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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

AGR-2 Compact 6-4-1 Post-Irradiation Examination Results

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety/heating testing, fuel performance modeling, and fission product transport (INL 2017). Several fuel irradiation experiments have been performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), and additional irradiations are planned. These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3T m ) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300°C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≃100) relative to the Low N specimen (ΔHV≃160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Accelerated radiation tolerance testing of Ti-based MAX phases

MAX phases have recently attracted significant attention for potential nuclear applications due to their novel properties such as unique hexagonal-compact nanolayered crystal structure, high-machinability due to lower hardness levels compared to conventional ceramics, and high-chemical inertness. In order for MAX phases to be used in nuclear reactors, two aspects deserve detailed investigations: (i) their phase stability at high-temperatures and (ii) microstructural defect formation and recovery induced by energetic particle irradiation. To date, degradation mechanisms of MAX phases at high-temperatures and following irradiation are largely unexplored fields of research. This work focuses on the evaluation of two Ti-based MAX phases – Ti 2 AlC and Ti 3 SiC 2 – within the context of extreme environments. To accomplish this, a one-of-a-kind comparison between neutron irradiations, performed over a decade of research at the High Flux Isotope Reactor (HFIR), and heavy-ion irradiations, carried out in situ in a Transmission Electron Microscope (TEM) at the MIAMI-2 facility, has been conducted. The results show Ti-based MAX phases are prone to accelerated decomposition under the conditions investigated. This questions the hypothesis that MAX phases exhibit high phase stability, especially when used in future nuclear energy systems where energetic particle irradiation is a dominating degradation mechanism.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Field-Scale Testing of a High-Efficiency Membrane Reactor (MR)—Adsorptive Reactor (AR) Process for H2 Generation and Pre-Combustion CO2 Capture

The study objective was to field-validate the technical feasibility of a membrane- and adsorption-enhanced water gas shift reaction process employing a carbon molecular sieve membrane (CMSM)-based membrane reactor (MR) followed by an adsorptive reactor (AR) for pre-combustion CO2 capture. The project was carried out in two different phases. In Phase I, the field-scale experimental MR-AR system was designed and constructed, the membranes, and adsorbents were prepared, and the unit was tested with simulated syngas to validate functionality. In Phase II, the unit was installed at the test site, field-tested using real syngas, and a technoeconomic analysis (TEA) of the technology was completed. All project milestones were met. Specifically, (i) high-performance CMSMs were prepared meeting the target H2 permeance (>1 m3/(m2.hbar) and H2/CO selectivity of >80 at temperatures of up to 300 °C and pressures of up to 25 bar with a <10% performance decline over the testing period; (ii) pelletized adsorbents were prepared for use in relevant conditions (250 °C < T < 450 °C, pressures up to 25 bar) with a working capacity of >2.5 wt.% and an attrition rate of <0.2; (iii) TEA showed that the MR-AR technology met the CO2 capture goals of 95% CO2 purity at a cost of electricity (COE) 30% less than baseline approaches.

42 ENGINEERING↗
DOE OSTI2023

Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Lach, Tim↗
DOE OSTI2025

Conceptual Designs for Irradiation Creep Testing of SiC in HFIR

Understanding irradiation creep of nuclear fuel cladding is important to properly size the initial fuel-cladding gap and understand when pellet-cladding contact is expected to occur due to a combination of fuel swelling and cladding creep-down. Irradiation creep also plays a role in relaxing stresses that develop in-pile. Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are the leading long-term accident-tolerant fuel cladding concept for light-water reactors (LWRs). Although some limited data are available regarding irradiation creep of the individual constituents (fibers, matrix), data regarding irradiation creep of SiC/SiC composites are currently insufficient. Additional data regarding irradiation creep compliance and the rupture lifetime (combination of creep and slow crack growth) are needed to understand material limitations. This work describes the design and development of two irradiation vehicles that are being pursued for testing SiC/SiC concepts in the High Flux Isotope Reactor (HFIR). The first is a passive experiment, referred to as the PRECISE experiment, that leverages the constant coolant pressure of HFIR to compress a metallic bellows and provide a well-characterized load to drive creep in a SiC/SiC dog bone specimen. The total creep strain would be quantified post-irradiation by measuring dimensional changes of the specimen length as well as local dimensional changes within the gauge region. Non-stressed specimens would also be irradiated under the same conditions to provide an indication of dimensional changes due to radiation-induced swelling in the absence of creep. A second, more complex experiment, referred to as the INSITE experiment, is being designed in parallel that would use pneumatics to pressurize a metal bellows and linear variable differential transformers (LVDTs) to measure the specimen displacement in situ during irradiation. Such an experiment would provide significantly more data regarding the evolution of the creep compliance as a function of dose and applied stress within a single experiment but would require significantly more development time and cost to execute. The primary concern with the INSITE experiment is the accuracy, reliability, and expected lifetime of the LVDTs during irradiation at elevated temperatures. Efforts are being made to adjust the experiment design and operating procedure to limit LVDT temperatures and mitigate or otherwise compensate for uncertainties due to factors such as temperature fluctuations, creep in the surrounding structural materials, and drift of the LVDTs. This work describes the experiment designs, thermal and structural analysis that were performed to ensure that the desired temperature and stress conditions can be achieved, some initial sensitivity analyses to predict the evolution of the radiation-induced specimen displacements, and potential sources of uncertainty in the measurements. Out-of-pile testing is being performed in parallel to confirm that the test trains achieve the expected stress states in the specimens and do not result in prohibitive stress concentrators (e.g., in the grip regions) that might risk pre-mature failure. The PRECISE experiments are proceeding toward fabrication and assembly with HFIR insertion planned during fiscal year 2026. The INSITE experiment is progressing toward out-of-pile demonstrations, which will provide more conclusive evidence regarding the feasibility of executing these tests in HFIR or whether alternative displacement monitoring techniques may need to be considered.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Role of Mg:Al ratio and Pt promotion on mitigating the deactivation of Ni-based methane steam reforming catalysts

Ni-based catalysts for CH 4 steam reforming require fine-tuning to withstand deactivation under dynamic operation conditions relevant to emerging H 2 -driven technologies. This study investigates the impact of the Mg:Al ratio and Pt presence in catalyst composition on the activity and stability of industrially relevant Ni-based mono- and bimetallic catalysts during simulated daily start-up and shut-down cycles in various gas atmospheres. The evolution of the catalyst structure during extensive testing procedures was investigated in detail by complementary electron microscopy, in situ/operando XAS and XRD. The results obtained revealed that catalyst deactivation is promoted at high Mg:Al ratios and especially affects the monometallic catalysts. By converting CH 4 at lower temperatures, Pt regulates the extent of Ni oxidation and its incorporation into MgO lattice. Further prevention of catalyst deactivation was achieved by optimizing the reactor shut-down procedure to minimize the simultaneous exposure to high temperatures and H 2 O vapors. By flushing the reactor with N 2 only around the reaction extinction temperature, a fraction of Ni species is maintained in metallic state, which is beneficial for the long-term activity and reaction operation economy.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Amine Oxidation Catalyzed by NO 2

Amine oxidation is a major risk in the deployment of amine scrubbing for CO2 capture from post-combustion flue gas. Oxidation probably occurs by dissolved oxygen at elevated temperature, Fe(III)/Fe(II) shuttling between the absorber and the stripper, and by NO2 in the absorber. Amine selection, dissolved oxygen removal, and solvent reclaiming mitigate oxidation due to dissolved oxygen and iron shuttle mechanism, but not by NO2. This work examines piperazine (PZ) oxidation by NO2 in a bench-scale high gas flow reactor (HGF). PZ is an effective second-generation solvent with high CO2 capacity, fast capture rate, good thermal stability, and good resistance to oxidation. Flue gas typically contains 0.5–5 ppm NO2 and 10–100 ppm NO. NO2 at 1 ppm in the flue gas appears to cause significant amine oxidation in pilot plant testing. PZ solution can easily absorb NO2, and it is hypothesized that NO2 can catalytically oxidize PZ through free-radical propagation. The issue may not be solved by a NO2 pre-scrubber since the coexistence of NO and O2 in the flue gas will keep producing non-negligible amounts of NO2 in the ductwork and the absorber after the NO2 incoming with the flue gas has been scrubbed. In addition, NO2 may not only oxidize PZ but also the intermediate degradation products, and it is unknown if the existence of dissolved iron influences the oxidation by NO2. Previous workers have also shown that NO2 is the stoichiometric source of nitrosamine. A comprehensive analysis of NO2 oxidation is needed to address solvent management and its health/environmental impact. The new high gas flow (HGF) reactor focuses only on the absorber environment and is simpler to interpret because it does not include oxidation at high temperatures and by dissolved metal catalysts from corrosion. This paper presents the results of HGF experiments and shows evidence to support a hypothesis of amine oxidation catalyzed by NO2, presents the synergic effect of NO2 and iron on amine oxidation, points out the different effects of NO2 in clean and degraded solvents, and demonstrates that NO2 not only can oxidize amines but also their degradation products.

NO2, iron, piperazine (PZ), solvent oxidation, sol↗
DOE OSTI2022

Summary of Geometric Parameters and Their Effects on Performance of U-10Mo Fuel Plates

A monolithic plate-type fuel system has been under development to convert high-performance test reactors from highly enriched uranium to low-enrichment uranium fuels and is now moving into the qualification phase, a predecessor to the timely conversion of the target reactors. To qualify this fuel system, the plates must meet safety standards and perform well in a reactor. The plates must maintain mechanical integrity, exhibit geometric stability, and have stable and predictable in-reactor behavior. The requirement to maintain mechanical integrity under normal operating conditions is primarily demonstrated by successful testing. However, each high-performance reactor employs a distinct design, resulting in distinct plate geometries, with unique features, attributes, irregularities, and tolerances. Due to the abundance of such distinct geometric varieties, a single “generic” plate geometry capturing all extremes is not achievable. It is also impractical to test each of these proposed designs in a reactor. This limitation necessitates cautious evaluations since the thermomechanical response of a plate with a certain geometry may not be representative for a plate with a significantly different geometry. To address concerns related to in-reactor plate performance, large set of sensitivity studies were performed. These parametric studies aimed to better understand irradiation performance, while evaluating the sensitivity of results to various modeling inputs, including geometric, operational, and material parameters. This work studied selected geometric parameters based on provided fuel specifications and performed a series of parametric simulations. The resulting temperature, displacement, and stress strains were comparatively evaluated to determine the effects of various geometric parameters. This draft provides a “high-level summary” of our parametric sensitivity studies and summarizes the key findings.

42 ENGINEERING↗
DOE OSTI2024

Single Primary Heat Extraction and Removal Emulator (SPHERE) Long Duration Testing

For heat-pipe cooled microreactor development, it is essential to understand the characteristics of heat pipes and how they function under a wide range of operating conditions. An important process for heat-pipe cooled microreactor development is the passive heat removal and how the performance may change over long periods of time. Increased experimental data provides additional information to assess the operational lifetime of alkali metal heat pipes. Idaho National Laboratory has completed testing over a long duration of a high-performance sodium filled heat pipe while monitoring axial temperature profile, power supplied by the heaters, and heat removed by a gas-gap calorimeter. The results from this testing can aid in heat pipe validation efforts.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Development of a Light-Trapping, Planar-Cavity Receiver for Enclosed Solar Particle Heating

Concentrating solar thermal power (CSP) technology development has recently focused on high efficiency power cycles and chemical reactions that require high operating temperatures. An increase in receiver operating temperatures relative to current commercial CSP technology is necessary to support more efficient, high temperature power cycles or thermochemical processes. Solar receivers operating with heat transfer media temperatures >700 degrees C face various challenges including thermal performance, scalability, thermal-mechanical issues, and receiver service life. In addition, heat transfer media selection for high-temperature processes has been moving towards inert solid particles in Gen3 CSP, and reactive solid media and/or gases in solar thermochemical processes. To overcome the shortcomings and limitations of conventional designs for molten salt or particle receivers and receiver reactors, we introduce a unique light-trapping, planar cavity receiver (LTPCR) configuration and will present the current development progresses to demonstrate receiver feasibility through modeling, testing, and prototype demonstration.

ENGINEERING,SOLAR ENERGY↗
DOE OSTI2025

Friction and wear of nuclear graphite exposed to molten FLiNaK salt environment

This report formally documents the completion of the Advanced Reactor Technologies Level 3 Milestone (M3TG-25OR0501104), “Initiate graphite wear studies of graphite samples exposed to molten salt environment,” due August 1, 2025. The report summarizes the ongoing activities aimed at characterizing the friction and wear behavior of graphite in molten LiF–NaF–KF (FLiNaK) salt environment. The wear and friction behavior of self-mated ET-10 nuclear graphite were tested at varying temperatures (550°C and 650°C), contact loads (40 and 80 N) and sliding speeds (1 and 10 mm/s) in a controlled argon environment. The results were compared to the wear and friction behavior of the same graphite material in a dry argon environment. This report presents initial studies on the friction and wear behavior of graphite in sliding contact mode, with future work focused on investigating the graphite matrix materials, as well as rolling and impact contact modes. The outcomes of this report could provide insights into assessing the integrity of the graphitic components of pebble-bed fluoride salt–cooled high-temperature reactors (PB-FHRs) to enhance their safe operation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Development of Surveillance Test Articles with Reduced Dimensions and Material Volumes to Support MSR Materials Degradation Management

This report details the efforts toward developing new surveillance test article designs with reduced dimensions and material volumes to support materials surveillance technology development for advanced reactors. Two fabrication methods for the surveillance test articles are described. Welded test articles were fabricated with 316H and A617 materials, and interlocking test articles were fabricated using A709 and titanium-zirconium-molybdenum (TZM). The preliminary results demonstrate the successful design and testing of the flat surveillance test articles. The report also describes ongoing efforts to use an induction heating test setup to increase the heat up and cool down rates in testing the surveillance test articles. A brief description of the planned FY-24 work is provided.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

U-gas +Cool GTL™ - A New Integrated Process for Direct Biomass Conversion to Liquid Fuels

The U-Gas gasification technology, marketed by Sungas, is a proven fluid bed gasification technology for biomass conversion. U-Gas has been tested on a wide variety of biomass feeds including wood and waste streams. The U-Gas Technology has successfully converted wood for more than 20 years in a gasification plant in Skive Denmark at the 150t/d size. GTI has also recently developed the Cool GTL process for converting biogas to high quality gasoline, jet, and diesel fuels. The technoeconomics and LCA for the combination of the two technologies results an integrated optimized process, which will be discussed. The Cool GTL process is still in development and the development should be complete by 2023. Cool GTL converts high CO2 content feeds to synthesis gas with a 2-2.4/1 H2/CO ratio suitable for fuels synthesis. The process is based on a novel, highly stabile CO2/steam reforming catalyst. The 2.0-2.4/1 H2/CO synthesis gas ratio is always achieved, independently of feed CO2 content, by adjusting the amount of water added according to the amount of CO2 in the feed. The reformer catalyst has been tested for 500 hours and shows no measurable deactivation. The product from the reformer then goes directly to a low temperature slurry bed Fischer Tropsch reactor with an integrated finishing hydroisomerization reactor which makes hydrocarbon liquids from the synthesis gas. With this configuration no wax is produced. The technoeconomics and carbon intensity for the integration of UGas and the Cool GTL processes will be compared to other biomass conversion approaches. The use of electrolysis to boost H2 in the feed and increase liquid production will also be analyzed as a potential process improvement with sensitivities to the cost of electricity and the use of renewable electricity.

09 BIOMASS FUELS↗
DOE OSTI2024

REBCO Insert Coils toward High-Field, Large-Bore Magnet for Quantum Physics Research

High-temperature superconducting (HTS) magnets offer a promising solution for generating high magnetic fields efficiently and economically, serving as a vital component in nextgeneration scientific instruments and carbon-neutral power systems, notably in cost-effective compact fusion reactors. These high magnetic fields play a pivotal role in advancing research on quantum materials, particularly in elucidating the intricate electronic states near magnetic phase transitions. Here, this paper delves into the research conducted at the Princeton Plasma Physics Laboratory, which is divided into three phases aimed at overcoming technical hurdles in creating large-bore, high-field HTS magnets for cutting-edge quantum physics research. In Phase 1, we designed, constructed, and tested a compact REBCO solenoidal magnet comprising six double-pancake coils featuring a 41.3-mm inner diameter and a 70.0-mm outer diameter. It adopts a no-insulation approach to ensure electrical and thermal stability. Successful testing in a saturated liquid nitrogen bath confirmed its capability to generate a 0.8-T field at 77 K and 2 T at 65 K, with ongoing integration into liquid helium testing. Building on this experience, Phase 2 involves the design of an HTS insert coil, intended to nest within a 12-T low-temperature superconducting (LTS) outsert to achieve a minimum of 20 T at 4.2 K. The focus of Phase 2 centers on addressing challenges posed by screening-current (SC) effects, particularly the associated stress/strain issues. Employing a numerical model that fully couples the SC and mechanical analysis, we discuss strategies for managing stress to mitigate the effects of SCinduced stress concentrations.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2022

High-performance and high-fidelity Monte Carlo solutions to the BEAVRS benchmark

The BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulation) benchmark is solved by PRAGMA, the GPU-based continuous energy Monte Carlo code. The resulting solutions are comprised of the detailed simulation results of two cycles, each of which consists of the zero power physics test (ZPPT) and the core depletion calculations. The ZPPT consists of characteristic parameters, such as critical boron concentration (CBC), control rod bank worth, isothermal temperature coefficients, and assembly-wise detector signal, which are compared with measured data provided by the benchmark administration. The core depletion calculations were performed for both the hot full power and the load follow modes, and the comparison was made with the measured or deduced CBCs and assembly-wise detector signals. In the load follow calculations, the operating power history was approximately applied to simulate the real operation as closely as possible. PRAGMA performed the various calculations with a tremendous number of histories ranging up to hundreds of millions per cycle, exploiting GPUs' massively parallel performance. The load follow run time was shorter than 16 hours on a single rack of computing nodes mounded with 24 gaming GPUs with a remarkable agreement with the measurements for most comparisons. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

A methodology for corrosion testing ODS steels in liquid tin under reactor conditions in HFIR

Liquid metal–based divertor concepts have promising attributes for application in fusion reactors. Liquid tin is being considered as one of those liquid metals due to its excellent thermophysical properties. However, tin is extremely corrosive to steels at elevated temperatures, and the coupled effects of corrosion with neutron irradiation for application in fusion reactors have not been quantified. Researchers at Oak Ridge National Laboratory (ORNL) have designed and irradiated a series of capsules in the High Flux Isotope Reactor (HFIR) to help address gaps in the literature with respect to coupled corrosion and neutron irradiation effects. The capsules are filled with solid tin designed to melt due to the gamma heating in HFIR and allow interaction with the specimens. The capsules contain a SiC thermometer for post-irradiation temperature verification. Five capsules were irradiated in HFIR for 10.5 days, accumulating 0.93 dpa in the FeCrAl specimens. Post-irradiation temperature verification was performed on the SiC thermometers, and temperatures were observed to be higher than what was expected from thermal models. An explanation for the model underprediction is given herein. In conclusion, the capsules described within this article are the first to irradiate liquid metal within the HFIR flux trap and demonstrated the safety basis for continued liquid coolant studies in HFIR.

FeCrAl↗
DOE OSTI2024

Analysis of Deformation and Fracture Mechanisms in the Harvested High-Dose Baffle-Former Bolt with Stress-Corrosion Cracks Formed While in Service

This report presents the results of advanced mechanical testing conducted on miniature tensile specimens excised from an irradiated baffle-former bolt, a commercial pressurized water reactor component. The specimens were extracted from the midsection of the bolt shank, where the estimated damage dose reached 23 displacements per atom. The mechanical testing included the following tests: (1) conventional tensile testing at room temperature, augmented by digital image correlation (DIC) to enable noncontact strain measurements, and (2) in situ tensile testing within a scanning electron microscopy (SEM) instrument equipped with energy-dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD) detectors to evaluate active deformation and fracture mechanisms. Additionally, SEM fractographic analysis revealed alterations in the fracture mechanism from a predominantly ductile fracture in nonirradiated steel to a mixed fracture mode (still dominating ductile and minor cleavage spots) in irradiated specimens derived from the baffle bolts. The findings indicate a complex strain localization behavior for in-service irradiated steel specimens. Beyond conventional necking at the macro scale and defect-free channel formation at the micro scale, DIC results identified the presence of deformation bands approximately 100 μm in width. These bands consist of chains of grains exhibiting elevated local strain levels and may be considered an intermediate or mesoscale level of strain localization. These bands become discernible near the yield stress as localized hot spots, areas of elevated strain, persisting throughout most of the experiment. The formation of mesoscale deformation bands as a strain localization mechanism may exacerbate the defect-free channel formation in irradiated materials, further promoting irradiation-assisted stress corrosion crack initiation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗