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Hawkins, Laura R.

Publications and source records attributed to Hawkins, Laura R..

6 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Summary Report on Ion Irradiation Study of Ceramic Coating on Suppressing FCCI

Advanced cladding is critical for advanced nuclear reactors with an enhanced performance in radiation tolerance and neutron transparency. Using advanced cladding will ensure the adequate thermal conductivity and mechanical stability of the cladding base material, corrosion resistance, high-temperature coolant compatibility of the cladding surface, and chemical stability in the cladding inner wall against fuel cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a modified surface, a clad base material, and a modified inner wall) promises to meet all these requirements. Initial research and development (R&D) regarding this innovative cladding seek to demonstrate the effectiveness of a thin ceramic coating in suppressing FCCI under ion irradiation to high dose. This United States (U.S.) Department of Energy (DOE)–Office of Nuclear Energy (NE) project report summarizes the results of recent ion irradiation studies of diffusion-couple samples with details of an ion irradiation experiment, the characterization of the interface microstructure of cerium (Ce)/titanium nitride (TiN)/oxide-dispersion strengthened (ODS), Ce/TiN/iron (Fe), Ce/ODS, and Ce/Fe samples irradiated with 80 million electron-volts (MeV) of xenon (Xe) ions to 100 displacements per atom (dpa) at 500?C using the Argonne Tandem Linac Accelerator System (ATLAS) facility at Argonne National Laboratory (ANL). The results of the ODS-substrate sample are encouraging and demonstrate the effectiveness of a thin ceramic coating on FCCI mitigation under irradiation. The results from an Fe-substrate sample revealed a complex microstructure, and the root cause is discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2023

A quantitative method to determine the region not influenced by injected interstitial and surface effects during void swelling in ion-irradiated metals

We propose and demonstrate a microstructurally-based experimental method to quantitively determine the depth regions in self-ion-irradiated metals that are affected by the injected interstitial effect and various surface effects, focusing on the choice of safe analysis zones to minimize the impact of these phenomena. The goal is to define the depth ranges where extracted data can be confidently applied to ion-neutron correlations for reactor application. Since ion energies in the range of 1–5 MeV are most frequently employed by the radiation effects community, irradiations were conducted at four energies in this range, all proceeding at 475 °C. The experiment was conducted on relatively pure single crystal iron to focus only on physical phenomena, avoiding the influence of possibly confounding chemical or segregation processes. Care was also taken to minimize the influence of other physical factors such as crystalline orientation. It was shown that, at 475 °C, ion energies of ≤1 MeV were too shallow in penetration and did not yield a safe depth range, but irradiations at 2.5 MeV and above yielded useful safe zones with predicted swelling behavior becoming independent of ion energy. The surface-affected zone width was found to be roughly twice that of the void-denuded zone width and to be independent of accumulated displacement dose. The largest injected-interstitial effect arises from the injected interstitial depression of void nucleation and growth. The interstitial-affected region starts at about one half of the projected range and does not show any “spreading” of its influence in depth as the peak damage level increases from 50 to 100 dpa. This study provides some confidence that enhances the credibility of ion simulation when applied to prediction of void swelling in neutron environments.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Status Report on Ion Irradiation Study of Ceramic Coating in Suppressing Fuel/Cladding Chemical Interaction

Advanced cladding is critical for the development of advanced nuclear reactors with enhanced performance in radiation tolerance and neutron transparency. It will ensure adequate thermal conductivity and mechanical stability of cladding base material, corrosion resistance and high temperature coolant compatibility of cladding surface. Finally, it provides chemical stability in the cladding inner wall against fuel/cladding chemical interaction (FCCI). An innovative cladding with a three-layer structure (i.e., a treated surface, a base material, and a treated inner-wall) promises to meet all the listed requirements. The initial work on innovative researching and developing these cladding materials seeks to demonstrate the effectiveness of a thin ceramic coating in suppressing FCCI through ion-irradiation. It consists of a study of Ce/TiN/Fe diffusion couples. This report provides an update on the status of ceramic-coating preparation and the testing of diffusion couple samples, the design of an ion-irradiation experiment, and characterization of the microstructure of a legacy-coupon sample coated with 1 µm of TiN and irradiated with 5 MeV Fe ions to 200 dpa at 500°C.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Designing Nuclear Fuels with a Multi-Principal Element Alloying Approach

Previous research has shown that multi-principal element alloys (MPEAs) using chromium, molybdenum, niobium, tantalum, titanium, vanadium, and zirconium can form stable body-centered-cubic (BCC) structures across a large temperature region (25°C to 1000°C). This is the same crystal structure as γ-uranium (U), which has shown desirable thermal and irradiation behavior in previous alloy fuel research. It is hypothesized then that the MPEA alloying approach can be used to produce a stable BCC uranium-bearing alloy and to retain its stability throughout anticipated operating regimes of power-producing reactors. Candidate elements were assessed using Monte Carlo N-Particle (MCNP) analysis to determine uranium densities necessary to make the alloy an economically viable fuel compared to conventional fuel forms. Following neutronic considerations, materials property databases and empirical predictors were used to determine the compositions with a high potential to form a BCC solid solution alloy. The final four alloys were MoNbTaU 2 , MoNbTiU 2 , NbTaTiU 2 , and NbTaVU 2 , which were cast using arc melting of raw elemental foils and chunks. Characterization of the fabricated alloys included scanning electron microscopy, X-ray diffraction, energy dispersive X-ray spectroscopy, and transmission electron microscopy. The results showed a two-phase system with a U-rich matrix phase surrounding the refractory precipitates. The U phase was found to contain varying concentrations of the alloying elements and was a BCC γ-U phase. These results warrant further research to identify ideal compositions for use as an advanced alloy fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Measurement of grain boundary strength of Inconel X-750 superalloy using in-situ micro-tensile testing techniques in FIB/SEM system

Grain boundaries (GBs), known as two-dimensional defects, are omnipresent in polycrystalline metallic alloys and thus influence a wide range of mechanical properties under different environmental conditions like irradiation and corrosion. Therefore, quantifying the strength of individual GBs is critical for understanding the degradation of mechanical properties of materials under different conditions. In this study we developed an efficient approach for the fabrication of micro-tensile specimens with a GB almost perpendicular to the tensile direction, which is expected to advance the development of individual GB tensile testing at micro or nanoscale in a wide scope of materials. Further, an in-situ cantilever micro-tensile testing method was developed and used to quantify the strength of a Σ3 GB in Inconel X-750 with the combination of finite element modeling. The average ultimate tensile strength (UTS) of a non-irradiated Σ3 GB is estimated at around 1.4 GPa, comparable to that of a neutron-irradiated Σ3 GB with a dose of ~1.5 dpa (1.3 GPa). Moreover, the in-situ push-to-pull micro-tensile testing technique developed in this work provides valuable insights into the high-angle GB deformation and fracture behavior. This method generates qualitatively similar ductility behavior before and after neutron irradiation as the bulk material testing. However, the ductility and UTS values obtained from this method are different from bulk measurements due to vastly different specimen dimensions.

36 MATERIALS SCIENCE↗