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Johnston, Brandon

Publications and source records attributed to Johnston, Brandon.

9 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

High Temperature Tensile Testing of Molybdenum Before and After Exposure in Flowing He at 600°C to 1000°C

ORNL supported NorthStar Medical Radioisotope’s accelerator-based concept to produce the medical isotope molybdenum-99 (Mo-99). The Mo targets are expected to experience in the accelerator temperature ranging from 200°C to 1000°C, and previous studies have demonstrated the embrittlement of Mo targets at room temperature when exposed to impure He at 800 to 1000°C. The goal of this project was to evaluate potential embrittlement at temperature ranging from 600°C to 1000°C for three Mo materials exposed at the same temperatures for 100h in high purity He (O 2 < 5ppm (µl/l)) with a flow rate of ~1m/s. Enriched Mo-100 (aMo) and conventional pressed and sintered Mo (UHP) disks were provided by NorthStar while low carbon arc cast rolled Mo, fabricated at ORNL, was used for comparison. The as fabricated rolled Mo exhibited superior tensile strength due to the alloy very fine grain size with limited ductility except at 1000°C with the elongation at rupture reaching 30%. The UHP Mo disks showed the lowest strength but highest ductility at 20-1000°C with significant strain hardening during tensile testing. The ductility of the aMo material was quite low at all temperatures due to a high density of voids at grain boundaries. Significant embrittlement of the aMo material was observed after exposure at 600°C to 1000°C, most likely because of oxygen diffusion and segregation at grain boundaries. No embrittlement was observed for the UHP Mo material, with excellent elongation at rupture for some of the specimens oxidized in He and tested at 600-1000°C. Significant variation in ductility from one UHP Mo specimen to another was, however, measured, both in the as fabricated and He-oxidized conditions. The rolled Mo was not affected by oxidation in He at 800°C, but a decrease in strength and increase in ductility was observed at 1000°C, likely due to the material recrystallization.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Effects of Cr/Zircaloy-4 coating qualities for enhanced accident tolerant fuel cladding

Cr-coated zirconium alloys represent a modern approach to enhance cladding safety during accident scenarios. Two high-power impulse magnetron sputtered Cr-coated Zry-4 systems were subjected to simulated loss-of-coolant accident conditions to investigate cladding performance. The first Cr-coating (4.8 µm thick) was deposited onto Zry-4 cladding and exhibited through-thickness cracking while the second Cr-coating (6.8 µm thick) was deposited with improved deposition parameters onto polished Zry-4 and exhibited no cracking. During burst testing, the coating with a higher density of defects failed to consistently reduce oxidation and exhibited similar burst behavior as Zry-4. In contrast, the second Cr-coating reduced ZrO 2 formation through formation of Cr 2 O 3 and displayed enhanced burst temperatures by ~80 °C compared to Zry-4. Utilizing an empirical relation for burst behavior of zirconium alloys, the 6.8 µm Cr/Zry-4 system displayed enhanced burst temperatures equivalent to an effective 0.464 mm increase in Zry-4 wall thickness, highlighting the value of continuous Cr coatings for accident scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Impact of annealing at 1000°C in flowing He with low O 2 concentration on the room temperature tensile properties of Mo alloys

NorthStar is developing an accelerator-based technology to produce the medical isotope Mo-99. A previous ORNL report (ORNL/SPR-2022/2490) studied the oxidation behavior of Mo in flowing He with low oxygen content at 800°C and 1000°C and it was shown that an O 2 concentration lower than ~5ppm (μ1/1) is required to achieve acceptable oxidation rates for wrought or powder metallurgy (PM) Mo. Oxidation may also impact the mechanical properties of Mo and this report summarizes tensile results at room temperature on wrought and PM Mo before and after exposure at 1000°C in flowing He with O 2 concentration varying from 1 to 15 ppm O 2 . It was found that oxidation at 1000°C results in a significant decrease of Mo strength and ductility with brittle intergranular fracture surfaces, likely due to the segregation of oxygen at grain boundaries. For the PM Mo material, the decrease in strength and ductility was also observed after annealing in vacuum at 1000°C, indicating that the diffusion of the oxygen trapped in the material pores to grain boundaries after 20-100h at 1000°C is sufficient to embrittle the material. Finally, initial results on actual aMo and high-density disks provided by NorthStar revealed the lack of ductility of the materials due to the specific disk microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Report Summarizing Progress in Digital Image Correlation Analysis of Burst Phenomenon

This report is an update on the progress made towards utilizing in-situ methods to observe and quantify Zry-4 deformation behavior during burst testing. A heavily modified burst test was developed that allows both digital image correlation and infrared thermography to be utilized during testing to quantify strain and axial thermal gradient data during burst. A variety of tests ranging from 6.2-15.9 MPa internal pressure and utilizing both constant and transient pressure conditions were conducted and presently reported on. The test modifications, which are detailed extensively in a previous report, are effectively found to have no discernible impact on burst temperatures when compared to historical Zircaloy burst data. To support future modeling studies, the IR thermography data was used to approximate the thermal gradient down the axial length at burst for all the claddings. Then, DIC is first correlated to thermocouple temperature for a 10.3 MPa Transient pressure test, and then DIC correlated to IR thermal gradient data down the axial length of a 8.2 MPa Transient pressure test at several points.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Development and Implementation of Digital Image Correlation for Out-of-Cell Burst Testing

This report details the progress made towards utilizing in-situ methods to observe and quantify unirradiated Zircaloy-4 (Zr4) fuel cladding deformation during out-of-cell design basis accident conditions, specifically during loss-of-coolant accident (LOCA) burst testing in the Severe Accident Test Station at Oak Ridge National Laboratory. Digital image correlation (DIC) was implemented to calculate cladding strain during burst and infrared (IR) thermography was used to attempt a correlation of strain with temperature and to investigate thermal gradients. For both techniques, early experimentation revealed experimental modifications were necessary for implementation. For DIC, this manifested as shortened cladding lengths which could be centered in front of the viewing chamber bored through the IR furnace side, and for IR thermography this meant conducting burst tests in air without a reaction tube and the incorporation of a SiC shell to eliminate direct IR reflections from the tungsten lamps. In the following, the technical issues and subsequent origin of these modifications are discussed, laying the framework for a proof-of-concept DIC/IR characterization of a burst test in the final section.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

FY21 burst activities with coated Zircaloy-4 under accident conditions

This report summarizes the results of Advanced Fuels Campaign (AFC) accident-tolerant fuel (ATF) burst activities. Nuclear service grade Zry-4 was procured and coated with a 7-micron thick Cr coating. The coating quality was investigated, and there were several defects at the Cr/Zry-4 interface due to the surface roughness of the as received Zry-4 tubing. To provide insight into the effect of coating defects on the cladding performance under accident scenarios, the unirradiated uncoated and coated material was tested under loss of coolant accident (LOCA) and pellet cladding mechanical interaction (PCMI) reactivity insertion accident (RIA) conditions. The defected coating appeared to have no impact on the cladding performance under these scenarios when compared to the as-received cladding material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High Temperature Steam Oxidation of Irradiated FeCrAl in the Severe Accident Test Station

FeCrAl-UO2 test capsules were fabricated at Oak Ridge National Laboratory (ORNL) and irradiated at the Advanced Test Reactor. Following irradiation, samples were sectioned from the irradiated rod and oxidation kinetics were evaluated to access the candidate cladding high-temperature oxidation performance following irradiation. The irradiation was performed at approximately 400°C to a burnup of 10 GWd/MT. The high-temperature oxidation tests were conducted in the ORNL Severe Accident Test Station at 1200 and 1300°C. Weight measurements were taken before and after oxidation testing. Cross-sections of the cladding were metallographically mounted and optical microscopy was performed. Measurements of the oxidation layer before and after high-temperature testing were collected. The results indicate the irradiated FeCrAl C35M alloy provided good thermal stability up to 1200°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Mechanisms of Water-Stimulated Mg 2+ Intercalation in Vanadium Oxide: Toward the Development of Hydrated Vanadium Oxide Cathodes for Mg Batteries

As lithium-ion batteries approach their theoretical limits for energy density, magnesium-ion batteries are emerging as a promising next-generation energy storage technology. However, progress in magnesium-ion battery research has been stymied by a lack of available high capacity cathode materials that can reversibly insert magnesium ions. Vanadium Oxide (V 2 O 5 ) has emerged as one of the more promising candidate cathode materials, owing to its high theoretical capacity, facile synthesis methods, and relatively high operating voltage. This review focuses on the outlook of hydrated V 2 O 5 structures as a high capacity cathode material for magnesium-ion batteries. In general, V 2 O 5 structures exhibit poor experimental capacity for magnesium-ion insertion due to sluggish magnesium-ion insertion kinetics and poor electronic conductivity. However, several decades ago, it was discovered that the addition of water to organic electrolytes significantly improves magnesium-ion insertion into V 2 O 5 . This review clarifies the various mechanisms that have been used to explain this observation, from charge shielding to proton insertion, and offers an alternative explanation that examines the possible role of structural hydroxyl groups on the V 2 O 5 surface. While the mechanism still needs to be further studied, this discovery fueled new research into V 2 O 5 electrodes that incorporate water directly as a structural element. The most promising of these hydrated V 2 O 5 materials, many of which incorporate conductive additives, nanostructured architectures, and thin film morphologies, are discussed. Ultimately, however, these hydrated V 2 O 5 structures still face a significant barrier to potential applications in magnesium-ion batteries. During full cell electrochemical cycling, these hydrated structures tend to leach water into the electrolyte and passivate the surface of the magnesium anode, leading to poor cycle life and low capacity retention. Recently, some promising strides have been made to remedy this problem, including the use of artificial solid electrolyte interphase layers as an anode protection scheme, but a call to action for more anode protection strategies that are compatible with trace water and magnesium metal is required.

25 ENERGY STORAGE↗