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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

An approach to combine neutron and ion irradiation data to accelerate material qualification for nuclear reactors

Next-generation nuclear power plants are generally characterized by higher operating temperatures, increased neutron fluences and energies, and distinct corrosive coolant environments versus the existing light water reactor fleet. Whether using existing materials in new environments, newly developed materials tailored for these environments, or new manufacturing methods, the traditional decades-long approach for materials qualification does not facilitate rapid deployment. Ion irradiation has demonstrated success in reproducing material microstructure and select property evolution resulting from neutron irradiation with three to four orders of magnitude reduction in time and cost, making it an ideal candidate for accelerated irradiation testing. Because microstructure has a large impact on bulk material properties, limited neutron irradiation data at lower damage levels can in principle be combined with accelerated ion testing results and modeling and simulation to form an accurate prediction of microstructure evolution and select properties under different neutron irradiation conditions and at higher damage levels. The objective of this work is to present a conceptual framework of specific steps to fulfill several technical challenges associated with qualifying materials for performance in radiation environments on an accelerated time frame. A brief review of the regulatory landscape for materials in nuclear environments is presented, followed by additional overviews to understand the current state of the art for correlation of materials properties across radiation environments using experimental and computational methodologies. Finally, the roles of academia, national laboratories, and industry in the advancement of this accelerated materials qualification framework are discussed as a path forward, with possible case studies presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at The LWR Relevant Temperature

Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) on the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Based on a comprehensive microstructure and property screening performed in Phase-1 of the ARRM program, a total of five alloys, together with 316L and X-750 as references, were down-selected for Phase-2 neutron irradiation studies. Because thermal aging could exert a synergistic effect on neutron irradiation due to the low neutron damage rate on the order of 10 –7 displacements per atom per second (dpa/s), Grade 92 (one of the five down-selected alloys) and two heats of 316L were selected in this task to study the effect of aging at 350°C for 12.5–12.7 kh on microstructure and mechanical properties. The aging time is approximately corresponding to 5 dpa neutron irradiation, which can be served as a reference for the 5-dpa-irradiated samples of the alloys to help understand the independent neutron damage influence on microstructure and mechanical properties. Optical microscopy and scanning electron microscopy were used for microstructural characterization. Hardness, tensile, Charpy impact toughness, and fracture toughness in the ductile regime of the aged samples were examined. The aging of Grade 92 led to the formation of many Laves phase in sizes of ~100–200 nm and resulted in some reduction in hardness and yield/ultimate tensile strength with some increases in uniform and total plastic elongations, which may have helped the enhancement of Charpy impact toughness, e.g., ~4 J increase in upper-shelf energy and 20.5°C reduction in ductile-brittle transition temperature compared with the unaged condition. The fracture toughness of the aged Grade 92 showed decent toughness of ~303 MPa√m (K Jq ) with ~87 tearing modulus at 22°C, which reduced to ~241 MPa√m (~20% reduction) with ~75 tearing modulus at 300°C. The two heats of 316L are differentiated by their amounts of δ-ferrite, i.e., ~1 vol% in heat T1103 and ~4 vol% in heat N5B8. The two heats of 316L were subjected to 15% cold work (CW) as this condition is often used in nuclear reactors to help trapping radiation-induced defects and thus postpone the steady state swelling stage. The aging did not result in any noticeable microstructure changes, except for possible segregations at grain boundaries that need to be further investigated. The aging resulted in slight reduction in hardness of 316L-T1103 and slight increase in hardness of 316L-N5B8, which is consistent with their tensile testing results. The 15%CW might have introduced some inhomogeneity, resulting in large standard deviations in hardness, which were reduced after the aging. The aged 316L followed the same trend as the unaged condition for the strength and elongation results from the tensile testing. Minimum elongations appeared within ~300–600°C, above which the elongations seem to increase. Unlike the negligible or minor changes in microstructure, hardness and tensile results of the two heats of 316L, their fracture toughness showed noticeable difference. The aged 316L-T1103 showed good toughness of ~350 MPa√m with ~80 tearing modulus at 22°C, which reduced to ~271 MPa√m (~23% reduction) with ~83 tearing modulus at 300°C. In contrast, the aged 316L-N5B8 showed lower toughness of ~244 MPa√m with ~53 tearing modulus at 22°C, which reduced to ~180 MPa√m (~26% reduction) with ~31 tearing modulus at 300°C. The preliminary results indicate that the presence of high volume of δ-ferrite would noticeable impair fracture toughness although it may not influence hardness and tensile properties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

A simplified approach for creep damage and remaining life calculation for materials surveillance in advanced reactors

High temperature reactor materials will experience a combination of mechanical degradation caused by creep and fatigue and environmental degradation caused by neutron irradiation and, in some cases, exposure to corrosive coolants. A materials surveillance program is one option for ensuring the safe, reliable operation of key reactor components under these conditions. Such a program would monitor the degradation in key material properties over time, using this data to predict how changes in the material properties affect components performance. This could then be used to make plant operational decisions and support eventual plant license extensions. Past work de-scribes such a material surveillance program based on passively actuated mechanical test articles that impose creep-fatigue type loading on test materials driven only by changes in temperature, for example those experienced by components under standard operating cycles. This report focuses on two aspects of the proposed material surveillance program: how to design the test articles to mimic the mechanical response of the corresponding component and how to use ex-situ test data to deter-mine the amount of damaged experienced by the test article (and hence the corresponding compo-nent) in service. Specifically, this report derives and validated simplified methods for both tasks, replacing earlier, more complicated approaches based on simulating the test article response and finding the best specimen design or current damage via complex numerical optimization. The sim-plified methods for each task developed here can be implemented in spreadsheet software and are simple enough for practical use in future operating plants. This report derives the methods (one for sizing and two options for damage inference), verifies the simplified approaches versus more so-phisticated methods, and compares the results of applying the simplified methods to previous re-sults using the complex numerical optimization approach. The report also provides a worked sam-pled problem applying the simplified techniques to a realistic high temperature reactor component.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Surface-Tailoring and Morphology Control as Strategies for Sustainable Development in Transport Sector

Surface wetting plays an important role in the corrosion protection processes of aerospace applications. Here, we demonstrate the use of ultrafast femtosecond (fs) laser processing techniques to tailor the wetting properties of aluminum (Al) substrates by creating diverse surface morphologies. Specifically, two distinct laser scanning methods—dot-hatching and cross-hatching—were employed to fabricate microstructures on the substrates. By varying the incident laser parameters, we confirmed that the resulting surface morphologies exhibit different wetting behaviors, spanning from hydrophilicity to hydrophobicity. Furthermore, time-resolved spreading tests validate that dynamic wetting behaviors can also be modified. This fs laser processing approach provides a straightforward, one-step fabrication method for effectively modifying the wetting properties of Al alloys.

de Almeida Prado, Luis Antonio Sanchez↗
DOE OSTI2024

Long-Term Performance of Reference Electrodes in Alkaline Radioactive Waste Storage Environments

Accurate measurements of corrosion potential are important for assessing the likelihood of internal localized corrosion and stress corrosion cracking of carbon steel tanks used for storing radioactive wastes. Reference electrodes in underground radioactive waste storage tanks are challenging to deploy, and more difficult to extract and replace frequently due to radiological exposure and disposal constraints. Hence, electrodes that exhibit stable performance over long periods of immersion in these waste environments are desirable. The present study evaluates the stability of reference electrodes used in radioactive waste storage tanks over a much longer period than previously studied. Long-term tests on Ag/AgCl and Hg/HgO reference electrodes were performed in nonradioactive simulants formulated from wastes stored at the Hanford site. Electrode degradation, which was studied by various in situ and ex situ evaluation techniques, was correlated to changes in electrode fill chemistry from waste intrusion via the porous frit junction. An intentional contamination study was performed to better understand and predict contamination effects on electrode potential drift.

Materials Science↗
DOE OSTI2021

Modeling Summary of ASNF in DOE Sealed Standard Canisters

A pathway for road-ready and final disposition packaging configurations for the aluminum-clad spent nuclear fuel (ANSF) fuel dictates storage within helium backfilled sealed DOE standard canisters. The typical packaging configuration for the 15-foot DOE canisters places 10 advance test reactor (ATR) elements within a Type 1a basket, and three baskets are loaded within each DOE canister. During in-reactor operations and cooling pond storage conditions, oxyhydroxide layers form on the surface of the aluminum clad fuel. These layers produce hydrogen gas over time due to the fuel’s radiation field. As part of the packing procedure, the ATR fuel should be dried to remove any residual physio-/chemi- sorbed water bound to the surface. The results of this modeling will include results at fully saturated and fully dried conditions. In addition, fuels that are currently stored at the Savannah River Site were also studied for their potential for hydrogen and pressure build up. These two additional fuels modeled were the Missouri University Research Reactor (MURR) fuel, which is packaged in the same configuration as the ATR, but with a 10-foot-tall DOE standard canister. This was selected due to its relatively high decay heat compared to other DOE-managed ASNF. The second additional fuel studied with the modeling effort was the High Flux Isotope Reactor (HFIR) fuel. This fuel is modeled as two separate DOE canisters with the inner and outer annulus split for storage. The HFIR was selected for study due to its high aluminum cladding surface area. In the associated experimental work, Task 2 - Oxyhydroxide Layer Radiolytic Gas Generation Resolution, additional tests were completed in a helium environment, and updated G-values for the radiolytic production of hydrogen from the oxyhydroxide layers were provided. These values were 2.92 ×10 -4 µmol/J at 50% relative humidity and 4.12 ×10 -4 µmol/J at 100% relative humidity. These values are lower than the value for Argon that was used in prior modeling results. In addition, prior modeling results have been completed with the G-value applied to just the mass of the corrosion layer, and this has been updated to apply the G-value for hydrogen generation to the full mass of the fuel. These two effects combine to show much smaller pressure and hydrogen build up for the sealed canister model. For a nominal scenario of stored ATR fuel, after 50 years the model results give a 1.36 atm total pressure, 7% mole percent hydrogen, for the upper decay heat, 1.51 atm total pressure, 16% mole percent hydrogen, and for upper decay heat with undried fuel 2.6 atm total pressure, 15% mole percent hydrogen. For the MURR nominal case, the model results give 1.34 atm and 6% hydrogen, for upper decay heat this gives 1.41 atm total pressure with 10.8 % hydrogen, and for upper decay heat with undried fuel, this gives 2.38 atm total pressure with 9.9% hydrogen. The nominal scenario for HFIR fuel gives 1.39 atm total pressure with 9.9% hydrogen, the upper decay heat case gives 1.43 atm with 12.1% hydrogen, and the upper decay heat with undried fuel gives 2.17 atm total pressure with 11.9% hydrogen. These results confirm the ATR scenario bounds the other intact ASNF modeled here. No case modeled yields significant oxygen, and the lower decay heat cases for all fuels modeled have hydrogen concentrations that are under the 4% flammability limit after 50 years of storage. In addition, the modeled pressures for all cases are all significantly below the 500-psi limit for the DOE standard sealed canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Editors’ Choice—Examining Performance and Durability of Anion Exchange Membrane Fuel Cells with Novel Spirocyclic Anion Exchange Membranes

A series of spirocyclic copolymer membranes with varying ion exchange capacities (IECs) were investigated to probe the impact of polymer properties on in situ fuel cell performance and stability. In-situ electrochemical tests and post-mortem electron microscopy analysis of cross-sectioned membrane electrode assemblies (MEAs) have been combined with voltage loss breakdown analysis to evaluate the performance and degradation of different MEAs, and to probe the catalyst morphology and electrode structure at different stages of operation. Voltage loss breakdown results show that membrane degradation and kinetic losses played only a minor role in observed performance degradation and that performance losses were primarily related to increasing mass transport losses. From microscopy studies, carbon corrosion and Pt nanoparticle growth were identified at both the cathode and anode although more pronounced on the cathode resulting in significant structural changes. The membrane with the lowest IEC (1.3 mmolg −1 ) demonstrated the lowest peak power density ~ 1.16 W cm −2 , however, it showed the most stable performance (constant 0.6 A cm −2 hold) with ~ 5% degradation over 540 h. Isolation of performance losses and microscopic analysis of electrodes for anion exchange membrane fuel cells has not been reported previously, and these results help identify critical performance degradation concerns.

25 ENERGY STORAGE↗
DOE OSTI2021

Evaluation of Formate Salt PCM’s for Latent Heat Thermal Energy Storage

This work examines formate salts as potential phase change materials (PCMs) for middle-high temperature (≤250 °C) latent heat thermal energy storage applications. The thermophysical properties of three formate salts were characterized: pure sodium formate and binary blends of sodium/potassium formate and sodium/calcium formate. The stability of formate PCM’s was evaluated by thermal cycling using differential scanning calorimetry where sodium formate and sodium/potassium formate appeared stable over 600 cycles, while sodium/calcium formate exhibited a monotonic decrease in heat of fusion over the test period. A longer test with sodium formate led to gas release and decomposition of the salt. FTIR analysis of the PCM showed degradation of formate to oxalate. T-history experiments with 50-g PCM quantities demonstrated a bulk supercooling of only 2–3 °C for these salts. Thermal conductivity enhancement of over 700% was achieved by embedding aluminum in the solid PCM. Finally, mild carbon steel was immersed in molten sodium formate for up to 2000 h. Sodium formate was found to be non-corrosive, as calculated by mass loss and confirmed by cross-sectional high-resolution microscopy. FTIR analysis of the PCM after 2000 h shows oxidation at the free surface, while the bulk PCM remained unchanged, further indicating a need to protect the formate from atmospheric exposure when used as a PCM.phase change materials; formate salts; latent heat thermal energy storage; thermal cycling; supercooling; thermal conductivity enhancement; corrosion

25 ENERGY STORAGE↗
DOE OSTI2022

MSR Salt Spill Accident Testing Using Eutectic NaCl-UCl 3

Assessing the potential consequences of identified accident scenarios is an essential part of the licensing process for a new nuclear reactor and is achieved by using accident progression models. A likely accident scenario that will be evaluated by developers of molten salt reactors (MSRs) is a spill of radionuclide-bearing fuel salt onto the reactor containment floor (i.e., a salt spill accident). Models to determine the consequences of a molten salt spill accident need to be parameterized and validated using experimental data, but little experimental effort has been dedicated to fill these data gaps to date. Specifically, data is needed to enable model development for individual processes (e.g., spreading, heat transfer, corrosion, radionuclide vaporization, and aerosol generation) that quantifies the sensitivities towards the initial conditions of the spill, the ambient environment, and the features of the containment. In addition, integrated experiments that simulate salt spill accidents will need to be conducted to provide insight into coupled processes and data for model validation, but these experiments will require the use of proven methods to quantify the processes under evaluation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Inverse Consequences of the SnO 2 Protection Layers on Pt/C Catalysts in Proton-Exchange Membrane Fuel Cells

Proton-exchange membrane fuel cells (PEMFCs) are promising energy-conversion systems, offering an appealing blend of high energy efficiency and low environmental impact. However, carbon corrosion of PEMFCs is known to significantly degrade their performance, remaining a critical challenge to overcome. In this study, we applied a Nb-doped SnO 2 (Nb-SnO 2 ) nanoparticle coating on Pt/C catalysts as a protective layer, with the Sn/C ratio in the precursors varying from 0.25:1 to 2.0:1. Contradictory behaviors of the coated Pt/C catalysts were observed at different Sn/C ratios. The Sn/C = 1.0 sample exhibited improved electrochemically active surface area retention after 500 cycles of accelerated stress testing (AST) but with more significant polarization and resistance increase observed in the polarization curves. In addition, agglomeration of Nb-SnO 2 particles was observed at a higher Sn/C ratio in the AST of a membrane electrode assembly, with less shrinkage of the total thickness of the Nb-SnO 2 -coated Pt/C electrode. We speculate that formation of Nb-SnO 2 agglomerates occurs once the protective layer is broken down or the unprotected carbon surface is corroded and that these Nb-SnO 2 agglomerates increase the tortuosity of the electron pathways and significantly increase the cell polarization.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2022

Multi-Modal Characterization of Nuclear Fuels and Materials at the Idaho National Laboratory Materials & Fuels Complex

The Idaho National Laboratory (INL) leads cutting-edge research pertaining to the advancement of nuclear reactor technologies, including nuclear fuels and materials. The INL Irradiated Materials Characterization Laboratory (IMCL), Electron Microscopy Laboratory (EML), and future Sample Preparation Laboratory (SPL) are available to the nuclear research community to assess the behavior of nuclear fuels and materials, efficiently and comprehensively characterizing from the engineering to atomistic scale. The IMCL is a unique, 12,000-square-foot facility located at the INL Materials and Fuels Complex designed for analysis of irradiated materials. The facility operates advanced characterization instruments that are sensitive to vibration, temperature, and electromagnetic interference in modular radiological shielding and confinement systems, granting researchers the ability to assess the microstructural, chemical, mechanical and thermophysical properties of nuclear materials, especially irradiated fuels. The EML is dedicated to advanced characterization of materials with optical and electron microscopy tools, including scanning electron microscopy/focused-ion beam (SEM/FIB) and transmission electron microscopy (TEM). Upon construction, the SPL will be a 3 story, 49,000 sq. ft facility, that is the most modern reactor structural materials testing and analysis facility in the world, designed to investigate reactor structural materials in support of life-extension programs and development of advanced reactor concepts, including mechanical testing and advanced characterization capabilities. This presentation will showcase some of the main capabilities available at both IMCL, EML, and SPL, specifically illustrating how these characterization techniques are incorporated into multi-modal characterization work scopes to elucidate the degradation of nuclear structural materials and irradiated fuels.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Second Annual Report on Development of Microwave Resonant Cavity Transducer for Fluid Flow Sensing

We are investigating a microwave resonant cavity transducer for flow sensing in the vessel of a high temperature fluid advanced reactor (AR), such as a molten salt cooled reactor (MSCR) or a sodium fast reactor (SFR). This transducer is a hollow metallic cylindrical cavity, with the flat wall of the cylinder flexible enough to undergo microscopic deflection due to dynamic fluid pressure. Membrane deflection leads to a shift in the resonant frequency, which can be detected with a spectrum analyzer. Because the transducer is intended for immersion in a high temperature corrosive fluid, understanding of material degradation is crucial for estimation of transducer performance lifetime, and development of measurement interpretation algorithms. We conducted a preliminary computational investigation of relevant damage mechanisms of a stainless steel 316 cylindrical resonator in FLiBe salt. The two main damage mechanisms, creep and corrosion, were modeled using multiphysics COMSOL software. Degradation was modeled for a temperature range 500°C to 700°C. Coupling of the damage mechanisms was not considered. These models predict significant inelastic deformation at most temperatures due to creep, and qualitatively predict chromium depletion both along the liquid/solid interface and along the grain boundaries. An algorithmic approach for compensation of these degradation effects during fluid flow measurements will be developed in the future work. To validate sensor physics, we have performed proof-of-principle test of flow sensing in water. For this test, we have developed a cylindrical resonator for K-band, which was machined from brass. The cavity was excited through WR-42 waveguide through a subwavelength hole on the side of the wall of the cylinder. To increase the spectral signal visibility, we developed a signal processing method for baseline subtraction. A flow loop for proof-of-principle test of transducer performance in water was assembled. A commercial flow meter was installed in the loop for reference measurements. Cylindrical cavity was excited in the TE 011 mode with resonant frequency f ≈ 17.8GHz. Frequency shift of cavity spectral response was obtained by gradually increasing water flow rate from 0 to 60gpm. Corresponding monotonic increase of resonant frequency shift by several MHz was observed. Approximate figure of merit of sensitivity to flow rate is 100KHz/GPM. In addition, we have identified an existing liquid sodium experimental setup for demonstration of flow sensing in environment similar to that of an advanced reactor. The setup consists of a cylindrical vessel and center feed line, where transducer inserted through the lid will measure velocity of the impinging liquid jet. As a calibration experiment, we have assembled a water vessel with center feed with the same dimensions as those of the liquid sodium setup. We have also developed and insertion probe consisting of a 50cm brass waveguide enclosed in protective SS316 tube. Using the water loop, we have demonstrated feasibility of sensing the impinging liquid jet in the vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

LandRAY Final Report

The overarching Project objective is to demonstrate the feasibility of using an innovative Power Take-Off (PTO) Module, to be used in the Columbia Power Technologies, Inc. (C·Power) utility-scale wave energy converter (WEC). The PTO Module uniquely combines a large-diameter, direct-drive, rotary permanent magnet generator; a patent-pending rail-bearing system; and a corrosion-resistant fiber-reinforced-plastic structure. The rail bearing system allows for: • A flexible non-rigid and low-cost generator frame as compared to conventional machines. • A reduction of air gap and consequently lower cost electromagnetic design • Ability to reduce costs by increasing diameter to much larger dimensions, while maintaining the small air gap. • A direct drive electro-mechanical solution that is the fastest and most precise approach for responding to the dynamic variability of forces from wave energy. The land-based test of the PTO Module (DDP1) will provide an important risk reduction phase, to demonstrate the reliability and feasibility of a large-diameter, small air-gap generator that is capable of delivering high efficiency, fault-tolerance and relatively simple "at-sea" maintenance capabilities – qualities that are essential for cost-effective wave energy devices. This is the first known test of such a system.

16 TIDAL AND WAVE POWER↗
DOE OSTI2022

CISCC Countermeasures for Spent Fuel Canisters in Concrete Casks

Aging management of spent fuel is an important issue worldwide. Chloride-induced stress corrosion cracking (CISCC) of welded canisters is particularly important for concrete casks, where it occurs in response to a combination of environmental, material, and residual stress factors. In this paper, we introduce a countermeasure to prevent the occurrence of CISCC by adjusting the flow rate of the cooling air into the concrete cask so that the canister surface temperature does not decrease below 100°C, which is a conservative deliquesce temperature threshold for CISCC occurrence. The flow adjustment device was installed in a cask model, and two tests were conducted under the heat rate conditions corresponding to those of an actual canister. In the first test, the device was operated under a condition where the canister surface temperature was below 100°C to verify whether it could increase the canister surface temperature to above 100°C. In the second test, the heat rate was decreased from a state where the canister surface temperature was above 100°C, to verify whether the device could maintain the canister surface temperature above 100°C. Good results were obtained in both cases. However, in the case of a low decay heat rate of spent fuel after extended dry storage, the temperature of the canister surface may decrease below 100°C even if no air supply is provided. Therefore, in this case, it may be necessary to take measures such as blocking air outlets.

Takeda, Hirofumi↗
DOE OSTI2022

Microstructural and nanomechanical studies of PVD Cr coatings on SiC for LWR fuel cladding applications

The microstructure and nanomechanical properties of pure Cr coatings deposited on SiC with various physical vapor deposition (PVD) methods have been investigated for the mitigation of hydrothermal corrosion of SiC-SiC f (SiC fiber-reinforced SiC matrix) composite fuel cladding in light water reactors (LWR). Cr coatings (4–7 μm thick) were deposited on unbiased SiC substrates using six variants of magnetron sputtering processes: (i) standard (S-DCMS), (ii) pulsed (P-DCMS), (iii) ion-assisted (I-DCMS), and (iv) pulsed ion-assisted direct current magnetron sputtering (PI-DCMS), (v) high-power impulse magnetron sputtering (HiPIMS), and (vi) bipolar HiPIMS (B-HiPIMS). Microstructural characterization and nanoindentation testing were used to evaluate effects of deposition technique on coating microstructure and nanomechanical properties. The application of positive reverse pulses to the sputter target induced coarsening of the columnar grain structure, while ion bombardment of substrates during deposition promoted densification of the coating microstructure. Both effects are exemplified in B-HiPIMS deposition, resulting in a high-density microstructure with compressive residual stress. Material pile-up around nano-indents and plastic work during nanoindentation were analyzed to understand deformation behavior of the coatings. Finally, the results suggest the B-HiPIMS process to be among the promising methods for the deposition of Cr coatings on SiC-SiC f for LWR cladding application.

36 MATERIALS SCIENCE↗
DOE OSTI2020

A Hybrid AI/ML and Computational Mechanics Based Approach for Time-Series State and Fatigue Life Estimation of Nuclear Reactor Components

Environmental fatigue modeling is a complex problem due to multiple failure modes and their intermixing. The failure modes are function of various underlying causes in addition to the corrosive effect of reactor coolant environment. Some of the major causes are time-dependence of material associated with cyclic loading, load sequence effect associated with random/variable amplitude loading, effect of strain amplitude and rates, effect of varying temperature (along both temporal and spatial directions) and the effect of mean strain and stress. The nonlinear intermixing of failure modes associated with above mentioned causing parameters makes the environmental fatigue modeling is a challenging task. Because of this challenge, fatigue is traditionally being modeled based on experimental data. However, test based empirical approach often requires hundreds of fatigue tests to model the above-mentioned intermixing failure causes even for a single material system. The problem is further exaggerated for reactor component made from multi-material systems such as made from both carbon and stainless-steel base metals and their similar and dissimilar metal welds. With the difficulty of conducting hundreds of fatigue tests to capture the above-mentioned intermixing failure causes, fatigue modeling approaches often depends on empirical models based on limited available test data such as available through ASME code and NUREG 6909. However, these limited test-data-based models may not be enough to accurately predict the life of reactor components. Accurate prediction of life of reactor component would become a necessity, particularly when the license of the reactors to be extended for long-term-operation (LTO) that is for well beyond its original design life of 40 years. The requirement of extending the license of reactor under LTO requires hundreds of fatigue tests to be conducted to understand the mechanism associated with the above-mentioned interdependent failure causes. However, conducting large number of fatigue tests is not a feasibility due to the cost involved. To address this issues Argonne National Laboratory (ANL) with the sponsorship of DOE Light Water Reactor Sustainability (LWRS) program trying to develop a hybrid predictive modeling approach. This is based on limited experiment-data, Artificial-intelligence (AI) – Machine-Learning (ML) - Deep-Learning (DL) based techniques and Multiphysics-computational-mechanics based modeling tools. The hybrid approach not-only can improve the accuracy of the existing stress analysis and fatigue modeling approach but also can reduce the over-dependency on test-based approach. Towards this goal following are some of the major contributions based on ANL’s FY-20 environmental fatigue modeling activities: 1) A cyclic plasticity material model database for 82/182 dissimilar metal weld, which can be readily shared with US nuclear industry and regulatory agency on request. 2) A well validated analytical modeling methodology to perform cycle-by-cycle stress prediction under both constant amplitude fatigue loading and variable amplitude fatigue loading (with load-sequence effect). 3) An AI/ML/DL based methodology to predict unmeasurable cyclic strain based on other available sensor signals. This type of approach can be used for estimating strain in real reactor components from other sensor readings. 4) An AI/ML based approach to improve the US capability on environmental fatigue testing. This is by improving ANL’s existing environmental fatigue testing capacity to conduct ASME required strain-controlled tests (by controlling strain amplitudes and its rate), while not measuring the strain (due to the difficulty of placing an extensometer in a narrow autoclave in a PWR-water-test system). 5) A simulation and experiment based probabilistic modeling methodology for time-series fatigue state and life estimation of reactor metal such as dissimilar metal weld.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Long-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at The LWR Relevant Temperature

Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer periods of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) initiated a collaboration with the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program on the Advanced Radiation Resistant Materials (ARRM) program, focused on developing and test degradation resistant alloys for light water reactor (LWR)- relevant environments. Based on a comprehensive microstructure and property screening, the ARRM program selected a total of five alloys (i.e., Grade 92, 310, 690, 718A, and 725), together with 316L and X-750 as references, for further investigations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Comprehensive Evaluation for Protective Coatings: Optical, Electrical, Photoelectrochemical, and Spectroscopic Characterizations

Numerous efficient semiconductors suffer from instability in aqueous electrolytes. Strategies utilizing protective coatings have thus been developed to protect these photoabsorbers against corrosion while synergistically improving charge separation and reaction kinetics. Recently, various photoelectrochemical (PEC) protective coatings have been reported with suitable electronic properties to ensure low charge transport loss and reveal the fundamental photoabsorber efficiency. However, protocols for studying the critical figures of merit for protective coatings have yet to be established. For this reason, we propose four criteria for evaluating the performance of a protective coating for PEC water-splitting: stability, conductivity, optical transparency, and energetic matching. We then propose a flow chart that summarizes the recommended testing protocols for quantifying these four performance metrics. In particular, we lay out the stepwise testing protocols to evaluate the energetics matching at a semiconductor/coating/(catalyst)/liquid interface. Finally, we provide an outlook for the future benchmarking needs for coatings.

36 MATERIALS SCIENCE↗