James I. Cole

Development of Diffusion Barrier Coatings for Mitigation of Fuel-Cladding Chemical Interactions

Fuel Cladding Chemical Interactions (FCCI) in a nuclear reactor occur due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and the formation of low melting point eutectic compounds. Deposition of diffusion barrier coatings of a thin oxide on the inner surface of the cladding can potentially reduce or delay the onset of FCCI. This study examines the feasibility of using nanofluid-based electrophoretic deposition (EPD) process to deposit coatings of titanium oxide, yttria-stabilized zirconia (YSZ) and vanadium oxide. The deposition parameters, including the nanofluid composition, current, and voltage were optimized for each coating material using test flat substrates of T91 ferritic-martensitic steel. Diffusion characteristics of the coatings were investigated by diffusion couple experiments using the fuel surrogate cerium. These diffusion couple studies performed in the temperature range of 560°C and 585°C showed that the oxide coatings significantly reduce the solid state inter-diffusion between cerium to steel.

Implementation of focused ion beam (FIB) system in characterization of nuclear fuels and materials

Beginning in 2007, a program was established at the Idaho National Laboratory to update key capabilities enabling microstructural and micro-chemical characterization of highly irradiated and/or radiologically contaminated nuclear fuels and materials at scales that previously had not been achieved for these types of materials. Such materials typically cannot be contact handled and pose unique hazards to instrument operators, facilities, and associated personnel. Over the ensuing years, techniques have been developed and operational experience gained that has enabled significant advancement in the ability to characterize a variety of fuel types including metallic, ceramic, and coated particle fuels, obtaining insights into in-reactor degradation phenomena not achievable by any other means. The following article describes insights gained, challenges encountered, and provides examples of unique results obtained in adapting dual beam FIB technology to nuclear fuels characterization.

High Resolution Transmission Electron Microscopy of Irradiation Damage in Inconel X-750

The effects of irradiation on Inconel® (Inconel is a registered trademark of Special Metals Corporation and its subsidiaries) X-750, have been shown to lead to embrittlement and intergranular fracture. This is now widely accepted to be a result of intergranular helium bubbles over the fluence range studied. This paper provides a quantitative assessment and a detailed discussion of the radiation-induced defects including; helium bubbles (size and density distribution, and grain boundary area fraction), dislocation loops and stacking fault tetrahedra, and the disordering and dissolution of secondary gamma prime precipitates. The microstructural evolution will be presented and discussed as a function of dose (from ~5.5 to ~80 dpa), helium concentration (~1300 to ~25,000 appm helium), and irradiation temperature (~120–280 to ~300–330 °C).

A Comparative Study of Welded ODS Cladding materials for AFCI/GNEP Applications

This research project involved working on the pressure resistance welding of oxide dispersion strengthened (ODS) alloys which will have a large role to play in advanced nuclear reactors. The project also demonstrated the research collaboration between four universities and one nation laboratory (Idaho National Laboratory) with participation from an industry for developing for ODS alloys. These alloys contain a high number density of very fine oxide particles that can impart high temperature strength and radiation damage resistance suitable for in-core applications in advanced reactors. The conventional fusion welding techniques tend to produce porosity-laden microstructure in the weld region and lead to the agglomeration and non-uniform distribution of the neededoxide particles. That is why two solid state welding methods - pressure resistance welding (PRW) and friction stir welding (FSW) - were chosen to be evaluated in this project. The proposal is expected to support the development of Advanced Burner Reactors (ABR) under the GNEP program (now incorporated in Fuel Cycle R&D program). The outcomes of the concluded research include training of graduate and undergraduate students and get them interested in nuclear related research.

Sample preparation artifacts in nuclear materials and mitigation strategies

Diverse microstructures form in nuclear materials upon exposure to radiation. The defects produced during irradiation of materials can alter their mechanical properties and lead to embrittlement of reactor structural materials during service life. Therefore, it is imperative to know various radiation effects in reactor materials since it can aid in understanding in-reactor degradation behavior, accounting for irradiation effects in design, and producing new generation radiation-tolerant materials. Characterization of radiation-induced changes in reactor materials at the nano and atomic scales is typically conducted in transmission electron microscopes (TEM). Three most commonly used sample preparation techniques include electro-polishing, broadbeam ion milling, and focused ion beam (FIB) approach. However, preparation of samples using conventional sample preparation techniques, such as electro-polishing and ion milling, requires close-in, hands-on manipulation of the sample for extended periods of time. This is not feasible with highly radioactive nuclear materials.

An Assessment of Layer Development at the Fuel/Cladding Interface During Irradiation of Metallic SFR Fuel Elements

To investigate fuel cladding chemical interaction in irradiated metallic nuclear fuels, diffusion couple experiments have been performed using prototypic metallic fuel alloys with additions of noble metal and lanthanide fission product components mated against stainless steel claddings. The developed interdiffusion zones have been characterized using SEM/EDS/WDS to determine the development of phases and the interdiffusion behavior of specific fuel, cladding, and fission product components. The formed diffusion structures have been compared to actual interaction zones that form in irradiated metallic SFR fuels. This paper discusses how the structures compare between the diffusion couple-generated interdiffusion zones and those that develop in irradiated metallic nuclear fuels. It was found that similarities exist between the phase development and interdiffusion behavior in the annealed diffusion couples and the irradiated fuels. Nd, Mo, and Ru, which were added to a fuel alloy to represent fission products that are present in irradiated metallic nuclear fuels, were found to exhibit interdiffusion behavior in annealed diffusion couples that was similar to what has been observed in actual irradiated metallic fuels. This was also true for the original fuel components U, Pu, and Zr, along with the cladding constituent Fe, Ni, and Cr.