into the secondary confinement (e.g. glovebox) atmosphere which is either exhausted or stripped based on the function of the secondary confinement (glovebox) system. Accelerator based processes for medical isotope production represent an atypical tritium contamination challenge. In medical isotope production, deuterium supply gas is ionized and accelerated to a tritium gas target to produce neutrons that are then used to produce the medical isotopes through additional nuclear fission reactions. To create large voltage differentials for accelerator operations, an electrical insulation medium is needed to prevent or rapidly quench electric discharges. A common electrical insulation medium utilized in accelerator applications is sulfur hexafluoride (SF 6 ) gas. SF 6 has a high dielectric strength and allows for the construction of smaller accelerator systems compared to other electrical insulation mediums such as air or dry nitrogen. Due to tritium permeation/diffusion through accelerator process and confinement materials, there is the possibility that tritium can contaminate the electrical insulation medium of the accelerator. Tritium contaminated SF 6 creates a material without any obvious processes for managing the contamination, reuse, or disposal of the used SF 6 . This document will discuss possible management strategies for tritium contaminated SF 6 for accelerator-based processes for Molybdenum-99 (Mo-99) production.

07 ISOTOPE AND RADIATION SOURCES↗