Engineering topics
Sridharan, Kumar
Publications and source records attributed to Sridharan, Kumar.
Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys
The development and understanding of the mechanical properties of neutron-irradiated FeCrAl alloys is increasingly a critical need as these alloys continue to become more mature for nuclear reactor applications. This study focuses on the mechanical properties of model FeCrAl alloys and of a commercial FeCrAl alloy neutron-irradiated to up to 13.8 displacements per atom (dpa) at irradiation temperatures between 320 and 382 °C. Tensile tests were completed at room temperature and at 320 °C, and a subset of fractured tensile specimens was examined by scanning electron microscopy. Results showed typical radiation hardening and embrittlement indicative of high chromium ferritic alloys with strong chromium composition dependencies at lower doses. At and above 7.0 dpa, the mechanical properties saturated for both the commercial and model FeCrAl alloys, although brittle cleavage fracture was observed at the highest dose in the model FeCrAl alloy with the highest chromium content (18 wt %). Finally, the results suggest the composition and microstructure of FeCrAl alloys plays a critical role in the mechanical response of FeCrAl alloys irradiated near temperatures relevant to light water reactors.
A combined APT and SANS investigation of α' phase precipitation in neutron-irradiated model FeCrAl alloys
Here, FeCrAl alloys are currently under consideration for accident-tolerant fuel cladding applications in light water reactors owing to their superior high-temperature oxidation and corrosion resistance compared to the Zr-based alloys currently employed. However, their performance could be limited by precipitation of a Cr-rich α' phase that tends to embrittle high-Cr ferritic Fe-based alloys. In this study, four FeCrAl model alloys with 10–18 at.% Cr and 5.8–9.3 at.% Al were neutron-irradiated to nominal damage doses up to 7.0 displacements per atom at a target temperature of 320 °C. Small angle neutron scattering techniques were coupled with atom probe tomography to assess the composition and morphology of the resulting α' precipitates. It was demonstrated that Al additions partially destabilize the α' phase, generally resulting in precipitates with lower Cr contents when compared with binary Fe-Cr systems. The precipitate morphology evolution with dose exhibited a transient coarsening regime akin to previously observed behavior in aged Fe-Cr alloys. Similar behavior to predictions of the LSW/UOKV models suggests that α' precipitation in irradiated FeCrAl is a diffusion-limited process with coarsening mechanisms similar to those in thermally aged high-Cr ferritic alloys.
Correlative Microscopy of Neutron-Irradiated Materials
A nuclear reactor core is a highly demanding environment that presents several unique challenges for materials performance. Materials in modern light water reactor (LWR) cores must survive several decades in high-temperature (300-350°C) aqueous corrosion conditions while being subject to large amounts of high-energy neutron irradiation. Next-generation reactor designs seek to use more corrosive coolants (e.g., molten salts) and even greater temperatures and neutron doses. The high amounts of disorder and unique crystallographic defects and microchemical segregation effects induced by radiation inevitably lead to property degradation of materials. Thus, maintaining structural integrity and safety margins over the course of the reactor's service life thus necessitates the ability to understand and predict these degradation phenomena in order to develop new, radiation-tolerant materials that can maintain the required performance in these extreme conditions.
Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report
A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.
Heterogeneous dislocation loop formation near grain boundaries in a neutron-irradiated commercial FeCrAl alloy
FeCrAl alloys are an attractive materials class for nuclear power applications due to their increased environmental compatibility over more traditional nuclear materials. Preliminary studies into the radiation tolerance of FeCrAl alloys under accelerated neutron testing between 300-400 °C have shown post-irradiation microstructures containing dislocation loops and Cr-rich ' phase. Although these initial works established the post-irradiation microstructures, little to no focus was applied towards the influence of pre-irradiation microstructures on this response. Here, a well annealed commercial FeCrAl alloy, Alkrothal 720, was neutron irradiated to 1.8 dpa at 382 °C and then the role of random high angle grain boundaries on the spatial distribution and size of dislocation loops, dislocation loops, and black dot damage was analyzed using on-zone scanning transmission electron microscopy. Results showed a clear heterogeneous dislocation loop formation with dislocation loops showing an increased number density and size, black dot damage showing a significant number density decrease, and an increased size of dislocation loops in the vicinity directly adjacent to the grain boundary. Lastly, these results suggest the importance of the pre-irradiation microstructure on the radiation tolerance of FeCrAl alloys.
Corrosion-induced microstructural developments in 316 stainless steel during exposure to molten Li 2 BeF 4 (FLiBe) salt
The microstructural evaluation and characterization of 316 stainless steel samples that were tested in molten Li 2 BeF 4 (FLiBe) salt were investigated in this study for evaluating its performance in high-temperature molten fluoride salts. Recently, 316 stainless steel and FLiBe salt are being actively considered as the main structural alloy and primary coolant of fluoride salt-cooled high-temperature reactor (FHR), a leading nuclear reactor concept for the next generation nuclear plants (NGNP). In support of the materials development for the FHR, high-temperature corrosion tests of 316 stainless steel in molten FLiBe salt at 700°C have been conducted in both bare graphite crucibles and 316 stainless steel-lined crucibles in an inert atmosphere for up to 3000 hours. The microstructure of the tested samples was comprehensively characterized using scanning electron microscopy (SEM) in conjunction with energy dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD), and scanning transmission electron microscopy (STEM) with EDS. In addition to the noticeable intergranular corrosion attack on surface, the corrosion in terms of the Cr depletion along high angle grain boundaries (15-180º) extended to 22µm in depth after 3000-hour exposure to molten FLiBe salt in graphite crucible. The coherent Σ3 grain boundary appeared high resistance to the Cr depletion. The substantial Cr depletion from the near-to-surface layer induced phase transformation from γ-martensite to α-ferrite phase (FeNi x ) during corrosion at 700ºC. Furthermore, the presence of graphite in the molten salt doubled the corrosion attack depth and led to the formation of round Mo2C, hexagonal Cr 7 C 3 and needle-like Al 4 C 3 phase within the alloy as deep as 50 µm after 3000-hour corrosion testing. Based on the microstructural analysis, the corrosion mechanisms of 316 stainless steel in molten FLiBe salt in different corrosion crucibles were illuminated through schematic diagrams. Additionally, a thermal diffusion controlled corrosion model was developed and validated by experimental data for predicting the long-term corrosion attack depth.
Corrosion of Alloy Haynes 230 in High Temperature Supercritical Carbon Dioxide with Oxygen Impurity Additions
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Observations of defect structure evolution in proton and Ni ion irradiated Ni-Cr binary alloys
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Development of Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding
There is considerable interest in the development of coatings for zirconium-alloy fuel cladding used in light water reactors (LWR) to improve their oxidation resistance during high temperature accident conditions and to enhance performance under normal operating conditions. These coatings should also enhance the heat transfer characteristics of the cladding surface so that heat from the fuel rod is efficiently transferred into the water coolant. One potential candidate coating material is zirconium-silicide. Transition metal silicides have been generally shown to have outstanding oxidation resistance. Zirconium-silicide is particularly attractive because Zr is strongly oxophilic, so that the Zr-silicide layer is expected to bond more strongly to the inner, native ZrO{sub 2} layer. The relative proximity of the free energy of formation of oxides of Zr and Si can lead to alternating multilayered oxide structures with radiation damage resistance. Additionally, oxidation of zirconium silicide can produce excellent corrosion resistant glassy zircon (ZrSiO{sub 4}) and silica (SiO{sub 2}), possessing self-healing qualities. This paper reports results of our results of high temperature oxidation tests of bulk zirconium-silicide and coatings of zirconium-silicide deposited on zirconium-alloy test flats and cladding tube configurations using plasma vapor deposition (PVD) process. We also report our initial results on the experimental facility for studying boiling characteristics of coated cladding tubes in prototypic pressurized water conditions. (authors)
Corrosion of Candidate Materials in Molten FLiBe Salt for Fluoride-Salt-Cooled High-Temperature Reactor (FHR)
Molten fluoride salts have attracted considerable attention because of their desirable thermophysical properties that lead to superior heat transfer. Recently, fluoride salts are being considered as coolants in the fluoride salt-cooled high temperature reactors (FHRs). The successes of the molten salt reactor experiment (MSRE) program in 1950's-70's at Oak Ridge National Laboratory (ORNL) has also driven the pursuit of molten salt nuclear reactor as one of the lead next generation nuclear reactor concepts. Unlike MSRE where the uranium fuel was dissolved in the molten fluoride salt, the FHR concept is expected to use molten salt (2LiF-BeF{sub 2}, FLiBe) only as coolant in conjunction with solid TRISO fuel particles immersed in the molten salt. One of the important challenges in FHR development is corrosion of structural materials in high-temperature molten 2LiF-BeF{sub 2} (FLiBe) salt. Currently, 316 stainless steel (nominal composition: Fe-12Ni-17Cr-2.5Mo-1Si-2Mn-0.03C) and Hastelloy N (nominal composition: Ni-7Cr-16Mo-5Fe-1Si-0.8Mn) are deemed as promising candidate alloys for the construction of some of the structural components of FHR. 316 stainless steel is ASME code certified for high temperatures (up to 760 deg. C) applications. Hastelloy N has been shown to have outstanding corrosion resistance in molten FLiBe but present code certification data limit the use of this alloy to only about 704 deg. C. Modifications of Hastelloy N with higher creep strength are presently being developed at ORNL. Other materials such graphite, SiC/SiCf and C/Cf composites, are also considered for structural applications in the FHR. The presence of these multiple materials in a molten FLiBe salt medium can further exacerbate and complicate corrosion effects. Therefore, it is necessary to understand the corrosion behavior of these materials in molten FLiBe salt. The corrosion behavior of Hastelloy N, 316 stainless steel, and CVD SiC and two types of SiC-SiC{sub f} composites have been tested in molten FLiBe salt at 700 deg. C. When tested in metal-lined crucibles both Hastelloy N and 316 stainless steel showed very good corrosion resistance. Samples tested in graphite crucibles showed greater propensity for corrosion and carburization was observed in the near surface regions of the alloys. CVD SiC showed outstanding corrosion resistance. However, the two SiC-SiC{sub f} composites exhibited corrosion attack implying that the manufacturing process may play a role in the corrosion of these composites. (authors)
Irradiation-enhanced α' precipitation in model FeCrAl alloys
We have irradiated the model FeCrAl alloys with varying compositions (Fe(10–18)Cr(10–6)Al at.%) with a neutron at ~ 320 to damage levels of ~ 7 displacements per atom (dpa) to investigate the compositional influence on the formation of irradiation-induced Cr-rich α' precipitates using atom probe tomography. In all alloys, significant number densities of these precipitates were observed. Cluster compositions were investigated and it was found that the average cluster Cr content ranged between 51.1 and 62.5 at.% dependent on initial compositions. Furthermore, this is significantly lower than the Cr-content of α' in binary FeCr alloys. As a result, significant partitioning of the Al from the α' precipitates was also observed.
Role of Defects in Swelling and Creep of Irradiated SiC
Silicon carbide is a promising cladding material because of its high strength and relatively good corrosion resistance. However, SiC is brittle and therefore SiC-based components need to be carefully designed to avoid cracking and failure by fracture. In design of SiC-based composites for nuclear reactor applications it is essential to take into account how mechanical properties are affected by radiation and temperature, or in other words, what strains and stresses develop in this material due to environmental conditions. While thermal strains in SiC can be predicted using classical theories, radiation-induced strains are much less understood. In particular, it is critical to correctly account for radiation swelling and radiation creep, which contribute significantly to dimensional instability of SiC under radiation. Swelling typically increases logarithmically with radiation dose and saturates at relatively low doses (damage levels of a few dpa). Consequently, swelling-induced stresses are likely to develop within a few months of operation of a reactor. Radiation-induced volume swelling in SiC can be as high as 2%, which is significantly higher than the cracking strain of 0.1% in SiC. Swelling-induced strains will lead to enormous stresses and fracture, unless these stresses can be relaxed via some other mechanism. An effective way to achieve stress relaxation is via radiation creep. Although it has been hypothesized that both radiation swelling and radiation creep are driven by formation of defect clusters, existing models for swelling and creep in SiC are limited by the lack of understanding of specific defects that form due to radiation in the range of temperatures relevant to fuel cladding in light water reactors (LWRs) (<1000°C). For example, defects that can be detected with traditional transmission electron microscopy (TEM) techniques account only for 10-45% of the swelling measured in irradiated SiC. Here, we have undertaken an integrated experimental and modeling effort to discover the previously invisible defects in irradiated SiC and to determine the contributions of these defects to radiation swelling. Knowledge of the most stable defect structures and the rate controlling processes during defect evolution is essential for development of predictive models for swelling and creep as a function of temperature and radiation dose. This research has been enabled by state-of-the-art imaging techniques, such as the aberration corrected scanning transmission electron microscopy (STEM) (FEI TITAN) closely coupled with multi-scale models of stable defect clusters and their evolution.
Complementary Techniques for Quantification of alpha' Phase Precipitation In Neutron-Irradiated Model Fe-Cr-Al Alloys
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Effect of carbon ion irradiation on Ag diffusion in SiC
Transport of Ag fission product through the silicon-carbide (SiC) diffusion barrier layer in TRISO fuel particles is of considerable interest given the application of this fuel type in high temperature gas-cooled reactor (HTGR) and other future reactor concepts. The reactor experiments indicate that radiation may play an important role in release of Ag; however so far the isolated effect of radiation on Ag diffusion has not been investigated in controlled laboratory experiments. In this study, we investigate the diffusion couples of Ag and polycrystalline 3C–SiC, as well as Ag and single crystalline 4H–SiC samples before and after irradiation with C 2+ ions. The diffusion couple samples were exposed to temperatures of 1500 °C, 1535 °C, and 1569 °C, and the ensuing diffusion profiles were analyzed by secondary ion mass spectrometry (SIMS). We found that diffusion coefficients calculated from these measurements indicate that Ag diffusion was greatly enhanced by carbon irradiation due to a combined effect of radiation damage on diffusion and the presence of grain boundaries in polycrystalline SiC samples.
Impact of corrosion test container material in molten fluorides
The effects of crucible material choice on alloy corrosion rates in immersion tests in molten LiF–NaF–KF (46.5–11.5-42 mol. %) salt held at 850 °C for 500 hrs are described. Four crucible materials were studied. Molten salt exposures of Incoloy-800H in graphite, Ni, Incoloy-800H, and pyrolytic boron nitride (PyBN) crucibles all led to weight-loss in the Incoloy-800H coupons. Alloy weight loss was ~30 times higher in the graphite and Ni crucibles in comparison to the Incoloy-800H and PyBN crucibles. It is hypothesized galvanic coupling between the alloy coupons and crucible materials contributed to the higher corrosion rates. Alloy salt immersion in graphite and Ni crucibles had similar weight-loss hypothesized to occur due to the rate limiting out diffusion of Cr in the alloys to the surface where it reacts with and dissolves into the molten salt, followed by the reduction of Cr from solution at the molten salt and graphite/Ni interfaces. As a result, both the graphite and the Ni crucibles provided sinks for the Cr, in the formation of a Ni–Cr alloy in the case of the Ni crucible, and Cr carbide in the case of the graphite crucible.
High-temperature corrosion of UNS N10003 in molten Li 2 BeF 4 (FLiBe) salt
Here, corrosion testing of Hastelloy N in molten fluoride salt was performed in purified molten 2 7 LiF-BeF 2 (66-34mol%) (FLiBe) salt at 700°C for 1000 hours, in pure nickel and graphite capsules. In the nickel capsule tests, the near-surface region of the alloy exhibited an about 200 nm porous structure, an approximately 3.5 μm chromium depleted region, and MoSi 2 precipitates. In tests performed in graphite capsules, the alloy samples gained weight due to the formation of a variety of Cr 3 C 2 , Cr 7 C 3 , Mo 2 C and Cr 23 C 6 , carbide phases on the surface and in the subsurface regions of the alloy. A Cr depleted region was observed in the near-surface region where Mo thermally diffused toward either surface or grain boundary, which induced approximately 1.4 μm Ni 3 Fe alloy layer in this region. The carbide containing layer extended to about 7 μm underneath the Ni 3 Fe layer. The presence of graphite dramatically changes the mechanisms of corrosion attack in Hastelloy N in molten FLiBe salt. Evaluated by in terms of the depth of attack, graphite clearly accelerates corrosion, but the results appear to indicate that the formation of Cr 23 C 6 phase might stabilize the Cr and mitigate its dissolution in molten FLiBe salt.