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Techniques that employ positron annihilation spectroscopy are powerful tools to investigate defect structures and concentrations in materials. A hindrance to experimental design and the interpretation of results lies in the lack of agreement in the literature concerning the proper form of the positron implantation profile, a function that determines the sensitivity range for all non-slow positron annihilation spectroscopy techniques. Employing the dominant 22 Na isotopic source, a positron implantation profile database of 270 common materials is published. The parameters for a novel implantation profile functional form providing superior agreement with simulation are derived. Finally, and most critically, an algorithm is presented and validated, which permits utilization of the published elemental implantation profile parameters to produce the positron implantation profile for any material of interest. This tool provides rapid calculation of the sensitivity range for all positron annihilation techniques, enabling more informed experimental design and more accurate knowledge of the spatial distribution of defects in materials.

Background: In response to supply shortages caused by the COVID-19 pandemic, N95 filtering facepiece respirators (FFRs or “masks”), which are typically single-use devices in healthcare settings, are routinely being used for prolonged periods and in some cases decontaminated under “reuse” and “extended use” policies. However, the reusability of N95 masks is limited by degradation of fit. Possible substitutes, such as KN95 masks meeting Chinese standards, frequently fail fit testing even when new. The purpose of this study was to develop an inexpensive frame for damaged and poorly fitting masks using readily available materials and 3D printing. Results:An iterative design process yielded a mask frame consisting of two 3D printed side pieces, malleable wire links that users press against their face, and cut lengths of elastic material that go around the head to hold the frame and mask in place. Volunteers (n = 45; average BMI = 25.4), underwent qualitative fit testing with and without mask frames wearing one or more of four different brands of FFRs conforming to US N95 or Chinese KN95 standards. Masks passed qualitative fit testing in the absence of a frame at rates varying from 48 to 94 % (depending on mask model). For individuals who underwent testing using respirators with broken or defective straps, 80–100 % (average 85 %) passed fit testing with mask frames. Among individuals who failed fit testing with a KN95, ~ 50 % passed testing by using a frame. Conclusions: Our study suggests that mask frames can prolong the lifespan of N95 and KN95 masks by serving as a substitute for broken or defective bands without adversely affecting fit. Use of frames made it possible for ~ 73 % of the test population to achieve a good fit based on qualitative and quantitative testing criteria, approaching the 85–90 % success rate observed for intact N95 masks. Frames therefore represent a simple and inexpensive way of expanding access to PPE and extending their useful life. For clinic

The initial stage of radiation defect creation has often been shown to follow a power law distribution at short time scales, recently so with tungsten, following many self-organizing patterns found in nature. The evolution of this damage, however, is dominated by interactions between defect clusters, as the coalescence of smaller defects into clusters depends on the balance between transport, absorption, and emission to/from existing clusters. The long-time evolution of radiation-induced defects in tungsten is studied with cluster dynamics parameterized with lower length scale simulations, and is shown to deviate from a power law size distribution. The effects of parameters such as dose rate and total dose, as parameters affecting the strength of the driving force for defect evolution, are also analyzed. Excellent agreement is achieved with regards to an experimentally measured defect size distribution at 30 K. This study provides another satisfactory explanation for experimental observations in addition to that of primary radiation damage, which should be reconciled with additional validation data.

Much of today's research in nuclear materials relies heavily on archived, historical specimens, as neutron irradiation facilities become ever more scarce. These materials are subject to many processes of stress- and irradiation-induced microstructural evolution, including those during and after irradiation. The latter of these, referring to specimens “naturally aged” in ambient laboratory conditions, receives far less attention. The long and slow set of rare defect migration and interaction events during natural aging can significantly change material properties over decadal timescales. This paper presents the results of natural aging carried out over 15 years on austenitic stainless steels from a BN-350 fast breeder reactor, each with its own irradiation, stress state, and natural aging history. Natural aging is shown to significantly reduce hardness in these steels by 10–25% and partially alleviate stress-induced hardening over this timescale, showing that materials evolve back towards equilibrium even at such a low temperature. The results in this study have significant implications to any nuclear materials research program which uses historical specimens from previous irradiations, challenging the commonly held assumption that materials “on the shelf” do not evolve. Keyword: Nuclear and High Energy Physics; General Materials Science; Nuclear Energy and Engineering, U.S. Nuclear Regulatory Commission (Grant NRC-HQ-84-15-G-0045)

Irradiation-induced void swelling remains a major challenge to nuclear reactor operation. Swelling may take years to initiate and often results in rapid material property degradation once started. Alloy development for advanced nuclear systems will require rapid characterization of the swelling breakaway dose in new alloys, yet this capability does not yet exist. We demonstrate that transient grating spectroscopy (TGS) can detect void swelling in single crystal copper via changes in surface acoustic wave (SAW) velocity. Scanning transmission electron microscopy (STEM) links the TGS-observed changes with void swelling-induced microstructural evolution. These results are considered in the context of previous work to suggest that in situ TGS will be able to rapidly determine when new bulk materials begin void swelling, shortening alloy development and testing times., United States. Department of Energy (Agreement DE-NA0002135)

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues., Electric Power Research Institute (Contract 10002739), Electric Power Research Institute (Contract 10004433), Electric Power Research Institute (Contract 10005086)

Understanding and predicting radiation damage is of central importance to develop radiation-tolerant structural materials for current and next-generation nuclear systems. Single-phase solid solution alloys constitute attractive choices due to their promising mechanical properties and radiation tolerance. Here, by examining radiation-induced defect production and evolution in single-phase Ni-Fe alloys, we show that radiation damage resistance directly correlates with thermodynamic mixing energy and heterogeneity of defect diffusion. We found that radiation damage in materials decreases linearly with lowering mixing energy, and the relationship holds true for all studied Ni-Fe compositions. The damage reduction with varying composition is further ascribed to the increasing heterogeneity of point defect migration across a complex potential energy landscape that enhances defect recombination. This new insight into the dynamical evolution of radiation defects points to a thermodynamic criterion for designing radiation-tolerant materials.

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud may cause serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each makes critical assumptions regarding how to treat the complex, porous microstructure of crud and its resultant effects on temperature, pressure, and crud chemistry. In this study, we demonstrate that crud is indeed a fractalline porous medium using flowing loop experiments, validating the most recent models of its effects on LWR fuel cladding. This crud is shown to match that in other LWR-prototypical facilities through a porosity-fractal dimension scaling law. Implications of this result range from post-mortem analysis of the effects of crud on reactor fuel performance, to utilizing crud's fractalline dimensions to quantify the effectiveness of anti-fouling measures., Electric Power Research Institute (Contract 10002739), Electric Power Research Institute (Contract 10004433), Electric Power Research Institute (Contract 10005086)

Full three dimensional (3D) simulations of ion implantation are necessary in a wide range of nanoscience and nanotechnology applications to capture the increasing effect of ion leakage out of surfaces. Using a recently developed 3D Monte Carlo simulation code IM3D, we first quantify the relative error of the 1D approach in three applications of nano-scale ion implantation: (1) nano-beam for nitrogen-vacancy (NV) center creation, (2) implantation of nanowires to fabricate p–n junctions, and (3) irradiation of nano-pillars for small-scale mechanical testing of irradiated materials. Because the 1D approach fails to consider the exchange and leakage of ions from boundaries, its relative error increases dramatically as the beam/target size shrinks. Lastly, the “Bragg peak” phenomenon, where the maximum radiation dose occurs at a finite depth away from the surface, relies on the assumption of broad beams. We discovered a topological transition of the point-defect or defect-cluster distribution isosurface when one varies the beam width, in agreement with a previous focused helium ion beam irradiation experiment. We conclude that full 3D simulations are necessary if either the beam or the target size is comparable or below the SRIM longitudinal ion range., National Science Foundation (U.S.) (Grant DMR-1120901), National Natural Science Foundation (China) (11475215), National Natural Science Foundation (China) (11775254), Gordon and Betty Moore Foundation, Chinese Academy of Sciences. Youth Innovation Promotion Association

The influence has been studied of thermo-mechanical treatment, sensitization conditions, and neutron irradiation on the pitting corrosion resistance of austenitic 316LN stainless steel variants in 10% FeCl[subscript 3]·6H[subscript 2]O at 22 °C. Variants of this steel were modified with additions of nitrogen, manganese, copper, and tungsten, as well as testing cast, cold-rolled, grain boundary engineered (GBE), and as-received variants. It was found that the 316LN steel variant with additions of 0.2% N and 2% Mn had the best pitting corrosion resistance of all studied conditions. When irradiated in a light water reactor (LWR) to a maximum fluence of 3 × 1017 n/cm[superscript 2] (E > 1.1 meV, Tirr < 50 °C), neutron irradiation surprisingly increased the resistance of GBE steels to pitting corrosion. An anisotropy of corrosion resistance of GBE and cold rolled steels was observed. Keywords: Austenitic stainless steel, Alloying, Nitrogen, Tungsten, Copper, Sensitization, Pitting corrosion, Grain boundary engineering

The elastic and thermal transport properties of opaque materials may be measured using transient grating spectroscopy (TGS) by inducing and monitoring periodic excitations in both reflectivity and surface displacement. The “phase grating” response encodes both properties of interest, but complicates quantitative analysis by convolving temperature dynamics with surface displacement dynamics. Thus, thermal transport characteristics are typically determined using the “amplitude grating” response to isolate the surface temperature dynamics. However, this signal character requires absolute heterodyne phase calibration and contains no elastic property information. Here, a method is developed by which phase grating TGS measurements may be consistently analyzed to determine thermal diffusivity with no prior knowledge of the expected properties. To demonstrate this ability, the wavelength-dependent 1D effective thermal diffusivity of pure germanium is measured using this type of response and found to be consistent with theoretical predictions made by solving the Boltzmann transport equation. This ability to determine the elastic and thermal properties from a single set of TGS measurements will be particularly advantageous for new in situ implementations of the technique being used to study dynamic materials systems., United States. National Nuclear Security Administration. Stewardship Science Graduate Fellowship (cooperative Agreement No. DE-NA0002135), SUTD-MIT International Design Centre (IDC), U.S. Nuclear Regulatory Commission (MIT Nuclear Education Faculty Development Program)

Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010., Cataloged from PDF version of thesis., Includes bibliographical references (p. 275-291)., A material that resists lead-bismuth eutectic (LBE) attack and retains its strength at 700°C would be an enabling technology for LBE-cooled reactors. No single alloy currently exists that can economically meet the required performance criteria of high strength and corrosion resistance. A Functionally Graded Composite (FGC) was created with layers engineered to perform these functions. F91 was chosen as the structural layer of the composite for its strength and radiation resistance. Fe-12Cr- 2Si, an alloy developed from previous work in the Fe-Cr-Si system, was chosen as the corrosion-resistant cladding layer because of its chemical similarity to F91 and its superior corrosion resistance in both oxidizing and reducing environments. Fe-12Cr-2Si experienced minimal corrosion due to its self-passivation in oxidizing and reducing environments. Extrapolated corrosion rates are below one micron per year at 700°C. Corrosion of F91 was faster, but predictable and manageable. Diffusion studies showed that 17 microns of the cladding layer will be diffusionally diluted during the three year life of fuel cladding. 33 microns must be accounted for during the sixty year life of coolant piping. 5 cm coolant piping and 6.35 mm fuel cladding were produced on a commercial scale by weld-overlaying Fe-12Cr-2Si onto F91 billets and co-extruding them, followed by pilgering. An ASME certified weld was performed followed by the prescribed quench-and-tempering heat treatment for F91. A minimal heat affected zone was observed, demonstrating field weldability. Finally, corrosion tests were performed on the fabricated FGC at 700°C after completely breaching the cladding in a small area to induce galvanic corrosion at the interface. None was observed. This FGC has significant impacts on LBE reactor design. The increases in outlet temperature and coolant velocity allow a large increase in power density, leading to either a smaller core for the same power rating or more power output for the same s, by Michael Philip Short., Ph.D.