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This is the poster our intern will present at AIM 2025 Conferences highlighting the data-driven representation of AFM data we established.
Reliable material performance is required for plasma-facing material (PFM) candidates. Previous research has shown that plasma and neutron radiation exposure induces microstructural changes in PFMs; changes in thermal and electrical conductivities and in material hardening and embrittlement were also observed after neutron irradiation. These material property changes will negatively impact the performance of the PFMs in a fusion reactor. Despite the well-known connection between material microstructure, properties, and performance, there is a need for validated modeling capabilities connecting PFM property degradation with microstructural evolution under fusion-relevant conditions. We are developing a simulation capability to couple plasma-induced microstructural evolution to material property degradation. Our approach relies on deliberate mapping between individual simulation models and experimental characterization for validation. The open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) software was used for this simulation capability development. A MOOSE phase-field model was coupled with the cluster dynamics code, Xolotl, to predict microstructural evolution. Microstructure characterization techniques, including scanning electron microscopy (SEM), transmission electron microscopy (TEM), and laser scanning confocal microscopy (LSCM) are used to validate these microstructural evolution simulations. Calculation of thermal and electrical conductivities with first principles simulations was performed for bulk material and for grain boundaries; these results are used within MOOSE models to calculate effective thermal and electrical conductivities as a function of grain characteristics. Thermoreflectance and four-probe techniques were employed to measure the thermal and electrical conductivities, respectively. A MOOSE crystal plasticity model was adapted to predict microstructure-sensitive deformation behavior, and X-ray diffraction (XRD) was used to collect bulk dislocation density data for validation. After individual simulation validation, these models are coupled to predict material property changes resulting from plasma exposure. We focused here on an experimental design to emphasize the separate effects of moderate thermal loads and plasma exposure using tungsten. Annealing of tungsten was performed under a protective environment for temperatures ranging from 500 C to 1500 C. The plasma exposure was completed in the Tritium Plasma Experiment at Idaho National Laboratory under a deuterium flux of 1e22 D/m^2-s. This incremental approach is employed to build confidence in the modeling capability: separate-effects tests ensure that the models capture key mechanisms from single environmental conditions before predicting PFM property degradation under combined loads. We will show our early results from coupling these simulation models to predict PFM property changes from microstructural evolution. Comparisons of the simulation results with preliminary validation data will be discussed.
Operational environments in generation IV reactors involve corrosive and irradiative conditions at elevated temperatures. Typical reactor operations consist of transients which impose cyclic loads on reactor components. These cyclic loads, combined with corrosive and irradiative environments, result in synergistic degradation of component materials. However, limited data exists on the coupled damage effects on materials for reactor environments. While the surrogate material surveillance concept has been used in light water reactors to assess irradiation damage, existing material surveillance technologies are not suitable for in-situ monitoring of coupled material degradation. The materials surveillance program focuses on material degradation management and the estimation of remaining life of reactor components through surveillance test articles. This paper presents the design and analysis methodology of a bi-metal surveillance test article, which uses difference in thermal expansion coefficient between two metals to induce in-situ cyclic loads. This report presents the work conducted in FY 25, to test the surveillance test article in air and salt environments. These test specimens were evaluated after thermal cycle exposure and remaining life is measured through creep test.
Reactor Pressure Vessels (RPVs) are critical components in nuclear reactors, housing the reactor core and coolant under extreme conditions of temperature, pressure, and radiation. These harsh environments contribute to the degradation of RPV materials over time, presenting challenges for extending reactor operations beyond their original design lifespans. The NRC Expanded Materials Degradation Assessment (EMDA) report volume 3 have been instrumental in guiding research to support extending the operational life of light water reactors (LWRs) up to 80 years. It provides a comprehensive framework to address technical challenges related to aging and degradation mechanisms in RPVs. The LWRS program has played a key role in advancing this research, supporting projects such as the UCSB ATR-2 Experiment, material testing from Zion and Palisades reactors, and the development of advanced mini-compact tension testing techniques. These efforts have been crucial in identifying and addressing gaps in our understanding of RPV aging, contributing to the successful subsequent license renewals of eight LWR units in the U.S. The EMDA report volume 3, built on the Phenomena Identification and Ranking Table (PIRT) analysis from earlier versions of the EPRI Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs), provides a detailed assessment of RPV degradation mechanisms. However, as EPRI has updated the MDM and IMT, it is important to revisit research priorities and methodologies to reflect these changes. The revised MDM and IMT may introduce new factors affecting long-term RPV performance and safety, highlighting the need for continued research and updated guidance to ensure the reliable and safe operation of reactors beyond 80 years.
While several decades of materials degradation research enabled the current US fleet of Light Water Reactors (LWR) to plan to produce electrical power through extended operations up to 80 years of plant life, it now seems appropriate to evaluate what issues, methods, and timelines need to be considered to meet the expected electric power demands for life beyond eighty (LBE) years. To address these questions, it will be necessary to review the electrical capacity and the projected LWR fleet capacity including new builds and advanced reactors. This effort should begin with a materials degradation assessment focused on known and possible unknown issues using a reduced Expanded Materials Degradation Assessment [1], which was employed to identify knowledge gaps for a second license renewal (60-80 yrs.). The focus of the LBE assessment should include: Establishing a timeline to initiate assessments of possible materials aging issues and to ensure sufficient time to evaluate materials degradation research gaps; assessing components and materials to be evaluated, e.g., metals (reactor pressure vessels and alloys within and outside the pressure boundary), concrete, cables, as well as mitigation methods, advanced monitoring, and validation with ex-service materials; initiating a LBE research plan focused on developing an expanded mechanistic understanding of materials degradation and refined models through Codes and Standards evaluation for use by the nuclear industry; and maintaining and strengthening the nuclear materials human knowledge base to address new and as yet unknown degradation modes.
Material degradation in Advance Test Reactors (ATR) is governed by irradiation, corrosion, elevated temperature exposure and cyclic mechanical creep-fatigue loads. This degradation information during reactor operation condition is limited. Hence, material damage monitoring is a key aspect of the design, analysis and licensing of ATR components. The idea is to monitor material component operation conditions of component by using a surveillance test article. This test article is fabricated with bi-metal configuration with two different thermal expansion coefficients, and design is motivated by Simplified Model Test (SMT) specimen which can capture structure-like mechanical response. Upon raising temperature of the bi-metal test article configuration, expansion mismatch results tensile load on specimen. Thus, temperature dependent passively actuated loading is achieved. The idea is to place this surveillance test article in reactor at location ‘x’ to surveil the mechanical response at critical location ‘y’. By calibrating the test article design, material degradation at critical location can be surveilled through assessing the degradation in surveillance test article. This study presents test article development with different material combinations and follow-up experimental testing work through passively loading test article with temperature history. The test article geometry and observed test results are presented in presentation slides.
It may be beneficial to use hermetic designs for supercritical CO2 (sCO2) cycles machinery as they would eliminate CO2 leakage through shaft end seals. This would reduce CO2 emissions and operating costs for makeup of lost process fluid. Those designs may replace traditional oil-lubricated bearings with actively controlled magnetic bearings operating at high temperatures in the process fluid environment (to reduce the need for active cooling). This paper investigates the material degradation of various types of magnets in CO2. Included are several permanent and soft magnetic materials (Alnico 9C, Alnico 5-7C, and 18-T550 grade SmCo, and Hiperco 50) with or without coatings (nickel plating or C5 coating). The materials were exposed to: (1) flowing gaseous CO2 at 1,022 °F (550 °C) and atmospheric pressure in a furnace and (2) sCO2 at 842 °F (450 °C) and 1,500 psi (103 bar) in an autoclave. The preliminary mass change measured after total exposures of 1,000 hours and 2,000 hours are included.
Lithium-ion batteries are widely used in applications from consumer electronic devices to stationary energy storage. Appropriate management of batteries is challenging due to limited data on their performance and materials degradation. Previous studies have focused on characterization of single cells under specific operating conditions. In the present work, commercial 18650 lithium-ion cells with LiNi x Mn y Co 1-x-y O 2 (NMC) and LiNi x Co y Al 1-x-y O 2 (NCA) positive electrodes were characterized by a wide range of electrochemical and materials techniques after cycling at 15, 25, or 35 °C to ∼80% capacity. The NCA cells exhibit weak temperature dependence in their cycle aging and materials degradation. The NMC cells exhibited increased capacity fade and materials degradation as ambient temperature decreased. All cells exhibited loss of lithium inventory as their primary degradation mode. However, the NCA cells only showed evidence of solid electrolyte interphase (SEI) growth whereas the NMC cells showed signs of Li plating at 15 °C, transitioning to SEI growth at 35 °C. The NMC cells displayed signs of loss of active material at the positive electrode at lower temperatures, suggesting that Li plating is correlated to additional processes that increase the rate of degradation. These results highlight the importance of avoiding broad generalizations about Li-ion battery temperature dependence.
Polysiloxane foams utilized for stockpile applications experience changes over time induced by exposure to thermal, radiative, and mechanical stresses. The interplay of these stresses can lead to chemical and/or physical changes in the polymer, such as compression set and material degradation over time. Understanding the mechanisms of material aging and identifying material degradation prior to failure are critical in maintaining the integrity of the nuclear stockpile. Our prior work using solid-state NMR spectroscopy has shown that we can observe signatures of thermal, radiative, and mechanical aging of silicone elastomers using differences in NMR parameters such as relaxation (T1 and T2). However, most of our measurements have been on bulk samples and have only provided an average aging behavior of each sample. Here, we have used magnetic resonance imaging (MRI) to provide relaxation-weighted images that are a metric of structural heterogeneity (crosslink density) in polymers across multiple length scales. In addition, the larger sample size that can be accommodated in animal MRI scanners has the potential to make the technique adaptable to detecting heterogeneities in whole parts.
Polysiloxane foams utilized for stockpile applications experience changes over time induced by exposure to thermal, radiative, and mechanical stresses. The interplay of these stresses can lead to chemical and/or physical changes in the polymer, such as compression set and material degradation over time. Understanding the mechanisms of material aging and identifying material degradation prior to failure are critical in maintaining the integrity of the nuclear stockpile. Our prior work using solid-state NMR spectroscopy has shown that we can observe signatures of thermal, radiative, and mechanical aging of silicone elastomers using differences in NMR parameters such as relaxation (T1 and T2). However, most of our measurements have been on bulk samples and have only provided an average aging behavior of each sample. Here, we have used magnetic resonance imaging (MRI) to provide relaxation-weighted images that are a metric of structural heterogeneity (cross-link density) in polymers across multiple length scales. In addition, the larger sample size that can be accommodated in animal MRI scanners has the potential to make the technique adaptable to detecting heterogeneities whole parts.
Historical and current data requirements for component qualification, as-manufactured graphite material properties, Irradiated & degraded material issues, code rules – Construction & Operation, status of current ASME code rules, progress in design rules, degradation, construction vs. operation, what should we be planning, and new technical areas getting started.
Lost circulation material (LCM) selection is critical to effectively and efficiently treating wellbore fluid losses in geothermal drilling where costs of treatment can be as much as 30% of the total drilling cost. We conducted several uniaxial compaction experiments on 10 different materials and several material mixtures to identify critical mechanical parameters of each. Materials degraded at 200°C were also investigated to understand how elevated temperatures in geothermal wells would degrade their compaction behavior. Granular materials tended to have lower compressibility and higher compression resistance, while more elongated and softer materials had less mechanical stiffness. Mixing materials tended to moderate the mechanical behaviors while heating universally increased the compaction of materials. Microscopy showed that particle strength tended to correlate positively with roundness and circularity and negatively with elongation of a material. Convexity of the degraded and undegraded materials showed heating may have increased the convexity or roughness of the individual particles. In conclusion, we concluded that granular materials are likely to provide the best seals in wells but that a mixture of size distribution, mechanical rigidity, and elongation is more likely to form a better seal for geothermal wells.
Microbially induced corrosion (MIC) focuses on the degradation of solid materials, such as glass or metal. Soil microbes are often associated with the corrosion of foreign objects in the rhizosphere. Paenibacillus polymyxa SCE2, a facultative anaerobic bacterium in soil, is of the same genus as bacteria found near nuclear waste disposal sites. Time-of-flight secondary ion mass spectrometry (ToF-SIMS) was used for imaging surface changes induced by P. polymyxa SCE2 cultured on two synthetic glass coupons to represent natural analogs of materials that were studied in relation to the vitrification of nuclear waste. Multimodal imaging was used to verify bacterial coverage across the glass surface after long-term growth. ToF-SIMS spectral analysis showed detection of glass component ions, such as silicon oxide (m/z – 59.96 SiO 2 – ) and aluminum oxide (m/z – 101.95 Al 2 O 3 – ), and biofilm’s extracellular polymeric substance (EPS) components, such as pentadecanoic acid (m/z – 241.22 C 15 H 29 O 2 – ) and sterol lipids (m/z – 311.16 C 20 H 23 O 3 – ). ToF-SIMS spectral, imaging, and depth profiling analyses showed that the glass rich in silica and other light elements (“granite glass”) had more “corrosion related” peaks than the glass that was less silica-rich and contained more iron (“dike glass”). Furthermore, these surface and interface compositional and spatial differences observed in the mass spectra and imaging were attributed to bacterial metabolism and an electron transfer mechanism influenced by morphological and compositional differences between the two types of glasses. ToF-SIMS is effective in studying microbial effects, bringing new molecular insights into MIC in a broader context of materials degradation.
This project focused on providing a fundamental physico-chemical understanding of the coupling mechanisms of corrosion- and radiation-induced degradation at material-salt interfaces in Ni-based alloys operating in emulated Molten Salt Reactor(MSR) environments through the use of a unique suite of aging experiments, in-situ nanoscale characterization experiments on these materials, and multi-physics computational models. The technical basis and capabilities described in this report bring us a step closer to accelerate the deployment of MSRs by closing knowledge gaps related to materials degradation in harsh environments.
To enable sustainable carbon-free fusion energy, managing reactor structural material degradation during normal operation as well as accident scenarios is vital. Tungsten (W) plasma-facing materials (PFMs) are susceptible to aggressive high-temperature oxidation during air-ingress fusion reactor accidents, yet there's a lack of oxidation kinetic data for irradiated tungsten. Here, in this study, we utilize atmospheric environmental transmission electron microscopy (ETEM) to present the first kinetic data for substrate-free W nanofuzz oxidation at 400 ºC and 500 ºC in 1 bar dry air. Comparison with pristine bulk W during the early parabolic stage suggests an irradiationdecelerated oxidation for W nanofuzz. Our time-resolved in-situ characterization reveals a durable amorphous surface oxide, likely promoted by high-flux He + irradiation-induced surface defects, serving as an effective passivating layer that impedes nanofuzz oxidation onset. This surface oxide layer also interfaces well with newly formed orthorhombic WO 3 , facilitated by stress relief through He bubble shrinkage, providing lasting passivating protection throughout the nanofuzz parabolic oxidation. This new finding challenges conventional notions of irradiation's negative impact on metal oxidation, and calls for advanced characterization to enhance our understanding of fusion energy materials degradation, informed by further accident modeling.
Components in operating commercial nuclear power plants must withstand very harsh environments that include extended time at neutron and gamma irradiation, stress, and temperature, as well as possible exposure to corrosive media. The many modes of materials degradation are complex and often include synergies between multiple environmental variables and conditions that vary depending on locations and materials. Understanding and managing materials degradation is a requirement for the continued safe and reliable operation of nuclear power plants.
The oxidation of carbon fibers at high temperatures is the primary degradation process in the thermal protection system of many hypersonic flight vehicles. Predicting the rate and the extent of oxidation is critical to ensure a safe and effective design. An oversized thermal protection system adds unnecessary mass, while an under-designed one risks system failure and mission loss. Resolving high-temperature material degradation due to oxidation has been a long-standing challenge in designing for re-entry flight environments. Using time-resolved in situ X-ray microtomography, the oxidation of carbon fibers at high temperatures is directly imaged, resolving the two limiting degradation regimes: diffusion- and reaction-limited. The ability to resolve material degradation in time at the sub-micron scale sheds light on the ablation phenomenon and enables predictions of material constitutive properties evolving in time, with profound implications on the ability to model the aerothermal response of heat shield materials in hostile environments.
This paper describes results from the mechanical evaluation of unirradiated SiC fiber–reinforced SiC matrix composite tubes under a controlled high-temperature steam environment. The experiments were aimed at identifying key material degradation behavior under environments relevant to loss-of-coolant accidents of light water reactors. Mechanical tests of the SiC composite tubes at 1000 °C under steam and inert environments were conducted using a unique test capability. Further, the material tested was a duplex tube with a thick monolithic SiC layer on the outer surface. The tubes were subjected to preloading at ~100 MPa in tension before exposure to high-temperature steam with up to 75% of the preload at a constant displacement. In the presence of matrix cracks, the steam exposure caused embrittlement of the SiC composite tubes and failure at a stress level below the pretest stress. The material degradation was explained by a fiber oxidation model, which can be applied to various SiC cladding concepts. The embrittlement could be a limiting factor for using SiC cladding subjected to loss-of-coolant accident conditions.