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Chen, Wei-Ying

Publications and source records attributed to Chen, Wei-Ying.

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Progress Report of Alloy 709 Performance in Liquid Sodium

With improved high-temperature properties, Alloy 709 shows great potential for advanced reactor applications. To understand the effects of sodium on A709, small tensile specimens were machined from two commercial heats and subject to sodium exposure in two forced convection loops operated by Argonne. Thermal aging experiments were also conducted in parallel to isolate any sodium-specific effects. An increasing trend in yield strength and a decrease trend in ultimate tensile strength can be seen with the increasing exposure time, suggesting microstructural changes at these temperatures. Nonetheless, the thermally aged and sodium exposed samples behaved similarly, implying no sodium-specific effect at these temperatures. In contrast, 316H SS exposed to sodium at 650°C showed a strong sodium effect. The different responses of A709 and 316H SS to sodium exposure is of interest and requires additional effort to understand. The cold traps of the sodium loop were replaced early in FY23. During a maintenance operation, an unexpected sodium leak incident occurred at one of the sodium loops, triggering an emergency response. An investigation into the incident revealed that the leak happened at a VCR fitting on the recently replaced cold trap. The metal gasket of the leaked VCR fitting was not compressed sufficiently. A dimensional mismatch between the new component and old piping configuration was responsible for a significant misalignment. A new cold trap with improved design is in production. The new design provides additional flexibility and an access point for post-installation helium leak check and for venting of trapped gas. These design features will make the system more error-tolerant and the installation quality verifiable. A thermodynamic analysis was performed to understand the carburization-decarburization behavior of A709 in sodium environment. The equilibrium carbon activity in the alloy was evaluated with M 23 C 6 , TiC, and NbC. The carburization-decarburization boundary was calculated with different methods developed previously. The results suggests that A709 will be carburized in SMT-1 and SMT-2, and can be decarburized in low-carbon-activity environments such as EBR-II at temperatures above 620-650°C. Nonetheless, A709 should be less likely to be decarburized than 316H SS when exposed to the same low-carbon-activity environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Effect of thermal aging on microstructure and stress corrosion cracking behavior of Alloy 152 weldment

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray conducted in small, 0.2 mm - step line scans in the high-deformation regions of the weld root – covering areas spanning from the weld heat affected zone (HAZ) in Alloy 690 to the weld and weld butter on LAS - did not show evidence of LRO in any of the three Alloy 152 heats aged to an equivalent of 60 years of service. Testing in a primary water environment of two heats of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloys. However, the SCC CGR response of the aged samples appears to show a deterioration in performance, confirming our previous observation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In-situ and ex-situ characterization of ion-irradiated AM materials

Additive manufacturing (AM) has attracted increasing attention in recent years as a new way of making high-quality components for nuclear reactors. While AM materials are compositionally similar to their conventionally produced counterparts, they do possess different microstructures, such as dislocation cells and chemical inhomogeneity, that can lead to different mechanical properties and performance behavior. In this study, the irradiation response of AM materials was investigated. In-situ and ex-situ ion irradiations were performed on AM316L and AM316H stainless steels (SS) at 300 and 600°C. The influence of the dislocation cell structure on the evolution of irradiation-induced dislocation loops was evident at 600ºC, but was much weaker at 300ºC. No voids were observed with the in-situ ion irradiation up to 10 dpa at both temperatures. Post-irradiation energy dispersive spectroscopy showed radiation-induced segregation (RIS) near grain boundaries and the formation of Cr-rich oxides throughout the matrix. The extent of segregation at dislocation cell walls varies with dose. Nanoindentation tests performed on the AM316L SS irradiated at 600ºC showed a complex dose dependence with softening at low doses and hardening at high doses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

In-situ TEM investigation of void swelling in nickel under irradiation with analysis aided by computer vision

Understanding the stability of irradiation-induced voids in materials is important for engineering material's swelling behavior under irradiation. In-situ TEM offers a spatial and temporal resolution that is suitable for investigating the evolution of voids under irradiation. However, the in-situ videos have often been too large to be analyzed manually, leaving the valuable data underutilized. For this work, we developed a deep learning-based semantic segmentation model to consistently study the growth and shrinkage of voids in nickel under 1 MeV krypton ion irradiation at various temperatures from 525 °C to 650 °C. With a foil thickness near 100 nm and ion flux of 6.3 × 10 11 ions∙ cm –2 ∙s –1 , the pre-existing voids, which were created beforehand by irradiation at 600 °C to 0.5 dpa, shrank at low temperatures and grew at high temperatures under irradiation, where the transition occurred at 575 °C (~0.5 T M ). The observed stability transition provided new insight for the shrinkage mechanism of voids under irradiation. In addition, an annealing experiment on nickel, previously irradiated at 600 °C to 3 dpa, was performed sequentially at 650 °C to 720 °C to reveal the shrinkage rate of void as a function of temperature and void size. The advantage of combining computer vision and in-situ TEM to obtain comprehensive void evolution was demonstrated.

36 MATERIALS SCIENCE↗
DOE OSTI2022

DefectTrack: a deep learning-based multi-object tracking algorithm for quantitative defect analysis of in-situ TEM videos in real-time

Abstract In-situ irradiation transmission electron microscopy (TEM) offers unique insights into the millisecond-timescale post-cascade process, such as the lifetime and thermal stability of defect clusters, vital to the mechanistic understanding of irradiation damage in nuclear materials. Converting in-situ irradiation TEM video data into meaningful information on defect cluster dynamic properties (e.g., lifetime) has become the major technical bottleneck. Here, we present a solution called the DefectTrack , the first dedicated deep learning-based one-shot multi-object tracking (MOT) model capable of tracking cascade-induced defect clusters in in-situ TEM videos in real-time. DefectTrack has achieved a Multi-Object Tracking Accuracy (MOTA) of 66.43% and a Mostly Tracked (MT) of 67.81% on the test set, which are comparable to state-of-the-art MOT algorithms. We discuss the MOT framework, model selection, training, and evaluation strategies for in-situ TEM applications. Further, we compare the DefectTrack with four human experts in quantifying defect cluster lifetime distributions using statistical tests and discuss the relationship between the material science domain metrics and MOT metrics. Our statistical evaluations on the defect lifetime distribution suggest that the DefectTrack outperforms human experts in accuracy and speed.

42 ENGINEERING↗
DOE OSTI2022

Effect of thermal aging on microstructure and crack growth response of Alloy 152 weld

Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 152 weldments during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni2Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 152 under accelerated thermal aging. The materials studied involved three heats of Alloy 152 used to produce a dissimilar metal weld (DMW) joining an Alloy 690 plate to an Alloy 533 low alloy steel (LAS) plate, thermally aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service). The microstructural characterization by means of synchrotron X-ray did not find evidence of LRO in any of the three heats aged to an equivalent of 60 years of service. Testing in a primary water environment of a heat of Alloy 152 aged at 370°C to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. However, the SCC CGR response of the aged sample appears to show a deterioration in performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Effects of Sodium Exposure on the Tensile Properties of Grade 91 steel

Two heats of G91 steel have been investigated in liquid sodium at 550- 650°C to understand their corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure experiments of subsized, sheet-type tensile specimens of G91 were performed in Argonne’s forced convection sodium materials testing loops, SMT-1 and SMT-2. Maximum exposure times of ~49,000 h at ~38,000 h, and ~20,000 h have been achieved at 550, 600, and 650°C, respectively. Thermal aging experiments of G91 steel were also conducted in parallel, and the results were compared with those of the sodium-exposed specimens to isolate the effect of thermal aging from the effect of sodium exposure. It was found that the yield stress and ultimate tensile strength of G91 were reduced by more than 50% after sodium exposures at 650°C due to decarburization. The reduction in tensile strength was accompanied by the dissolution of M 23 C 6 carbides and excessive grain growth. The effect of sodium at 550 and 600°C was insignificant for exposure times up to ~49,000 h and ~38,000 h, respectively. The effect of sodium on the microstructural stability and tensile properties of G91 is largely dependent on the carbon mass transfer in the alloy-sodium system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Unveiling the interaction of nanopatterned void superlattices with irradiation cascades

Nanopatterned microstructures in materials can have a profound impact on materials’ physical and chemical properties. While voids are typically considered as detrimental defects in irradiated materials, the patterning of nanoscale voids causes the formation of void superlattices and provides a highly efficient mechanism for gas storage. Despite the important applications of nanopatterned defect superlattices, how they degrade under irradiation remains unclear. Here we provide direct observation of the evolution of void superlattices under irradiation and elucidate the interaction of void superlattices with irradiation cascades. We reveal that the instability of void superlattices under irradiation is caused by heterogenous void shrinkage and demonstrate the imperative role of mixed 1D/3D diffusion of self-interstitial atoms and injected inert gas atoms on void shrinkage and void superlattice instability. Understanding the degradation mechanisms of nanopatterned microstructures is essential to designing damage-tolerant materials and broadening their applications in extreme environments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Carburization and decarburization behavior of Grade 91 ferritic-martensitic steel in liquid sodium environments

Here, this paper presents a study of carbon transfer and its effect on microstructure and tensile properties of Grade 91 (G91) ferritic-martensitic steel exposed to sodium at 550–650 °C. Sodium exposure tests were conducted in Argonne's forced convection sodium loops up to exposure times of ~40,000 h. Thermal aging study of G91 steel was conducted in parallel to isolate the thermal aging effect from the sodium effect. It was found that sodium exposures at 650 °C dissolved M 23 C 6 carbides, eliminated the martensite subgrain structure resulting in excessive grain growth and reduced the tensile strength by >50%, while sodium exposures at 550 and 600 °C had an insignificant effect on its microstructure and tensile properties. These effects were attributed to the carburization/decarburization process of G91 steel in sodium environments. Carbon concentrations in sodium were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8–1.2 ppm in the SMT-1 loop and 0.3–0.7 ppm in the SMT-2 loop. Thermodynamic analysis of the carburization – decarburization process was conducted for G91 steels exposed in sodium environments. The carbon activity-concentration relationship for G91 was evaluated by considering four phases in G91, i.e. bcc ferrite, M 23 C 6 , NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M 23 C 6 carbides at high carbon activities, while NbC and VC carbides dominated the process at low carbon activities. The calculated carburization-decarburization boundary showed that G91 would undergo decarburization at 650 °C and carburization at 550 °C in the sodium loop environments, which was consistent with our experimental observations. This experimental and theoretical analysis provided a basis for predicting the effect of carbon transfer on the integrity of reactor components in sodium environments and for the design of new alloys used in sodium-cooled fast reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2022

In-situ irradiation-induced studies of grain growth kinetics of nanocrystalline UO 2

The thermal conductivity of UO 2 fuel needs to be high enough to dissipate the heat generated from the fission reaction. Since grain size affects thermal conductivity and grain size can evolve with irradiation, it is critical to understand in-reactor UO 2 grain growth. Most studies of grain growth in UO 2 are based on thermally driven processes at elevated temperatures. However, studies have shown that grain growth can occur even at cryogenic temperatures by ballistic processes. Such irradiation-induced grain growth in UO 2 is yet to be studied. Advanced in-situ Kr ion irradiation and transmission electron microscopy were systematically performed on nanocrystalline UO 2 thin films at temperatures ranging from 50 K to 1073 K; grain growth was observed at all temperatures. A combination of manual and machine learning techniques was used to measure and plot grain size evolution against irradiation fluence at various irradiation temperatures. The machine learning method has significantly improved the analysis efficiency and reduced human labors. The grain diameter data were fitted using classical grain growth and thermal spike models to describe grain growth kinetics with and without irradiation effect. Grain growth during low temperature irradiation (≤ 475 K) can be well described by the thermal spike model. Above 475 K, there were additional thermally assisted processes that further accelerate the grain growth. At the highest irradiation temperature about 1075 K, both irradiation-induced dislocation loops and cavities/bubbles were observed to form in the UO 2 . In this report, the effects of irradiation-induced defects on grain growth kinetics are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Radiation induced amorphization of carbides in additively manufactured and conventional ferritic-martensitic steels: In-situ experiments on extraction replicas

In-situ irradiations using 1 MeV Kr 2+ ions in a transmission electron microscope were performed on extraction replica samples containing M 23 C 6 carbides and MX carbonitrides from two ferritic-martensitic (FM) steels: (i) 9%Cr-1%W-TaV based Eurofer97 and (ii) 9%Cr-1%Mo-VNb (all in wt.%) based additively manufactured (AM) Grade91 steel. The irradiations were performed between 100 and 773 K up to a maximum dose of ~2.4 displacements per atom (dpa). The M 23 C 6 carbides are highly susceptible to radiation induced amorphization (RIA), while the MX type nanoprecipitates are highly amorphization resistant across the entire irradiation temperature range. Between 100 and 423 K, RIA of M 23 C 6 carbides occurs very rapidly with critical amorphization doses ranging between ~0.35 and 0.9 dpa, increasing to higher doses at 573 K. Complete amorphization of the M 23 C 6 carbides up to doses of ~2.4 dpa is not possible at 773 K under the present irradiation conditions. The critical amorphization dose of M 23 C 6 carbides increases nearly exponentially with irradiation temperature. The critical temperature for the crystalline-to-amorphous phase transformation (T c→a ) of M 23 C 6 carbides irradiated as replica samples, i.e. without the surrounding metal matrix, was estimated to be ~812 K. Comparing the present results with neutron irradiation data on bulk samples reveals a decrease in the critical amorphization dose for M 23 C 6 and an increase of T c→a , highlighting the effect of dose rate on the amorphization behavior that is qualitatively consistent with literature on other non-metals. Changes in the minor chemistry of M 23 C 6 , such as presence or absence of W, V, Nb, Mo, seems to have little effect on the amorphization behavior. Here, in-situ irradiations on extraction replica samples provide a novel pathway to explore radiation tolerance of nanoprecipitates in nuclear structural materials

36 MATERIALS SCIENCE↗
DOE OSTI2021

Effect of heat treatment on creep behavior of 316 L stainless steel manufactured by laser powder bed fusion

The objective of this study is to understand the thermal stability of microstructure and its effect on the creep behavior of additively-manufactured 316L stainless steel (AM 316L SS). Creep specimens were fabricated from rods printed by a laser powder bed fusion process. Six different heat treatments, namely 650 °C/1 h, 700 °C/1 h, 750 °C/1 h, 800 °C/1 h, 900 °C/1 h and 1050 °C/1 h were applied to the creep specimens. The heat-treated specimens were creep-tested under the same condition, 550 °C/275 MPa to evaluate the effect of post-build heat treatment on the creep behavior of AM 316L SS. In the temperature range of 650-750 °C, dislocation density within cells, and cell size and wall thickness were affected by the heat treatment, while elemental segregation at boundaries remained unchanged and Mn-enriched Si oxide particles remained stable. In the temperature range of 750-900 °C, concurrent changes were observed in dislocation cell structures, elemental segregation and oxide particles. The 1050 °C-heat treatment removed cell structures and boundary solute segregations, leading to formation of equiaxed grains; Mn-enriched Si oxide particles in the as-built specimen were replaced by more stable Mn-enriched Cr oxides. The creep life of AM 316L SS increased after the heat treatment at 650 °C and then decreased with increasing temperature up to 900 °C where cell/subgrain structures still existed. The creep rate and creep elongation followed an opposite trend. The creep rate can be correlated with the structural parameters of cell/subgrain size and cell wall/boundary thickness and can be rationalized by different strengthening mechanisms of dislocation cells and subgrains. The heat treatment affected strongly the primary and secondary creep but had a minimal effect on the tertiary creep of AM 316L SS.

316L stainless steel↗
DOE OSTI2021

Helium bubble formation in nickel under in-situ krypton and helium ions dual-beam irradiation

Helium has profound effects on the microstructure and mechanical property of materials used in nuclear power systems. Here to understand the interaction between helium and irradiation-induced defects and its effect on bubble formation, single-beam and dual-beam irradiations on nickel were performed with 16 keV helium ions and 1 MeV krypton ions under in-situ TEM observation. For 1 MeV krypton single-beam irradiation, voids were observed at 600ºC but not at 500ºC and 700ºC. For helium single-beam irradiation at 500ºC, helium moved freely and formed bubble-loop complexes. For dual-beam irradiation at 500ºC, helium was trapped by vacancies and formed bubbles homogeneously. The structure of dislocation loops was also affected by the presence of helium. For krypton single-beam irradiation at 500ºC, the microstructure was {100} periodic wall of faulted and perfect dislocation loops. For dual-beam irradiation at 500ºC, periodic wall was not observed, and dislocation loops were predominantly perfect. For single-beam helium at 500ºC, dislocation loops were produced by athermal process of SIAs ejection or loop punching mechanisms. The interplay between helium and vacancies played an important role on the evolution of helium bubbles and defect microstructures.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Performance of Pulsed Thermal Tomography Imaging with Machine Learning-Based Classification of Defects in Additively Manufactured Structures

Additive manufacturing (AM) is an emerging method for cost-efficient fabrication of complex topology nuclear reactor parts from high-strength corrosion resistance alloys, such as stainless steel and Inconel. AM of metallic structures for nuclear energy applications is currently based on laser powder bed fusion (LPBF) process. Some of the challenges with using LPBF method for nuclear manufacturing include the possibility of introducing pores into metallic structures. Integrity of AM structures needs to be evaluated nondestructively because material flaws could lead to premature failures in high temperature nuclear reactor environment. Currently, there exist limited capabilities to evaluate actual AM structures non-destructively. Pulsed Thermal Tomography Imaging (PTT) provides a capability for non-destructive evaluation (NDE) of subsurface defects in arbitrary size structures. The PTT method is based on recording material surface temperature transients with infrared (IR) camera following thermal pulse delivered on material surface with flash light. The PTT method has advantages for NDE of actual AM structures because the method involves one-sided non-contact measurements and fast processing of large sample areas captured in one image. Following initial qualification of an AM component for deployment in a nuclear reactor, a PTT system can also be used for in-service nondestructive evaluation (NDE) applications. In this report, we describe recent progress in enhancing PTT capabilities in detecting and visualizing microscopic defects in metallic specimens. The thermal tomography (TT) algorithm obtains depth reconstructions of spatial effusivity from the data cube of sequentially recorded surface temperatures. However, interpretation of TT images is non-trivial because of blurring of images with increasing depth. To address this challenge, we have developed a deep learning convolutional neural network (CNN) to classify size and orientation subsurface defects in simulated TT images. CNN is trained on a database of TT images created for a set of simulated metallic structures with elliptical subsurface voids. Test of CNN performance demonstrate the ability to classify radii and angular orientation of subsurface defects in TT images. In addition, we have shown that CNN trained on elliptical defects is capable of classifying irregular-shaped defects obtained from scanning electron microscopy (SEM) of stainless steel sections printed with LPBF.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Effect of thermal aging and irradiation on microstructure and crack growth response of Alloy 690

Nickel-base Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys and are expected to receive initial operating licenses for 60 years. Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear plants, and possibly spent fuel storage containers. The objective of this research is to understand the microstructural changes occurring in high-Cr, Ni-based Alloy 690 during long time exposure to the reactor operating temperatures, and the effect of these changes on the service performance. One area of particular concern is the potential for long range ordering (LRO), i.e. formation of the intermetallic Ni 2 Cr phase under prolonged exposure to reactor temperatures and/or irradiation, which can increase strength, decrease ductility, and cause dimensional changes or lead to in-service embrittlement of components made with these alloys. Hence, this research focused on the microstructural evolution and the SCC response of Alloy 690 under accelerated thermal aging and irradiation conditions. The materials studied involved Alloy 690 from two sources: (i) Alloy 690 plate from an Alloy 152 dissimilar metal weld (DMW) joining the Alloy 690 plate to Alloy 533 LAS, aged at three different temperatures (370°C, 400°C and 450°C) for up to 75,000h (equivalent to 60 years of service), and (ii) specimens neutron-irradiated in the BOR-60 reactor up to 40 dpa. For aged Alloy 690 specimens, hardness was found to increase with aging time, however, the microstructural characterization by means of synchrotron X-ray did not find evidence of LRO. The microstructural characterization of neutron-irradiated specimens by TEM found no evidence of LRO either. Testing in a primary water environment of Alloy 690 specimens aged to a 60-year service equivalent revealed a fatigue and corrosion fatigue crack growth responses similar to those measured on the un-aged alloy. The SCC CGR response was also low. Overall, the two Alloy 690 heats investigated in this work, aged up to 60-year service equivalents or exposed to neutron irradiation up to 40 dpa, did not exhibit a deterioration in microstructure or performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

IVEM Contributions to CINR Penn State MT17-12797: In situ Ion Irradiation to Add Irradiation Assisted Grain Growth to the MARMOT Tool

In-situ TEM experiments were performed at the IVEM-Tandem Facility in Argonne National Laboratory to study the irradiation-induced grain growth of nanocrystalline UO 2 and CeO 2 thin films using 1 MeV Kr ions at various temperatures from 50K to 1073K. The grain growth kinetics was studied under both irradiation and thermal annealing conditions. Bright field and dark field TEM images and diffraction patterns were systematically recorded as a function of temperature, irradiation dose and annealing time. A comprehensive grain growth data of UO 2 was collected for further analysis and for the implementation to the MARMOT tool.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Location-Dependent Mechanical Property Evaluation on Additively Manufacture Materials

This report summarizes research activities conducted at Argonne National Laboratory in support of the development, qualification and certification of additively manufactured (AM) metallic components to allow for innovative reactor design and licensing for the Transformational Challenge Reactor (TCR). Focus is on the evaluation of high temperature mechanical properties including creep and fatigue of 316L stainless steel manufactured by the laser powder bed fusion process. Creep behavior of AM 316L stainless steel was evaluated for rods printed by either or a combination of both lasers of the dual-laser system of a Concept Laser-M2 printer to examine the consistency across the build plate. Creep tests were conducted at temperatures of 550, 575, 600, and 650°C and stresses between 175 and 300 MPa using ASTM standard-sized specimens. The effect of heat treatments at different temperatures on creep properties of AM 316L SS was examined to understand the processing-microstructure-property relationship. Fatigue properties of AM 316L SS were investigated for two print geometries (rod and plate) and in two build orientations of printed plates. Locations of specimens in the build were carefully tracked such that the testing data can be directly related to location-specific in situ data to establish links between printing process, post-printing treatment, microstructure and mechanical properties. The work is to support the development of a digital platform informed approach to AM component qualification and certification for nuclear applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗