Selective autophagy is a conserved subcellular process that maintains the health of eukaryotic cells by targeting damaged or toxic cytoplasmic components to the vacuole/lysosome for degradation. A key player in the initiation of selective autophagy in S. Cerevisiae (baker's yeast) is a large adapter protein called Atg11. Atg11 has multiple predicted coiled-coil domains and intrinsically disordered regions, is known to dimerize, and binds and organizes other essential components of the autophagosome formation machinery, including Atg1 and Atg9. We performed systematic directed mutagenesis on the coiled-coil 2 domain of Atg11 in order to map which residues were required for its structure and function. Using yeast-2-hybrid and coimmunoprecipitation, we found only three residues to be critical: I562, Y565, and I569. Mutation of any of these, but especially Y565, could interfere with Atg11 dimerization and block its interaction with Atg1 and Atg9, thereby inactivating selective autophagy.

U-Pu-Zr metallic fuels are important fuel candidates for future advanced and/or test reactors. To better understand the fuel performance and guide future fuel design, a U-20Pu-10Zr (in weight) metallic fuel irradiated in the Experimental Breeder Reactor-Ⅱ was revisited using advanced electron microscopies. This fuel cross section was irradiated to a burnup of 6.15 % at a cladding temperature of ∼500°C. Several transmission electron microscopy samples were extracted from different but representative radial locations of the fuel cross section using focused ion beam technique. Phase identification and chemical analysis with sub-micron spatial resolution were carried out using a scanning transmission electron microscopy. Multiple phenomena that are critical to fuel performance, such as Zr redistribution, were revealed in an unpreceded highly detailed manner. An improved understanding of fuel reconstruction in U-Pu-Zr metallic fuel under reactor irradiation is provided.

Understanding swelling behavior in monolithic, uranium-molybdenum, plate-type fuel is necessary to qualify the fuel for reactor use and for the conversion of high performance research reactors from highly enriched to low-enriched uranium. Multiple mechanisms influence plate dimensional stability, including solid and gaseous fission-product induced swelling, irradiation-assisted creep, fuel-phase transformation, and interaction-layer formation. Separating these phenomena remains a challenge, and current models do not appear to adequately predict experimental results at higher fission densities where it is most critical. To mitigate the mechanistic uncertainty, post-irradiation profilometry is used to increase the number of data points available over a range of in-reactor irradiation experiments conducted at the Advanced Test Reactor, and to provide a statistical precedent for a swelling-behavior model. This work establishes a predictive swelling correlation as a function of fission density, with associated confidence and prediction bounds, by analyzing more than 18,000 thickness data points collected on 74 irradiated U-10Mo monolithic fuel test plates over a range of irradiation conditions.

Abstract Uranium mononitride (UN) with 5 wt.% uranium dioxide (UO2) is used as a model system to study the phase and defect evolution under proton irradiation in nitride-oxide composite. Phase composition, crystallographic orientation relationships (ORs) and dislocation loops were characterized using X-ray diffraction, transmission electron microscopy, and energy dispersive X-ray spectroscopy techniques. Proton-irradiation at elevated temperatures promoted the transformation of UN into uranium sesquinitride (U2N3) and UO2 phases. U2N3 and UO2 formed a fully coherent structure with two ORs: {002}U2N3‖{002}UO2 and [001]U2N3‖[001]UO2; U2N3{101}‖UO2{101} and U2N3[101]‖UO2[101] due to low lattice misfit (2.3%) and low interfacial energy (127 mJ/m2). Observed oxidation of UN and coherent interface are consistent with density-functional theory calculations which suggest lower energy for oxidized configuration and low energy of the interface. The dislocation loops grew while their number density decreased with the temperature and dose. The loop size was over three times larger in two nitride phases than that in UO2, while the number density was one order of magnitude higher in UO2 than in nitride phases. Loop density and diameter were analyzed using a kinetic rate theory that considers stoichiometric loop evolution. This analysis led to the conclusion in all compounds loop growth is governed by mobility of uranium interstitials, and enabled measurement of diffusion coefficients of uranium interstitials and non-metal interstitials and vacancies. This analysis provided a comparative study of early stage of microstructure evolution under irradiation which has implications for use of this mixture as advanced fuel in nuclear energy systems.

This article on research reactor fuel irradiation performance reviews the overall irradiation performance requirements a fuel system must meet to demonstrate suitability for regulatory qualification and safe use in research and test reactors. The criteria that must be evaluated to demonstrate the requirements are met and the principal phenomena that lead to fuel properties degradation during service life are broadly discussed in the context of the reactors operational and safety envelope.

Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.

Three dimensional (3D) X-ray microscopy, also referred to as X-ray computed tomography, is a data-rich characterization technique that can provide surface and subsurface spatial information in a non-destructive manner. 1-3 However, 3D X-ray imaging of nuclear fuel is a challenging endeavor. 4 X-ray attenuation of high-Z materials, such as uranium, can be significant. This limits the practical maximum X-ray path length (i.e., sample size) to a few millimeters at most. This physical limitation impedes the nondestructive microstructural imaging of real-world samples, such as fuel rods. However, scaling nuclear fuel samples to the appropriate size allows for X-ray path lengths that are suitable for high resolution, laboratory-based 3D X-ray microscopy. transmissions. This presentation will give an overview of laboratory-based 3D X-ray microscopy of appropriately sized, unirradiated U-10Zr fuel cylinders as part of a Nuclear Science User Facility-funded collaborative project to validate Idaho National Laboratory’s MARMOT mesoscale fuel performance code. The challenges of imaging these samples, and their solutions, will be addressed. These include the development of a novel, GPU-accelerated post-reconstruction cupping artifact correction method and the implementation of pattern recognition-based segmentation methods 5, 6 using open-source image processing software 7, 8 .

Uranium mononitride (UN) with 5 wt.% uranium dioxide (UO2) is used as a model system to study the phase and defect evolution under proton irradiation in nitride-oxide composite. Phase composition, crystallographic orientation relationships (ORs) and dislocation loops were characterized using X-ray diffraction, transmission electron microscopy, and energy dispersive X-ray spectroscopy techniques. Proton-irradiation at elevated temperatures promoted the transformation of UN into uranium sesquinitride (U2N3) and UO2 phases. U2N3 and UO2 formed a fully coherent structure with two ORs: {002}U2N3‖{002}UO2 and [001]U2N3‖[001]UO2; U2N3{101}‖UO2{101} and U2N3[101]‖UO2[101] due to low lattice misfit (2.3%) and low interfacial energy (127 mJ/m2). Observed oxidation of UN and coherent interface are consistent with density-functional theory calculations which suggest lower energy for oxidized configuration and low energy of the interface. The dislocation loops grew while their number density decreased with the temperature and dose. The loop size was over three times larger in two nitride phases than that in UO2, while the number density was one order of magnitude higher in UO2 than in nitride phases. Loop density and diameter were analyzed using a kinetic rate theory that considers stoichiometric loop evolution. This analysis led to the conclusion in all compounds loop growth is governed by mobility of uranium interstitials, and enabled measurement of diffusion coefficients of uranium interstitials and non-metal interstitials and vacancies. This analysis provided a comparative study of early stage of microstructure evolution under irradiation which has implications for use of this mixture as advanced fuel in nuclear energy systems.

Approximately 250 research and test reactors are used for materials research, education and training, neutron scattering, isotope production, and materials research in 55 countries. This class of nuclear reactors operate over an enormous range of core power densities, ranging from near zero to more than 30 times higher than conventional pressurized water reactors. Because research and test reactors are designed to meet a wide range of needs, a broad spectrum of fuel system designs and material chemistries are used to meet reactor performance objectives; this is most apparent in the class of high-performance research reactors. High-performance research reactors are designed with the overarching goal of producing a sustained high neutron flux focused toward a specific purpose; for example, neutron scattering, material testing, or isotope production. To achieve this goal requires fuel systems designed to operate at power levels and fission densities that exceed the requirements of commercial power reactors. The complex irradiation-enhanced chemical evolution of the fuel with respect to the cladding, noble fission gases, and solid fission products determines the performance and lifetime of these fuels.

A boron-modified molybdenum silicide material having the composition comprising about 80 to about 90 weight % Mo, about 10 to about 20 weight % Si, and about 0.1 to about 2 weight % B and a multiphase microstructure including Mo.sub.5 Si.sub.3 phase as at least one microstructural component effective to impart good high temperature creep resistance. The boron-modified molybdenum silicide material is fabricated into such products as electrical components, such as resistors and interconnects, that exhibit oxidation resistance to withstand high temperatures in service in air as a result of electrical power dissipation, electrical resistance heating elements that can withstand high temperatures in service in air and other oxygenbearing atmospheres and can span greater distances than MoSi.sub.2 heating elements due to improved creep resistance, and high temperature structural members and other fabricated components that can withstand high temperatures in service in air or other oxygen-bearing atmospheres while retaining creep resistance associated with Mo.sub.5 Si.sub.3 for structural integrity.

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 degrees C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to 20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice. (C) 2017 Elsevier B.V. All rights reserved.

Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

Autophagy is a dynamic process by which cytosolic material is sequestered into double membrane vesicles called autophagosomes and then sent to the vacuole / lysosome for degradation. Selective autophagy, which targets specific unwanted components such as a damaged mitochondria, protein aggregates and pathogens, plays central roles in innate and adaptive immunity, regulation of inflammatory responses, pathogen clearance and protection against Parkinson's and Alzheimer's diseases. Autophagy is performed by dozens of autophagy‐related proteins (Atg) that assemble in a hierarchal fashion to form the pre‐autophagosomal structure (PAS). Atg11, a protein crucial for selective autophagy in the yeast Saccharomyces Cerevisiae , contains two coiled‐coil domains, CC2 and CC3, which are the primary binding sites used to organize the autophagy initiation complex and recruit Atg9‐containing vesicles to the PAS. It is not known whether Atg11 performs both of these processes simultaneously, or whether there is competition between its binding partners that would suggest a sequential model of action. Here we use a random mutagenesis / yeast‐2‐hybrid screen to determine the specific amino acid residues in the Atg11 CC2 region that are necessary for the interaction with Atg9 in order to more precisely map its binding site. Knowledge of the specific binding sites of all of Atg11's partners will help us to determine how Atg11 organizes these proteins to promote the creation of the autophagosome and accomplish selective autophagy. Support or Funding Information NSF RUI Grant #1613653 “Scaffold or Assembly Line: How Does Atg11 Organize its Binding Partners for the Initiation of Selective Autophagy?” to S. K. Backues

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or similar to 5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup. (C) 2017 Elsevier B.V. All rights reserved.

This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

The core components of nuclear reactors (e.g., fuel assemblies, spacer grids, control rods) encounter harsh environments due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of nuclear power plants; post-irradiation examination (PIE) can reveal information about the integrity of these components. Neutron computed tomography (CT) is one important PIE measurement tool for nondestructively evaluating the structural integrity of these items. CT typically requires many projections to be acquired from different view angles, after which a mathematical algorithm is used for image reconstruction. However, when working with heavily irradiated materials and irradiated nuclear fuel, obtaining many projections is laborious and expensive. Image reconstruction from a smaller number of projections has been explored to achieve faster and more cost-efficient PIE. Classical reconstruction methods (e.g., filtered backprojection), unfortunately, do not typically offer stable reconstructions from a highly asymmetric, few-projection data set and often create severe streaking artifacts. We propose an iterative reconstruction technique to reconstruct curved, plate-type nuclear fuel assemblies using limited-angle CT. The performance of the proposed method is assessed using simulated data and validated through real projections. We also discuss the systematic strategy for establishing the conditions of reconstructions and finding the optimal imaging parameters for reconstructions of the fuel assemblies from few projections using limited-angle CT. Results show that a fuel assembly can be reconstructed using limited-angle CT if 36 or more projections are taken from a particular direction with 1 degrees angular increment.

Abstract The corrosion resistance of cerium silicide, a surrogate of uranium silicide, is investigated to gain insight into the reaction of uranium silicide with water. As-received and proton-irradiated Ce3Si2, CeSi2, and CeSi1.x monolithic pellets are subjected to corrosion tests in water at 300°C and 9 MPa for up to 48 h. Results show that an oxide layer composed of Ce4.67 (SiO4)3O forms on the surface of all samples, and it grows thicker with extended exposure times. Irradiated samples corrode to a greater extent than their unirradiated counterparts, which is mainly a result of the existing post-irradiation cerium oxide and the presence of ion-induced defects. Most of the Ce3Si2 samples crack (as-received) or fracture (ion-irradiated) during testing, which is due to the brittleness of the samples and oxide erosion/spallation that occur during testing.

A monolithic fuel design based on U-Mo alloy has been selected as the fuel type for conversion of United States’ high-performance research reactors (USHPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184kJ/mole, consistent with a previous diffusion-couple study. The phases in the U–Mo/Zr interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion-couple study.