Reactor Components Research Articles

Development and validation of transient Thermal-Hydraulic model for SFR steam generators

The steam generator (SG) is an important component in sodium-cooled fast reactors (SFRs). It transfers the heat from sodium systems to the balance of plant (BOP), producing steam for electricity production. Any disturbances originating from the BOP propagate to the nuclear systems through SG. Hence, it is especially important to model and predict the thermal–hydraulic behavior of SG during such transients. Towards this, a thermal–hydraulic model of SG of a typical medium-sized SFR has been developed using general-purpose system simulation code – FLOWNEX. Usually, the moving boundary method is used to model SGs as it involves phase change and multiple heat transfer regimes. However, in the current model, the fixed mesh approach, generally used for modeling pipelines, single-phase heat exchangers, etc. in plant dynamic codes, has been used to model SGs. In this paper, the modeling details and the solution procedures used in the FLOWNEX are discussed in detail. The steady-state results of the model developed have been compared with the results from in-house code DESOPT. The results were in good agreement at various power levels as well as in off-design conditions. The transient performance of the model has been validated against an in-house experiment carried out in Steam Generator Test Facility (SGTF), IGCAR, India, and a steam generator shutdown experiment carried out in Energy Technology Engineering Center (ETEC), Santa Susana Field Laboratory (SSFL), USA. The transient results were also in good agreement, thereby increasing the confidence level of the predictions. The model developed will be integrated to plant dynamic codes containing other system models and will be used to carry out system-level simulations in the future.

Efficient Sensitivity Analysis for Enhanced Heat Transfer Performance of Heat Sink with Swirl Flow Structure under One-Side Heating

Excellent heat transfer performance has increasingly become a key issue that needs to be solved urgently in the development process of large-scale fusion equipment. The study of heat transfer performance improvement to scientifically and reasonably determine the design parameters of the high heat flow (HHF) components of fusion reactors based on the efficient in-depth analysis of the heat transfer mechanism and its sensitive factors is of great significance. In this paper, a liquid-vapor two-phase flow model with subcooled boiling for a large length-diameter ratio swirl tube structure in the HHF calorimeter component is proposed to analyze the effects of key design parameters (such as inlet temperature of cooling water flow, swirl tube structure parameters, etc.) on its heat transfer performance. Then, considering the high computational cost of the liquid-vapor two-phase flow model, and in order to improve the efficiency of the sensitivity analysis of these design parameters, the polynomial response surface surrogate model of heat transfer performance function was constructed based on Latin hypercube sampling. On this basis, by combining the proposed surrogate model, the sensitivity index of each design parameter could be obtained efficiently using the Sobol global sensitivity analysis method. This method could greatly improve the calculation efficiency of the design parameter sensitivity analysis of HHF components in the fusion reactor, which provides vital guidance for the subsequent rapid design optimization of related components.

Effect of rare earth oxide doping on microstructure and mechanical property of Mo-30 W solid solution alloy

Oxide dispersion-strengthened (ODS) molybdenum‑tungsten (MoW) alloys are ideal materials for plasma-facing components of fusion reactors. In this paper, the effects of rare earth oxides (RExOy) Y2O3, CeO2 and La2O3 on the sintering behavior, microstructural and mechanical properties of MoW solid solution alloys were investigated. The initial alloy powders were produced by spray drying combined with the two-step reduction method (composed of carbothermal reduction and deep hydrogen reduction steps), and the solid solution alloy with good mechanical properties was prepared by subsequent sintering. The results showed that the doping of rare earth oxides suppressed the grain growth of MoW alloy, and thus improved the mechanical properties. Compared with CeO2 and Y2O3, the doping of La2O3 is more favorable to obtain dense and fine-grained MoW solid-solution alloys. The doping of La2O3 produced a larger Zener force and interfacial drag force, which effectively hindered the grain boundary migration and refined the grains. The microhardness and bending strength of Mo-30 W-1La2O3 alloy reached the maximum values of 557 HV and 771 MPa.

SIMULASI ANALISIS VELOCITY ALIRAN FLUIDA PADA TANGKI REAKTOR NUKLIR MENGGUNAKAN METODE COMPUTATIONAL FLUID DYNAMICS (CFD)

Center for Applied Nuclear Science and Technology (Pusat Sains dan Teknologi Nuklir Terapan /PSTNT) is a work unit that is under and directly responsible to the Deputy for Nuclear Science and Applications of Nuclear Technology, National Nuclear Energy Agency. In PSTNT Bandung, the function of the tank is as a container to place the reactor components, the reactor shield for the heat exchanger and as a place for nuclear reactions to take place. In nuclear reactor tank flow distribution ocures in various velocity.The purpose of this practical work is to analyze the velocity of fluid flow in the reactor tank from the inlet flow to the bottom surface of the reactor and from the bottom surface to the top surface of the TRIGA 2000 reactor. The method used in this practical work is the Computational Fluid Dynamics (CFD) Velocity method. CFD is a method for simulating fluid flow. The method is processed numerically and then digitally modeled. In practice, the authors modeled and studied the conceptual design of the reactor system using CFD based on Gambit and Fluent applications in a reactor tank. GAMBIT is an application produced by fluent inc. which is useful for making a model and discrete (meshing) to be analyzed by numerical methods. Fluent is an application that can solve fluid flow cases with the results obtained after discretizing the GAMBIT application. The results by CFD shows the distribution of fluid flow from the inlet pipe to the bottom surface of the reactor tank then the flow rises to the top surface to the outlet pipe. The velocity of the fluid flow from the inlet flow to the bottom surface was getting smaller, and the smallest value was at the bottom of the nuclear reactor, by 0.164 m/s and the flow distribution from the bottom surface to the top surface of the velocity value in a nuclear reactor would be bigger by the largest value being on the top surface of the nuclear reactor is 2.9 m/s.

Temperature dependence of the deuterium concentration in C-Si codeposits

• Sputter deposition has been used to create SiC codeposits at temperatures 300 – 900 K. • The deuterium content was measured by thermal desorption spectroscopy. • The deuterium content was found to be very similar to the D-content of carbon deposits created in the same way. • The temperature dependence followed very closely the trends observed for plasma-created codeposits, with the D content decreasing above 400 K. Silicon carbide is a low-Z material seen as a possible alternative to graphite for some plasma-facing components in magnetic confinement fusion reactors. In this study, we have looked at the deuterium content of C-Si codeposits created by sputter deposition at temperatures 300 – 900 K. As measured by thermal desorption spectroscopy, the D content of such deposits was found to be very similar to the D-content of carbon deposits created in the same way. The temperature dependence followed very closely the trends observed for plasma-created codeposits, with the D content decreasing above 400 K. This suggests that the tritium retention associated with SiC first wall materials would be very similar to that of a graphite first wall.

Comparison of the deuterium permeability of copper, CuCrZr, and Cu layers

In order to estimate the fuel loss in ITER and further future fusion devices, the deuterium permeation through different wall and structural materials are studied. In order to determine the effective permeability, gas-driven deuterium permeation measurements are performed on Cu and ITER grade CuCrZr. The obtained permeabilites for Cu and ITER grade CuCrZr are very similar and in agreement to literature values for Cu. For a better estimation for fusion reactor components, combined material samples are studied. Cu layers were applied on steel substrates by magnetron sputter deposition. With these studies, the influence of interfaces and microstructure on the hydrogen permeation is investigated. Our study reveals that in the case of Cu layered steel substrates the influence of the interface on the permeation flux is minor compared to the influence of the microstructure on the permeability. The Cu layer permeability is around one order of magnitude smaller than the Cu bulk permeability in the temperature range between 300 ∘ C and 550 ∘ C . • Study of the deuterium permeation through first wall materials of fusion devices • Cu, CuCrZr-IG, and Cu layers on ITER steel are prepared, characterized and analyzed • The influence of an interface on the permeation flux was studied • The permeability of Cu layers is one order of magnitude lower than of bulk Cu • The reduction of the permeability is mainly due to the microstructure of the layer

Diagnostic Technique Characterizing Neutron Irradiation Effects on Additively Manufactured Inconel 625 Eliminating the Need for Remote Handling

Experimentally characterizing radioactive materials can be time consuming and expensive. This is mainly due to the size requirements of inspected specimens. Due to the growing interest in using additively manufactured components in next-generation reactors, there is an urgent need to develop new accelerated testing techniques with regard to characterizing radiation damage. This will ensure a more timely certification of the unique material structures inherent to additively manufactured parts. In this study, we investigate a means to reduce the time investment, and thus the human exposure to radioactive specimens in need of experimental characterization. We determine the feasibility of using ultra-small specimens in lieu of much larger specimens to characterize bulk material properties before and after irradiation. Experiments were conducted to investigate this technique and compare it to conventional bulk irradiations and characterization activities. It was found that discernable radiation damage existed in the ultra-small specimens even after relatively short neutron irradiation times. The results also demonstrate decreased radiation hardening in additive manufactured Inconel 625 material relative to its wrought forms.

Effect of dose rate on dislocation loop evolution in tungsten: Combination of defect generation rate and elastic interaction

Ion irradiation is often used to simulate neutron irradiation in reactor components, while the dose rate discrepancy makes it difficult to directly compare ion irradiation with neutron irradiation. A better understanding of dose rate effect is thus necessary to emulate neutron irradiation effect with ions. Here, based on in-situ 30 keV He+ irradiation in tungsten in TEM, we find that defect generation rate and elastic interaction are two key factors controlling dislocation loop evolution at different dose rates. By adjusting defect generation rate, the elastic interaction between the loops will play different roles. Finally, a schematical illustration is given to explain the different in loop density, size and spatial distribution caused by the combination of these two factors. Our results provide the underlying insights for understanding the dose rate effects of energetic ion irradiation.

Grain Boundary Migration as a Self-Healing Mechanism of Tungsten at High Temperature

The tungsten components in nuclear fusion reactors need to withstand the radiation cascade damage caused by the neutron bombardment of high temperature and high throughput fusion reaction during service. These damages are mainly present as a high concentration of point defects and clusters, which lead to a series of problems such as irradiation-hardening and decreased thermal conductivity of materials. In this study, molecular dynamics simulations are carried out to study the dynamic interaction between grain boundaries and the void in tungsten at high temperatures (T > 2500 K). Different interatomic potentials of W were tested, and the most appropriate one was selected by the thermodynamic and kinetic properties of W. Simulation results show that the dynamic migration of grain boundary can absorb the void, but the absorption efficiency of grain boundaries is sensitive to their structural characteristics, where the high-angle GBs are more absorptive to the void than the low-angle GBs. It is found that the void absorption cannot be completely attributed to the thermal diffusion mechanism during the GB-void interaction; the dynamic migration of high-angle GBs can significantly accelerate the void absorption. This study reveals a GB migration-induced self-healing mechanism of W at high temperatures.

Nonlinear MHD modeling of n = 1 RMP-induced pedestal transport and mode coupling effects on ELM suppression in KSTAR

Fully suppressing edge-localized modes (ELMs), e.g., with resonant magnetic perturbations (RMPs), is essential to reach and sustain high-performance steady-state H-mode plasmas because large ELMs can significantly reduce the lifetime of divertor components in future tokamak reactors. RMP-driven ELM suppression in KSTAR has been modeled by coupling the neoclassical transport code PENTRC to the nonlinear 3D MHD code JOREK. We have found that the radial transport from the combined effects of the kink-peeling, tearing response, and neoclassical toroidal viscosity can explain the pedestal degradation observed in experiments. In addition, it has been found that the RMP response can increase the inter-ELM heat flux on the lower outer divertor by redistributing the heat transport between the divertor plates. In addition to the degraded pedestal, ELM suppression is also attributable to the RMP-induced mode interactions. While the linear stability of peeling-ballooning mode (PBMs) improves owing to the degraded pedestal, the PBM and RMP interaction increases the spectral transfer between edge harmonics, preventing catastrophic growth and the crash of unstable modes. Here, it turns out that the magnetic islands near the pedestal top can play a vital role in mediating the mode interactions.

Dynamic responses of submerged components in advanced reactors: experimental and numerical studies

The seismic design of an advanced nuclear reactor must consider the interaction of vessel internal components with the surrounding coolant: fluid–structure interaction (FSI). Available analytical solutions for FSI of submerged components do not accommodate multiple-component, intense seismic inputs and complex reactor and internal geometries. Physical testing of reactor vessels and internals to inform seismic design is impractical and cost-prohibitive, leaving the use of verified and validated, robust numerical models as the only plausible option for analysis and design. Physical data that could be used for validating such numerical models for multi-component shaking are not available. This article describes an experimental program performed on a 6-degree-of-freedom earthquake simulator to generate data that could support validation of seismic FSI numerical models for submerged components in commercial finite element codes. A scaled model of a base-supported reactor vessel with simplified representations of submerged internals was tested to generate submerged-component response histories for a range of seismic inputs. The generated data were used to validate numerical models in the finite element code LS-DYNA. Numerical models were validated for calculation of hydrodynamic pressure responses on internals, in-water frequencies of internals, and acceleration responses of internals. The generated data and the analysis recommendations could aid engineering analysts designing submerged components and systems for seismic effects.

Simulation of Neutron and Gamma Dose Rate of The TRIGA 2000 Reactor Using Monte Carlo Method

The neutron and gamma radiation doses were calculated from the operation of the 1 MW TRIGA 2000 Reactor in a simulation using the Monte Carlo method with MCNPX and PHITS program. Simulation is done by modelling the geometry of the reactor component materials and running it on a computer. The radiation source in the form of a fission reaction in the reactor core has been simulated using MCNPX to produce a dose of neutron and gamma radiation in the TRIGA 2000 core. Attenuation of neutron and gamma radiation by the reactor building is simulated using the PHITS code so that the neutron and gamma dose rates are obtained on the source, y, and z from reactor core. Interpolation of dose rate curves on large material thicknesses was carried out with the TVL neutron and gamma values of the simulation results for each reactor material. The simulation results shows that the gamma neutron dose rate outside the TRIGA 2000 reactor building is still below the dose limit value for radiation workers.

Coolant Channel Design for Additively Manufactured Reactor Cores

ABSTRACT Additive manufacturing (AM) methods are currently being explored for applications in nuclear reactors to make advanced reactors more efficient, safe, and reliable. The Transformational Challenge Reactor (TCR) program has explored AM for nuclear by designing a high-temperature gas reactor (HTGR) using an AM silicon carbide fuel form with uranium nitride–tristructural isotropic fuel. This work details the design process for the TCR fuel form’s coolant channels using computational fluid dynamics models with conjugate heat transfer. Additionally, this work discusses how these models were interfaced with other design teams, project milestones, and the agile design method used to mature the reactor design. The methodology deployed was able to create a channel design with lower maximum fuel temperatures and thermal stresses in the fuel form over traditional channel designs that can be manufactured subtractively. These results were achieved with only small manufacturing penalties. Results are discussed and presented on lessons learned for designing AM components for nuclear reactors. Finally, areas of opportunity are discussed for advanced design tools to further automate design activities and optimize reactors with fewer built-in assumptions.

T-REX: Numerical tool for tungsten damage assessment for DEMO

One of the R&D focus in the European fusion energy program is to establish a physical and technological basis for reliable power exhaust during entire operational situations of a DEMOstrational power plant. Following the design adopted for ITER, the baseline Plasma Facing Component for DEMO divertor is made of tungsten as armor material. The tungsten divertor target are the most thermally loaded in-vessel components in fusion reactor. The PFCs lifetime is affected by material degradations under the different loadings including High Heat Flux (HHF) and neutron irradiation. It has been reported a loss of mechanical properties due to softening (restoration/recrystallization). Latest finite element modeling developments allow to estimate component lifetime under HHF loading taking into account the progressive mechanical properties change due to softening. We propose here to take into account the influence of thermal and neutron loadings on the tungsten mechanical properties change for PFC lifetime assessment. For this, up to date mechanical properties of tungsten, softened tungsten, and neutron-irradiated tungsten are used as input data in the tool named T-REX (for Tungsten-Response based on EXperimental data). This tool is implemented as a user subroutine in a finite elements code. The influence of the neutron irradiation is studied as function of the dpa dose, the incident heat flux intensity and the current defined divertor component geometries (ITER and DEMO). With the study performed, it is shown that the geometry may be optimized to reduce the damage induced by plasticity. The numerical results also highlights that the plastic strain increment obtained after each thermal cycle reduces with the dpa dose.

Electronic energy loss and ion velocity correlation effects in track production in swift-ion-irradiated LiNbO3: A quantitative assessment between structural damage morphology and energy deposition

• Essential factors including electronic energy loss and ion velocity co-affect latent track damage. • Damage morphologies from spherical defects to discontinuous and continuous tracks are confirmed. • Atomic energy deposition and temperature evolution are determined by inelastic thermal spike model. • A quantitative relationship between ion irradiation condition and track damage response is proposed. • The work provides a deeper insight for understanding ion-solid interactions and damage behaviors. The primary motivation for studying how irradiation modifies the structures and properties of solid materials involves the understanding of undesirable phenomena, including irradiation-induced degradation of components in nuclear reactors and space exploration, and beneficial applications, including material performance tailoring through ion beam modification and defect engineering. In this work, the formation mechanism of latent tracks with different damage morphologies in LiNbO 3 crystals under 0.09–6.17 MeV/u ion irradiation with an electronic energy loss from 2.6–13.2 keV/nm is analyzed by experimental characterizations and numerical calculations. Irradiation-induced damage is preliminarily evaluated via the prism coupling technique to analyze the correlation between the dark-mode spectra and energy loss profiles of irradiated regions. Under the irradiation conditions of different ion velocities and electronic energy losses, different damage morphologies, from individual spherical defects to discontinuous and continuous tracks, are experimentally characterized. During ion penetration process, the ion velocity determines the spatiotemporal distribution of deposited irradiation energy induced by electronic energy loss, meaning that the two essential factors including electronic energy loss and ion velocity co-affect the track damage. The inelastic thermal spike model is used to numerically calculate the spatiotemporal evolutions of energy deposition and the corresponding atomic temperature under different irradiation conditions, and a quantitative relationship is proposed by comparison with corresponding experimentally observed track damage morphologies. The obtained quantitative relationship between irradiation conditions and track damage provides deep insight and guidance for understanding the damage behavior of crystal materials in extreme radiation environments and selecting irradiation parameters, including ion species and energies, for ion beam technique application in atomic-level defect manipulation, material modification, and micro/nanofabrication.

Revealing the residual stress distribution in laser welded Eurofer97 steel by neutron diffraction and Bragg edge imaging

• Through-thickness residual stress distribution in three dimensions is evaluated. • High-resolution residual strain is mapped by neutron Bragg edge imaging. • Location-dependant reference lattice spacing is first applied on neutron imaging. • Correlation between microstructure, residual stress and micro-hardness is studied. Eurofer97 steel is a primary structural material for applications in fusion reactors. Laser welding is a promising technique to join Eurofer97 plasma-facing components and overcome remote handling and maintenance challenges. The interaction of the induced residual stress and the heterogeneous microstructure degrades the mechanical performance of such fusion components. The present study investigates the distribution of residual stress in as-welded and post-heat treated Eurofer97 joints. The mechanistic connections between microstructure, material properties, and residual stress are also studied. Neutron diffraction is used to study the through-thickness residual stress distribution in three directions, and neutron Bragg edge imaging (NBEI) is applied to study the residual strain in high spatial resolution. The microstructures and micro-hardness are characterised by electron backscatter diffraction and nanoindentation, respectively. The M-shaped residual stress distribution through the thickness of the as-welded weldment is observed by neutron diffraction line scans over a region of 1.41 × 10 mm 2 . These profiles are cross-validated over a larger area (∼56 × 40 mm 2 ) with the higher spatial resolution by NBEI. The micro-hardness value in the fusion zone of the as-welded sample almost doubles from 2.75 ± 0.09 GPa to 5.06 ± 0.29 GPa due to a combination of residual stress and cooling-induced martensite. Conventional post weld heat treatment (PWHT) is shown to release ∼ 90% of the residual stress but not fully restore the microstructure. By comparing its hardness with that of stress-free samples, it is found that the microstructure is the primary contribution to the hardening. This study provides insight into the prediction of structural integrity for critical structural components of fusion reactors.

A study of isothermal mechanical fatigue of welded reactor steel (15H2MFA)

As far as the reactor pressure vessel steel is concerned, the isothermal fatigue study of welded joints is helpful to protect the pressure vessel from blasting and save the life of its operators. Furthermore, the study helps to improve the future design of the reactor components and also enthuse to develop new materials. In material science, the fatigue of metal is quite difficult to deal with, and that of welded joint is an even more complex phenomenon. This complexity arises from different main factors: thermal cycles during welding, which will strongly affect the base material and induce residual stresses and distortion, the fusion process with the filler metal leads to heterogeneity to microstructure which will affect or change the original mechanical properties and chemical composition through out of the welded joint. In this study, isothermal mechanical fatigue of welded reactor steel has been contemplated. The samples of the welded joint used in this study are the ‘15H2MFA steel type of VVER-440’ reactor pressure vessel operating at Paks Nuclear Power Plant, Hungary. The reactor vessel is made of low-alloy heat-resistant steel and austenitic steel with a corrosion-resistant inner (clad) surface.

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

Carburization and decarburization behavior of Grade 91 ferritic-martensitic steel in liquid sodium environments

This paper presents a study of carbon transfer and its effect on microstructure and tensile properties of Grade 91 (G91) ferritic-martensitic steel exposed to sodium at 550–650 °C. Sodium exposure tests were conducted in Argonne's forced convection sodium loops up to exposure times of ∼40,000 h. Thermal aging study of G91 steel was conducted in parallel to isolate the thermal aging effect from the sodium effect. It was found that sodium exposures at 650 °C dissolved M23C6 carbides, eliminated the martensite subgrain structure resulting in excessive grain growth and reduced the tensile strength by >50%, while sodium exposures at 550 and 600 °C had an insignificant effect on its microstructure and tensile properties. These effects were attributed to the carburization/decarburization process of G91 steel in sodium environments.Carbon concentrations in sodium were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8–1.2 ppm in the SMT-1 loop and 0.3–0.7 ppm in the SMT-2 loop. Thermodynamic analysis of the carburization – decarburization process was conducted for G91 steels exposed in sodium environments. The carbon activity-concentration relationship for G91 was evaluated by considering four phases in G91, i.e. bcc ferrite, M23C6, NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M23C6 carbides at high carbon activities, while NbC and VC carbides dominated the process at low carbon activities. The calculated carburization-decarburization boundary showed that G91 would undergo decarburization at 650 °C and carburization at 550 °C in the sodium loop environments, which was consistent with our experimental observations. This experimental and theoretical analysis provided a basis for predicting the effect of carbon transfer on the integrity of reactor components in sodium environments and for the design of new alloys used in sodium-cooled fast reactors.

Additively manufactured strain sensors for in-pile applications

Accurate, real-time monitoring of strain in fuel, cladding, and structural components of nuclear reactors is critical to better understand radiation induced phenomena during reactor tests and operations. The data provided is crucial to verify physics-based, multiscale modeling and simulation efforts, which aim to shorten the timeline for the development of new nuclear materials. Resistive strain gauges have limited performance during in-pile experiments due to the harsh operating conditions and limited physical space between fuel and cladding components. In this work, aerosol jet printing using silver nanoparticle inks was used to fabricate interdigitated electrode capacitive strain gauges on aluminum alloy 6061 tensile specimens. To simulate the temperatures of a traditional light water reactor, the capacitive strain gauges were tested with a mechanical test frame up to 300 °C and compared to commercially available bondable resistive strain gauges. The printed capacitive strain gauges exhibited a gauge factor of 1.0 and showed higher reproducibility and predictability of strain sensing performance than the resistive strain gauge. The results demonstrate the potential of aerosol jet printing to fabricate strain sensors with predictable performance and reduced invasiveness for high-temperature applications with confined spacing.