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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Report on FY 2021 testing in support of integrating Alloy 800H and 2.25Cr-1Mo for EPP analysis

The use of simplified Elastic-Perfectly Plastic (EPP) design methods avoids the complexities and limitations of the elastic analysis approach in evaluating strain limits and creep-fatigue damage for Class A metallic coolant boundary components in ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B. No stress classification and linearization are required in the EPP methods. Further, the strain limits and creep-fatigue damage evaluation procedures based on elastic analysis results were developed for the reactor operating conditions where uncoupled, rate-independent plasticity and stationary creep can be used to approximate the deformation in the component adequately. But at very high temperatures, the deformation is rate-dependent, and it can only be approximated adequately by unified viscoplastic model. Thus, for example, the elastic analysis approach cannot be used to evaluate strain limits and creep-fatigue damage for Alloy 617 above 650 °C. The EPP methods remove the temperature restriction. They can be used to evaluate strain limits and creep-fatigue damage for the full range of temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Fabrication and Testing of Two Passively Actuated Creep-Fatigue Surveillance Test Articles

This report describes the fabrication, testing, and analysis of two families of passively-actuated creep-fatigue test articles. These test articles induce cyclic mechanical load in a test section actuated solely through a mismatch in the thermal expansion coefficient of two materials and driven by a change in the specimen temperature. Test articles of this type could provide material degradation data for a materials surveillance program in future operating molten salt reactors, where variations in the reactor temperature caused by normal or upset cycles would provide the temperature change in the sample. The purpose of the current experimental campaign is to demonstrate the feasibility of fabricating realistically-sized test articles, test the articles under repeated thermal cycling to assess their robustness, particularly the reliability of the bimetallic welds, and collect strain data from instrumented samples to validate the methods used to size the test articles to match key features of the mechanical response of operating reactor components. The thermal cycling tests are ongoing but the test data to date suggests the current test article design is viable, robust, and can be successfully designed to match a target mechanical response. Finally, the report describes preliminary work on welding refractory alloys to 316H stainless steel. Replacing the test article driver material, currently Alloy 617, with a refractory would increase the coefficient of thermal expansion mismatch and support further miniaturization of the test articles.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Design and Optimization of a Gas-Cooled, Airfoil Fin Microchannel Heat Exchanger

High-performance microchannel heat exchangers are needed to supply heat for power conversion for nuclear microreactors. An airfoil fin microchannel design, constructed of Alloy 617 with helium as the working fluid, was analyzed and optimized using a design of experiments with artificial intelligence and machine learning techniques. The use of airfoil fins offers the potential to reduce pressure drop across the heat exchanger, as compared to other types of channel configurations. A framework for topology optimization of airfoil fin PCHEs has been developed that can be readily extended to different fin sizes and shapes, as well as different inlet and operating conditions, materials of construction, and working fluids. An optimization procedure was developed that employs computational fluid dynamics for a set of design points identified using Latin hypercube sampling. STAR-CCM+ was used to analyze a simplified two-channel configuration where five parameters were varied – inlet angle, fin scale, extent of staggering, transverse and longitudinal pitches. Two methods were compared for generating surrogate models – a 5D polynomial and a regression neural network. A response surface approximation was created from the surrogate models and input to a genetic algorithm. The genetic algorithm identified a set of optimal points on the Pareto front. The optimal geometry was found across six channel Reynolds numbers ranging from 1000 to 5000 to analyze how varying inlet conditions affects the optimal design. A set of optimal designs that maximizes heat transfer and minimizes pressure drop was identified, and a thermal stress analysis was performed on the optimal design. This work has developed a digital framework for the expedient topology design and evaluation of PCHE designs for gas-cooled microreactor applications. Correlations for the Nusselt number and Darcy friction factor were developed that can be useful for thermal hydraulic analyses using system codes. A thermal stress analysis was conducted and a brief discussion of the status of code cases of PCHEs for nuclear applications is given. Testing and thermomechanical modeling is needed to facilitate future code compliance of PCHEs for high pressure and high temperature applications.

42 ENGINEERING↗
DOE OSTI2023

ASME Section III, Division 5, High Temperature Reactors

Section III, Division 5 stands ready to support near-term deployment of advanced reactors. Progress on improving design rules, extending design lifetimes, adding more materials, and adding advanced component fabrication methods. Lessons-learned from Alloy 617 Code Case effort have streamlined the balloting workflow for Class A material code cases. After requester submitted material design parameters and supporting data package to ASME, Division 5 could turn around a material code case in about three Code Week cycles (less than a year). Data requirements for new materials are described in Division 5, “Nonmandatory Appendix HBB-Y, Guidelines For Design Data Needs For New Materials”.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A617 Notch Effect and Crack Growth Testing Status

This is an update on the Alloy 617 notched specimen creep testing and crack growth rate testing performed in FY23. Long-term notch testing continues, with some new results on double V-notch specimens. Creep-fatigue crack growth rate testing has begun and some preliminary results are shown.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Experimental and Analytical Verification of ASME Section IiII Division 5 Creep-Fatigue Design Rules

The continuous advancement of structural materials and the growing demands for more reliable and economical structural components in high-temperature reactor applications have necessitated the development of comprehensive design methodologies and design rules. Mechanical degradation of structural components at elevated temperatures subjected to cyclic deformation is controlled by the creep-fatigue damage. Over the past few decades, diligent research efforts have been dedicated to refining the development of elevated temperature design rules in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 5 and to develop conservative design rules that can effectively guard against the risk of creep-fatigue failure. In ASME Section III, Division 5, for a design to pass the creep-fatigue acceptance criteria, creep damage and fatigue damage are evaluated separately, and these damages must not violate the bi-linear creep-fatigue interaction diagram, i.e., the so-called D-diagram. The creep-fatigue damage evaluation procedure assumes that the effects of the actual cyclic loading sequence can be bounded by assuming that the individual loading cycles are uniformly distributed throughout the component design life. In this study, creep-fatigue experiments with variable amplitudes and loading sequencies were designed and performed on Alloy 617 at high temperatures. The results were analyzed to evaluate the loading history effect on creep-fatigue damage accumulation and to verify the assumptions for the creep-fatigue evaluation design rules.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Cluster Dynamics Modeling Needs for the Advanced Materials and Manufacturing Technologies Program

This milestone report aims to identify and assess the cluster dynamics (CD) modeling requirements within the Department of Energy's Office of Nuclear Energy (DOE-NE) Advanced Materials and Manufacturing Technologies (AMMT) program and to communicate these needs to the DOE-NE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The goal is to ensure NEAMS is well-informed about the CD modeling requirements to support AMMT's mission of accelerating the development, qualification, demonstration, and deployment of advanced structural materials and manufacturing for nuclear energy applications. CD modeling is an essential tool for predicting the degradation of structural materials under irradiation, which is a key component of AMMT's accelerated qualification process. The AMMT program focuses on both additively manufactured and wrought structural alloys, such as laser powder-bed fusion 316H austenitic stainless steel, alloy 709, Haynes 244, and alloy 617. These materials require a generalized CD modeling framework to facilitate rapid model development and computational simulation. A flexible, generalized CD software, similar to the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element framework, would enable modeling of various cluster types, including defect clusters, defect-solute clusters, and multicomponent clusters, incorporating thermodynamics and kinetics parameters. Radiation effects, microstructural feature evolution, and multi-dimensional modeling are critical considerations for the CD model. The usability of the CD code should allow for easy modification and coupling with MOOSE-based simulations. Additionally, the software should adhere to Nuclear Quality Assurance-1 standards, include a testing suite for verification and validation, and be version-controlled within a national laboratory-managed Git repository. Benchmark problems are needed to assess code predictions and performance.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

High Temperature Gas Cooled Reactor Materials

Overview of high temperature materials testing within the ART Gas Cooled Reactor Program. Covers 800H weldment creep testing and Alloys 617 and 709 creep and creep-fatigue crack growth rate testing.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Corrosion of SS 316, Inconel 617, Inconel 625 in molten chloride salt with emphasis on the speciation of Cr, Ni, and Co in molten salt matrix by XAFS

Dr. Hartmann’ report summarizes the corrosion behavior of materials used in the construction and operation of molten salt reactors which is is of great interest to the molten salt reactor (MSR) community. In the present study we intent to decipher the very elemental phenomena on the corrosion of austenitic chromium-containing stainless steel SS316, as well as Inconel 625 and ASME-coded high temperature Alloy (Inconel) 617. We initiated and are in continuation of conducting ampoule type corrosion testing at 550 °C and 650 °C to determine chromium diffusivity and the related activation energy as well as the speciation of the corrosion products.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Elucidating the corrosion mechanism of commercial Ni-based Superalloys in UCl3 containing-chloride Molten Salt Systems

Molten salt reactors (MSRs) have gained renewed interest, providing several advantages over their predecessors, including the capability to consume spent fuels, enhancing the environmental sustainability of the uranium fuel cycle. For example, molten chloride fast reactors (MCFRs) can reach criticality with molten chloride spent fuel containing high concentrations of impurities, such as actinide products like uranium chloride (UCl3). However, the redox potential of chloride molten salt fuels may change in the presence of these impurities, dictating their corrosivity and in turn the corrosion performance of structural components, such as those constructed from nickel (Ni)-based alloys. The purpose of this investigation is to assess the extent of corrosion of Ni-based alloy, Inconel 617, when exposed to UCl3-LiCl-KCl eutectic salt. Inconel 617 one of only six structural materials that are fully qualified by the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for high-temperature nuclear reactor components, making it a technologically mature material to consider for constructing MCFRs. Inconel 617 specimens were submerged in a static LiCl-KCl-UCl3 eutectic salt mixture heated at 700 C for 1000 h under an inert atmosphere. Upon completion, the extent of corrosion was analyzed through a multi-modal characterization approach spanning the engineering to nanoscale, employing computed tomography, focused-ion beam, and transmission electron microscopy techniques. Results from this investigation will enhance our understanding of property-to-performance relationships of candidate structural materials for MSRs with respect to corrosion resistance and interactions between the salt and alloy interface.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Assessing the Impact of Molten Halide Salts on Creep of Structural Alloys at 650°-750°C

There is a critical lack of data on the mechanical behavior of candidate structural materials for advanced nuclear reactors under molten halide salt environments. Limited legacy data from the molten salt reactor experiment (MSRE) program showed a significant reduction in creep rupture strength of a Ni-base alloy in molten fluoride salt. With ongoing efforts to commercialize different molten salt reactor concepts, the industry can considerably benefit from quantitative information on the impact of molten halide salts on the engineering properties such as creep and fatigue strength of materials of interest. The present work aims to assess the role of molten salt corrosion on the creep behavior of three alloys 316H, 617 and 282 at 650-816 °C. Creep tests were conducted in fluoride (FLiNaK) and chloride (NaCl-MgCl2) salts. Initial results from the ongoing testing will be presented which suggest that the molten salt environment caused a 25-50% reduction in creep rupture lifetime compared to air exposures. Physics-based corrosion and creep models were employed to gain some insights into the potential degradation mechanisms.

Pillai, Rishi↗
DOE OSTI2024

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↗
NASA NTRS1993

Effect of Long-Term 1093 K Exposure to Air or Vacuum on the Structure of Several Wrought Superalloys

Long-term 1093 K heat treatments of three commercial superalloy sheet materials were undertaken in air -IAI and vacuum. With either exposure, significant precipitation of second phases occurred in the Co-base Haynes(R) Alloy 188 (HA 188) and the Ni-base Haynes(R) Alloy 230 (HA 230); however, much less precipitation was found in the exposed Ni-base alloy Inconel(R) 617 (IN 617). Although some grain growth occurred in HA 198, no changes in the grain size of either HA 230 or IN 617 were observed after 22,500 h at temperature. Oxidation during air heat treatments led to weight gain due to the formation of chromia + spinel scales and surface-connected grain boundary pits/oxides in all three superalloys. Both the weight gain and depth of intergranular attack were dependent on the square root of time, which is indicative of diffusion-controlled phenomena. Because many alloy samples had neighbors in close proximity, most vacuum heat treated specimens did not suffer significant loss of volatile elements. However, some exposed samples were subjected to unrestricted vacuum heat treatments, allowing estimates of volatilization to be made. Based on the data for HA 188, the weight loss during 1093 K vacuum exposure was diffusion controlled once the inhibiting effects of surface films on the as-received alloys were broken down.

J.D. Whittenberger↗
DOE OSTI2024

High-Temperature Oxidation Behavior of Wrought and Additive Manufactured Ni-based Alloys in Direct-Fired Supercritical CO2 Power Cycle Environments

Materials selection is a key concern for corrosion resistance in elevated temperature and pressure direct-fired supercritical CO2 (sCO2) power cycles. Herein, the wrought product form of alloy H282, as well as several additional nickel-based commercial alloys (230, 263, 617, 625, 740H), were exposed to direct-fired supercritical CO2 conditions (95CO2 – 4H2O – 1O2) at 750 °C and 20 MPa for over 2,000 h. Additionally, additive manufactured (AM) 282 produced by laser powder bed fusion (LPBF) were also included in the exposures. This paper focuses primarily on the differences observed in the degradation behavior of the wrought and additively manufactured materials. AM samples in the as-printed state were found to form less protective oxide scales resulting in higher oxidation rates compared to wrought samples. However, applying a 600 grit surface finish to the AM samples modified the oxidation behavior to closely resemble that of the wrought material. The AM samples formed Mo-rich carbides in the alloy beneath the oxide following sCO2 exposure, implying potentially higher levels of carbon uptake relative to the wrought material. Conversely, the AM samples seemed to be less affected by water-assisted oxide volatilization compared to the wrought material. The results are discussed in terms of the potential compatibility issues that may arise when using Ni-based alloys in the hot portions of direct-fired sCO2 power cycles, particularly in the case of thin-walled components.

Carney, Casey↗
DOE OSTI2023

High-Temperature Corrosion of Chromia-Forming Ni-Based Alloys in CO 2 Containing Impurities

Future technologies require structural materials resistant to environmental degradation in high-temperature CO 2 -rich environments. In this work, we exposed several commercially available Ni-based alloys (230, 263, 282, 617, 625, and 740H) to atmospheric pressure gases intended to simulate the compositions expected in future direct-fired supercritical CO 2 power cycles. The alloys were exposed to 95% CO 2 + 4% H 2 O + 1% O 2 and the same gas containing 0.1% SO 2 at temperatures of 600, 650, 700, 750, and 800 °C for 2500 h. With minor exceptions, chromia scales formed on all alloys at all temperatures in the SO 2 -free gas, yielding parabolic growth rates that followed an Arrhenius temperature dependence. Behavior in the SO 2 -containing gas was more complex. Generally, the alloys performed well at temperatures of 650, 750, and 800 °C. While some alloys further performed relatively well across the whole temperature range, several of the alloys experienced chromia failure resulting in non-protective duplex oxide scales and high oxidation rates, at temperatures of 600 and 700 °C. Deviation from protective behavior was associated with internal sulfide formation and, additionally for the case of 600 °C, external sulfate formation. Extensive carburization accompanied growth of the non-protective duplex oxide scales, which made it more difficult for the alloy to recover after initiation of the sulfur-induced accelerated corrosion process. The thermodynamic and kinetic factors influencing the accelerated corrosion in the presence of sulfur are discussed. The results suggest that caution is required when assessing compatibility of Ni-based alloys for CO 2 -based systems when sulfur-based impurities are expected.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermo-mechanical behavior of hypoeutectic Ni-Y-Zr alloys

Microstructure refinement and optimized alloying can improve metallic alloy performance: stable nanocrystalline (NC) alloys with immiscible second phases, e.g., Cu-Ta, are stronger than unstable NC alloys and their coarse-grained (CG) counterparts, but higher melting point matrices are needed. Hypoeutectic, CG Ni-Y-Zr alloys were produced via arc-melting to explore their potential as high-performance materials. Microstructures were studied to determine phases present, local composition and length scales, while heat treatments allowed investigating microstructural stability. Alloys had a stable, hierarchical microstructure with ~250 nm ultrafine eutectic, ~10 µm dendritic arm spacing and ~1 mm grain size. Hardness and uniaxial compression tests revealed that mechanical properties of Ni-0.5Y-1.8Zr (in wt%) were comparable to Inconel 617 despite the small alloying additions, due to its hierarchical microstructure. Here, uniaxial compression at 600 °C showed that ternary alloys outperformed Ni-Zr and Ni-Y binary alloys in flow stress and hardening rates, which indicates that the Ni 17 Y 2 phase was an effective reinforcement for the eutectic, which supplemented the matrix hardening due to increased solubility of Zr. Results suggest that ternary Ni-Y-Zr alloys hold significant promise for high temperature applications.

36 MATERIALS SCIENCE↗