Microstructural Responses of Alloy 625 and Alloy 625 Plus Under Ion and Proton Irradiations
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Materials Science↗
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Alloy 625 Qualification Pathway for ASME Section III Division 5 Class A construction
The American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III Division 5 provides construction and inspection rules to ensure the safety of nuclear components operating at elevated temperatures, defined as those operating above 700°F (370°C) or 800°F (425°C), depending on the material type. At present, there are six materials approved for Class A (high safety significance) component construction under ASME BPVC Section III Division 5. Out of these six materials, five materials are steels or iron-based alloys and one material, Alloy 617, is a nickel alloy. Nickel alloys are stronger than steels; however, Alloy 617 consists of 10-15% cobalt, and activation is a concern under irradiation. Therefore, there is a need to qualify new nickel alloy with lower cobalt content. Alloy 625 is one candidate which has similar mechanical properties at elevated temperature compared to Alloy 617 and has a maximum of 1% cobalt. Although Alloy 625 material properties have been developed in the past to support allowable stress development under ASME Section II, new test results would be needed to qualify this material for elevated temperature component construction under ASME Section III Division 5. The purpose of this report is to describe a path towards developing a Nuclear Code Case to qualify Alloy 625 (UNS N06625; Grade 1 and Grade 2) for elevated temperature nuclear use in accordance with the ASME BPVC rules. This report reviews the existing database on Alloy 625, presents proposed test campaign, and discusses potential paths for accelerating the accelerated material qualification process.
36 - MATERIALS SCIENCE↗
Comparing structure-property evolution for PM-HIP and forged alloy 625 irradiated with neutrons to 1 dpa
The nuclear power industry has growing interest in qualifying powder metallurgy with hot isostatic pressing (PM-HIP) to replace traditional alloy fabrication methods for reactor structural components. But there is little known about the response of PM-HIP alloys to reactor conditions. This study directly compares the response of PM-HIP to forged Ni-base Alloy 625 under neutron irradiation doses ~0.5–1 displacements per atom (dpa) at temperatures ranging ~321–385 °C. Post-irradiation examination involves microstructure characterization, ASTM E8 uniaxial tensile testing, and fractography. Up through 1 dpa, PM-HIP Alloy 625 appears more resistant to irradiation-induced cavity nucleation than its forged counterpart, and consequently experiences significantly less hardening. This observed difference in performance can be explained by the higher initial dislocation density of the forged material, which represents an interstitial-biased sink that leaves a vacancy supersaturation to nucleate cavities. These findings show promise for qualification of PM-HIP Alloy 625 for nuclear applications, although higher dose studies are needed to assess the steady-state irradiated microstructure.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Corrosion property of Alloy 625 in Molten FLiNaK salt according to the Tellurium Concentrations
For this work, the corrosion properties of Alloy 625 have been studied in molten FLiNaK salt for 48 h at 700 °C as a function of tellurium (Te) concentrations using immersion tests and electrochemical methods. Under specific Te concentrations, we found that Te-induced corrosion can be suppressed and help protect the alloy against corrosion. The sample after the immersion test containing 0.1 wt.% Te (0.1 wt.% Te sample) had the lowest corrosion depth and mass loss, coupled with the highest charge transfer resistance, obtained using electrochemical impedance spectroscopy (EIS). On the other hand, the alloy with Te content above 0.1 wt.% exhibited severe corrosion penetration and lower charge transfer resistance in FLiNaK. Based on thermodynamical and electrochemical analyses, a corrosion mechanism of Alloy 625 in the molten salt containing Te was theorized involving the formation of stable corrosion products such as Ni/Fe rich layer depleted in Cr, nickel telluride, and chromium oxide.
Hong, Minsung↗
Evaluation of HCF strength of Alloy 625 with non-optimum additive manufacturing process parameters
An extensive experimental program was undertaken to generate the process-structure-HCF property data for additive-manufactured Alloy 625. Round tensile and flat dog-bone HCF specimens were fabricated using a laser powder bed fusion process. Several builds were fabricated with different processing parameters. The fatigue strengths were determined using a step-test method and fracture surfaces were characterized to identify the critical microstructure feature associated with the formation of the fatigue crack. Finally, a fracture mechanics-based methodology using fatigue crack growth data measured on a wrought alloy is shown to provide conservative estimates of the HCF lives for the additive-manufactured alloy.
36 MATERIALS SCIENCE↗
Solid-State Transformation of an Additive Manufactured Inconel 625 Alloy at 700 °C
Inconel 625, a nickel-based superalloy, has drawn much attention in the emerging field of additive manufacturing (AM) because of its excellent weldability and resistance to hot cracking. The extreme processing condition of AM often introduces enormous residual stress (hundreds of MPa to GPa) in the as-fabricated parts, which requires stress-relief heat treatment to remove or reduce the internal stresses. Typical residual stress heat treatment for AM Inconel 625, conducted at 800 °C or 870 °C, introduces a substantial precipitation of the δ phase, a deleterious intermetallic phase. In this work, we used synchrotron-based in situ scattering and diffraction methods and ex situ electron microscopy to investigate the solid-state transformation of an AM Inconel 625 at 700 °C. Our results show that while the δ phase still precipitates from the matrix at this temperature, its precipitation rate and size at a given time are both smaller when compared with their counterparts during typical heat treatment temperatures of 800 °C and 870 °C. A comparison with thermodynamic modeling predictions elucidates these experimental findings. Our work provides the rigorous microstructural kinetics data required to explore the feasibility of a promising lower-temperature stress-relief heat treatment for AM Inconel 625. The combined methodology is readily extendable to investigate the solid-state transformation of other AM alloys.
36 MATERIALS SCIENCE↗
Experimental Investigations into the Corrosion of Alloy 625 Using NaCl-PuCl3 Molten Salt in a Natural Circulation Microloop
Molten salt reactors (MSRs) can potentially revolutionize the nuclear industry by providing a path to a near-zero nuclear waste fuel cycle, contributing to more sustainable energy sources. As a plethora of MSR developers in the United States work toward an aggressive commercialization timeline, many of their fueled-salts—notably, chloride-based compositions—have limited operational testing with nuclear material. Licensing and operating these reactors require an understanding of corrosion effects on reactor materials of construction under operational conditions. The TerraPower Molten Chloride Fast Reactor (MCFR) is a liquid-fueled chloride-salt fast reactor which has received notable interest from the utility sector based on its desirable economic characteristics. The reactor operates at low pressure but does not require the use of highly reactive chemicals, leading to a reduced use of concrete and steel during construction. Additionally, liquid fuel allows for inherently stable behavior and natural circulation during a loss-of-site-power scenario. MCFR can be refueled while operating which makes it compatible with variable generation sources such as wind and solar. MCFR is a breed-and-burn in-situ reactor that does not implement any chemical processing or separations in the fuel cycle. Only mechanical filtration of noble metals and off-gassing of noble gases are utilized while the actinides stay mixed with the fuel at all times. The MCFR will require technology development to reach commercialization. With a breed-and-burn in-situ reactor like MCFR, the transmutation of fertile U-238 to fissile Pu-239 allows for much greater fuel utilization.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Laser Powder Bed Fusion Manufactured Alloys 625 and 282 and Properties Relevant to Advanced Nuclear Components
Under the Advanced Materials and Manufacturing Technologies program, two Ni-based alloys fabricated by laser powder bed fusion (LPBF) have been evaluated: γ′-strengthened Haynes 282 and solution-strengthened Inconel 625. Large printing defects were observed in the LPBF 282 alloy fabricated using a Renishaw AM250 machine, likely due to particle spattering during printing. Annealing at 1,180°C for 1 h followed by 4 h at 800°C in a high density of 50 nm γ′-strengthening precipitates as well as partial recrystallization resulted in a bimodal grain distribution. Creep testing performed at 750°C revealed lower creep life and ductility for the LPBF 282 compared with wrought 282. X-ray computed tomography combined with optical and scanning electron microscopy microstructural characterization revealed crack formation during creep testing, initiated either from printing defects or from creep cavitation at grain boundaries. Printing defects were the likely reason for the lower creep performance of LPBF 282 and could be suppressed by optimizing the build configuration. Printing 625 on an EOS M290 machine using the recommended EOS parameters resulted in a very low defect density. Superior creep strength at 725°C was observed for the as-printed LPBF 625 along the build direction when compared with wrought 625. No defect larger than ~50 µm could be detected by X-ray computed tomography in the as-printed conditions and after creep testing for 1,000 h at 725°C and 150 MPa. Small, needle-like δ-phase precipitates were observed after creep testing for 100 h at 725°C and 200 MPa. The rapid formation of the γ phase is directly related to the initial Nb and Mo segregation in the LPBF 625 cell walls, and its effect on the alloy’s high-temperature performance will be evaluated. Solution annealing was carried out at 1,150°C for 1 h and led to full recrystallization of the alloy. A decrease in strength and increase in ductility were observed after solution annealing, and the alloy recrystallization resulted in isotropic tensile properties at room temperature, contrary to what was observed for the as-printed LPBF 625.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Microstructure and Residual Stress in Functionally Graded 316L Stainless Steel/Inconel 625 Alloys Fabricated by Direct Energy Deposition
Functionally graded materials (FGMs) provide a unique solution to combine distinct properties within a single component to enhance its overall performance. Understanding the microstructure and residual stress distribution is particularly important as the material dissimilarity in FGMs can result in high residual stress that affects the mechanical integrity of structural components. Here, this work aims to study the microstructure and residual stress of additively manufactured FGMs and the effects of isothermal annealing. In this study, 316L stainless steel/Inconel 625 FGMs were built by direct energy deposition, and characterized by optical and electron microscopy as well as synchrotron-based X-ray diffraction. Our study reveals that thermal annealing at 500°C for 3 h has minimal effects on the microstructure and chemistry of the graded materials, but effectively relieves the residual stress and leads to a uniform distribution of residual stress.
36 MATERIALS SCIENCE↗
Comparison of Neutron Radiation Response of Conventionally Wrought Versus Additive Manufacturing Production Methods in Alloy 625
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Nuclear Science & Technology↗
Multi material/functionally graded wire arc additive manufacturing of high strength steel valves clad with nickel alloy 625 used for oil extraction
This Phase 1 technical collaboration project (ORNL/TM-2022/2683) began on August 1st, 2020 and concluded on July 31st, 2022. The collaboration partner Schlumberger is a large business enterprise assisting oil companies to extract oil from land and subsea cost competitively and safely. It is based in Houston, Texas USA. In this work as study of additive deposition of AWS EF100S-G and AWS ERNiCrMo-3 wires via wire-arc process was conducted.
36 MATERIALS SCIENCE↗
Effect of Water Vapor on Lifetime of 625 and 120 Foils During Oxidation Between 650 and 800 °C
In this work, the oxidation behavior of alloy 625 and 120 foils was studied at 650, 700 and 800 °C in dry air and flowing air + 10% H 2 O up to 10,000 h (alloy 625) or 30,000 h (alloy 120). The effect of water vapor on Cr loss was investigated. Manganese and iron in the 120 foil induced faster Cr depletion in the foil and breakdown of the Cr 2 O 3 scale into Fe and Cr-rich oxide compared to the alloy 625 foil. In the latter case, the presence of Nb led to the formation of Nb and Cr-rich oxide after breakdown of the Cr 2 O 3 scale. Simultaneous Cr loss due to oxidation and volatilization of the Cr 2 O 3 oxide scale was predicted and compared to experimental Cr loss measurements for exposures up to 30,000 h.
36 MATERIALS SCIENCE↗
Evaluation of Directed Energy Deposition and Laser Powder Bed Fusion Nickel-Based Alloys Process Application Envelopes Based on Performance, Process Economics, Supply Chain Risks, and Reactor-Specific Targeted Components
The goal of the Advanced Materials and Manufacturing Technologies (AMMT) program is to accelerate the deployment of new materials and manufacturing technologies into advanced nuclear-related systems. Although 316H stainless steel fabricated by laser powder bed fusion (LPBF) has already been identified as an alloy that could have a significant effect on various reactor technologies, many other materials and manufacturing techniques are being evaluated. Nickel-based alloys typically offer higher-temperature capabilities compared with advanced stainless steels, and previous reports looked at three Ni-based alloy categories: low-Co alloys with a potential use close to the reactor core; high-temperature, high-strength alloys; and molten salt–compatible alloys. Due to the increasing interest in the low Co 625 alloy for nuclear applications, the alloy was compared at the Oak Ridge National Laboratory (ORNL) with 617, 230, two high-temperature, high-strength solution-strengthened alloys. Hot cracking could not be suppressed for alloy 230, and it was shown that these cracks, which were elongated along the build direction (BD), had a drastic effect on the ductility of alloy 230 at room temperature when specimens were machined perpendicular to the BD. Cracking was not as significant for alloy 617, but still led to significant variation in ductility at room temperature along the build direction. On the contrary, LPBF printing of crack-free alloy 625 was achieved using similar printing parameters, and the alloy looks like a very promising candidate for various reactor technologies.
36 MATERIALS SCIENCE↗
Magnetic susceptibility of Inconel alloys 718, 625, and 600 at cryogenic temperatures
After a hydrogen fuel bleed valve problem on the Discovery Space Shuttle was traced to the strong magnetization of Inconel 718 in the armature of the linear variable differential transformer near liquid hydrogen temperatures, the ac magnetic susceptibility of three samples of Inconel 718 of slightly different compositions, one sample of Inconel 625, and on sample of Inconel 600 were measured as a function of temperature. Inconel 718 alloys are found to exhibit a spin glass state below 16 K. Inconel 600 exhibits three different magnetic phases, the lowest-temperature state (below 6 K) being somewhat similar to that of Inconel 718. The magnetic states of the Inconel alloys and their magnetic susceptibilities appear to be strongly dependent on the exact composition of the alloy.
Goldberg, Ira B.↗
Near-Field Corrosion Interactions between Glass and Corrosion Resistant Alloys
This study explores the corrosion interactions between model nuclear waste glass materials and corrosion resistant alloys, under conditions that simulate the near field of a nuclear waste repository. The interactions between the corrosion of stainless steel (SS) 316, alloy G30, or alloy 625, and international simple glass or soda-lime silica glass are systematically studied. The dissimilar materials were exposed in close proximity to each other in different electrolytes at 90 oC. After exposure, the glass surface exposed near metals showed different regimes of corrosion, with distinct surface morphologies and chemistries that were likely affected by the local environment created by the localized corrosion of metals. Surface and solution analyses showed that the corrosion rate of glass was enhanced by the presence of metals. Infrared spectroscopy data suggested the local build-up of stresses in the contact area of glass, which may lead to the mechanical instability of the glass alteration layer. On the other hand, the effect of glass on metal corrosion is strongly dependent on the leaching solution. In electrolytes containing abundant aggressive anions such as Cl-, glass seems to suppress the localized corrosion of SS by the precipitation of a Si-rich surface film that protects the SS substrate from solutions. However, in less aggressive electrolytes, the corrosion rate of SS was increased by the presence of glass corrosion products. Overall, our study showed that the hidden and localized damage on glass in contact with metals may enhance the release of incorporated radionuclides compared to typical uniform glass corrosion. Therefore, more attention should be paid to the corrosion interactions between metals and glass under repository conditions.
Guo, Xiaolei↗
Investigation into Interfacial Mixing Behavior of Blown Powder Deposited Inconel 625-Copper Alloy Bimetallic for Improvement of Bimetallic Joint Strength
Use of laser blown powder, directed energy deposition (DED) provides a method for fabrication of bimetallic functional materials for high heat flux applications. However, this relatively new process currently has a low technology readiness level (TRL). Key to increasing the TRL is a better understanding of the factors that control mixing of the bi-metallic materials at the interface. The resulting interfaces in specimens obtained from three vendors were characterized using optical and electron microscopy with electron dispersion spectroscopy in addition to mechanical testing. Based on the range of process parameters and resulting interfaces, a model is being developed to predict interfacial mixing for development of specifications for improved and repeatable microstructures.
Additive Manufacturing↗