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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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Fracture Toughness Characterization of Generation II FeCrAl Alloys after ~18 dpa Irradiation

FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding applications due to their excellent corrosion resistance to the elevated temperature steam environment. Currently, the handbook on FeCrAl material properties contains only limited data regarding the fracture toughness properties of any FeCrAl alloy. This includes alloys currently under investigation within the Advanced Fuels Campaign (AFC) at Oak Ridge National Laboratory (ORNL). In this project, a series of irradiation capsules have been irradiated in the High Flux Isotope Reactor (HFIR) at ORNL with two Generation II FeCrAl candidate alloys, i.e., C06M and C36M, to assess the fracture response of these alloys after neutron irradiation. These alloys represent the “book-end” compositions for C26M, the alloy currently being developed as the leading candidate for LWR cladding. A total of six irradiation capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. These damage doses represent the expected middle and end of life damage levels for typical LWR cladding while the irradiation temperature regimes will provide insight into the role of varying microstructural features on the fracture toughness properties of neutron irradiated FeCrAl alloys. To date, irradiation of all capsules has been completed in HFIR. This report summarizes the latest results of microhardness and fracture toughness PIE for the 16 dpa capsules (FCAB2, FCAB4, and FCAB6), for which the measured irradiation conditions were: 204°C/17.6dpa, 343°C/18.3dpa, and 507°C/18.6dpa. The main conclusions of this study can be summarized as follows: 1) After the 204°C/17.6dpa irradiation, both C06M and C36M exhibited significant irradiation hardening and embrittlement 2) After the 343°C/18.3dpa irradiation, both C06M and C36M exhibited small irradiation hardening without irradiation embrittlement 3) After the 507°C/18.6dpa irradiation, both C06M and C36M exhibited irradiation softening without irradiation embrittlement 4) Comparing the microhardness and Master Curve reference temperature T 0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M. This should be mainly due to a flat response of the Master Curve reference temperature T 0q to the irradiations at 166-204°C and 315-343°C ranges 5) C06M showed a lower T 0q , meaning better toughness, than C36M at the unirradiated condition and such trend was kept even after neutron irradiation except for the 166-204°C irradiation where both materials had similar T 0q . 6) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.

36 MATERIALS SCIENCE↗
DOE OSTI2024

High Temperature Tensile Testing of Molybdenum Before and After Exposure in Flowing He at 600°C to 1000°C

ORNL supported NorthStar Medical Radioisotope’s accelerator-based concept to produce the medical isotope molybdenum-99 (Mo-99). The Mo targets are expected to experience in the accelerator temperature ranging from 200°C to 1000°C, and previous studies have demonstrated the embrittlement of Mo targets at room temperature when exposed to impure He at 800 to 1000°C. The goal of this project was to evaluate potential embrittlement at temperature ranging from 600°C to 1000°C for three Mo materials exposed at the same temperatures for 100h in high purity He (O 2 < 5ppm (µl/l)) with a flow rate of ~1m/s. Enriched Mo-100 (aMo) and conventional pressed and sintered Mo (UHP) disks were provided by NorthStar while low carbon arc cast rolled Mo, fabricated at ORNL, was used for comparison. The as fabricated rolled Mo exhibited superior tensile strength due to the alloy very fine grain size with limited ductility except at 1000°C with the elongation at rupture reaching 30%. The UHP Mo disks showed the lowest strength but highest ductility at 20-1000°C with significant strain hardening during tensile testing. The ductility of the aMo material was quite low at all temperatures due to a high density of voids at grain boundaries. Significant embrittlement of the aMo material was observed after exposure at 600°C to 1000°C, most likely because of oxygen diffusion and segregation at grain boundaries. No embrittlement was observed for the UHP Mo material, with excellent elongation at rupture for some of the specimens oxidized in He and tested at 600-1000°C. Significant variation in ductility from one UHP Mo specimen to another was, however, measured, both in the as fabricated and He-oxidized conditions. The rolled Mo was not affected by oxidation in He at 800°C, but a decrease in strength and increase in ductility was observed at 1000°C, likely due to the material recrystallization.

36 MATERIALS SCIENCE↗
NASA NTRS1972

Effect of aging at 1040 C (1900 F) on the ductility and structure of a tantalum alloy, T-111

The post-aging embrittlement of T-111 (tantalum - 8-percent tungsten - 2-percent hafnium) following exposure for up to about 10,000 hours at 1040 C in either vacuum or liquid lithium was investigated for sheet and tubing samples. This thermal aging was shown to greatly increase the sensitivity of T-111 to hydrogen embrittlement during subsequent room temperature specimen processing or testing. The hydrogen embrittlement problem can be avoided by preventing exposure to the T-111 to moisture during post-aging processing or testing. Aging at 1040 C also resulted in formation of HfO2 particles at grain boundaries, which may contribute to the observed embrittlement.

Watson, G. K.↗
NASA NTRS1975

Effects of long-term aging on ductility of the columbium alloys C-103, Cb-1Zr, and Cb-752 and the molybdenum alloy Mo-TZM

A program was conducted to determine if aging embrittlement occurs in the columbium alloys C-103, CB-1Zr, and Cb-752 or in the molybdenum alloy Mo-TZM. Results showed that aging embrittlement does not occur in C-103, Cb-1Zr, or Mo-TZM during long-term (1000 hr) aging at temperatures in the range 700 to 1025 C. In contrast, aging embrittlement did occur in the Cb-752 alloy after similar aging at 900 C. A critical combination of the solute additions W and Zr in Cb-752 led to Zr segregation at grain boundaries during long-term aging. This segregation subsequently resulted in embrittlement as indicated by an increase in the ductile-brittle transition temperature from below -1960 C to about -150 C.

Stephens, J. R.↗
NASA NTRS1977

Effects of long-term aging on ductility and microstructure of Cb and Mo alloys

A program was conducted to determine if aging embrittlement occurs in columbium-base alloys C-103, Cb-1Zr, and Cb-752 or in a molybdenum alloy, Mo-TZM. Results showed that aging embrittlement does not occur in C-103, Cb-1Zr, or Mo-TZM during long-term (1000 hour) aging at temperatures in the range of 700-1025 C. In comparison, aging embrittlement occurred in the Cb-752 alloy after similar aging at 900 C. A critical combination of the solute additions W and Zr led to Zr segregation at grain boundaries during long-term aging which subsequently resulted in embrittlement as indicated by an increase in the ductile-brittle transition temperature from below -196 C to about -150 C.

Stephens, J. R.↗
NASA NTRS1979

The combined influence of chemical, metallurgical and mechanical factors on environment assisted cracking

The principal aim of the paper is to re-emphasize and focus on both the multidisciplinary nature of the environment assisted cracking or embrittlement phenomenon. The multiplicity of factors involved in the embrittlement process is indicated, the mutual dependence of these factors and the influences of mechanical and environmental conditions are considered, and the interactions of various factors in determining the overall embrittlement response are discussed. The need for an interdisciplinary approach for resolving the major differences and for understanding embrittlement is outlined.

Williams, D. P., III↗
NASA NTRS1985

Oxidation Behavior of Glassy Alloys

The oxidation behavior of high temperature glassy alloys produced by rapid solidification processing is investigated and the effects of processing and composition on oxidation behavior is studied. Glassy Ta-44.5at%Ir, Ta-40at%Ir-10at%B and Nb-45at%Ir oxidized rapidly at 700 to 800 C at an oxygen partial pressure of .001 atm. The alloys were embrittled during the oxidation process. No apparent oxidation or embrittlement of the Ta-Ir alloy occurred after oxidation for 4h at 500 C at an oxygen partial pressure of .001 atm. Embrittlement occurred, however, after 100h of exposure under the latter conditions. Alloy embrittlement is associated with the partial or full conversion of the metallic glass to a mixture of crystalline beta-Ta2O5 and metallic iridium. Hot compaction of glassy alloys of this type must be limited to relatively low temperatures (approx. 500 C) and short times at the low temperatures unless extremely low oxygen partial pressures can be achieved during the compaction process.

Yurek, G.↗
NASA NTRS1999

Effects of Internal and External Hydrogen on Inconel 718

Internal hydrogen embrittlement (IHE) and hydrogen environment embrittlement (HEE) tensile and bend crack growth tests were performed on Inconel 718. For the IHE tests, the specimens were precharged to approximately 90 ppm hydrogen by exposure to 34.5 MPa H2 at 650 C. The HEE tests were performed in 34.5 MPa H2. Parameters evaluated were test temperature, strain rate for smooth and notch specimen geometries. The strain rate effect was very significant at ambient temperature for both IHE and HEE and decreased with increasing temperatures. For IHE, the strain rate effect was neglible at 260'C, and for HEE the strain rate effect was neglible at 400 C. At low temperatures, IHE was more severe than HEE, and at high temperatures HEE was more severe than IHE with a cross over temperature about 350 C. At 350 C, the equilibrium hydrogen concentration in Inconel 718 is about 50% lower than the hydrogen content of the precharged IHE specimens. Dislocation hydrogen sweeping of surface absorbed hydrogen was the likely transport mechanism for increasing the hydrogen concentration in the HEE tests sufficiently to produce the same degree of embrittlement as that of the more highly hydrogen charged IHE specimens. The main IHE fracture characteristic was formation of large, brittle flat facets, which decreased with increasing test temperature. The IHE fracture matrix surrounding the large facets ranged between brittle fine faceted to microvoid ductility depending upon strain rate, specimen geometry as well as temperature. The HEE fractures were characteristically fine featured, transgranular and brittle with a significant portion forming a "saw tooth" crystallographic pattern. Both IHE and HEE fractures were predominantly along the {1 1 1) slip and twin boundaries. With respect to embrittlement mechanism, it was postulated that dislocation hydrogen sweeping and hydrogen enhanced localized plasticity were active in HEE and IHE for concentrating hydrogen along (1 1 1) slip and twin planes. Final brittle failure occurred by hydrogen induced planer decohesion.

Walter, R. J.↗
NASA NTRS2003

Thermal Contributions to the Degradation of Ground Laboratory and Space Irradiated Teflon

The aluminized Teflon fluorinated ethylene propylene outer most layer of the multilayer insulation blankets on the Hubble Space Telescope has become embrittled resulting in severe on-orbit cracking. During the second servicing mission, a piece of aluminized fluorinated ethylene propylene was retrieved that had curled, exposing the back-surface aluminum to space. This extremely embrittled piece reached 200 C on-orbit, 150 C higher than the nominal temperature extreme. Therefore, experiments have been conducted to determine the effect of heating on the degradation of fluorinated ethylene propylene that has been irradiated in a ground laboratory facility or in space on Hubble Space Telescope. Samples of pristine, x-ray irradiated and Hubble Space Telescope retrieved fluorinated ethylene propylene were heat treated from 50 to 200 C at 25 C intervals in a high vacuum facility and tensile tested. Density measurements were also obtained. Results indicate that heating does not embrittle non-irradiated Teflon. However, there is a significant dependence of the embrittlement of irradiated Teflon on heating temperature, with near complete loss of elongation at 100 C and higher. Rate of degradation changes, which were consistent with the glass transition temperature for fluorinated ethylene propylene, were present in the data. The results support chain scission as the primary mechanism of degradation of fluorinated ethylene propylene on Hubble Space Telescope, and indicate the significance of the on-orbit temperature of fluorinated ethylene propylene with respect to its degradation in the space environment.

deGroh, Kim K.↗
NASA NTRS2000

Insights Developed Into the Damage Mechanism of Teflon FEP Thermal Control Material on the Hubble Space Telescope

Metalized Teflon FEP (DuPont; fluorinated ethylene propylene) thermal control material on the Hubble Space Telescope (HST) has been found to degrade in the space environment. Teflon FEP thermal control blankets retrieved during the first servicing mission were found to be embrittled on solar-facing surfaces and to contain microscopic cracks (the FEP surface is exposed to the space environment). During the second servicing mission, astronauts noticed that the FEP outer layer of the multilayer insulation blanketing covering the telescope was cracked in many locations. Large cracks were observed on the light shield, forward shell, and equipment bays. A tightly curled piece of cracked FEP from the light shield was retrieved during the second mission. This piece was severely embrittled, as witnessed by ground testing. A Failure Review Board was organized by NASA Goddard Space Flight Center to determine the mechanism causing the multilayer insulation degradation. This board included members of the Electro-Physics Branch of the NASA Glenn Research Center at Lewis Field. Density measurements of the retrieved materials obtained under the review board's investigations indicated that FEP from the first servicing mission was essentially unchanged from pristine FEP but that the second servicing mission FEP had increased in density in comparison to pristine FEP (ref. 1). The results were consistent with crystallinity measurements taken using x-ray diffraction and with results from solid-state nuclear magnetic resonance tests (see the table and ref. 1). Because the second servicing mission FEP was embrittled and its density and crystallinity had increased in comparison to pristine FEP, board researchers expected that the first servicing mission FEP, which was also embrittled, would also have increased in crystallinity and density, but it did not. Because the retrieved second servicing mission material curled while in space, it experienced a higher temperature extreme during thermal cycling (estimated at 200 C) than the first servicing mission material (estimated at 50 C). Therefore, Glenn initiated and conducted an investigation of the effects of heating pristine FEP and FEP that had been exposed on the Hubble Space Telescope. Samples of pristine and first and second servicing mission FEP were heated to 200 C and evaluated for changes in density and morphology. We hoped that the results would help explain why FEP degrades in the Hubble Space Telescope space environment.

deGroh, Kim K.↗
DOE OSTI2023

Machine Learning Prediction of Fracture Toughness in Hydrogen-charged Stainless Steels

Austenitic stainless steels are structural materials utilized in tritium gas pressure boundaries since they are resistant to hydrogen isotope embrittlement [1-3]. However, exposure to tritium over long periods of time leads to tritium uptake which decays to result in helium ingrowth. This helium ingrowth results in further embrittlement effects which are synergistic with that from the hydrogen isotope [4]. Therefore, it is important for tritium facilities to understand the material limitations of stainless steel in this environment. The Savannah River National Laboratory (SRNL) has available a large experimental data set of austenitic stainless steels which have been exposed to tritium environments for various lengths of time. With the availability of this data set, machine learning (ML) algorithms provide an opportunity to model the embrittlement of stainless steel due to the algorithm’s ability to identify patterns in data sets that are difficult and costly to identify in other manners [5]. Ultimately, the amount and quality of the available data is one defining force in the ability of a ML model to accurately predict the desired outputs. The models developed herein will illustrate the ability for the various algorithms to predict the change in fracture toughness in stainless steels due to hydrogen-isotope embrittlement.

Hoar, Eric T.↗
DOE OSTI2025

Direct observation of strain-enhanced hydrogen segregation and failure at high-angle grain boundaries in nickel

Understanding the mechanisms underlying hydrogen embrittlement remains difficult, even in single-element metals. Both microstructure and stress state influence hydrogen distribution in metals and alloys, which impacts deformation and failure. Here, in this work, we use in-situ Kelvin probe force microscopy (KPFM) to monitor the hydrogen distribution in pure nickel over time at 1.1 % and 3.5 % strain. The sample strained to 3.5 % results in preferential hydrogen segregation to high-angle grain boundaries whereas the sample strained to 1.1 % does not exhibit preferential hydrogen segregation. Optical digital image correlation (DIC) shows that hydrogen charging results in both localized and reduced strains during tensile testing of a notched sample. Later stages of deformation and failure (e.g. microcracking) are studied by using in-situ transmission X-ray microscopy (TXM). TXM reveals nanoscale structural changes to a propagating crack in a hydrogen environment. Localized void growth and secondary cracking occur at grain boundaries near the primary crack front. Correlative electron backscattering diffraction (EBSD) is used to relate the cracking at grain boundaries to the hydrogen segregation observed in KPFM. These findings are unified in a proposed hydrogen embrittlement mechanism that describes the interaction of hydrogen with grain boundaries, and the role of grain boundaries in hydrogen embrittlement.

Grain boundary fracture↗
DOE OSTI2025

Ductility evaluation of nitrided Zircaloy-4 cladding by ring compression testing

Here, this study evaluated the embrittlement of Zircaloy-4 cladding by dry air and dry nitrogen environments as a function of process temperature and duration of exposure. A Netzsch thermogravimetric analyzer was used to measure the mass gain of cladding sections over time with specified temperature and gas flow programs. Oxidation of Zircaloy in dry air produced a brittle ZrO 2 layer with ductile unreacted metal. In contrast, nitridation of Zircaloy had a slower reaction rate but produced a brittle cladding that was more easily breached during ring compression tests. The ring compression tests demonstrated that ductility of Zircaloy cladding decreases as nitrogen pickup increases; a ductile-to-brittle transition occurs near a mass gain of 2.1 wt %. Nitridation showed minimal temperature dependency in ductility degradation, whereas oxidation showed significant temperature dependency in ductility degradation, with severe embrittlement occurring at 1200 °C. Optical micrographs confirmed the formation of uniform ZrN layers (approximately 12 µm thick) on the cladding surface, indicating that cladding embrittlement is primarily the result of nitrogen diffusion rather than metal layer thinning.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Harvesting Reactor Pressure Vessel Beltline Material from the Decommissioned Zion Nuclear Power Plant Unit 1

The decommissioning of the Zion Nuclear Power Plant (NPP) provided a unique opportunity to harvest and study service-aged reactor pressure vessel (RPV) beltline materials. This work, conducted through the U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) Program, aims to improve the understanding of radiation-induced embrittlement to support extended nuclear plant operations. Material segments containing the Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B Heat B7835-1 base metal, obtained from the intermediate shell region with a peak fluence of 0.7 × 10 19 n/cm 2 (E > 1.0 MeV), were extracted, cut into blocks, and machined into test specimens for mechanical and microstructural characterization. The segmentation process involved oxy-propane torch-cutting, followed by precision machining using wire saws and electrical discharge machining (EDM). A chemical composition analysis confirmed the expected variations in alloying elements, with copper levels being notably higher in the weld metal. The harvested specimens enable a detailed evaluation of through-wall embrittlement gradients, a comparison with the existing surveillance data, and the validation of predictive embrittlement models. This study provides critical data for assessing long-term reactor vessel integrity, informing aging-management strategies, and supporting regulatory decisions to extend the life of nuclear plants. This article is a revised and expanded version of a paper entitled, “Current Status of the Characterization of RPV Materials Harvested from the Decommissioned Zion Unit 1 Nuclear Power Plant”, PVP2017-65090, which was accepted and presented at the ASME 2017 Pressure Vessels and Piping Conference, Waikoloa, HI, USA, 16–20 July 2017.

harvesting beltline material↗
DOE OSTI2021

Enhanced work hardening from oxygen-stabilized ω precipitates in an aged metastable β Ti-Nb alloy

High levels of oxygen in solid solution in Ti alloys are considered detrimental to mechanical properties because of embrittlement concerns. In metastable β titanium alloys, the formation of isothermal ω precipitates is also known to cause severe embrittlement and ductility reduction. However, oxygen has been shown to partition to the ω phase during ageing, and this partitioning behavior may potentially impact ω’s mechanical contribution. Using micropillar compression, we compared the deformation behavior of Ti-20Nb (at. %) with oxygen-stabilized ω precipitates to the behavior of oxygen-free specimens. The oxygen-stabilized microstructures showed increased compressive yield strength and enhanced work hardening behavior compared to oxygen-free specimens. In the absence of oxygen, the compressed pillars showed slip band formation and catastrophic failure, and transmission electron microscopy imaging revealed that ω precipitates were sheared within the continuous deformation channels resulting in slip localization. In contrast, oxygen-stabilized ω precipitates were harder to shear and the formation of continuous deformation channels was suppressed during compression, leading to improved work hardening behavior up to 15% strain. Furthermore, this counter-intuitive role of oxygen may offer design strategies to address the significant embrittlement and loss of ductility observed for ω-strengthened β Ti alloys without oxygen and avenues to expand the use of β Ti alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2026

3D probabilistic fracture mechanics / computational fluid dynamics simulation of a reactor pressure vessel under transient conditions

Reactor pressure vessels (RPVs) are safety-critical light-water-reactor components that, under irradiation, experience long-term material degradation in the form of embrittlement. This can increase their susceptibility to fracture under thermal-shock conditions, which could occur during off-normal transients such as loss-of-coolant accidents (LOCAs). During a LOCA, the most severe conditions for the RPV occur when emergency core cooling water is injected through the cold legs into the water-and-steam-filled RPV. The rapid cooling of the downcomer and internal RPV surface causes decreased temperature and elevated thermally driven tensile stresses in the RPV wall. This, combined with long-term material embrittlement, may cause fracture initiation at pre-existing flaws, challenging the integrity of the RPV. Assessing RPV integrity during transients with a large spatial variation in the coolant temperature requires a modeling approach that considers the effects of spatially varying coolant temperature on the fracture probability of a population of flaws distributed throughout the RPV, accounting for spatially varying embrittlement. Here, the present study addresses this need by demonstrating first-of-its-kind coupling of a high-fidelity 3D computational fluid dynamics code with 3D probabilistic fracture mechanics This was accomplished using representative models of a pressurized-water reactor subjected to small- and medium-break LOCA conditions, both of which can result in large spatial temperature variations. While the observed impact of accounting for 3D effects was minimal under the small-break LOCA this study indicates a significant increase in the probability of fracture initiation under the medium-break LOCA when 3D effects are considered, relative to a spatially uniform cooling scenario.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

One dislocation at a time

The direct observation of enhanced dislocation mobility in iron by in situ electron microscopy offers key insights and adds to the ongoing debate on the mechanisms of hydrogen embrittlement. The tiniest atom in nature, hydrogen, is pervasive and sneaky. When hydrogen atoms, either from gas or liquid sources, come into contact with a metal, they rush into the solid and squeeze through the narrow spaces between the host atoms. Being chemically active, atomic hydrogen can potentially bind to either host or impurity atoms, forming hydrides and thus changing a material’s properties. Metals, which are well known and widely used in industry for their strength and ductility, can become dangerously brittle when exposed to hydrogen. Furthermore, this phenomenon is known as hydrogen embrittlement and can lead to catastrophic failure of a load-bearing part. For example, the hydrogen produced when water molecules break apart in coolants in a nuclear reactor can cause sudden failure of the pressure vessel. In another example, although blending hydrogen in natural-gas pipelines provides a promising pathway to transition into the hydrogen economy, it could lower the fatigue resistance of the pipeline steel, making it more susceptible to crack growth due to cyclic loading induced by pressure fluctuations in the pipeline. Remarkably, perhaps as a result of a multitude of recognized failure mechanisms in which hydrogen can potentially participate, it still remains unknown exactly why hydrogen makes a metal brittle. Now, writing in Nature Materials, Longchao Huang and colleagues present unambiguous experimental observations of hydrogen-enhanced dislocation mobility in iron, offering key insights into the hydrogen embrittlement mechanisms.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Characterization and Quantification of Radiation-Induced Clusters/Precipitates in RPV Steels Using STEM-EDS and Machine Learning

Over the operational lifespan of a nuclear reactor, reactor pressure vessel (RPV) steels are subjected to significant neutron irradiation, resulting in complex microstructural changes and the consequent degradation of mechanical properties. Various physically motivated correlation models have been developed to predict neutron irradiation-induced embrittlement of RPVs under different irradiation conditions. However, the efficient and accurate characterizations and quantification of radiation-induced clusters in RPVs are still challenging, which will affect the precision of the predictive models for embrittlement of RPV components. In the DOE Visiting Faculty Program (VFP) research work at Oak Ridge National Lab (ORNL), I integrate machine learning to aid Scanning Transmission Electron Microscopy – Energy Dispersive X-ray Spectroscopy (STEM-EDS) analyses, which improve the characterization and quantification of radiation-induced clusters in RPV steels, thereby enabling more accurate predictions of material behavior under irradiation. The surveillance base- and welded- RPV steels were annealed at various temperatures of 340 °C, 450 °C and 500 °C for up to 168 hours, respectively. Afterwards, I have characterized radiation-induced clusters using advanced STEM-EDS techniques and subsequently applying machine learning algorithms to analyze and refine STEM-EDS datasets, enhancing the quantification of clusters compositions and distributions. In the end, an efficient workflow for integrating STEM-EDS data analysis with machine learning to address challenges including noise reduction has been developed. The completion of this VFP work will support bridge critical gaps in the accurate quantification of radiation-induced clusters in RPV steels using STEM-EDS and support the development of more precise models for predicting RPV embrittlement in the Light Water Reactor Sustainability program supported by Department of Energy and enhancing the collaboration between ORNL and Alred University. The outcome of the VFP project will leverage a few research papers submission to peer-reviewed journals in the relevant scientific field and a few oral presentations at national and international conferences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗