Engineering topics

Sevigny, Gary J.

Publications and source records attributed to Sevigny, Gary J..

8 records

DOE OSTI2023

Betavoltaics with absorber layer containing coated scintillating particles

A beta-voltaic device made up of silica covered scintillating particles incorporated within an isotope absorbing layer to produce an improved power source. Lost beta particles are converted to UV light which is also converted to power in a beta-voltaic converter. The addition of the scintillating particles effectively increases the power efficiency of a BV device while maintaining the slim profile and smaller size of the power source. This arrangement makes possible implementation in space, defense, intelligence, medical implants, marine biology and other applications.

Hubbard, Lance R.↗

DOE OSTI2023

Cerium oxide impact on fresh and hardened properties of cementitious materials

Cements used in global nuclear waste treatment, storage, and disposal require thorough testing of fresh and cured properties to demonstrate regulatory compliance. Using actual radioactive wastes is often cost-prohibitive during early scoping tests necessitating the use of non-radiological surrogates. Cerium (Ce) is often used as a surrogate for actinides uranium (U) and plutonium (Pu) in spent nuclear fuel, yet few studies have explored how Ce impacts the properties used to qualify a cement for use. Here, this work compares the fresh and cured properties of three classes of cements – ordinary portland cements, ultra-high performance cement composites, and magnesium phosphate cements – each with and without ceria (CeO 2 ). Due to CeO 2 sorption of water (54 vol%) that effectively decreases the water available to hydrate and form binding cement phases, variations in key cement properties were detected that suggest surrogate CeO 2 –cementitious matrices may not be representative of matrix behaviors when using U and Pu sources.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Accelerated beta radiation aging of interlayer titanium nitride in gallium nitride contacts

Currently, there is no good way to determine the influence of the radiation on the aging of betavoltaic electrical contacts. Here, this work tested a method to accelerate the aging of the contacts inside a betavoltaic by enhancing energy deposition within the interfacial region of interest. An electrical contact of gold-titanium on gallium nitride was aged by exposure to tritium betas and electrons in a beamline. The interface stoichiometry was compared to the electrical performance of the contact. The method to age the betavoltaic component could help predict lifetime performance of the internal electrical contacts and provide assurance of deployment reliability.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Tritium Betavoltaic Powered Sensor Platforms: Power Augmentation with Scintillating Particles

Betavoltaics (BV) are long-life power sources that typically convert beta particle radiation into electricity. Largely, the radioactive decays within the source go unharvested by the device. This work seeks to augment the power generation of BV devices by integration of scintillating particles within the radiative getter to convert beta emission which would otherwise not leave the getter into usable light for power generation. Silica-covered barium fluoride scintillating particles were integrated into a tritiated water getter. Power generation was increased from 100s of nW to µW levels with the addition of 0.2 wt. % particles into the getter. Nanowatt-scale sensor platforms were demonstrated with the µW BV devices and the maximum possible lifetime of such platforms was estimated. As this technique enables higher power density BV devices from conventional Si semiconductors (compared to wider bandgap BVs), the further implementation may lower the barrier-to-deployment of these long-life power/sensor platforms.

42 ENGINEERING↗

DOE OSTI2021

Evidence of lithium mobility under neutron irradiation

Understanding the evolution of intermetallic materials in high-radiation environments is of great importance for fusion science and national security. Tritium ( 3 H) and lithium ( 6 Li, 7 Li) transport within neutron irradiated claddings coated with aluminide (FeAl3) has been investigated using state-of-the-art multimodal imaging. Specifically, scanning electron microscopy – focused ion beam (SEM-FIB) was used to prepare irradiated coating lift-out samples for follow-on microanalysis. Here, scanning transmission electron microscopy (STEM) was used to acquire atomic-scale information on the carbonaceous structure and elemental mapping. Atomic force microscopy (AFM) was used to determine lift-out dimensions nondestructively. Time-of-flight secondary ion mass spectrometry (ToF-SIMS) spectral and depth profiling show unexpected lithium isotopic distributions in the irradiated cladding, raising an important evidence of possible light isotope mobility within the cladding under elevated temperature and irradiation. ToF-SIMS three-dimensional chemical mappings of mid- and bottom-cladding coatings show light isotopic (e.g., 3 H, 6 Li, 7 Li) distributions in the irradiated coating and give new insights into the fundamental mechanism related to transport mechanisms within the cladding. Multimodal imaging is power to link microstructure, chemistry, and nanoscale defects that impact reliability of these materials. Furthermore, chemical mapping offers observations of the microstructural evolution due to irradiation and provides insights into unexpected material transport under extreme conditions.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Neutron Irradiation Induced Changes in Isotopic Abundance of 6Li and 3D Nanoscale Distribution of Tritium in LiAlO2 Pellets Analyzed by Atom Probe Tomography

Tritium, a radioactive isotope of hydrogen is produced by neutron irradiation of 6Li enriched LiAlO2 pellets in Tritium producing burnable absorber rods (TPBARs). Three-dimensional nanoscale mapping of Tritium in irradiated pellets is critical to understand its spatial variation and its association with various microstructural features in the pellets. Spatially resolved analysis of 6Li isotopic enrichment in such ceramic pellets before and after irradiation can provide insights to heterogeneities in Li isotopic distribution. However, Li being a light isotope, its spatially resolved compositional or isotopic analysis is a challenging task to most analytical methods. Here we used atom probe tomography, to evaluate the Li isotopic enrichment in both the LiAlO¬2 matrix and the Li deficient secondary phase LiAl5O8 before irradiation. These results were then compared with 6Li isotopic ratio measured by APT analysis of neutron irradiated pellets to clearly demonstrate the unique capability of APT to quantitatively analyze the isotopic enrichment of 6Li as well as for other light elements present in these materials. Evidence for heterogenous nanoscale distribution of 3H and O3H within irradiated pellet microstructure is also provided, which is attributed to 3H trapping in irradiation induced vacancy clusters or voids within the LiAlO2 matrix. The characterization results given here, now prove the feasibility for using APT for analyzing nanoscale spatially resolved isotopic enrichment of Li as well as other light elements such as Tritium in TPBARs. This work also paves way to adopt APT for analyzing light elements such as Li and tritium in materials used for nuclear fusion reactor applications, geochemistry, astrophysics as well as for energy storage applications.

Devaraj, Arun↗

DOE OSTI2020

Vitrification of High-Cr Glass in Research-Scale Melter

This test report describes the experimental results from a small-scale test using the research-scale melter (RSM) at Pacific Northwest National Laboratory to demonstrate processing of a high-Cr simulated feed stream, designated HLW-HCr-16. The RSM is a small, joule-heated melter capable of processing melter feed continuously. The melter is equipped with Inconel ® 693 electrodes, Monofrax ® K-3 refractory, and an Inconel 690 pour spout. An electric kiln surrounds the melter body and minimizes heat loss from the melter body during operation. The RSM is equipped with an offgas treatment system that employs quenching, wet scrubbing, and high-efficiency mist elimination. The glass-discharge section is heated to facilitate pouring of the glass. The melter is fitted with a melt cavity that is ~25 cm (10 in.) in diameter with a nominal glass depth of 8.9 cm (3.5 in.). The melter was operated with a target glass temperature of 1150°C and target plenum temperature between 550°C and 700°C for this test. The air bubbling rate was 4.2 L/min. Overall, during the continuous operation of the melter for ~ 103 hours, ~ 141 kg of glass was produced. At the conclusion of the test, the melter and exhaust lines were visually inspected for particulate deposition and corrosion. Entrained material had adhered to the underside of the melter lid and to the exhaust piping. Enrichments in elements such as Cl, F, B, K, Li, P, Na, and S were measured in these deposits through inductively coupled plasma–optical emission spectroscopy and X-ray fluorescence analysis. When the melter electrodes and air bubbler tube were removed from the glass in the RSM, the electrodes appeared discolored, but no significant loss of metal was observed. The processing of a high-Cr simulant, HLW-HCr-16, in the RSM produced glass at an average rate of 1.36 kg/h, equaling a melter-surface-area normalized glass generation rate of 654 kg/day/m 2 . The resulting glass met the toxicity characteristic leaching procedure requirement. Test results of crystallinity, electrical conductivity, and viscosity showed good processing properties of this high-Cr high-level waste glass. RSM offgas was also sampled and analyzed at periodic intervals during steady-state operating conditions. The total decontamination factor averaged by four sampling periods was 134. The concentrations of CO and NO x in emissions were 237 to 422 and 69 to 94 parts per million by volume, respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Probing the Radial Chemistry of Getter Components in Light Water Reactors via Controlled Electrochemical Dissolution

Getters are among the key functional components in the Tritium Producing Burnable Absorber Rods (TPBARs) of Light Water Reactors, and are used to capture the released tritium gas. They are cylindrical tubes comprised of the alloy zircaloy-4 which is coated with nickel plating both internally and externally. Their exposure to tritium as well as the irradiation environments in the light water reactors changes their structure, chemical composition and chemistry. Understanding the radial tritium distribution is key to gaining insight into the evolution of new chemistry upon irradiation in order to predict getter performance. Consequently, efforts are underway to devise a method that can allow selective probing into the getter core radially to probe the chemistry akin to carefully and selectively peeling of the layers of an onion. We have recently devised an electrochemical dissolution technique for controlled radial dissolution of zirconium based liners; this work describes the first proof-of-concept demonstration of the applicability of this technique on the controlled radial dissolution of nickel plated getters. The technique is complemented with periodic analysis of the electrolyte solution to measure the rates of release of the various elemental components and therefore glean information on their radial distribution across the getter; further complementary microscopy on the exposed surfaces allows us insight into the chemistry at the getter core.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗