Engineering topics

Mouche, Peter A.

Publications and source records attributed to Mouche, Peter A..

8 records

All sources NASA NTRS DOE OSTI

DOE OSTI2023

High-Energy X-ray Diffraction Microscopy for Nuclear Forensics (FY23 Project Report)

36 MATERIALS SCIENCE↗

DOE OSTI2022

High-Energy X-ray Diffraction Microscopy for Nuclear Forensics FY2022 Project Report

Morphological information on nuclear material has been identified using visible light and scanning electron microscopy. These identify qualitative differences in particle morphology. Three-dimensional imaging of materials through alternating scanning electron microscopy imaging and focused ion beam milling has also been used. Unfortunately, these techniques are time- and labor-intensive, with significant sample preparation required and lengthy analysis times. Further, the resulting 3D images are qualitative, require manual identification, and do not capture statistically-representative populations. High energy X-ray 3D imaging using a direct-beam or diffracted-beam (High-Energy Diffraction Microscopy) have been developed at the Advanced Photon Source and can produce quantitative information on grains (phase, location, etc.) and pores (size distribution, sphericity) in a material. These techniques require only minutes to characterize a sample volume and are non-destructive, thus suitable for a wide range of existing samples and for confirmatory analyses to be carried out using conventional microscopy techniques. In this first year of the project, all uranium oxide samples were synthesized and characterized using conventional analyses by the analytical chemistry laboratory. Conventional analysis methods included powder x-ray diffraction, scanning electron microscopy, impurity analysis via inductively coupled plasma mass spectrometry, and infrared spectroscopy. Impurity analysis shows a drop in boron content from UO 3 to the lowest U 3 O 8 calcination temperature, but otherwise no appreciable difference in any sample. Analysis of diffraction data shows a flip of peaks from UO 3 dominated for the 600 °C calcined sample to U 3 O 8 dominated at 700 °C and 800 °C. Analysis of scanning electron microscopy images shows that with increased calcination temperature the size distribution of particles seems to increase and broaden. Both of these last findings are in line with previously published data, though this work used significantly fewer particles to simply show similar trends instead of getting truly quantitative particle analysis. Infrared analysis similarly shows ingrowth of U 3 O 8 as calcination temperature is increased, along with depression of peaks associated with UO 3 and water. Samples were prepared for analysis at the Advanced Photon Source at beamline 1-ID. It is anticipated that analysis will occur in November of 2022. AI/ML techniques to de-noise data coming out of 1-ID during the analyses was also developed during this time using previously gathered data. Preliminary results using a self-supervision technique called Noise2Selfshow good de-noising of data. Once the uranium oxide samples are analyzed, real data will be used to test the de-noising and other AI/ML techniques that may be developed in the second year of the project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Effects of sample bias on adhesion of magnetron sputtered Cr coatings on SiC

Swelling of SiC at 300 C due to in-service neutron irradiation causes tensile residual stresses in coatings which are expected to adversely affect the performance of coated SiC composite fuel cladding for light water reactors. Matching the coating swelling with the substrate, a solution common for thermal expansion, is not practical in the case of neutron irradiation. Biasing samples during magnetron sputtering deposition induces compressive residual stress which may counteract this. In this study, chromium coatings were deposited on SiC by DC magnetron sputtering with no external heating at bias voltages of –50V, –75V, and –100V. The effects of the bias voltage on morphology, residual stress, microstrain, texture, and adhesion are shown. The low deposition temperature resulted in the coating microstructure evolution following an energetic particle bombardment dominated trend. Additionally, at the two lower bias voltages knock-on implantation dominated increasing the residual stress and microstrain while at the highest bias voltage, thermal spike migration allowed for defect relaxation. When the knock-on induced compressive residual stress exceeded 0.8 GPa microcrack formation in the SiC substrate decreased coating adhesion. While no microcracks formed at the lowest bias voltage, insufficient atomic mobility during coating growth lead to voids forming in the coating. A balance is needed to form void-free coatings that have high compressive residual stress.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Air oxidation of yttrium hydride as a high temperature moderator for thermal neutron spectrum fission reactors

Yttrium hydride (YH x ) is an attractive moderator material for thermal neutron spectrum fission reactors requiring a small reactor core volume and has been selected as the neutron moderator for the Transformational Challenge Reactor (TCR), an advanced gas-cooled microreactor. Before YH x can be used in this application, it is important to understand the material response to off-normal conditions. In the present study, 550–650 °C isothermal dry air oxidation was performed to simulate a depressurized loss of force circulation (DLOFC) event. The oxidation was performed using thermogravimetric analysis (TGA) on bulk crack-free YHx coupons. Oxidation studies were also performed on Y coupons to elucidate the impact of H on oxidation. Both the chemistry and distribution of processing impurities were found to strongly affect oxidization behavior on a batch-to-batch basis. Regardless of batch, YHx oxidized at a significantly lower rate than Y at all temperatures, and the lower rate was directly correlated with increased hydride content. Metallic Y exhibited complex exponential kinetics, whereas YH x also exhibited complex kinetics but gained considerably less mass. According to literature reports on protonic and native-ion conductivities of Y 2 O 3 and mass spectrometry analysis of gaseous reaction products formed during the oxidation of YH x , a mechanism for the reduced oxidation rate of yttrium hydride is suggested.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Adhesion, structure, and mechanical properties of Cr HiPIMS and cathodic arc deposited coatings on SiC

Chromium deposited by high-power impulse magnetron sputtering (HIPIMS) versus cathodic arc (CA) processes exhibits very different mechanical properties. Combining the two can result in a single-phase with superior performance that can be tailored for use in coating SiC for advanced nuclear fuel cladding. Coating morphology, residual stress, elemental depth profiles, and mechanical testing by scratch, pull-off adhesion, and microcantilevers are shown for HiPIMS, CA, and combined coatings. CA coatings were likely to spall and had lower adhesion strength due to tensile residual stresses but had desirable material properties. Additionally, by depositing an initial layer of Cr by HiPIMS followed by a layer of Cr by CA, a more adherent coating was achieved and some of the stress issues with CA morphology were resolved. Combined coatings withstood at least 80 MPa in pull-off adhesion tests and had maximum failure stress values of 5GPa in microcantilever tests. These results were better than either individual deposition method and point to a hybrid approach being a path forward for a more robust coating.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Development of Standardized Property Requirements, Measurement Methods, and Reporting Guidance for Coatings

The diverse research programs presently in progress around the world have advanced a large number of coated cladding concepts to various stages of maturity. Industry teams have developed leading coating compositions and geometries, but significant disparity exists in the importance placed on different coating properties, as well as the applicability and standardization of the various methods used to measure these properties. This milestone report provides an initial assessment of the most important coating properties and the experimental methods available for their determination to lay the foundation for standardized reporting and development of comprehensive datasets for modeling efforts. This report has determined that the residual stress of the coating is the most crucial property to investigate, yet it has received minimal attention to date due to a knowledge gap about the growth and formation of the brittle Laves phase under reactor conditions. The significance of residual stress of the coating is further enhanced by its connection and interdependence with other important properties like creep and coating failure mechanisms. This milestone proposes an importance rating for residual stress properties and their associated measurement techniques to fill in knowledge. Furthermore, techniques to investigate the outlined properties are discussed, and their capabilities and limitations are illustrated to provide guidance on their use.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Evaluation of the effects of neutron irradiation on first-generation corrosion mitigation coatings on SiC for accident-tolerant fuel cladding

In this work, high purity SiC and SiC/SiC composites coated with commercial TiN, Cr, CrN, or CrN/Cr multilayer coatings were irradiated in Ar or flowing PWR water in the Massachusetts Institute of Technology Nuclear Reactor Laboratory (MITR). Irradiation in Ar was performed in the core. In the water environment, identical samples were placed in one of three different locations: in-core, providing exposure to neutron damage and radiolysis-affected water; above-core, where samples were exposed to radiolysis-affected water but not neutron damage, or outside of the core, where samples were exposed to the coolant water without the effects of radiation. Radiation in Ar revealed significant cracking of all but the TiN coatings, attributed to differential swelling between the coating and substrate. Lattice swelling was not observed in any of the coatings, but 0.2% void swelling was observed in the Cr coating. All of the coatings failed during water exposures in the core. Likewise, CrN/Cr spalled in each condition. Cr was protective, except under radiation damage as a result of cracking, and TiN severely degraded in the core with no coating was found following exposure. A SiC/coating ATF cladding system is anticipated to perform adequately following improvements in coating ductility and purity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Stability of zirconium hydride

This reports complete Milestone M2.2.1 Stability of zirconium hydride: Experimental investigation of the hydrogen concentration of encapsulated zirconium hydride before and after joining. In this milestone, we first report the successful fabrication of crack-free bulk zirconium hydride by using the ORNL static hydriding system. The fabricated hydride was characterized by using X-ray diffraction (XRD) and X-ray computed tomography (XCT). XRD results showed that the δ-phase zirconium hydride, α-zirconium, and ε-phase hydride were present. XCT confirmed the absence of the internal cracking. The hydrogen concentration was determined by the weight change before and after hydrogen loading, given the fact that the oxide phase was not found during XRD analysis. Two metallic cladding materials, i.e., molybdenum and FeCrAl, were selected to encapsulate the fabricated zirconium hydride pellets. Demonstration capsules were designed and machined. Mo crucible was successfully sealed by using a combination of electron beam welding and gas tungsten arc welding. Due to the undesired neutronics property of Mo, further investigation of this cladding material was not pursued. Electron beam welding and laser beam welding were employed to seal the FeCrAl capsules. XRD were performed on the zirconium hydride before and after the sealing. The results showed that the welding process has negligible impact on the hydrogen content in the FeCrAl-canned zirconium hydride. The presented work fulfilled the proposed work under this milestone and paves a pathway to fabricate single-phase zirconium hydride and to develop robust cladding materials to can bulk zirconium hydride.

36 MATERIALS SCIENCE↗