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Irradiation effects of high energy protons studied on silver-cadmium and nickel-cadmium cells containing battery electrodes and potassium hydroxide electrolytes
Test facilities for studying electron irradiation effects on lithium-doped silicon solar cells
FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding applications due to their excellent corrosion resistance to the elevated temperature steam environment. Currently, the handbook on FeCrAl material properties contains only limited data regarding the fracture toughness properties of any FeCrAl alloy. This includes alloys currently under investigation within the Advanced Fuels Campaign (AFC) at Oak Ridge National Laboratory (ORNL). In this project, a series of irradiation capsules have been irradiated in the High Flux Isotope Reactor (HFIR) at ORNL with two Generation II FeCrAl candidate alloys, i.e., C06M and C36M, to assess the fracture response of these alloys after neutron irradiation. These alloys represent the “book-end” compositions for C26M, the alloy currently being developed as the leading candidate for LWR cladding. A total of six irradiation 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. These damage doses represent the expected middle and end of life damage levels for typical LWR cladding while the irradiation temperature regimes will provide insight into the role of varying microstructural features on the fracture toughness properties of neutron irradiated FeCrAl alloys. To date, irradiation of all capsules has been completed in HFIR. This report summarizes the latest results of microhardness and fracture toughness PIE for the 16 dpa capsules (FCAB2, FCAB4, and FCAB6), for which the measured irradiation conditions were: 204°C/17.6dpa, 343°C/18.3dpa, and 507°C/18.6dpa. The main conclusions of this study can be summarized as follows: 1) After the 204°C/17.6dpa irradiation, both C06M and C36M exhibited significant irradiation hardening and embrittlement 2) After the 343°C/18.3dpa irradiation, both C06M and C36M exhibited small irradiation hardening without irradiation embrittlement 3) After the 507°C/18.6dpa irradiation, both C06M and C36M exhibited irradiation softening without irradiation embrittlement 4) Comparing the microhardness and Master Curve reference temperature T 0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M. This should be mainly due to a flat response of the Master Curve reference temperature T 0q to the irradiations at 166-204°C and 315-343°C ranges 5) C06M showed a lower T 0q , 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 where both materials had similar T 0q . 6) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.
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.
Advancements were made in automating microscopy-based material characterization, particularly in studying irradiation effects on additively manufactured (AM) materials using machine learning (ML) and computer vision (CV). These automation efforts address the challenges of analyzing complex microstructures, accelerating the detection of irradiation-induced defects. Two CV models were developed at Argonne National Laboratory (ANL) to enhance transmission electron microscopy (TEM) analysis of irradiated AM 316 stainless steel. The first model focused on the detection of irradiation-induced dislocation loops, which contribute to material hardening and embrittlement. These loops, categorized as faulted or perfect, were automatically detected and classified using a Mask R-CNN model trained on TEM images from both in-situ and ex-situ ion irradiation experiments. The model achieved high accuracy, with precision, recall, and F1 scores of 0.839, 0.734, and 0.776, respectively, demonstrating its effectiveness in analyzing dislocation loops in irradiated AM materials. The second CV model was developed to analyze the size and wall thickness of dislocation cells in laser powder bed fusion (LPBF) 316 stainless steel. Using a U-Net++ architecture with EfficientNet as the encoder, the model was trained on TEM images to segment and measure cell size and wall thickness.
A high efficient In0.48Ga0.52P/In0.01Ga0.99As/Ge triple junction solar cell has been developed for application in space and terrestrial concentrator PV system [1-3]. Recently, a high conversion efficiency of 31.5% (AM1.5G) has been obtained in InGaP/(In)GaAs/Ge triple junction solar cell, and as a new top cell material of triple junction cells, (Al)InGaP [1] has been proposed to improve the open-circuit voltage (Voc) because it shows a higher Voc of 1.5V while maintaining the same short-circuit current (ISC) as a conventional InGaP top cell under AM1.5G conditions as seen in figure 1 (a). Moreover, the spectral response of 1.96eV AlInGaP cell with a thickness of 2.5..m shows a higher response in the long wavelength region, compared with that of 1.87eV InGaP cell with 0.6..m thickness, as shown in figure 1 (b). Its development will realize next generation multijunction (MJ) solar cells such as a lattice mismatched AlInGaP/InGaAs/Ge 3-junction and lattice matched AlInGaP/GaAs/InGaAsN/Ge 4-junction solar cells. Figure 2 shows the super high-efficiency MJ solar cell structures and wide band spectral response by MJ solar cells under AM1.5G conditions. For realizing high efficient MJ space solar cells, the higher radiation-resistance under the electron or proton irradiation is required. The irradiation studies for a conventional top cell InGaP have been widely done [4-6], but little irradiation work has been performed on AlInGaP solar cells. Recently, we made the first reports of 1 MeV electron or 30 keV proton irradiation effects on AlInGaP solar cells, and evaluated the defects generated by the irradiation [7,8]. The present study describes the recovery of 1 MeV electron / 30 keV proton irradiation-induced defects in n+p- AlInGaP solar cells by minority-carrier injection enhanced annealing or isochronal annealing. The origins of irradiation-induced defects observed by deep level transient spectroscopy (DLTS) measurements are discussed.
Low dose reactor irradiation effect on temperature dependence of dynamic modulus and internal friction of as-deposited pyrolytic graphite
Neutron irradiation effect on space charge limited current of electrons in high purity silicon, noting SCLC sensitivity in detecting traps and changes
Subthreshold electron irradiation effects on surface and bulk electrical properties of Li-doped and Li-free silicon photovoltaic cells
The effect of irradiation on the light transmittance of ethylene-propylene-A (FEP-A) was investigated by measuring the short-circuit current of the cells after each of several electron dose increments, immediately after total irradiation, and 16 hours after total irradiation. Results indicated no apparent overall loss in transmission due to irradiation of FEP-A. However, filter wheel measurements revealed a 'darkening' of the FEP-A at the blue end of the spectrum. Some embrittlement of the FEP-A occurred at an accumulated dose of 67,500,000 rads.
The concrete biological shield (CBS) of light-water reactors protects workers and the surrounding environment by absorbing neutron and gamma irradiation emitted from the reactor core. The radiation dose increases with the CBS’s operational time and, in the long term, becomes significant enough to raise the question of irradiation effects on concrete—and particularly on the structural integrity of the CBS. Irradiation-induced damage has been identified as one of the main degradation mechanisms in the CBS. Neutron radiation causes the swelling of aggregate-forming minerals at different rates and amplitudes depending on the mineral’s nature. Silicate-bearing minerals such as quartz are particularly sensitive to neutron radiation and experience up to 17.8% volumetric expansion. Aggregates comprise several minerals with different orientations and are, therefore, subject to cracking as a result of mismatch strains. Additionally, the swelling of aggregates creates significant stresses in the surrounding cement paste matrix, which also results in crack formation. In parallel with the collection of characterization and irradiation test data, development of modeling and simulation tools for irradiated concrete is ongoing with the support of the US Department of Energy Office of Nuclear Energy’s Light Water Reactor Sustainability (LWRS) program. This effort resulted in the development and application of the fast-Fourier transform (FFT)–based code Microstructure-Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) at Oak Ridge National Laboratory.
X-ray irradiation effect on glycogen content of rat brain
Radiation-enhanced precipitation of Cr-rich α’ in irradiated Fe-Cr alloys, which results in hardening and embrittlement, depends on the irradiating particle and the displacement per atom (dpa) rate. Here, we utilize a Cahn-Hilliard phase-field based approach, that includes simple models for nucleation, irradiating particle and rate dependent radiation-enhanced diffusion and cascade mixing to simulate α’ evolution under neutrons, heavy ions, and electron irradiations. Different irradiating particles manifest very different cascade mixing efficiencies. The model was calibrated using neutron data. For cascade inducing neutron/heavy-ion dpa rates at 300 °C between 10-8 and 10-6 dpa/s the model predicts approximately constant number density, decreasing radius, decreasing α’ Cr composition, and lower α’ volume fraction. The model then predicts a dramatic transition to no α’ formation above approximately 10-5 dpa/s, while electron irradiation, with weak mixing, had little effect at dpa rates up to 10-3 dpa/s. These model predictions are consistent with experiments. We explain the results in terms of the flux dependence of the radiation enhanced diffusion, cascade mixing, and their ratio, which all vary significantly in relevant flux ranges for neutron and cascade inducing ion irradiations. These results show that both cascade mixing and radiation enhanced diffusion must be accounted for when attempting to emulate neutron-irradiation effects using accelerated ion irradiations.
Even though nanocrystalline materials (20-100 nm) present an unprecedented potential, scientific knowledge related to the effect of neutron irradiation on the mechanical properties and microstructure is still scarce. Most of the past studies were conducted using ion irradiation which may not have the same effect as neutron irradiation because of the smaller irradiation volume and the higher dose rate. To reach a firm conclusion on the potential of nanocrystalline materials for nuclear reactor applications, extensive study of model metals with different stacking fault energy (SFE) is required to elucidate their behavior in radiation environments. Nanocrystalline copper and nickel are typically chosen because they are commonly used as model FCC metals in studies of radiation effects. Nickel is an FCC metal with a high stacking-fault energy (~125 mJ/m 2 ) compared to copper (~45 mJ/m 2 ). Hence, microcrystalline and nanocrystalline nickel samples were irradiated in the INL’s Advanced Test Reactor (ATR), as a part of FY08 North Carolina State University NSUF Irradiation Experiment# 96 to evaluate the irradiation behavior of these materials. The objective of this FY20 NSUF project# 19122 is to perform PIE (at PNNL NSUF facility) on previously ATR-neutron irradiated (1.2 and 2.6 dpa; 80-89°C) nanocrystalline and microcrystalline nickel samples to investigate the changes in mechanical properties and microstructures and evaluate whether nanocrystalline nickel is relatively more radiation resistant compared to conventional microcrystalline nickel. To perform PIE at PNNL, sixteen neutron irradiated specimens (microcrystalline and nanocrystalline) were transferred from the NSUF Nuclear Fuels and Materials Library at INL. Experimental techniques such as SEM/EBSD, XRD, TEM, Vickers microhardness and tensile testing were employed to characterize the effect of neutron irradiation on the microstructure and mechanical properties of nanocrystalline nickel and compared them with corresponding characteristics of microcrystalline nickel.
Particles comprised of a thermal neutron absorbing material (HfAl3) are dispersed in a metal matrix material with high thermal conductivity (aluminum) to conduct the heat generated by neutron capture away from the fuel and materials. This metal matrix composite is very promising for use as a conduction-cooled neutron absorber and has the potential to be useful as a shroud or heat sink for testing advanced fast reactor fuels and materials in an existing thermal reactor. To design and optimize the absorber block system for advanced reactor designs, fundamental understanding of the irradiation effect on material properties is necessary. This dataset contains TEM characterization results of neutron irradiated HfAl3-Al composite specimens. It is focused on the irradiation induced defects (dislocation lines and loops) characterization using the on-zone axis bright field STEM technique. This data was collected using a FEI Tecnai G2 F30 S/TEM at the Microscopy and Characterization Suite (MaCS), Center for Advanced Energy Studies (CAES).
Particles comprised of a thermal neutron absorbing material (HfAl3) are dispersed in a metal matrix material with high thermal conductivity (aluminum) to conduct the heat generated by neutron capture away from the fuel and materials. This metal matrix composite is very promising for use as a conduction-cooled neutron absorber and has the potential to be useful as a shroud or heat sink for testing advanced fast reactor fuels and materials in an existing thermal reactor. To design and optimize the absorber block system for advanced reactor designs, fundamental understanding of the irradiation effect on material properties is necessary. This dataset contains TEM characterization results of neutron irradiated HfAl3-Al composite specimens. It is focused on the irradiation induced defects (dislocation lines and loops) characterization using the on-zone axis bright field STEM technique. This data was collected using a FEI Tecnai G2 F30 S/TEM at the Microscopy and Characterization Suite (MaCS), Center for Advanced Energy Studies (CAES).
Environmental facility for real-time study of electron irradiation effects on Li-doped p-n silicon solar cells
Several empirical models of sunspot and facular irradiance effects were tested by assessing the degree of correlation between variations in the total solar irradiance, as measured by the active cavity radiometer irradiance monitor on the SMM and the measures of magnetic activity on the solar disk. This was done by analyzing images made during 21 days between June 20 and July 14, 1988. The paper also describes the instruments and the methods used to gather the ground-based photometric images, as well as the analysis procedure.