Mitch Meyer

Materials challenges for nuclear systems

The safe and economical operation of any nuclear power system relies to a great extent, on the success of the fuel and the materials of construction. During the lifetime of a nuclear power system which currently can be as long as 60 years, the materials are subject to high temperature, a corrosive environment, and damage from high-energy particles released during fission. The fuel which provides the power for the reactor has a much shorter life but is subject to the same types of harsh environments. This article reviews the environments in which fuels and materials from current and proposed nuclear systems operate and then describes how the creation of the Advanced Test Reactor National Scientific User Facility is allowing researchers from across the United States to test their ideas for improved fuels and materials.

Microstructure Evolution in ZrC Irradiated with Kr ions

The gas-cooled fast reactor (GFR) is one of six concepts for the Generation-IV nuclear energy system. The fuel for the GFR requires both a high heavy metal loading and the ability to withstand temperatures up to 1600°C during a loss of coolant accident. ZrC is among the few potential refractory ceramic materials with necessary properties to be considered as matrix materials for a dispersed carbide fuel. The radiation response of ZrC to high dose and temperature is a critical research need. This work investigated the microstructure of ZrC irradiated with 1 MeV Kr ions to doses of 10 and 30 dpa at 27°C and 10 and 70 dpa at 800°C with a damage rate approximately 3.0 × 10−3 dpa/s. No radiation-induced amorphization was found. A lattice expansion of approximately 7 % was observed for ZrC irradiated to 70 dpa at 800°C.

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.

RERTR-12 Post-irradiation Examination Summary Report

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding. These characterization activities are tailored specifically to define: • The mechanical response of fuel meat, cladding, and interlayers, including diffusion barrier integrity • Whether geometry is stable and predictable; that changes in channel gap do not compromise ability to cool fuel • That fuel performance is known and predictable • A limited set of physical properties that are important for the analysis of fuel burnup limits • Whether swelling is stable and predictable.

Thermal Properties Measurement Report

The Thermal Properties Measurement Report summarizes the research, development, installation, and initial use of significant experimental thermal property characterization capabilities at the INL in FY 2015. These new capabilities were used to characterize a U3Si2 (candidate Accident Tolerant) fuel sample fabricated at the INL. The ability to perform measurements at various length scales is important and provides additional data that is not currently in the literature. However, the real value of the data will be in accomplishing a phenomenological understanding of the thermal conductivity in fuels and the ties to predictive modeling. Thus, the MARMOT advanced modeling and simulation capability was utilized to illustrate how the microstructural data can be modeled and compared with bulk characterization data. A scientific method was established for thermal property measurement capability on irradiated nuclear fuel samples, which will be installed in the Irradiated Material Characterization Laboratory (IMCL).

AECL/U.S. INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors Fuel Requirements and Down-Select Report

This report documents the first milestone of the International Nuclear Energy Research Initiative (INERI) U.S./Euratom Joint Proposal 1.8 entitled “Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Light-Water Reactors.” The milestone represents the assessment and preliminary study of a variety of fuels that hold promise as transmutation and minor actinide burning fuel compositions for light-water reactors. The most promising fuels of interest to the participants on this INERI program have been selected for further study. These fuel compositions are discussed in this report.

U.S./EURATOM INERI - Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in LWRs -- Fuel Requirements and Down-Select Report

This report documents the first milestone of the International Nuclear Energy Research Initiative (INERI) U.S./Canada Joint Proposal entitled “Development of Inert Matrix Fuels for Plutonium and Minor Actinide Management in Power Reactors.” The milestone represents the assessment and preliminary study of a variety of fuels that hold promise as transmutation and minor actinide burning fuel compositions for light water reactors. The most promising fuels of interest to the participants on this INERI program have been selected for further study. These fuel compositions are discussed in this report.