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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Dataset of tensile properties for sub-sized specimens of nuclear structural materials

Mechanical testing with sub-sized specimens plays an important role in the nuclear industry, facilitating tests in confined experimental spaces with lower irradiation levels and accelerating the qualification of new materials. The reduced size of specimens results in different material behavior at the microscale, mesoscale, and macroscale, in comparison to standard-sized specimens, which is referred to as the “specimen size effect.” Although analytical models have been proposed to correlate the properties of sub-sized specimens to standard-sized specimens, these models lack broad applicability across different materials and testing conditions. The objective of this study is to create the first large public dataset of tensile properties for sub-sized specimens used in nuclear structural materials. We performed an extensive literature review of relevant publications and extracted over 1,000 tensile testing records comprising 55 columns including material type and composition, manufacturing information, irradiation conditions, specimen dimensions, and tensile properties. The dataset can serve as a valuable resource to investigate the specimen size effect and develop computational methods to correlate the tensile properties of sub-sized specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Machine Learning based Correlation of the Mechanical Properties of Sub-sized and Standard-sized Specimens

Mechanical testing with sub-sized specimens is essential in the nuclear industry, offering the ability to conduct tests in confined spaces with lower irradiation and expediting material qualification. However, smaller specimens exhibit different material behavior across scales, a phenomenon known as the "specimen size effect". In this study, we compiled over 1,000 tensile testing records, covering 54 parameters such as material type, composition, manufacturing details, irradiation conditions, specimen dimensions, and tensile properties through a comprehensive literature review. We focus on correlating sub-sized and standard specimens’ tensile mechanical properties on SS316 alloy, which has the most extensive dataset available. We explore ML-based models and uncertainty quantification for tensile properties, analyze key factors influencing these properties, and compare the effectiveness of ML models with existing analytical methods in addressing the specimen size effect.

tensile properties↗
DOE OSTI2025

Enhancing Post-Irradiation Examination (PIE) Efficiency with High-Throughput Experiments

Nuclear materials testing is essential for understanding changes in reactor environments, but traditional methods are costly and time-consuming. High-Throughput Experimentation (HTE) enables rapid screening and optimization of material properties by significantly increasing the number of specimens in the experiment. In this work, HTE methods were employed to irradiate novel materials at the Advanced Test Reactor (ATR) at Idaho National Laboratory. A total of 367 sub-sized tensile specimens were irradiated using the standard capsule design. To facilitate HTE, automated systems such as the Automated Mechanical Testing System were implemented for mechanical testing within hot cell environments, thereby reducing the need for manual interaction with hot cell manipulators and significantly lowering both the cost and duration of experiments. Digital image correlation (DIC) was used to measure displacement during mechanical testing, providing precise, non-contact measurement capabilities. The methodologies developed in this research will support the advancement of HTE techniques for future nuclear materials testing experiments at ATR. This approach allows for more efficient and cost-effective evaluation of new materials, ultimately leading to improved reactor performance, enhanced safety, and extended longevity within the nuclear industry.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Production of Creep-Resistant FeCrAl-ODS

Four ODS FeCrAl alloys with base alloy composition Fe-16Cr-4Al-2Mo-0.2Y (wt.%) were mechanically alloyed with minor additions of CeO 2 and Y 2 O 3 additions and were successfully extruded into rectangular-shaped bars. Two oxide compositions were explored, which were designated NC with oxide addition of 0.5CeO 2 plus 0.2Yin solution of the base alloy powder and PC with additions of 0.5CeO 2 and 0.7Y 2 O 3 to elevate the O level. The microstructure of both alloys revealed by SEM and EBSD to consist of very small grains with sizes of ~450-520 nm and weak texture components. Sub-size tensile specimens fabricated from both alloys were tensile tested from room temperature to 800ºC. The results for both alloys from room temperature to 400ºC showed very high strengths, but low ductilities, especially uniform strain. From 500ºC to 800ºC, the strengths decreased rapidly, and the ductilities increased significantly. The reason for the rapid decreases in tensile strengths above 500ºC for both alloys is not clear, but will require advanced characterization techniques including atom probe tomography, to determine the size and number density of the nano-size oxide particles, and TEM/EFTEM, for investigating the dispersion of nano-size oxide particles on grain boundaries to assess the Zener grain boundary pinning effect, which will hinder grain coarsening at high temperatures that may enhance the Hall Petch strengthening mechanism.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mechanical Property Measurements of HFIR-irradiated FM and ODS Alloys

This report provides a summary of the post-irradiation examination of the tensile properties of three ODS ferritic alloys, 14YWT-SM13, 9YWTV-PM2 and OFRAC, and two HT9 ferritic/martensitic steels, one with low N content and the other with high N content. The neutron irradiations were conducted in the flux trap of the High Flux Isotope Reactor as part of the A dditive M anufacturing, O xide Dispersion Strengthened Alloys and W rought Steels (AMOW) irradiation campaign. Sub-sized tensile specimens fabricated from each material were exposed to nominal target temperatures and neutron doses of 300°C, 385°C, and 525°C and 8 dpa, 16 dpa, and 50 dpa, respectively.

14YWT-SM13, 9YWTV-PM2, and OFRAC, and two HT9 ferr↗
DOE OSTI2025

Mechanical Property Measurements of HFIR-irradiated FM and ODS Alloys

This report provides a summary of the post-irradiation examination of the tensile properties of three ODS ferritic alloys, 14YWT-SM13, 9YWTV-PM2 and OFRAC, and two HT9 ferritic/martensitic steels, one with low N content and the other with high N content. The neutron irradiations were conducted in the flux trap of the High Flux Isotope Reactor as part of the Additive Manufacturing, Oxide Dispersion Strengthened Alloys and Wrought Steels (AMOW) irradiation campaign. Sub-sized tensile specimens fabricated from each material were exposed to nominal target temperatures and neutron doses of 300°C, 385°C, and 525°C and 8 dpa, 16 dpa, and 50 dpa, respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2026

On the roles of welding residual stresses in determination of fracture toughness in austenitic stainless steel SUS 304 pipeline girth welds

Welding residual stresses especially the high tensile stresses are proved to have negative impacts on the fatigue and fracture behaviors of welded structures. In this study, a virtual fabrication of test specimens from welding process to specimen preparation was carried out by numerical simulation. An austenitic stainless steel multi-pass pipe welding was simulated by transient thermal–mechanical finite element analysis, the residual stresses were then mapped into the test specimen to evaluate fracture toughness. The findings in this study confirmed that, residual stress can be high in a sub-sized compact tensile specimen, which may accelerate or hinder the crack propagation during actual fatigue and fracture tests as reported in recent years. The influence of the cutting location and orientation of the specimen on fracture performance was investigated systematically to provide a fundamental understanding of welding residual stress and necessary insights into the specimen preparation procedure. Considering the limitation of measuring techniques and the complexity of the stress distribution, the developed numerical model can be a very useful tool to elucidate the stress evolution and quantify the effect of remaining welding stress on fracture toughness.

Fracture behavior↗
DOE OSTI2025

Printed circuit primary heat exchanger for power conversion in heat pipe microreactor applications

Here, in the present study, a model was developed for a recuperated air Brayton cycle for heat pipe microreactors, and the effect of the primary heat exchanger type assessed. An annular flow heat exchanger and a printed circuit heat exchanger (PCHE) were evaluated. A sub-sized diffusion bonded, PCHE test specimen was manufactured and tested with nitrogen as the working fluid to validate and improve the PCHE model. A comparison of the thermal cycle efficiency for the Brayton cycle coupled to a heat pipe microreactor using each of the two primary heat exchangers showed competitive performance for both options, with achievable cycle efficiencies of 34 % and 35 % for the annular and printed circuit heat exchangers, respectively. The impact of heat pipe length and additional reheating stages on cycle efficiency was studied. Increasing the heat pipe condenser length from 0.8 to 2.0 m resulted in an increase in cycle efficiency of 3.5 % and 3.2 % for the annular and printed circuit heat exchangers, respectively. The use of additional reheating stages showed a reduction in cycle efficiency since the heat pipe surface area limits the primary heat exchanger size and effectiveness. These results highlight the importance of the primary heat exchanger on the performance of a heat pipe microreactor. Overall, the cycle model, testing, and heat exchanger model showed the potential for the PCHE to improve the economics of a heat pipe microreactor relative to the annular flow heat exchanger, which also opens the possibility to use different heat transfer fluids such as supercritical CO 2 .

PCHE↗
DOE OSTI2026

Mechanical Behavior of Neutron Irradiated Refractory Multi-Principal Element Alloys Processed via Spark Plasma Sintering

The search for advanced materials capable of withstanding the extreme conditions of Generation IV reactors is a critical area in materials science research. These reactors operate under severe environments, including high temperatures, corrosion, stress, and irradiation damage. Consequently, there is a need for innovative alloy systems to ensure the reliability and longevity of proposed Generation IV reactor components. Refractory multi-principal-element alloys (RMPEA) have emerged as a promising candidate due to their exceptional properties. These alloys, characterized by their composition of multiple principal elements in near-equiatomic ratios, exhibit superior resistance to irradiation damage, reduced void swelling, enhanced microstructural stability, and minimal irradiation-induced hardening. While initial studies on RMPEAs have shown promising results, most research has been limited to thin films, nanocrystalline microstructures, and ion irradiation, which do not accurately represent the behavior of bulk materials. To address this gap, our research focused on the neutron irradiation of bulk RMPEAs. We aim to conduct comprehensive post-irradiation examinations (PIE) of RMPEAs irradiated at the Advanced Test Reactor at Idaho National Laboratory. The RMPEAs were synthesized using spark plasma sintering (SPS) with mechanically alloyed metallurgical powder. The RMPEA specimens are a MoNbTi alloy system with additions of -Zr, and -ZrV. Furthermore, PIE consisted of mechanical testing and advanced materials characterization. The mechanical testing consisted of sub-sized tensile testing, micro- and nano- indentation. Microstructural characterization included scanning electron microscopy and transmission electron microscopy. Mechanical testing coupled with advanced microscopy techniques provides insight into phase morphology and its effects on the mechanical properties of the RMPEA specimens. The results indicate that both pristine and irradiated RMPEA specimens exhibited brittle behavior during tensile testing, which can be attributed to their heterogeneous microstructure. The SPS manufacturing process did not include any post treatment, which resulted in a heterogeneous microstructure. Energy-dispersive X-ray spectroscopy revealed the presence of intermetallic such as laves phases within the microstructure. Specifically, Ti-rich precipitates were observed in the MoNbTi specimen, while Mo-rich precipitates were found in the MoNbTiZrV specimen. Nano-hardness testing of pristine samples showed that the laves phases exhibited higher hardness values compared to the matrix phase, suggesting that precipitate hardening is likely the dominant hardening mechanism in these specimens. The results from this work will be used to build a finite element model to predict mechanical behavior of future MPEA compositions. Thus, enabling for a streamlined approach to developing novel MPEAs for the nuclear industry.

36 - MATERIALS SCIENCE↗