B.D. Miller

SCANNING ELECTRON MICROSCOPY ANALYSIS OF FUEL/MATRIX INTERACTION LAYERS IN HIGHLY-IRRADIATED U-Mo DISPERSION FUEL PLATES WITH Al AND Al–Si ALLOY MATRICES

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U?7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U?7Mo dispersion fuel elements with pure Al, Al?2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U?7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U?7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al?Si matrices.

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.

The Role of Grain Size on Neutron Irradiation Response of Nanocrystalline Copper

The role of grain size on the developed microstructure and mechanical properties of neutron irradiated nanocrystalline copper was investigated by comparing the radiation response of material to the conventional micrograined counterpart. Nanocrystalline (nc) and micrograined (MG) copper samples were subjected to a range of neutron exposure levels from 0.0034 to 2 dpa. At all damage levels, the response of MG-copper was governed by radiation hardening manifested by an increase in strength with accompanying ductility loss. Conversely, the response of nc-copper to neutron irradiation exhibited a dependence on the damage level. At low damage levels, grain growth was the primary response, with radiation hardening and embrittlement becoming the dominant responses with increasing damage levels. Annealing experiments revealed that grain growth in nc-copper is composed of both thermally-activated and irradiation-induced components. Tensile tests revealed minimal change in the source hardening component of the yield stress in MG-copper, while the source hardening component was found to decrease with increasing radiation exposure in nc-copper.

Analysis and comparison of focused ion beam milling and vibratory polishing sample surface preparation methods for porosity study of U-Mo plate fuel for research and test reactors

Uranium-Molybdenum (U-Mo) low enriched uranium (LEU) fuels are a promising candidate for the replacement of high enriched uranium (HEU) fuels currently in use in a high power research and test reactors around the world. Contemporary U-Mo fuel sample preparation uses focused ion beam (FIB) methods for analysis of fission gas porosity. However, FIB possess several drawbacks, including reduced area of analysis, curtaining effects, and increased FIB operation time and cost. Vibratory polishing is a well understood method for preparing large sample surfaces with very high surface quality. In this research, fission gas porosity image analysis results are compared between samples prepared using vibratory polishing and FIB milling to assess the effectiveness of vibratory polishing for irradiated fuel sample preparation. Scanning electron microscopy (SEM) imaging was performed on sections of irradiated U-Mo fuel plates and the micrographs were analyzed using a fission gas pore identification and measurement script written in MatLab. Results showed that the vibratory polishing method is preferentially removing material around the edges of the pores, causing the pores to become larger and more rounded, leading to overestimation of the fission gas porosity size. Whereas, FIB preparation tends to underestimate due to poor micrograph quality and surface damage leading to inaccurate segmentations. Despite the aforementioned drawbacks, vibratory polishing remains a valid method for porosity analysis sample preparation, however, improvements should be made to reduce the preferential removal of material surrounding pores in order to minimize the error in the porosity measurements.

Direct alpha spectrometry as a screening method for assaying thick, highly-radioactive materials

Single-crystal (sc) chemical vapor deposition (CVD) prepared diamond semiconductor detectors have been used as alpha spectrometers to analyze untreated, thick-matrix materials with high beta radiation emission rates. Tests using a scCVD diamond detector to assay alpha emissions from high burn-up irradiated nuclear fuel show the ability to produce reasonable thick-matrix alpha spectra when the beta radiation field exceeds the alpha radiation field by up to six orders of magnitude. Simple semi-quantitative analyses of thick matrix spectra have been used to confirm the presence of higher-energy alpha-emitting actinides (plutonium and curium) when in the presence of lower-energy alpha emitting actinides (uranium).

Irradiation effects in Generation IV nuclear reactor materials

Generation IV reactor structural materials will be exposed to high doses and temperatures during reactor operation that may lead to irradiation-induced degradation. This degradation will differ from that seen in light water reactors and therefore understanding mechanisms controlling material performance during irradiation is critical for evaluating the viability of Generation IV nuclear reactor concepts. This chapter discusses irradiation effects and microstructural changes that affect mechanical properties and dimensional stability of Generation IV reactor materials.

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.

Characterization of Phases Formed Between U-Pu-X Fuels and Fe-Based Cladding

Uranium-plutonium-zirconium (U-Pu-Zr) and uranium-plutonium-molybdenum (U-Pu-Mo) fuels, known for their high burnup and good thermal response, have been considered as candidate fuels for advanced fast reactors. During their lifetime in the reactor, irradiation in combination with high temperatures can result in swelling of the fuel and its interaction with the cladding. As a result of the complex fuel-cladding chemical interaction (FCCI), integrity of fuel and cladding could be compromised and therefore should be comprehensively examined. As part of the fuel cycle research and development (FCRD) program, formation of intermetallic phases within fuel-cladding interaction zones was investigated in scanning electron microscope (SEM) and transmission electron microscope (TEM).

Microstructural Characterization of the Irradiated Nuclear Fuels

The microstructural characterization using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) for the irradiated fuels played an important role to the understanding of fuel performance. Significant progresses have been made in recent years on SEM and TEM work for fuel development in reduced-enrichment for research and test reactors (RERTR) program [1, 2]. It is extremely challenging to prepare the samples from the highly radioactive irradiated fuel for high resolution microscopy analysis. For the complex microstructure of irradiated fuels, the traditional mechanical polishing tends to produce a smeared and disturbed surface making it difficult to reveal the original microstructure in SEM while the traditional TEM sample preparation often limits the ability to access the areas of interest for detailed analysis. The new development using the focused-ion-beam (FIB) lift-out and polishing technical at the Idaho National Laboratory (INL) demonstrated the great advantage in microstructural characterization for the irradiated nuclear fuels.