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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Post-Irradiation Fracture Toughness Characterization of Generation II FeCrAl Alloys

FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding application due to their exceptional resistance to oxidation in elevated temperature steam environments. Currently, limited fracture toughness data are available for the FeCrAl alloys, including the FeCrAl alloys newly developed at Oak Ridge National Laboratory (ORNL) under the U.S. Department of Energy’s Advanced Fuels Campaign (AFC) program. In this study, two Generation II candidate FeCrAl alloys, i.e., C06M (81.8Fe-10Cr-6Al-0.03Y-2Mo-0.2Si) and C36M (78.8Fe-13Cr-6Al-0.03Y-2Mo-0.2Si), were irradiated in the High Flux Isotope Reactor (HFIR) at ORNL to assess the fracture characteristics of these alloys after neutron irradiation. A total of six rabbit capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. Post-irradiation fracture toughness testing was performed following the Master Curve method in the ASTM E1921 standard. The main findings of this study are:1) Both the C06M and C36M alloys exhibited a similar response to irradiation concerning irradiation hardening and embrittlement.2) The irradiation temperature played different roles in terms of irradiation hardening and embrittlement for both C06M and C36M: after irradiation between 166°C and 204°C, both materials exhibited significant irradiation hardening and embrittlement; after irradiation between 315°C and 343°C, both materials showed small irradiation hardening without irradiation embrittlement. After irradiation between 501°C and 507°C, however, the irradiation softening without irradiation embrittlement was observed in both materials.3) Comparing the microhardness and Master Curve reference temperature T0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M steels. This should be mainly due to a flat response of the Master Curve reference temperature T0q to the irradiations at 166–204°C and 315–343°C ranges4) C06M showed a lower T0q, meaning better toughness, than C36M at the unirradiated condition, and such trend was kept even after neutron irradiation except for the 166–204°C irradiation after which both materials had similar T0q.5) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.

Chen, Xiang↗
NASA NTRSDate not supplied

Unraveling Solar Wind Space Weathering of Carbon-Rich Asterods: Low-Flux VS. High-Flux Ion Irradiation of Murchison

Introduction: Space weathering processesalter the microstructural, compositional, and optical properties of airless planetary surfaces.Solar wind space weatheringforms ion-damaged rims within the outermost ~100 nm of regolith grains through atomic-displacements and implantation of energetic H+and He+ions[1]. Transmission electron microscope(TEM)studies of space weathered Itokawa and lunar olivine particles reveal predominantly nanocrystalline solar wind-damaged rims with only minor regions of amorphization[2-5]. This finding differs from laboratoryion irradiation experiments in which samples become amorphous at fluences 2-3 orders of magnitude lower than the independentlyconstrained fluences for still-crystalline natural samples [6-7].Ion flux is generally considered to have a secondaryeffectcompared to fluence in these irradiation experiments. However, solar wind irradiation of natural samples occurs at a flux that is ~4-5 orders of magnitude lower than whatisused inlaboratory simulations (e.g., [6,8-9]). Constraining how the very low ion flux of solar wind influences the microstructural, chemical, and spectral alteration of silicate minerals will improve our understanding of solar wind space weathering and aid in the analysis of returned samples from the Hayabusa2 and OSIRIS-REx missions. Here we report results from low-and high-flux ion irradiation experiments on the Murchison (CM2)meteoritewithcomparable total fluences. Methods: We performed low-flux and high-flux He+(4 keV) and H+(1 keV) irradiation experiments on separate, dry-cut Murchison slabs. Low-flux He+irradiation reached a total fluence of 2.1 x 1016ions/cm2(~400 years of exposure at Bennu) using a flux of 3.6 x 1011ions/cm2/s; low-flux H+irradiation reached a total fluence of 3.9 x 1016ions/cm2(~30 years) using a flux of 6.6 x 1011ions/cm2/s. High-flux He+irradiation reached a total fluence of 2.0 x 1016ions/cm2(~400 years)using a flux of 9.1 x 1012ions/cm2/s; high-flux H+irradiation reached a total fluence of 5.8 x 1016ions/cm2(~50 years)using flux of 8.4 x 1012ions/cm2/s. By irradiating to similar total fluences in both the low-flux and high-flux experiments, we can better isolate and evaluatethe effects of ion flux. Differences between unirradiatedand irradiated surfaces are determined using threeanalytical techniques:(1) in situX-ray photoelectron spectroscopy(XPS) reveals changes in surface chemistry,(2) visible to near-infrared spectroscopy (VNIR; 0.35-2.5 μm) highlights changes in spectraltrends, such as albedo and slope, and (3) TEM and energy-dispersive X-ray spectroscopy (EDS) identifieschanges in nanoscale structure and elemental composition within olivine and matrix electron-transparent thin sections prepared using focused ion beam scanning electron microscopy (FIB-SEM). Results and Discussion: XPS analyses of the low-fluxand high-fluxHe+-and H+-irradiated samples showa minorreduction in surface carbon contentand chemicalreduction ofFe3+to Fe2+. The latter result is consistent with the higher Fe2+/Fe3+ratios observed in phyllosilicate-rich space weathered rims of Ryugu grains [10].VNIR analyses suggest that the low-flux H+-irradiated spectrum has a slightly bluer slope (over 0.65-2.5 μm) compared to the unirradiated VNIR spectrum from [9] while the low-flux He+-irradiated spectral slope is comparable to the unirradiated surface. Both the high-flux He+-and H+-irradiated VNIR spectra exhibitslightlyhigheralbedos at shorter wavelengths relativeto their unirradiated counterparts, with He+irradiation yielding a greater change. Thus far, the low-flux He+-irradiated matrix and olivine FIBsections as well as the low-flux H+-irradiated olivine FIBsection have been examined with TEM. The surfaces of all three samplesshowless vesiculation than high-flux/high-fluenceirradiated matrix and olivine samples from [9]. The depth of phyllosilicate amorphization in thelow-fluxHe+-irradiated matrix FIBsection varieslaterally, with crystalline domains occurring as shallow as ~25 nm in some regions. Thelow-flux He+-irradiated olivine FIBsection has acontinuous,~40-90 nm thick ion-affected layer with polycrystalline and amorphous microstructures. The low-flux H+-irradiated olivine sampleexhibits a distinct, ~10-40 nm ion-affected layer whose uppermost ~20 nmin thicker regions is predominantly amorphous.Continuing to compare results from this work to those from [9] will provide a comprehensive understanding of how ion flux impacts the microstructural, chemical, and optical modification of various mineral phases and, in turn,help identify space weathering features from parent body alteration in Hayabusa2 and OSIRIS-REx returned samples.

D L Laczniak↗
DOE OSTI2024

Discerning the effect of various irradiation modes on the corrosion of Zircaloy-4

Using proton irradiation, this study investigates the individual influence of several factors on the corrosion kinetics of Zircaloy-4 in a hydrogenated water environment simulating a Pressurized Water Reactor (PWR). Using both simultaneous irradiation-corrosion and autoclave corrosion, we separately examine: (i) the effect of pre-irradiation on modifying the structure of the material, (ii) the impact of irradiation on creating defects in the growing oxide layer during corrosion, and (iii) the influence of irradiation on increasing the corrosion potential through radiolysis during corrosion. To replicate neutron-irradiated microstructure, two proton pre-irradiation schedules were employed: Schedule 1 (isothermal irradiation at 350 °C to 5 dpa) to simulate high-temperature PWR conditions, and Schedule 2 (two-step process: irradiation to 2.5 dpa at -10 °C followed by 2.5 dpa at 350 °C) to simulate lower temperature PWR and Boiling Water Reactor (BWR) conditions. Long-term autoclave corrosion testing over 360 days at 320 °C revealed no significant difference between unirradiated samples and those pre-irradiated according to either schedule, with all samples exhibiting sub-cubic kinetics within the pre-transition regime. Irradiated samples underwent Simultaneous Irradiation Corrosion (SIC) tests, corroding in 320 °C water while being irradiated with protons. Corrosion was found to accelerate in all SIC-tested samples relative to autoclave conditions, with the greatest increase observed in non-pre-irradiated samples. Pre-irradiation with either schedule resulted in a slower corrosion rate compared to non-pre-irradiated regions under SIC conditions. The degree of radiolysis observed in the SIC tests surpassed typical PWR conditions, approaching levels found in BWRs. Radiolysis products were identified as the primary contributors to accelerated corrosion, corroborated by radiolysis bar tests. Furthermore, these findings underscore the intricate interactions between irradiation, corrosion, and water chemistry in determining Zircaloy-4 corrosion kinetics within nuclear reactor environments.

Pressurized Water Reactor↗
DOE OSTI2021

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR) 5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of the development and qualification of tri-structural isotropic (TRISO) low enriched fuel for use in high-temperature gas-cooled reactors (HTGR). The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR 5/6/7 experiment were to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel, beyond normal operating conditions. 3. Provide irradiated-fuel performance data and irradiated-fuel samples for post irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective is to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR 5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast (NE) flux trap of the ATR with a planned duration of 500 effective full-power days (EFPDs). The NE flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B?168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 EFPDs. Final burnup values, on a per-compact basis, ranged from 5.66% to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples (TC) had failed (the bottom three capsules lost all TCs). During the first five cycles (162B ? 165A), fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8?10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to the maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particle failures in Capsule 3, which was considered possible at the start of the irradiation, since the experiment was designed to operate beyond the high-temperature gas-cooled reactor (HTGR) normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

AGR-5/6/7 Irradiation Test Final As-run Report

This document presents the as-run analysis of the Advanced Gas Reactor (AGR)-5/6/7 irradiation experiment. AGR-5/6/7 is the last of a series of experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory in support of the development and qualification of tri-structural isotropic low-enriched fuel for use in high-temperature gas-cooled reactors. The test train contained five separate capsules that were independently controlled and monitored. Each capsule contained multiple 24.91-mm-long and 12.25-mm-dimeter compacts filled with low-enriched uranium carbide/oxide tri-structural isotropic fuel particles. The objectives of the AGR-5/6/7 experiment were to: • Irradiate reference-design fuel particles to support fuel qualification. • Establish operating margins for the fuel, beyond normal operating conditions. • Provide irradiated-fuel performance data and irradiated-fuel samples for post-irradiation examination and safety testing. The primary objective of the AGR-5/6 test (Capsules 1, 2, 4, and 5) was to verify the successful performance of the reference-design fuel under normal operating conditions. The AGR-7 test (Capsule 3) was designed to explore fuel performance at higher temperatures. Its primary objective was to demonstrate the capability of the fuel to withstand conditions beyond normal operating conditions, in support of plant design and licensing. AGR-5/6/7 will also provide irradiated-fuel performance data based on the fission gas release from particles during irradiation. To achieve the test objectives, the AGR-5/6/7 experiment was irradiated in the northeast flux trap of the ATR with a planned duration of 500 effective full-power days. The northeast flux trap was selected because its larger diameter provided greater flexibility for test-train design compared to the Large B positions used for the AGR-1 and AGR-2 irradiations, significantly enhancing test capabilities for the combined irradiation campaigns. Due to delays in the ATR schedule, the AGR-5/6/7 irradiation was significantly shorter than the originally planned 13-cycle schedule. Irradiation began on February 16, 2018 and ended on July 22, 2020, spanning nine ATR cycles (162B–168A) over two and a half years. Thus, the AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 effective full-power days. Final burnup values, on a per-compact basis, ranged from 5.66 to 15.26% fissions per initial heavy metal atom, while fast fluence values ranged from 1.62 to 5.55 × 1025 n/m2 (E >0.18 MeV). Time-averaged volume-averaged fuel temperatures on a capsule basis at the end of irradiation ranged from 756°C in Capsule 5 to 1313°C in Capsule 3 excluding days with significantly lower temperature during the two short powered axial locator mechanism cycles, 163A and 167A. By the end of irradiation, 48 out of 54 installed thermocouples had failed (the bottom three capsules lost all thermocouples). During the first five cycles (162B – 165A), the fission-gas isotope release-rate-to-birthrate (R/B) ratios were stable in the 10-8–10-6 range, and no in-pile particle failures were observed based on the gross gamma counts. During this time, the high exposed kernel fraction and high fuel particle temperatures in Capsule 1 led to a maximum R/B value of around 2 ? 10-6 for Kr-85m. The fission gas release in all capsules started to increase from the second half of Cycle 166A, when a large number of in-pile particle failures occurred in Capsule 1 and a gas line problem in this capsule caused fission gas leakage at various degrees into the other four capsules. This gas line problem also prevented a fission gas release measurement for Capsule 1 during the last three cycles due to gas flow isolation. By the end of irradiation, it is estimated that approximately 15 particles failed in Capsule 3, which was considered possible because the experiment was designed to operate beyond the high-temperature gas-cooled reactor normal operating temperature range. A few hundred in-pile particle failures were estimated for Capsule 1 by the end of Cycle 166A, but the total number of failures is unknown due to the lack of fission gas release data in the later cycles. Additionally, four potential in-pile failures were identified for Capsule 2 during the last cycle, Cycle 168A. In contrast, no in-pile failures were identified in the top two capsules (4 and 5) based on the absence of the typical spikes in gross gamma counts and low failure estimates using the AGR-3/4 R/B per exposed kernel model.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1996

Early Energetic Particle Irradiation of the HED Parent Body Regolith

Previous studies have shown that many individual grains within the dark phase of the Kapoeta howardite were irradiated with energetic particles while residing on the surface of the early HED regolith. Particle tracks in these grains vary in density by more than an order of magnitude and undoubtedly were formed by energetic heavy (Fe) ions associated with early solar flares. Early Irradiation of HED Regolith: Concentrations of excess Ne alone are not sufficient to decide between competing galactic and solar irradiation models. However, from recent studies of depth samples of oriented lunar rocks, we have shown that the cosmogenic 21-Ne/22-Ne ratio produced in feldspar differs substantially between Galactic Cosmic Radiation (GCR) and solar protons, and that this difference is exactly that predicted from cross-section data. Using Ne literature data and new isotopic data we obtained on acid-etched, separated feldspar from both the light and dark phases of Kapoeta, we derive 21-Ne/22-Ne = 0.80 for the recent GCR irradiation and 21-Ne/22-Ne = 0.68 for the early regolith irradiation. This derived ratio indicates that the early Ne production in the regolith occurred by both galactic and solar protons. If we adopt a likely one-component regolith model in which all grains were exposed to galactic protons but individual grains had variable exposure to solar protons, we estimate that this early GCR irradiation lasted for about 3-6 m.y. More complex two-component regolith models involving separate solar and galactic irradiation would permit this GCR age to be longer. Higher-energy solar protons would permit the GCR to be longer. Higher-energy solar protons would permit the GCR age to be shorter. Further, cosmogenic 126(Xe) in Kapoeta dark is no more than a factor of about 2 higher than that observed in Kapoeta light. Because 126(Xe) can only be formed by galactic protons and not solar protons, these data support a short GCR irradiation for the HED regolith. This would also be the maximum time peRiod for the solar irradiation. Various asteroidal regolith models, based on Monte Carlo modeling of impact rates as a function of size and on irradiation features of meteorites, predict surface exposure times of about 0.1 to 10 m.y., and depend on such factors as gravity, rock mechanical properties, and micrometeoroid flux. Because the depth at which solar Fe tracks are produced (is much less than 1 micrometer) is much less than the depth at which Solar Cosmic Rays (SCR) Ne is produced (about 1 cm), for a reasonably well-stirred HED regolith the "surface exposure time" for SCR 21-Ne production should be significantly longer than that for solar tracks and some other surface irradiation features. Enhanced Solar Proton Irradiation: For bulk samples of Kapoeta dark feldspar and a one-component regolith model, the derived ratio of 21-Ne/22-Ne = 0.68 implies that the early production ratio of SCR 21-Ne to GCR 21-Ne was about 0.5-1.5. This ratio is independent of any assumptions about the fraction of dark grains that are irradiated or of the variability in the degree of solar irradiation among grains. The 21-Ne SCR/GCR ratio indirectly derived from bulk Kapoeta pyroxene is somewhat larger, as is the ratio derived for simple two-component regolith models. Individual feldspar grains that were extensively solar irradiated would require even larger 21-Ne SCR/GCR production ratios. In contrast, the theoretical SCR/GCR production ratio for lunar feldspar with 0 g/CM2 shield ing is is less than or equal to 2, and the lowest ratio observed in near-surface samples of lunar anorthosites is less than or equal to 1. Considering the greater solar distance of Vesta (compared to the Moon), the likelihood that SCR 21-Ne was acquired under some shielding where production rates are lower, and the likelihood that the exposure time to galactic protons exceeded the exposure time to solar protons because of their very different penetration depths, the 21-Ne SCR/GCR production ratio on the HED parent body was probably < 0.1. The relatively large difference between the derived 21-Ne SCR/GCR ratio in Kapoeta dark feldspar and the estimated production ratio strongly indicates that the early solar irradiation involved a flux -20-50x the recent solar flux. This enhanced proton flux was probably associated with an overall greater solar activity in the first approximately 10(exp 7) to 10(exp 8) years of solar history.

Bogard, D. D.↗
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 OSTI2021

Data Report on Post-Irradiation Dimensional Change of AGC-1 Samples

This report documents the measured post irradiation dimensional change in the AGC-1 samples. The AGC-1 capsule is the first in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. AGC-1 irradiation began September 5, 2009 in the Advanced Test Reactor (ATR) and was completed on January 8, 2011. The capsule was cooled for 3 months in the ATR Canal, and then shipped to MFC in April 2011 for disassembly and sample extraction. After extraction the samples were shipped to the INL Research Center (IRC) for initial post-irradiation examination (PIE) and storage in the irradiated graphite vault. The AGC-1 capsule design contained “matched pair” samples to ascertain the irradiation-induced dimensional changes and levels of creep experienced in different graphite types. The irradiation-induced dimensional changes and creep levels are determined by comparing the total dimensional change for stressed and unstressed samples of the same type of graphite exposed to the same dose levels and at similar temperatures. Under irradiation creep (i.e. permanent strain due to irradiation, stress, and temperature) the stressed samples should demonstrate more dimensional change than the unstressed samples. This additional dimensional change in the stressed samples is designated as “irradiation-induced creep” in graphite. The data are further presented using the parameters influencing dimensional change in graphite; levels of induced stress, temperature, graphite type, and dose. However, the AGC-1 post-irradiation examination is a significant endeavor and this data report serves to provide irradiation-induced dimensional change data for AGC capsule design refinement as well as a status on the progress of the PIE activities. The dimensional changes of both the samples and graphite body are very important to the design of the future AGC capsules (AGC-3 through AGC-6) and are provided as soon as the data are available in order to determine whether design changes to the next capsule are required. A complete evaluation of the irradiation-induced dimensional change data will be performed for a final AGC-1 PIE report that will include full analysis of pre- and post-irradiation data, with verified AGC-1 irradiation conditions of temperature and dose.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Plasticity in irradiated FeCrAl nanopillars investigated using discrete dislocation dynamics

Here, in this paper, we investigate plasticity in irradiated FeCrAl nanopillars using discrete dislocation dynamics simulations (DDD), with comparisons to transmission electron microscopic (TEM) in situ tensile tests of ion- and neutron-irradiated commercial C35M FeCrAl alloy. The effects of irradiation-induced defects, such as a/2<111> and a<100> type loops and composition fluctuations representative of phase separation in irradiated FeCrAl alloys, are investigated separately as well as superposed together in simulations. We explore the effects of defects on the stress-strain behavior, specifically yield strength and hardening response, of FeCrAl nanopillars. Our simulations confirm the widely accepted fact that irradiated alloys exhibit a stress-strain response with higher yield strength as compared to unirradiated alloys. However, our DDD calculations reveal an atypical superposition of the hardening contributions due to composition inhomogeneity and irradiation loops wherein composition inhomogeneity annihilates the hardening due to irradiation loops at small scales. As a result, we observe that the yield strength in irradiated alloys, after taking into consideration the effects of both composition inhomogeneity and irradiation loops, is smaller than the yield strength of the alloys with only irradiation loops and is approximately same for the alloy with composition inhomogeneity alone. This is referred to as “destructive interference” between the hardening contributions due to composition fluctuations and irradiation loops in the paper. We also identify this destructive interference in the superposition in our parallel TEM in situ tensile tests on unirradiated, ion-irradiated, and neutron-irradiated C35M FeCrAl alloy. This destructive interference in the hardening contributions contrasts with the dispersed barrier hardening (DBH) models widely utilized by the experimental community to model the hardening contributions due to different irradiation induced defects. The effects of the loading orientations on the yield strength and hardening are investigated and the mechanisms for the hardening in irradiated FeCrAl alloys are also reported.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Emulation of neutron damage with proton irradiation and its effects on microstructure and microchemistry of Zircaloy-4

This work assesses the potential of proton irradiation to simulate the neutron damage to the matrix and laves phase Zr(Fe,Cr) 2 precipitates in Zircaloy-4. Isothermal proton irradiation has been performed on Zircaloy-4 samples at irradiation temperatures ranging from 250 to 350 °C. Two-step proton irradiation was also performed to enhance the amorphization of and iron loss from the laves phase Zr(Fe,Cr) 2 precipitates. The irradiated microstructures, including dislocation loops and rafts near SPPs, were observed in proton irradiated Zircaloy-4, which are consistent with neutron irradiated material at a similar damage level. The amount of irradiation-induced hardening after proton irradiation was similar to post neutron irradiated data. The significant amorphization of the SPPs and concurrent Fe redistribution observed on neutron irradiated materials can be effectively emulated using a two-step proton irradiation on Zircaloy-4. Hence, the neutron irradiation effect on Zircaloy-4 can be mostly captured using the two-step proton irradiation described in this study.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Investigation of ion irradiation effects on mineral analogues of concrete aggregates

Irradiation can cause prominent damage to reactor concrete aggregates leading to amorphization, strength and modulus decrease, radiation induced volume expansion (RIVE) and micro-cracking, which limits their long-term performance. Here, to develop an improved understanding of irradiation effects in concrete, three mineral analogues of concrete aggregates (limestone, marble and quartzite) were irradiated by 5.5 MeV He ions and 13 MeV Ni ions to surface doses of 0.011 displacements per atom (dpa) and 0.23 dpa, respectively, at room temperature. The two different ion species allow irradiation spectrum effects (ionizing and displacive) to be examined. Irradiation induced cracks were observed in He irradiated limestone and marble, and Ni irradiated quartzite. Full amorphization was observed in Ni irradiated quartzite with 14.3 % RIVE, and ∼25 % hardness and modulus decrease, while almost no change was observed in He irradiated quartzite except 4.35 % RIVE, revealing a possible ionization enhanced diffusion effect for high energy light ions. Furthermore, partial amorphization was observed in Ni irradiated marble and limestone matrix with a 12 % hardness decrease in marble while no amorphization was observed for He irradiation with a 20 % hardness increase in limestone matrix. The role of knock-on damage and irradiation spectrum on amorphization, volumetric expansion and mechanical property changes are discussed. Moreover, the onset and critical doses for amorphization and RIVE in quartz are obtained for ion irradiations at room temperature. The dose dependence of RIVE exhibits a delay compared to the amorphization behavior. The superior irradiation resistance of calcite phase compared to quartz phase implies there could be advantages to using calcareous aggregates and lowering the usage of siliceous aggregates for concrete in nuclear power plants for extended operation beyond 60 years. However, other effects such as corrosion, aging and reactions during severe accidents should also be considered, and further investigations are needed.

Amorphous↗
DOE OSTI2022

Comparison and calibration of dose delivered by 137 Cs and x-ray irradiators in mice

Objective. The Office of Radiological Security, U.S. Department of Energy's National Nuclear Security Administration, is implementing a radiological risk reduction program which seeks to minimize or eliminate the use of high activity radiological sources, including 137 Cs, by replacing them with non-radioisotopic technologies, such as x-ray irradiators. The main goal of this paper is to evaluate the equivalence of the dose delivered by gamma- and x-ray irradiators in mice using experimental measurements and Monte Carlo simulations. We also propose a novel biophantom as an in situ dose calibration method. Approach. We irradiated mouse carcasses and 3D-printed mouse biophantoms in a 137 Cs irradiator (Mark I-68) and an x-ray irradiator (X-Rad320) at three voltages (160 kVp, 225 kVp and 320 kVp) and measured the delivered radiation dose. A Geant4-based Monte Carlo model was developed and validated to provide a comprehensive picture of gamma- and x-ray irradiation in mice. Main Results. Our Monte Carlo model predicts a uniform dose delivered in soft-tissue for all the explored irradiation programs and in agreement with the absolute dose measurements. Our Monte Carlo model shows an energy-dependent difference between dose in bone and in soft tissue that decreases as photon energy increases. Dose rate depends on irradiator and photon energy. We observed a deviation of the measured dose from the target value of up to –9% for the Mark I-68, and up to 35% for the X-Rad320. The dose measured in the 3D-printed phantoms are equivalent to that in the carcasses within 6% uncertainty. Significance. Our results suggest that 320 kVp irradiation is a good candidate to substitute 137 Cs irradiation barring a few caveats. There is a significant difference between measured and targeted doses for x-ray irradiation that suggests a strong need for in situ calibration, which can be achieved with 3D-printed mouse biophantoms. A dose correction is necessary for bone doses, which can be provided by a Monte Carlo calculation. Lastly, the biological implications of the differences in dose rates and dose per photon for the different irradiation methods should be carefully assessed for each small-animal irradiation experiment.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

Mechanical behavior of additively manufactured and wrought 316L stainless steels before and after neutron irradiation

Fabrication of nuclear reactor components using additive manufacturing (AM) methods is now a practical option since the AM technologies have advanced to allow for building of complex parts with high quality materials. To assess the mechanical performance of printed components in reactor-relevant conditions and to build a property database for the AM 316L stainless steel (SS), mechanical testing and characterization were performed before and after neutron irradiation. In this work, miniature tensile specimens were irradiated at the High Flux Isotope Reactor (HFIR) to 0.2 and 2 displacements per atom (dpa) at 300 and 600°C. The AM 316L SS was tested in the as-built, stress-relieved, and solution-annealed conditions, and the wrought (WT) 316L SS in solution-annealed condition as a reference alloy. The baseline test result showed that the AM 316L SS, regardless of the post-build heat treatment, had higher strength than the WT 316L SS, but similar ductility. Post-irradiation tensile testing was conducted at RT, 300°C, and 500°C for selected irradiation conditions. Neutron irradiation induced significant changes in the mechanical behavior of the AM stainless steels, including both hardening and softening. Although the as-built 316L steel after 300°C irradiation showed necking just after yielding, the overall property changes of the as-printed alloy became less significant after 600°C irradiation. Irradiation-induced ductilization was also observed after the higher temperature irradiation. In general, the strength change was smaller in the relatively stronger as-built and stress-relieved AM SSs than in the solution-annealed AM and WT SSs. These relatively lower strength 316L SSs overall retained higher ductility in the irradiation conditions tested, but the stronger 316L SSs demonstrated a similar level of ductility after the higher temperature (600°C) irradiation. It is a positive assessment for the AM 316L materials that no embrittlement was observed within the test and irradiation conditions of the experiment.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Post-irradiation examination of AGR-3/4 TRISO fuel compacts using three-dimensional X-ray computed tomography

The AGR-3/4 irradiation tests combined the third and fourth planned irradiation experiments in the US Department of Energy’s Advanced Gas Reactor (AGR) testing campaign of tri-structural isotropic (TRISO) fuel compacts. In this article, we present post-irradiation examination (PIE) using X-ray computed tomography (XCT) of two unirradiated and two irradiated compacts from the AGR-3/4 irradiation tests. The irradiated compacts studied (compact 7–1 and compact 12–4) represent the upper and lower limit of burnup within the AGR-3/4 irradiation experiment. This article presents a detailed quantitative analysis on the post-irradiation structure of TRISO fuel compacts. Various quantitative parameters including shape, size, and packing of kernels, and their spatial distribution, were utilized to gain insights into the structural changes caused by irradiation. The equivalent diameter and sphericity were found to increase and decrease, respectively, in irradiated compact 7–1 due to its higher burnup. Nearest neighbor distance between fuel kernels decreased after irradiation, suggesting irradiation-induced shrinkage of graphitic matrix. Furthermore, each compact in AGR-3/4 irradiation tests contained 20 designed-to-fail (DTF) fuel particles that were meant to act as a source of fission product release to the experiment test train. Furthermore, in the present work, all DTF fuel particles in the four compacts studied were identified, and it was found that they exhibited larger kernel swelling in compact 12–4 and smaller kernel swelling in compact 7–1, compared to the driver particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1995

Long-term total solar irradiance variability during sunspot cycle 22

Total solar irradiance measurements from the 1984-1993 Earth Radiation Budget Satellite (ERBS) active cavity radiometer and 1978-1993 Nimbus 7 transfer cavity radiometer spacecraft experiments are analyzed to detect the presence of 11-, 22-, and 80-year irradiance variability components. The analyses confirmed the existence of a significant 11-year irradiance variability component, associated with solar magnetic activity and the sunspot cycle. The analyses also suggest the presence of a 22- or 80-year variability component. The earlier Nimbus 7 and Solar Maximum Mission (SMM) spacecraft irradiance measurements decreased approximately 1.2 and 1.3 W/sq m, respectively, between 1980 and 1986. The Nimbus 7 values increased 1.2 W/sq m between 1986 and 1989. The ERBS irradiance measurements increased 1.3 W/sq m during 1986-1989, and then decreased 0.4 W/sq m (at an annual rate of 0.14 W/sq. m/yr) during 1990-1993. Considering the correlations between ERBS, Nimbus 7, and SMM irradiance trends and solar magnetic activity, the total solar irradiance should decrease to minimum levels by 1997 as solar activity decreases to minimum levels, and then increase to maximum levels by the year 2000 as solar activity rises. The ERBS measurements yielded 165.4 +/- 0.7 W/sq m as the mean irradiance value with measurement accuracies and precisions of 0.2% and 0.02%, respectively. The ERBS mean irradiance value is within 0.2% of the 1367.4, 1365.9, and 1366.9 W/sq m mean values for the SMM, Upper Atmosphere Research Satellite (UARS), and Space Shuttle Atmospheric Laboratory for Applications and Science (ATLAS 1) Solar Constant (SOLCON) active cavity radiometer spacecraft experiments, respectively. The Nimbus 7 measurements yielded 1372.1 W/sq m as the mean value with a measurement accuracy of 0.5%. Empirical irradiance model fits, based upon 10.7 -cm solar radio flux (F10) and photometric sunspot index (PSI), were used to assess the quality of the ERBS, Numbus 7, SMM, and the UARS irradiance data sets and to identify irradiance variability trends which may be caused by drifts or shifts in the spacecraft sensor responses. Comparisons among the fits and measured irradiances indicate that the Nimbus 7 radiometer response shifted by a total of 0.8 W/sq m between September 1989 and April 1990 and that the ERBS and UARS radiometers each drifted approximately 0.5 W/sq m during the first 5 months in orbit.

Lee, Robert B., III↗
NASA NTRS1996

The effect of in vivo and in vitro irradiation (25 Gy) on the subsequent in vitro growth of satellite cells

The effect of in vivo and in vitro irradiation on subsequent satellite cell growth, in vitro, was investigated to ascertain the ability of a 25 Gy dose to inhibit satellite cell proliferation. Satellite cells were isolated from the left (irradiated) and right (non-irradiated) Pectoralis thoracicus of two-week-old tom turkeys 16 h (n=3) and seven weeks (n=2) after the left Pectoralis thoracicus had been irradiated (25 Gy). Satellite cells isolated from the irradiated and non-irradiated muscles exhibited similar (P>0.10) in vitro proliferation indicating that a population of satellite cells survived an in vivo dose of 25 Gy. In additional experiments, satellite cell cultures derived from tom turkey Pectoralis thoracicus were irradiated (25 Gy) in vitro. The number of satellite cells did not (P>0.05) increase in irradiated cultures for 134 h following irradiation, while satellite cells in non-irradiated cultures proliferated (P<0.05) over this time. At later time periods, satellite cell number increased (P<0.05) in irradiated cultures indicating that a population of satellite cells survived irradiation. The results of these in vitro experiments suggest that a 25 Gy dose of irradiation does not abolish satellite cell divisions in the turkey Pectoralis thoracicus.

NASA Discipline Musculoskeletal↗
DOE OSTI2023

Microstructural evolution and transmutation in tungsten under ion and neutron irradiation

This study aims to compare the effects of neutron and self-ion irradiation on the mechanical properties and microstructural evolution in W. Neutron irradiation at the HFR reactor to 1.67 dpa at 800 °C resulted in the formation of large Re and Os rich clusters and voids. The post-irradiation composition was measured using APT and verfified against FISPACT modelling. The measured Re and Os concentration was used to create alloys with equivalent concentrations of Re and Os. These alloys were exposed to self-ion irradiation to a peak dose of 1.7 dpa at 800 °C. APT showed that self-ion irradiation leads to the formation of small Os clusters, wheras under neutron irradiation large Re/Os clusters form. Voids are formed by both ion and neutron irradiation, but the voids formed by neutron irradiation are larger. By comparing the behaviour of W-1.4Re and W-1.4Re-0.1Os, suppression of Re cluster formation was observed. Irradiation hardening was measured using nanoindentation and was found to be 2.7 GPa, after neutron irradiation and 1.6 GPa and 0.6 GPa for the self-ion irradiated W-1.4Re and W-1.4Re-0.1Os. The higher hardening is attributed to the barrier strength of large voids and Re/Os clusters that are observed after neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Irradiation creep measurement and microstructural analysis of chromium nitride–coated zirconium alloy using pressurized tubes

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. Here, to accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300°C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8 $\times$ 10 21 n/cm 2 fast fluence (E n > 10 MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

36 MATERIALS SCIENCE↗