Advanced synchrotron characterization techniques for fusion materials science
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99 GENERAL AND MISCELLANEOUS↗
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Phase instabilities in austenitic steels during particle bombardment at high and low dose rates
Disruption of phase stability by energetic particle bombardment is a major challenge in designing advanced radiation-tolerant alloys and ion beam processing of nanocomposites. Particularly, ballistic dissolution susceptibility of different solute nanocluster species in alloys is poorly understood. Here, low dose rate neutron irradiations were conducted on a Fe-Cr-Ni based austenitic steel in the BOR-60 reactor (9.4 × 10 -7 dpa/s, 318 °C) followed by accelerated dose rate ion irradiations at multiple temperatures ( 10 -3 dpa/s, 380 – 420 °C). Using atom probe tomography, the stability of radiation-enhanced Cu-rich and radiation-induced Ni-Si-Mn-rich nanoclusters was evaluated. During neutron irradiation, Cu-rich clusters nucleated with their core concentrations progressively increasing with dose, while Ni-Si-Mn-rich clusters formed and evolved into G-phase precipitates. Ion irradiations dramatically altered the nanoclusters. Cu-rich clusters were ballistically dissolved, but Ni-Si-Mn-rich clusters remained stable and coarsened with dose at 400 and 420 °C, highlighting that different nanocluster species in a single microstructure can have innately distinct ballistic dissolution susceptibilities. Solute-specific recoil rates were incorporated into the Heinig precipitate stability model, which shows that in addition to radiation-enhanced diffusion, recovery from ballistic dissolution depends on solute concentration gradient near cluster interfaces. Finally, the combined experimental-modeling study quantified the critical temperatures and damage rates where ballistic dissolution dominates for each cluster species.
36 MATERIALS SCIENCE↗
Ar-39 recoil losses and presolar ages in Allende inclusions
Argon analyses were performed for five coarse-grained, Ca-Al-rich inclusions from the Allende meteorite. The samples were neutron-irradiated in evacuated ampuoles, and the Ar in the ampuoles and in the samples was analyzed. Up to 60 percent of Ar-39 was lost from the samples into the ampuoles due to recoil during neutron-irradiation; this loss resulted in substantial increases in the apparent Ar-40-Ar-39 ages of the samples. Substantial amounts of trapped Ar-36 of unknown origin were found in the inclusions, and the presence of trapped Ar-40 could not be ruled out. It is inferred that the interpretation of high Ar-40/K-40 ratios as very high (presolar) ages may be subject to question. It is noted that the degrees of Ar-39 recoil loss reported is comparable to recoil loss in terrestrial rocks and cannot be attributed to KCl contamination of the samples.
Villa, I. M.↗
Cavity evolution and void swelling in dual ion irradiated tempered martensitic steels
Microstructural evolutions in two variants of the F82H 8 wt%Cr tempered martensitic steel under dual (He + plus Fe 3+ ) ion irradiations at 500 °C have been characterized by transmission electron microscopy. Emphasis is on understanding the combined effects of displacements per atom (dpa) and the helium/dpa ratio (He/dpa) on cavity evolution and void swelling (f v ). The dual ion database includes 231 alloy-dpa-He/dpa conditions for irradiations up to ~ 82 dpa and 3700 appm helium. Here, we hypothesize that the large “scatter” observed in the f v data is primarily due to local microstructural variations, especially those that shield regions at shallower depths from the effects of injected interstitial atoms, which would otherwise reduce f v . More generally, our hypothesis is that the highest f v data, at a specified dpa and He/dpa, provides the most appropriate estimate of swelling for fusion neutron irradiation conditions. The dose dependence of f v can be described by an incubation dpa i , marking the onset of significant void growth, and a post-incubation swelling rate, f v ’ = Δf v /Δdpa. Both dpa i and f v ’ decrease with increasing He/dpa. The f v ’ data are quantitatively consistent with a simple model for defect partitioning between evolving bubble, void, and dislocation sinks strengths and biases. Most notably, the model correctly predicts both the incubation dpa i , and the subsequent f v ’, in spite of the fact that they reflect somewhat different physics. Our analysis also suggests that defect recombination is modest at 500 °C. Along with other considerations, this supports the hypothesis that, even at high dpa rates, dual ion irradiations can be used to reasonably emulate void nucleation and growth for a range of fusion-relevant neutron irradiation service conditions. A companion paper incorporates the results of this study in a reduced-order, data-driven void swelling model.
36 MATERIALS SCIENCE↗
Simultaneous exposure to neutron radiation and hydrogen environment: Effect on hydrogen retention
Here, this study demonstrates a novel methodology that simultaneously exposes materials to both neutron irradiation and a hydrogen environment, enabling study on in pile interaction between radiation-induced defects and hydrogen in tungsten. The hydrogen environment was established by releasing hydrogen from vanadium hydride within an irradiation capsule. The hydrogen pressure during irradiation was estimated as 14.2 Torr. Post-irradiation thermal desorption spectroscopy analysis showed increased hydrogen retention in irradiated tungsten compared to unirradiated samples. Two dominant desorption peaks at ∼470 °C and ∼700 °C were observed in irradiated and unirradiated tungsten samples, suggesting similar trapping mechanisms between the two samples. However, an additional desorption peak at ∼800 °C in irradiated tungsten from the hydride capsule suggests an interplay between irradiation induced defects and hydrogen exposure, leading to the formation of additional hydrogen trapping sites. These findings provide critical insights into hydrogen retention in tungsten under fusion-relevant conditions and stimulate future studies to investigate the synergistic effects of neutron irradiation and hydrogen exposure.
08 HYDROGEN↗
Radiation-Induced Changes in Single Crystal Calcite and Dolomite: Mineral Analogues of Light Water Reactor, Nuclear Power Plant Concrete Aggregates
Two analogues of aggregates (calcite and dolomite) found in nuclear power plant (NPP) concrete have been neutron irradiated to fluences up to 2 × 10 20 n/cm 2 at 52 °C and 4 × 10 19 n/cm 2 at 95 °C (E > 0.1 MeV). X-ray diffraction studies showed that these aging conditions affected crystallography of both calcite and dolomite, with both showing a growth in their unit cells compared to that of the unirradiated samples. Only dolomite samples showed a continuous increase of its lattice parameters with the increase in irradiation. In calcites, the lattice growth with irradiation dissipated after a fluence of 5 × 10 18 n/cm 2 . Radiation-induced disordering of the crystal structure was observed in both materials, while a healing effect of irradiation temperature was mainly observed in calcites. Dislocation densities calculated by using XRD data also supported these observations. Insights made into the unit cells of both minerals by using Fourier synthesis showed that the Mg sites are more prone to the neutron irradiation than the Ca sites, resulting in larger radiation effects and higher disruption in the crystal structure of dolomite than that of calcite. Furthermore, other than the polycrystalline characteristics observed by using TEM, these minerals did not show complete amorphization or significant structural decomposition at the studied irradiation fluences.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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↗
Effects of sample bias on adhesion of magnetron sputtered Cr coatings on SiC
Swelling of SiC at 300 C due to in-service neutron irradiation causes tensile residual stresses in coatings which are expected to adversely affect the performance of coated SiC composite fuel cladding for light water reactors. Matching the coating swelling with the substrate, a solution common for thermal expansion, is not practical in the case of neutron irradiation. Biasing samples during magnetron sputtering deposition induces compressive residual stress which may counteract this. In this study, chromium coatings were deposited on SiC by DC magnetron sputtering with no external heating at bias voltages of –50V, –75V, and –100V. The effects of the bias voltage on morphology, residual stress, microstrain, texture, and adhesion are shown. The low deposition temperature resulted in the coating microstructure evolution following an energetic particle bombardment dominated trend. Additionally, at the two lower bias voltages knock-on implantation dominated increasing the residual stress and microstrain while at the highest bias voltage, thermal spike migration allowed for defect relaxation. When the knock-on induced compressive residual stress exceeded 0.8 GPa microcrack formation in the SiC substrate decreased coating adhesion. While no microcracks formed at the lowest bias voltage, insufficient atomic mobility during coating growth lead to voids forming in the coating. A balance is needed to form void-free coatings that have high compressive residual stress.
36 MATERIALS SCIENCE↗
Accelerated radiation tolerance testing of Ti-based MAX phases
MAX phases have recently attracted significant attention for potential nuclear applications due to their novel properties such as unique hexagonal-compact nanolayered crystal structure, high-machinability due to lower hardness levels compared to conventional ceramics, and high-chemical inertness. In order for MAX phases to be used in nuclear reactors, two aspects deserve detailed investigations: (i) their phase stability at high-temperatures and (ii) microstructural defect formation and recovery induced by energetic particle irradiation. To date, degradation mechanisms of MAX phases at high-temperatures and following irradiation are largely unexplored fields of research. This work focuses on the evaluation of two Ti-based MAX phases – Ti 2 AlC and Ti 3 SiC 2 – within the context of extreme environments. To accomplish this, a one-of-a-kind comparison between neutron irradiations, performed over a decade of research at the High Flux Isotope Reactor (HFIR), and heavy-ion irradiations, carried out in situ in a Transmission Electron Microscope (TEM) at the MIAMI-2 facility, has been conducted. The results show Ti-based MAX phases are prone to accelerated decomposition under the conditions investigated. This questions the hypothesis that MAX phases exhibit high phase stability, especially when used in future nuclear energy systems where energetic particle irradiation is a dominating degradation mechanism.
36 MATERIALS SCIENCE↗
Report on the Integration of Experimental and Modeling Data for Initial Equivalence Study of Microstructural Evolution in Irradiated LPBF 316SS
Advanced materials and manufacturing technologies are poised to improve the safety and design characteristics of nuclear technologies and meet US energy, environmental, and economic needs. In particular, metal additive manufacturing (AM) provides an opportunity to produce novel materials and component geometries, but their use is not without hurdles arising from the inherent microstructure variability that can result from the layer-by-layer build approach. Given the greater possible microstructure variability in AM materials—and the dearth of materials test reactors—it is impractical to rely solely on neutron irradiation studies to produce data for materials qualification for every possibility. This work within the Advanced Materials and Manufacturing Technologies (AMMT) Environmental Effects technical area contributes to the rapid qualification framework by developing a science-driven framework for the accelerated qualification of materials for nuclear environments. A key product of the Environmental Effects technical area of the AMMT program is the Licensing Approach with Ions and Neutrons (LAIN). This approach recognizes that whether using existing materials in new environments, newly developed materials tailored for these environments, or new manufacturing methods, the traditional decades-long approach for materials qualification does not facilitate rapid deployment. In FY 2023, the AMMT program presented a conceptual framework of specific steps to fulfill several technical challenges associated with qualifying materials for performance in radiation environments on an accelerated time frame informed by the state of the art in materials science and a review of the current regulatory landscape. The objective of this section of the Environmental Effects technical area is to critically evaluate and refine the proposed qualification framework presented under AMMT by integrating the research results of the neutron irradiations, the ion irradiations, and modeling efforts. These ongoing efforts span across Argonne National Laboratory (ANL), Idaho National Laboratory (INL), and Oak Ridge National Laboratory (ORNL) and are closely coordinated.
36 MATERIALS SCIENCE↗
Development of a Framework and Methodology for an Advanced Reactor Materials Environmental Effects Design Guide
Advanced non-light-water reactor components may operate at elevated temperature while experiencing cyclic loading, significant neutron irradiation, and exposure to reactor coolant. ASME Boiler and Pressure Vessel Code, Section III, Division 5, provides design rules for elevated-temperature service but does not include specific procedures to account for environmental effects on material properties. This report develops an initial framework and methodology for an Environmental Effects Design Guide (EEDG) focused on neutron irradiation; coolant-environment effects are reserved for future work. The proposed approach treats irradiation as a property-based overlay on the existing Division 5 design process, with two routes: a sparse-data route applying two reduction factors — FCR on creep-rupture strength and FF on fatigue life — for the creep-fatigue evaluations that typically control the design of advanced high-temperature reactor components, and a fuller framework developing the property-to-rule chain across the four Division 5 checks (primary load, strain limits and ratcheting, creep-fatigue, and buckling), together with swelling and weldments as scope items. Both routes are scoped by an in-pile qualification that restricts the use of post-irradiation-examination-derived properties in regimes where an in-pile mechanism could control the design outcome. Illustrative outputs derived on a compiled annealed Type 316 database — FCR ≈ 0.78–0.86 and FF ≈ 0.4 — demonstrate the calculation method within that specific dataset. The framework is an initial, testable design-rule concept; it identifies a practical path for preliminary design evaluations under sparse data and the material data and testing needed to develop the framework further.
Barua, Bipul (ORCID:0000000247184113)↗
Improved Understanding of Zircaloy Hydrogen Pickup Mechanisms in BWRs. NSUF Report (Final Report)
Using an instrumented autoclave operating at 288°C and 1100 psi, electrical impedance spectroscopy (EIS) measurements were taken from three neutron irradiated fuel channel samples (Zircaloy-2 and Zircaloy-4). The three high priority fuel channel samples were tested at 288°C using a simulated PWR water with the AC frequency range extending down through 10 -4 Hz, with corresponding DC measurements to validate the resistance measurements. EIS measurements were also taken at room temperature using a simulated PWR water from four neutron irradiated fuel channel samples (Zircaloy-2 and Zircaloy-4). and four water rod samples (Zircaloy-2). EIS measurements were also performed on unirradiated Zircaloy-2 tubes that were previously oxidized in autoclaves at various temperatures for different time durations to obtain oxide layers of different thicknesses. These experiments proved to be more challenging than anticipated because of the small size of the samples, which required development of fixtures to allow valid testing of the materials, some of which are radioactive, under pressure and temperature. Equivalent circuits were used to model the data. Some samples resulted in Nyquist plots where the arcs were too close to each other resulting in the inability to separate the different oxide morphological thicknesses. Microstructural characterization in the form of optical microscopy, scanning electron microscopy and (scanning) transmission electron microscopy was performed to document the microstructural features of the oxidized Zr-2 and Zr-4 samples (oxide and metal/oxide interface) and to interpret the EIS results. At the start of this work, it was assumed that despite the laminar cracks and porosity visible in both optical and SEM, the oxide scale is dense near the metal, effectively preventing the water from contacting the metal directly. While it is generally accepted that the large laminar cracks pores allow the water to percolate into the oxide film, it is unknown just how deep it can go, and how interconnected is the porosity, which will dictate how much the porosity is connected electrically to the water within the oxide scale. The EIS results demonstrated that there is in fact a distribution of oxide thicknesses over the various samples, and within each material condition, that are communicating to the liquid. This is derived from the fact that Nyquist plots have an arc radius whose center point is often below the Z’ axis, implying that instead of a single oxide thickness, these oxides have a distribution of thicknesses. This distribution of thicknesses produces arcs that are not clearly separable in most of the samples.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
FeCrAl fuel/clad chemical interaction in light water reactor environments
This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO 2 fuel were irradiated at a temperature of ~300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. Finally, the study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO 2 interface than C35M due to the higher bulk Cr:Al ratio.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
COMSOL Results for the Nominal Steady-State Operation of the Proposed 95-MW LEU Silicide Core for HFIR Conversion
Engineering design studies are being performed to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel at Oak Ridge National Laboratory. This activity is sponsored by the Office of Reactor Conversion and Uranium Supply (ORCUS) under the auspices of the US Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. HFIR is a very high flux, pressurized, light water–cooled and moderated, flux trap–type research reactor with a core made of involute shaped U 3 O 8 /Al cermet fuel plates and coolant channels. HFIR currently operates at a thermal power of 85 MW and supports key national and international missions in neutron scattering, isotope production, materials/fuels irradiation, neutron activation analysis, gamma irradiation, and neutrino research. Advanced multiphysics computational fluid dynamics models have been developed in the COMSOL Multiphysics software to simulate the steady-state operating conditions for the proposed low-and high-density LEU U 3 Si 2 -Al (uranium silicide dispersion) fuel designs. The COMSOL models for HFIR inner and outer fuel element models incorporate various essential inputs and physics such as spatially dependent nuclear heat deposition, multilayer heat conduction, conjugate heat transfer, turbulent flows (using Reynolds-averaged Navier Stokes turbulence models), structural mechanics (thermal–structural interactions and fuel swelling), and oxide layer build-up. This report presents the best-estimate thermal hydraulics results for the low- and high-density optimized silicide LEU core designs at 95 MW steady-state nominal operation.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Amorphization and siliconization of silicon carbide as a first wall material
The understanding and prediction of silicon carbide (SiC) material evolution exposed to SOL plasma conditions is of prime interest because SiC represents a promising main chamber wall plasma-facing material for next-step fusion devices (low hydrogenic diffusion, good mechanical and thermal properties under neutron irradiation). Gross and net Si erosion rates from SiC surfaces in contact with a well-diagnosed L-mode plasma in the DIII-D tokamak have been simulated and the surface concentrations of impurities have been tracked as a function of time. Coupled simulation of surface model and impurity transport demonstrates amorphization of crystalline SiC exposed to L-mode plasma due to the accumulation of displacement damages under ion irradiation. This affects the lifetime of SiC plasma facing components. Surface evolution is tightly coupled to impurity transport in the plasma and therefore needs to be integrated with impurity transport simulations to effectively predict Si erosion rates and sub-surface concentrations as a function of time. The simulation workflow couples a semi-analytical surface model to the impurity transport code GITR. The surface model is a homogeneous mixed-material model that tracks physical & chemical sputtering and reflection of impurities. Gross erosion is primarily influenced by the background plasma parameters and redeposition patterns are mainly influenced by the prompt redeposition due to the gyro-orbits of impurity ions. Although crystalline form of SiC is preferable for fusion wall applications because of resistance to neutron irradiation, this work indicates that crystalline SiC will undergo amorphization under D plasma contact with implications of higher sputtering and fuel retention. These results direct us to explore the effects of amorphization on crystalline SiC and further the physics basis of SiC usage as first wall material for fusion environments.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Fiber/matrix debonding evaluation of SiCf/SiC composites using micropillar compression technique
Third-generation silicon carbide (SiC) composites reinforced by SiC fibers (Hi-Nicalon S [HNS] and Tyranno SA3 [SA3]) are attractive for use in next generation reactors owing to their high strength and chemical inertness at high temperatures, as well as enhanced radiation tolerance under neutron irradiation environments. To optimize composite performance, the interfacial mechanical properties of chemical vapor–infiltrated (CVI) SiC f /SiC composites are investigated in this effort by using a slant interface micropillar compression testing procedure. The micropillar test specimens, containing an inclined pyrolytic carbon (PyC) interphase, are prepared using a focused ion beam. The novel microcompression testing successfully quantifies the debond shear strength and internal friction coefficient of micropillar test samples by using the Mohr-Coulomb formulation. Furthermore, according to four types of SiCf/SiC composite microcompression test results, interfacial properties and debond mechanisms are significantly affected by the PyC layer thickness, the local bonding mechanism of the PyC interphase on the SiC fiber surface, and the surface roughness of fibers. Regardless of PyC thicknesses, SA3-reinforced CVI SiC f /SiC composites are found to have much higher debond shear strengths than HNS-reinforced SiC f /SiC CVI composites. By using this micropillar compression technique alongside analytical methods, we uncover new understandings of PyC interface properties. Additionally, the micropillar test results obtained are correlated with macroscopic mechanical properties of neutron-irradiated CVI SiCf/SiC composites.
42 ENGINEERING↗
The effect of phosphorus on precipitation in irradiated reactor pressure vessel (RPV) steels
Embrittlement of light water reactor pressure vessels (RPV) steels by fast neutron irradiation may limit extended nuclear plant life. Embrittlement, which is manifested as increases in various indexes of a ductile to brittle transition temperatures (ΔT), is primarily due to hardening by nanoscale precipitates containing Cu, Ni, Mn, and Si, which form under irradiation. In addition to these elements, P has also been found to play a role in embrittlement. While only slightly enriched in the precipitates, hardening and embrittlement increase with trace P concentrations in low-Cu steels. Here, we characterize the individual and synergistic irradiation precipitation and hardening mechanisms in a series of RPV steels containing no to low-Cu and with systematic variations in Ni and P. In this study, the steels were irradiated to a fluence of ~ 1.38 x 10 20 n/cm 2 at ~ 292 °C in the UCSB ATR-2 experiment. In Cu-free medium and high-Ni RPV steels, atom probe tomography shows that P and Ni promotes precipitation of P-Mn-Si-Ni and Mn-Si-Ni precipitates, respectively. The precipitate microstructure correlates with the observed irradiation hardening.
36 MATERIALS SCIENCE↗
Characterization of Fe-Cr alloys irradiated by neutrons at intermediate temperature
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Bhattacharyya, Dhriti↗