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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Irradiation Effects on Stability of δ-UZr2 phase in U-50 wt% Zr Alloy

U-50wt%Zr is a candidate metallic nuclear fuel with potential application in light water reactors due to its excellent thermal properties and high radiation tolerance. The Zr-rich UZr fuels possess greater swelling resistance and fission gas release characteristics compared with U-rich UZr fuels. In this current study, the δ-phase U-50wt%Zr is proton irradiated at room temperature to 1 displacement per atom (dpa) to provide insights on phase stability under irradiation conditions. High resolution characterization of Transmission Electron Microscopy (TEM) and Atom Probe Tomography (APT) characterization techniques are used to elucidate microstructural changes due to irradiation. TEM and APT results show highly oriented bcc β-Zr-rich platelet precipitates nucleating adjacent to α-U phases inside the UZr 2 matrix. Formation of this platelet morphology is characteristic of Widmänstatten structure which can be attributed to a variety of factors such as differences in thermal expansion coefficient between the phases, grain size, alloy composition, and cooling rate. The phases present are distinctly different than those observed through in situ annealing, but irradiation accelerates diffusion and phase separation kinetics. These microstructural changes in U-50wt%Zr are different from those achieved by pure thermodynamic or high temperature heavy ion irradiation experiments. Finally, our work, together with previous ones, highlight the necessity to study the U-Zr phase diagram under non-equilibrium thermodynamics conditions to support this material's deployment as a viable nuclear fuel form.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Multi–length scale characterization of point defects in thermally oxidized, proton irradiated iron oxides

A key for the success of safe nuclear power generation system is to consider structural materials that are economical, meet mechanical property needs, possess good corrosion resistance, and are radiation tolerant. Nevertheless, fundamental insights that elucidate the details of radiation damage on materials corrosion performance are lacking. This includes the behavior of surface oxides which often regulate corrosion. For example, it is unclear how non-equilibrium point defects, oxide structure, mass transport in oxides, and subsequent oxidation behavior are altered by the radiation. Here, in this work, some of the effects of proton irradiation on the attributes of point defects, iron oxide microstructures, and the physical nature of the oxidation product were correlated with corrosion behavior. Iron oxides, fabricated by thermal oxidation in air at 400°C and 800°C for 1 h, were subjected to 200 keV, 0.03 dpa (displacements per atom) of proton irradiation, and subjected to corrosion reactivity assessment using AC and DC electrochemical methods. Experimental methods that target materials information at different length scales, such as positron annihilation spectroscopy (atomistic), transmission electron microscopy (mesoscopic), and electrochemical methods (macroscopic), were coupled to shed light on the impact of radiation-induced defect modifications and structural changes in oxides on corrosion reactivity which ultimately affects durability in harsh environments.

36 MATERIALS SCIENCE↗
DOE OSTI2025

In situ self-ion (Fe + ) irradiation of ODS-FeCrAl alloy fuel cladding materials with different Cr contents: The early stages of Cr-rich α’ phase precipitation

Oxide-dispersion-strengthened FeCrAl (ODS-FeCrAl) alloys are candidate accident-tolerant fuel cladding materials for light water reactors because they demonstrate satisfactory resistance to materials degradation effects such as high-temperature oxidation, radiation-induced swelling, and creep. Their perspective deployment to market is challenged, however, by their inherent susceptibility to irradiation embrittlement caused by the precipitation of the brittle Cr-rich α’ phase at relatively low temperatures (≤475 °C). This work used in situ self-ion irradiation (150 keV Fe + ) in a transmission electron microscope to elucidate the early stages of Cr-rich α’ phase precipitation in three candidate ODS-FeCrAl alloy fuel cladding materials with different Cr contents (10, 12, and 20 wt.%) and microstructures. The early stages of the process resulting in the precipitation of the Cr-rich α’ phase in these three ODS-FeCrAl alloys under Fe + irradiation were investigated at room temperature and 300 °C up to total fluences of 1.7 × 10 15 ions·cm -2 (2 dpa) and 3.4 × 10 15 ions·cm -2 (4 dpa), using three damage dose rates (5 × 10 –5 , 3.3 × 10 –4 , and 2 × 10 –3 dpa·s -1 ). Post-irradiation examination via scanning transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy suggested that the precipitation of the Cr-rich α’ phase might be promoted by the phase separation of the alloy matrix into Cr-rich and Fe-rich regions. Interestingly, oxygen impurities segregated preferentially in the Cr-rich regions, possibly promoting the radiation-assisted formation of the Cr-rich α’ phase. α’ phase precipitation was more pronounced at room temperature when compared to 300 °C, and it was clearly promoted by the progressive increase in the Cr content of the ODS-FeCrAl alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Beam-water Interaction System for PFAS Treatment

Per- and polyfluoroalkyl substances (PFAS) are persistent environmental contaminants that resist conventional water treatment due to the strength of their carbon fluorine bonds. Electron-beam irradiation has shown potential for PFAS degradation; however, its effectiveness is limited by shallow energy penetration, restricting treatment to the upper region of the water surface. This project addresses this limitation through the design of a compact mixing chamber and recirculation system to enhance surface renewal and improve treatment efficiency. The system, constrained to approximately 10 cm 20 cm 20 cm, was developed using computer-aided design, computational fluid dynamics (CFD), and MATLAB-based hydraulic analysis to evaluate flow behavior, head loss, and pump compatibility. A perforated-baffle geometry and controlled flow range of 1 3 GPM were implemented to promote uniform mixing while maintaining a stable free surface for electron-beam exposure. Modeling results indicate improved flow distribution and reduced dead zones, supporting more consistent interaction between contaminated water and the beam. The final design enhances hydrodynamic performance within system constraints, contributing to scalable electron-beam water treatment technologies for improved PFAS degradation.

Ramos, Gerardo [Northern Illinois U.]↗
DOE OSTI2023

Microstructural Characterization of Ion Irradiated ODS MA956 Alloy

Oxide Dispersion Strengthened (ODS) alloys are considered as promising nuclear materials for the next generation advanced nuclear reactors due to their improved elevated temperature mechanical properties and radiation damage resistance. MA956 is a commercial ODS alloy with superior oxidation resistance and creep strength at elevated temperatures. The improved properties for ODS alloys can be attributed to the presence of dispersed nanosized oxide particles. These oxide particles can act as barriers to dislocation motion and provide additional sinks and sites for point defect recombination. Hence, their stability can greatly affect the properties of the irradiated ODS alloys. Besides, irradiation will induce defects into the ODS alloys, and these defects can develop further as irradiation dose increases. Furthermore, the defects propagation can significantly undermine the mechanical properties of the bulk materials, resulting in irradiation hardening and lower ductility. Therefore, it is necessary to have a good understanding of the defect’s evolution with irradiation. Hence, efforts are ongoing to understand the effect of ion irradiation on ODS MA956 with emphasis on oxide particles and evolution of defects as a function of dose (2.5, 50 and 100 dpa).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Interplay of negative electronic compressibility and capacitance enhancement in lightly-doped metal oxide Bi 0.95 La 0.05 FeO 3 by quantum capacitance model

Light-sensitive capacitance variation of Bi 0.95 La 0.05 FeO 3 (BLFO) ceramics has been studied under violet to UV irradiation. The reversible capacitance enhancement up to 21% under 405 nm violet laser irradiation has been observed, suggesting a possible degree of freedom to dynamically control this in high dielectric materials for light-sensitive capacitance applications. By using ultraviolet photoemission spectroscopy (UPS), we show here that exposure of BLFO surfaces to UV light induces a counterintuitive shift of the O 2p valence state to lower binding energy of up to 243 meV which is a direct signature of negative electronic compressibility (NEC). A decrease of BLFO electrical resistance agrees strongly with the UPS data suggesting the creation of a thin conductive layer on its insulating bulk under light irradiation. By exploiting the quantum capacitance model, we find that the negative quantum capacitance due to this NEC effect plays an important role in this capacitance enhancement.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Value Proposition of UV-Absorbers in PV Module Encapsulation

Various common crystalline silicon cell technologies were exposed to UVA radiation (1.24 Wm-2 nm-1 at 340 nm peak) on the front and back faces at 45 degrees Celsius for periods up to 3000 h, representing about 3 y of solar exposure in Phoenix, Arizona, USA. The resulting degradation of the open-circuit voltage and short-circuit current is presented. Of the various cell types examined, significant levels of degradation were seen in all cases. Less degradation was generally found after UV irradiation of cell fronts and older cell types, whereas a bifacial PERC type exposed on the rear showed about 25% degradation in short circuit current, attributable to lack of a diffused surface field. Selected cells were exposed to UV irradiation with the addition of long pass UV filters to replicate UV-absorbers in encapsulants. Modern cell designs are sensitive to UV-ID because of reduced or eliminated front and back surface field and increased dependence on high quality surface passivation. Single transformation of the independent variable (t, kW h/m^2) could be used to achieve a linear model of the data to extrapolate to 50 y. Solar Advisor Model (SAM) shows appropriate filtering of UV-irradiation can improve LCOE and net present value of plant. Some advanced cell types are seen to be UV-resistant (cell level solutions also exist). Solutions therefore exist on the cell, glass, and encapsulant level. Changes over time in each of these would also need to be considered (solarization, encapsulant browning...).

ENGINEERING,SOLAR ENERGY↗
DOE OSTI2022

The mechanism behind the high radiation tolerance of Fe-Cr alloys

With the great demand for high radiation tolerant materials for advanced nuclear energy technologies, Fe-Cr alloys are at the forefront with long standing validated performance. Yet, the real mechanism behind their high radiation resistance is in question and understanding the effect of varying Cr percentage is a grand challenge limiting further improvements. Here we applied depth resolved atomic scale probe of defects to uncover the real mechanism on how Cr improves radiation resistance and explain the controversial impact of increasing Cr percentage. By combining depth-resolved positron annihilation lifetime spectroscopy and Doppler broadening spectroscopy we investigated the effect of Cr alloying on the formation and evolution of atomic size clusters induced by ion irradiation in Fe. We also used atom probe tomography to investigate the possible presence of Cr clusters or a’ phase with high Cr composition. The study reveals that the well-known resistance to radiation in Fe-Cr alloys arises from the stabilization of vacancy clusters around Cr atoms which act as sinks for radiation-induced defects. Thus, Cr atoms do not provide a direct sink for interstitials; rather defect complexes for that consist of Cr atoms and vacancies in turn act as sinks for irradiation-induced vacancies and interstitials. Most importantly, we find that lower amounts of Cr create smaller, uniformly distributed defect clusters that act as efficient sinks for radiation damage, but larger quantities of Cr form a defect structure that is less homogenous in size and spatial distribution, resulting in less efficient damage recombination. No evidence of phase a’ was found before or after irradiation, which indicates that it does not contribute to the observed radiation tolerance.

36 MATERIALS SCIENCE↗
DOE OSTI2023

AGC-2 Specimen Post Irradiation Data Package Report

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time of flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A standard capsule design for structural material testing in the Advanced Test Reactor

Nuclear materials testing is commonly completed in various material test and research reactors throughout the world, including the Advanced Test Reactor (ATR), but the current capsule design, analysis, and fabrication process can take years to complete. To decrease the costs and time associated with materials testing, a standard capsule has been designed which houses various specimen geometries and allows irradiation in virtually any ATR test position. The standard capsule features locking end caps which connect and lock together to form the capsule stack, eliminating the need for a basket and maximizing the quantity of specimens which are contained within the capsule. A customizable internal gas gap provides thermal resistance between the specimens and reactor coolant, making specimen temperatures from approximately 370 to 1000 K achievable. The flexibility of the capsule design allows experimenters to choose irradiation positions based off desired neutron flux, with typical fluences per cycle ranging from 8.8x10 19 to 2.3x10 21 n/cm 2 , depending on experiment position. Here this paper presents and discusses the standard capsule design and analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Gold Nanorod-Affibody Conjugates Mediate Cancer Photothermal Therapy under 808 nm LED Irradiation

Current HER2-targeted therapies including monoclonal antibodies, antibody-drug conjugates (ADCs) and small-molecule tyrosine kinase inhibitors face limitations such as hepatotoxicity, development of treatment resistance, and subsequent disease relapse. To overcome these challenges, this study combines the specificity of a HER2-targeting affibody molecule (Z HER2:2891 -Cys) with the photothermal properties of gold nanorods (AuNRs), forming bioconjugates (Affi–AuNRs) for selective treatment of HER2-positive cancer cells. Affi–AuNRs were fabricated via a two-step surface modification: (i) ligand exchange to displace cetyltrimethylammonium bromide (CTAB) with poly(ethylene glycol) methyl ether thiol (mPEG-SH), and (ii) covalent attachment of the affibody molecule via Au–S chemistry. Affi-AuNRs exhibited a photothermal conversion efficiency of 64 ± 5%, comparable to that of CTAB-stabilized AuNRs (59 ± 4%). Confocal microscopy confirmed that Affi-AuNRs are selectively internalized by HER2-positive SKOV-3 cells, with minimal uptake by HEK293T cells, which have low HER2 expression. Upon exposure to 808 nm near-infrared (NIR) LED irradiation (408 mW cm –2 , 15 min), Affi-AuNRs significantly reduced SKOV-3 cell viability by 85 ± 9% (p < 0.001) relative to untreated controls with no detectable cytotoxicity in HEK293T cells. These results demonstrate HER2-selective photothermal cytotoxicity mediated by Affi–AuNRs under defined 808 nm LED irradiation conditions, supporting their continued development as nanotheranostic tools that integrate affibody-mediated specificity with plasmonic heating.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Thermocouple and RTD Drift Due to Neutron Irradiation and Temperature for Space Nuclear Applications: A Review

Temperature measurement is a key factor in the safe and reliable operation of nuclear reactors. Thermocouples and resistance temperature detectors (RTDs) are among the most common and reliable tools deployed for temperature measurement, and some are capable of operating in extreme environments such as very high temperature and neutron radiation fields anticipated in nuclear thermal propulsion (NTP) or fission surface power. These devices are known to exhibit drifting behavior in their temperature readings throughout their operational lifetimes. Here, to help select appropriate instrumentation, it is important to identify the factors that influence drift as well as typical behavior in these environments. This work is focused on identifying contributing factors in the performance of thermocouples and RTDs. An extensive literature review is included that discusses several experimental results in different environments. Although individual instrument performance may vary with construction methods, materials selection, and vendors, this review aims to assist in the selection process and identify typical expected behaviors of several devices.

Chesser, Ryan J. [Oak Ridge National Laboratory (O↗
DOE OSTI2023

Structural Tolerance of Zirconium Diboride under Electron Irradiation through in-situ Convergent Beam Electron Diffraction and Energy-dispersive X-ray Spectroscopy

Zirconium diboride (ZrB 2 ) is a ultra-high temperature ceramic with high melting temperature (>3000 K), super strength and hardness, good thermal conductivity, and excellent resistance to plasma erosion and chemical erosion, being widely applied in hypersonic flight, arc-electrodes, refractory materials, and plasma-facing materials in fusion reactors. Therefore, the structural stability of the material is essential for the applications. Up to now, the material was investigated under neutron irradiation and high-energy ion irradiations. It was reported that the neutron irradiation caused swelling and catastrophic cracking. High-energy ions (30 keV He + ) changed the surface morphology and caused the bright blue and brown colors on ZrB 2 surfaces between 920 K and 1120 K. The electron irradiation tolerance of the material has not been reported yet. Here we used convergent beam electron diffraction (CBED) and energy-dispersive X-ray spectroscopy (EDS) to investigate structural tolerances of ZrB 2 crystals under high-energy electrons at room temperature under vacuum. Finally, the investigation would benefit the understanding of its structural tolerance at microscales.

42 ENGINEERING↗
DOE OSTI2025

Suppression of irradiation hardening in tungsten-coated ferritic steel for fusion reactor blanket applications

W-coated reduced activation ferritic steels have been developed for use as plasma facing components in fusion reactor blankets, offering excellent sputtering resistance and structural strength. Previous high-temperature coating methods, such as diffusion bonding and brazing, caused interfacial deterioration due to thermal stress from mismatched thermal expansion between W and reduced activation ferritic steel. To address this, underwater explosive welding was introduced as a high-velocity cold process that joins dissimilar materials while maintaining a strong, thin interface without the thermal issues associated with traditional methods. In this study, the effects of neutron irradiation on the hardness and microstructure in W-coated F82H reduced activation ferritic steel (W/F82H) joined by underwater explosive welding are investigated. Following neutron irradiation at 290 °C, irradiation hardening is suppressed in W, F82H, and their interface within the W/F82H material. Furthermore, microstructural observations indicate that the recovery of work hardening and relaxation of elastic strain introduced during coating significantly contribute to the suppression of irradiation hardening in W/F82H. In conclusion, W/F82H exhibits significantly suppressed irradiation hardening compared with those in stand-alone materials. This suppression is explained by residual stress from thermal expansion mismatch and the unique microstructure at the interface. These results provide valuable insights for the development of more durable materials in nuclear fusion applications.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Phase stability and microstructure of neutron-irradiated substoichiometric yttrium dihydrides

The impact of the neutron-displacement damage on phase stability and microstructure of substoichiometric yttrium dihydrides (YH x , x <2) were investigated to assess their use as solid moderator in high-temperature nuclear reactors. YH x specimens were, thus, subjected to neutron irradiations in the range of 0.1–2 displacements per yttrium atom (dpa-Y) in the temperature range of 536–878°C at the Oak Ridge National Laboratory's (ORNL's) High Flux Isotope Reactor (HFIR). YH x specimens were initially prepared at stoichiometry (H/Y) ratios of 1.69 and 1.83. HFIR-irradiated specimens were characterized by variety of techniques to investigate H retention characteristics including dimensional analysis, optical microscopy, scanning electron microscopy electron back scatter diffraction (EBSD), transmission electron microscopy, thermal desorption spectroscopy (TDS), and high-energy x-ray diffraction (HE-XRD) characterizations. Overall, YH x exhibited notable structural and phase stability under short-term neutron-irradiation, except for the samples with significant silicon carbide (SiC) interaction at high doses and temperatures. Basic dimensional and mass measurements were misleading for accurate assessment of H retention, as confirmed by EBSD phase maps, XRD line profiles, and TDS signals. Thus, it was discussed that a robust H retention metric is needed to assess irradiated hydrides. Further, nanoscale cavities were observed as a result of the neutron irradiation in all samples. Although no clear impact of dose and irradiation temperature was determined, the initial H/Y ratio had an impact on the cavity number density where low H/Y specimens had high-resistance to cavity formation. The Y-vacancy cluster formation at the collision stage of the displacement cascade and their stabilization by H were considered to be the likely underlying mechanisms for the observed cavity microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Comparing Outdoor to Indoor Performance for Bifacial Modules Affected by Polarization-Type Potential-Induced Degradation

Bifacial photovoltaic (PV) modules have the advantage of using light reflected off of the ground to contribute to power production. Predicting the energy gain is challenging and requires complex models to do so accurately. Often, module degradation over time is neglected in models for the sake of simplicity or is underestimated. Comparing outdoor and indoor current–voltage (I–V) performance for bifacial modules is more challenging than for monofacial modules, as there are additional variables to consider such as rear albedo non-uniformity, cell mismatch, and their effects on temperature. This challenge is compounded when heterogeneous degradation modes occur, such as polarization-type potential-induced degradation (PID-p). To examine the effects of PID-p on I–V predictions using an empirical data-driven approach, 16 bifacial PERC modules are installed outdoors on racks with different albedo conditions. A subset is exposed to high-voltage biases of −1500 V or +1500 V. Outdoor data are traced at irradiance ranges of 150–250 W/m 2 , 500–600 W/m 2 , and 900–1000 W/m 2 . These curves are corrected using control module temperature, wire resistivity, and module resistance measured indoors. We examine several methods to transform indoor I–V curves to accurately, and more simply than existing methods, approximate outdoor performance for bifacial modules without and with varying levels of PID-p degradation. This way, bifacial performance modeling can be more accessible and informed by fielded, degraded modules. Distributions of percent errors between indoor and outdoor performance parameters and Mean Absolute Percent Errors (MAPEs) are used to assess method quality. Results including low-irradiance data (150–250 W/m 2 ) are discussed but are filtered for quantifying method quality as these data introduce substantial errors. The method with the most optimal tradeoff between low MAPE and analysis simplicity involves measuring the front side of a module indoors at an irradiance equal to plane-of-array irradiance plus the product of module bifaciality and albedo irradiance. This method gives MAPE values of 1–6.5% for non-degraded and 1.6–5.9% for PID-p degraded module performance.

14 SOLAR ENERGY↗
DOE OSTI2022

Structural and thermodynamic evolution of an amorphous SiOC ceramic after swift heavy ion irradiation

Radiation induced structural and energetic changes have been widely studied in crystalline materials, but studied to a much lesser extent in amorphous solids. Using a combination of neutron pair distribution function (PDF) analysis and high temperature oxide melt solution calorimetry, we obtained critical insights into the structural and energetic evolution in a swift heavy ion irradiated amorphous SiOC polymer-derived ceramic. The radiation modified SiOC structure becomes energetically less stable by 24.0±1.6 kJ/mol compared to the unirradiated structure. This destabilization is related to decreased connectivity of the Si-O-C network and destruction of free carbon. In comparison with unirradiated SiOC, the irradiated structure is more likely to become phase separated during subsequent thermal annealing. Here, our study has important implications for evaluating amorphous SiOC ceramic as a possible radiation resistant structure for nuclear applications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Evaluations of Silicon Carbide Temperature Monitors

The effect of irradiation on materials properties is an important field of study for materials usage in both fission and fusion systems for energy production. Neutron flux and energy spectrums are well understood; however, the irradiation temperature can be more difficult to determine. Since the early 1960s, Silicon Carbide (SiC) has been used as a passive post-irradiation temperature monitor because the irradiation defects anneal out above the irradiation temperature. Irradiation temperature is determined by measuring a property change after isochronal annealing or during a continuously monitored annealing process. There are many properties that may be measured including electrical resistivity, bulk density, dimensions, thermal diffusivity, or lattice spacing. Electrical resistivity is accepted as a robust measurement technique; however, such method is time-consuming since the steps involved must be performed in a serial manner. The use of thermal expansion from continuous dilatometry is an automated process requiring minimal setup time. As part of a Nuclear Science User Facilities (NSUF) project, low dose silicon carbide monitors were irradiated in the BR2 material test reactor at SCK, Belgium, and were then evaluated at the Idaho National Laboratory (INL) High Temperature Test Laboratory (HTTL) to determine their peak temperature achieved during irradiation. The technical significance of this work was that the monitors were irradiated to a dose that was significantly less than recommended in published literature. This paper will discuss the evaluation processes available at HTTL and the performance of the low dose silicon carbide temperature monitors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗