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At least 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Phase stability and microstructure of neutron-irradiated substoichiometric yttrium dihydrides

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

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

In situ high-energy X-ray diffraction study of tensile deformation of neutron-irradiated polycrystalline Fe-9%Cr alloy

The effect of neutron irradiation on tensile deformation of a Fe-9wt.%Cr alloy was investigated using in situ high-energy synchrotron X-ray diffraction during room-temperature uniaxial tensile tests. New insights into the deformation mechanisms were obtained through the measurements of lattice strain evolution and the analysis of diffraction peak broadening using the modified Williamson-Hall method. Two neutron-irradiated specimens, one irradiated at 300 °C to 0.01 dpa and the other at 450 °C to 0.01dpa, were tested along with an unirradiated specimen. The macroscopic stress–strain curves of the irradiated specimens showed increased strength, reduced ductility and work-hardening exponent compared to the unirradiated specimen. The evolutions of the lattice strain, the dislocation density and the coherent scattering domain size in the deformation process revealed different roles of the submicroscopic defects in the 300°C/0.01 dpa specimen and the TEM-visible nanometer-sized dislocation loops in the 450°C/0.01 dpa specimen: submicroscopic defects extended the linear work hardening stage (stage II) to a higher strain, while irradiation-induced dislocation loops were more effective in dislocation pinning. Lastly, while the work hardening rate of stage II was unaffected by irradiation, significant dynamic recovery in stage III in the irradiated specimens led to the early onset of necking without stage IV as observed in the unirradiated specimen.

36 MATERIALS SCIENCE↗
DOE OSTI2018

In-situ high-energy X-ray characterization of neutron irradiated HT-UPS stainless steel under tensile deformation

Here, the tensile deformation behavior of a high-temperature, ultrafine-precipitate strengthened (HT-UPS) stainless steel was characterized in-situ with high-energy X-ray diffraction at 20 and 400 °C. The HT-UPS samples were neutron irradiated to 3 dpa at 400 °C. Significant irradiation hardening and ductility loss were observed at both temperatures. Lattice strain evolutions of the irradiated samples showed a strong linear response up to near the onset of the macroscopic yield, in contrast to the unirradiated HT-UPS which showed a pronounced non-linear behavior well below the macroscopic yield. While the room temperature diffraction elastic moduli in the longitudinal direction increased after irradiation, the 400 degrees C moduli were similar before and after irradiation. The evolution of the {200} lattice strain parallel to the loading axis (ε L {200} ) showed unique characteristics: in the plastic regime, the evolution of (ε L {200} ) after yield is temperature-dependent in the unirradiated specimens but temperature-independent in the irradiated specimens; and the value of (ε L {200} ) at the yield is an irradiation-sensitive, temperature-independent . parameter. The evolution of (ε L {200} ) corresponds well with the dislocation density evolution, and is an effective probe of the deformation-induced long-range internal stresses in the HT-UPS steel.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Transmutation-induced precipitation in tungsten irradiated with a mixed energy neutron spectrum

Transmutation-induced precipitation in neutron-irradiated tungsten is an important performance concern for its application as plasma facing material in fusion reactors. In this study, segregation and precipitation of transmutant elements in single crystal and polycrystal tungsten irradiated at 460–1100 °C to 0.02–2.4 displacements per atom (dpa) in the High Flux Isotope Reactor were investigated using transmission electron microscopy. The results indicated that nanoscale W-Re-Os clusters were identified in the low dose regime from 0.02 to 0.44 dpa with irradiation temperature lower than 800 °C while acicular-shape precipitates formed when irradiation dose is higher than 1.5 dpa. A tentative roadmap of the kinetics process of the transmutation-induced precipitation in neutron-irradiated tungsten is presented characterizing the defect features (i.e., W-Re-Os clusters and precipitates) consisting of transmutant elements in tungsten irradiated to various doses. All TEM-visible voids were associated with the acicular-shape precipitates. Voids were formed prior to the formation of acicular-shape precipitates and act as strong trapping sites for mobile species involved in the precipitation together with dislocations. Thermal stability of W-Re-Os clusters was assessed by performing a 2-h anneal at 1200 °C on tungsten irradiated to 0.44 dpa at 705 °C. In conclusion, the kinetics process of transmutant elements and radiation defects are discussed to reveal the underlying mechanisms controlling the formation of precipitates in tungsten.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiation induced void spheroidization, shrinkage and migration in Cu at elevated temperatures: An in situ study

We show understanding the void evolution in irradiation environment is of great interest and significance, as irradiation-induced voids typically lead to pronounced volumetric swelling and degradation of mechanical properties. In situ studies on the irradiation response of nanovoids at elevated temperature remain limited. In this work, we performed systematic in situ 1 MeV Kr ++ irradiations on Cu with nanovoids in a transmission electron microscope up to 350 °C. The in situ studies revealed intriguing void spheroidization, shrinkage and migration. Furthermore, the morphology evolution and migration of nanovoids showed a strong dependence on irradiation temperature and initial void size. Post-irradiation analyses identified defect clusters in the form of stacking fault tetrahedrons, and the remaining large faceted nanovoids. The underlying mechanisms of irradiation-induced void spheroidization and shrinkage were discussed based on phase-field modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effect of heavy ion irradiation dose rate and temperature on α' precipitation in high purity Fe-18%Cr alloy

Cr-rich alpha prime precipitates (CrRP) induce hardening and embrittlement of FeCr alloys, but the kinetics of CrRP formation due to particle irradiation are not well understood. In this study, Fe18wt.%Cr alloy in solid solution state and pre-aged to produce relatively coarse CrRP was irradiated with 8 MeV Fe ions. The irradiation conditions involved two midrange doses of 0.37 and 3.7 displacements per atom (dpa), a wide range of dose rates (10 -5 –10 -3 dpa/s) and temperatures (300–450 °C). The distributions of CrRP after irradiation were studied with atom probe tomography (APT). The critical irradiation conditions to suppress CrRP formation were identified as 300 °C and 10-3 dpa/s; CrRP formation occurred readily at lower dose rates or higher temperatures. From 0.37 to 3.7 dpa, CrRP were observed to slightly grow at 350 °C and strongly coarsen at 450 °C. Specimens with pre-existing CrRP evolved into a similar precipitate distribution as detected after ion irradiation on solution annealed specimens at 300–350 °C to 0.37 dpa, indicating that the precipitate microstructure approaches a quasi-equilibrium for doses <1 dpa. Limited shrinking of pre-existing CrRP was observed after irradiation at 450 °C to 0.37 dpa, indicating a higher recovery rate at this temperature. Additionally, the evolution of CrRP is quantitatively explained by employing corrections to the historic Nelson-Hudson-Mazey precipitate stability model, and a radiation modified precipitation mechanism is proposed to account for the competition between radiation enhanced diffusion and ballistic dissolution which results in the modifications on both size and solute concentration of CrRP.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Irradiation-induced amorphization of UO 2 films prepared by spraying-assisted combustion synthesis

Spraying-assisted combustion synthesis with uranyl nitrate – acetylacetone – 2-methoxyethanol solutions was used to prepare UO 2 films on an aluminum alloy substrate. The tuning of the spraying parameters and annealing temperatures allowed the preparation of UO 2 films with thicknesses varying from 10–300 nm and 5–10 nm UO 2 grain size. High-resolution electron microscopy and X-ray photoelectron spectroscopy showed that increasing the annealing temperature promotes Mg diffusion from the substrate into the films. The incorporation of Mg reduced the overall crystallinity of the films. The irradiation with Ar 2+ ions (1.7 MeV energy and a fluence of 2 × 10 16 ions/cm 2 ) did not degrade the quality of the films. However, the Mg content significantly influenced the irradiation-induced restructuring of the UO 2 films. Irradiated films with low or no Mg content exhibit high crystallinity, and the UO 2 /Al interfacial layer becomes highly porous. Films with higher Mg content are mostly amorphized after irradiation. The origin of irradiation-induced amorphization was related to the formation of Mg y U 1-y O 2±x solid solutions. Chemically complex, pore-free, and amorphous Mg-Al-O film/substrate interfacial layers enable continuous Mg diffusion during irradiation. As a result, the gradual increase in Mg amounts triggers irradiation-induced precipitation of a crystalline MgO-rich phase within the amorphous films.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Carbon-Ion Irradiation Suppresses Migration and Invasiveness of Human Pancreatic Carcinoma Cells MIAPaCa-2 via Rac1 and RhoA Degradation

Purpose: To investigate the mechanisms underlying the inhibition of cancer cell migration and invasion by carbon (C)-ion irradiation. Methods and Materials: Human pancreatic cancer cells MIAPaCa-2, AsPC-1, and BxPC-3 were treated by x-ray (4 Gy) or C-ion (0.5, 1, 2, or 4 Gy) irradiation, and their migration and invasion were assessed 2 days later. The levels of guanosine triphosphate (GTP)-bound Rac1 and RhoA were determined by the active GTPase pull-down assay with or without a proteasome inhibitor, and the binding of E3 ubiquitin ligase to GTP-bound Rac1 was examined by immunoprecipitation. Results: Carbon-ion irradiation reduced the levels of GTP-bound Rac1 and RhoA, 2 major regulators of cell motility, in MIAPaCa-2 cells and GTP-bound Rac1 in AsPC-1 and BxPC-3 cells. Proteasome inhibition reversed the effect, indicating that C-ion irradiation induced Rac1 and RhoA degradation via the ubiquitin (Ub)-proteasome pathway. E3 Ub ligase X-linked inhibitor of apoptosis protein (XIAP), which directly targets Rac1, was selectively induced in C-ion–irradiated MIAPaCa-2 cells and coprecipitated with GTP-bound Rac1 in C-ion–irradiated cells, which was associated with Rac1 ubiquitination. Cell migration and invasion reduced by C-ion radiation were restored by short interfering RNA–mediated XIAP knockdown, indicating that XIAP is involved in C-ion–induced inhibition of cell motility. Conclusion: In contrast to x-ray irradiation, C-ion treatment inhibited the activity of Rac1 and RhoA in MIAPaCa-2 cells and Rac1 in AsPC-1 and BxPC-3 cells via Ub-mediated proteasomal degradation, thereby blocking the motility of these pancreatic cancer cells.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2025

Lattice response to ambient self-irradiation of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures

Self-irradiation in plutonium is a continuous process that introduces radiation damage ranging from point defects to nanoscale radiation damage as well as decay daughter product impurities that require to be accommodated by the host material. These changes are affecting the structure and properties of the material by lattice swelling, mechanical strengthening, and other alterations. X-ray diffraction and dilatometry studies have shown that the room temperature self-irradiation of δ-phase Pu-Ga alloys (2 and 7 at%Ga) results in swelling of the lattice that saturates after 0.1–0.2 displacement per atom (dpa) of accumulated lattice damage. Here, for a better understanding of these self- irradiation processes, defect accumulation and damage evolution, this time-of-flight neutron diffraction study investigates the impact of ambient self-irradiation effects on the structural evolution of δ-phase 239 Pu-Ga alloys exposed to cryogenic temperatures (<120 K), associated with a varied history of storage time and thermal treatments. Two key behaviors were observed in the lattice evolution after exposure to cryogenic temperatures with respect to room-temperature accumulated self-irradiation damage. Samples with no self- irradiation damage accumulation under ambient conditions showed swelling of the lattice at the end of the measurement, while samples that accumulated significant self-irradiation damage under ambient conditions showed only a partial recovery of the lattice parameter value upon completion of the measurements (cooling followed by heating back to room temperature). Possible causes of the behavior are proposed, and their associated effects are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Mechanical property degradation of high crystalline SiC fiber–reinforced SiC matrix composite neutron irradiated to ~100 displacements per atom

For the development of silicon carbide (SiC) materials for next-generation nuclear structural applications, degradation of material properties under intense neutron irradiation is a critical feasibility issue. This paper evaluated the mechanical properties and microstructure of a chemical vapor infiltrated SiC matrix composite, reinforced with a multi-layer SiC/pyrolytic carbon–coated Hi-Nicalon TM Type S SiC fiber, following neutron irradiation at 319 and 629 °C to ~100 displacements per atom. Both the proportional limit stress and ultimate flexural strength were significantly degraded as a result of irradiation at both temperatures. After irradiation at 319 °C, the quasi-ductile fracture behavior of the nonirradiated composite became brittle, a result that was explained by a loss of functionality of the fiber/matrix interface associated with the disappearance of the interphase due to irradiation. Finally, the specimens irradiated at 629 °C showed increased apparent failure strain because the fiber/matrix interphase was weakened by irradiation-induced partial debonding.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Progress in development of SiC-based joints resistant to neutron irradiation

This study fills a knowledge gap regarding neutron-irradiation resistance of SiC joints for nuclear applications, by investigating high-dose neutron irradiation effects on the strength of selected joints and low-dose neutron irradiation effects on recently developed joints fabricated by state of the art processing methods. The joining methods used for the high-dose radiation study included pressure-assisted liquid-phase sintering (LPS) of SiC nanopowder, pressureless calcia-alumina glass ceramics joining, and reaction sintering of Ti-Si-C powders with hot-pressing. The joints were neutron-irradiated at 530 °C to 20 displacements per atom (dpa). Other joining methods included low-pressure LPS of cold-pressed SiC green body, pressureless reaction sintered Ti-Si-C powder joint, spark plasma–sintered Ti diffusion bond, and hot-pressed Ti diffusion bond, which were irradiated at ~500 °C to ~2 dpa. There was no notable degradation of torsional strengths of the joints following the high-dose irradiation. The irradiation-induced degradation at low neutron dose was highly dependent on joint type.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Investigation of $\gamma$-irradiated polyvinylidene fluoride and its acute toxicity

Polyvinylidene fluoride (PVDF) is an industrial polymer with many applications. The EPR spectrum of polyvinylidene fluoride (PVDF) irradiated by γ-rays at low temperature (-196°C) provides evidence for the presence of radicals formed by loss of a fluorine atom from PVDF, and polyenyl radicals. Polyenyl radicals dominate in the polymer irradiated at room temperature. Peroxide radicals formed as a result of oxidation of primary free radicals by atmospheric oxygen are observed in the EPR spectrum of a polymer irradiated in air, or when air is introduced to a polymer irradiated in a vacuum. Carbonization and oxidation of γ-irradiated PVDF were observed in the XPS spectrum. A model for polymer degradation based on correlated G3(MP2) molecular orbital theory electronic structure calculations is described in this report. The presence of HF in the products and a low yield of products with the composition CxFyHz indicate the absence of noticeable thermal destruction of the main polymer chain upon heating to 220°C of both the initial and γ-irradiated PVDF. Although any HF released during the PVDF dehydrofluorination would be toxic, a study of the acute toxicity of γ-irradiated PVDF showed that the polymer does not exhibit toxicity after a single intragastric route of administration to mice.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Prediction of Thermal Conductivity for Irradiated SiC/SiC Composites by Informing Continuum Models with Molecular Dynamics Data

This article proposes a new method to estimate the thermal conductivity of SiC/SiC composites subjected to neutron irradiation. The modeling method bridges different scales from the atomic scale to the scale of a 2D SiC/SiC composite. First, it studies the irradiation-induced point defects in perfect crystalline SiC using molecular dynamics (MD) simulations to compute the defect thermal resistance as a function of vacancy concentration and irradiation dose. The concept of defect thermal resistance is explored explicitly in the MD data using vacancy concentrations and thermal conductivity decrements due to phonon scattering. Point defect-induced swelling for chemical vapor deposited (CVD) SiC as a function of irradiation dose is approximated by scaling the corresponding MD results for perfect crystal β-SiC to experimental data for CVD-SiC at various temperatures. The computed thermal defect resistance, thermal conductivity as a function of grain size, and definition of defect thermal resistance are used to compute the thermal conductivities of CVD-SiC, isothermal chemical vapor infiltrated (ICVI) SiC and nearly-stoichiometric SiC fibers. The computed fiber and ICVI-SiC matrix thermal conductivities are then used as input for an Eshelby-Mori-Tanaka approach to compute the thermal conductivities of 2D SiC/SiC composites subjected to neutron irradiation within the same irradiation doses. Predicted thermal conductivities for an irradiated Tyranno-SA/ICVI-SiC composite are found to be comparable to available experimental data for a similar composite ICVI-processed with these fibers.

SiC/SiC composites, thermal conductivity, fiber/ma↗
DOE OSTI2016

A mechanistic model for depth-dependent hardness of ion irradiated metals

A mechanistic model was developed for modeling the depth-dependent hardness in ion irradiated metallic materials. The model is capable of capturing the indentation size effect, ion irradiation induced damage gradient effect, and effect of unirradiated region acting as a soft substrate. A procedure was developed and described in detail to parametrize the model based on experimentally obtained hardness vs. indentation depth curves. Very good agreement was observed between our model predictions and experimental data of several different stainless steels subjected to various ion irradiation conditions. Additionally, two hardening mechanisms are revealed in the new model. One is the well-known indentation size effect arising from the creation of geometrically necessary dislocations as the indenter pierces into the materials. The other is the irradiation hardening due to the presence of irradiation-induced defects. As a function of indentation depth h, the hardening due to indentation size effect is described by $\bar{h}^*/h$, while the hardening due to irradiation first follows a power law form $ph^n$, then changes to Z/h - Q/h 3 , where $\bar{h}^*$, P, n, Z and Q > 0 are constants. This transition occurs at the indentation depth when the plastic zone reaches the end of the irradiated layer.

36 MATERIALS SCIENCE↗
DOE OSTI2018

High-energy x-ray diffraction microscopy study of deformation microstructures in neutron-irradiated polycrystalline Fe-9%Cr

In this paper, we used far-field high-energy x-ray diffraction microscopy (FF-HEDM) to measure in 3D the spatial distributions of the sizes, orientations, and residual strains of grains and subgrains formed by room-temperature tensile deformation in a neutron-irradiated (450 °C, 0.01dpa) Fe-9Cr specimen and its unirradiated counterpart. We found that neutron irradiation under this condition alone had no effect on the grain size distribution. After deformation, grains fragmented into subgrains in both unirradiated and irradiated specimens: the irradiated specimen contained a few large subgrains which co-existed with many smaller subgrains, while the unirradiated specimen contained small subgrains with a relatively uniform size distribution. Prior to deformation, the irradiated specimen had higher residual strain spread compared to its unirradiated counterpart, while after deformation to the maximum uniform elongation, the strain distributions among subgrains were similar between the unirradiated and irradiated specimens. The FF-HEDM measurements provide new insight into the effects of neutron irradiation on the mechanical response of Fe-Cr ferritic alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

This research demonstrated the first time that synchrotron X-ray micro-computed tomography (μ-CT) was utilized to analyze the three-dimensional (3-D) morphology of the microstructure of irradiated nuclear fuel. Both absorption contrast and propagation-based phase contrast enhanced μ-CT techniques were evaluated. Porosity and phase region volumes were determined in neutron irradiated uranium 10 wt.% zirconium (U-10Zr) using propagation-based phase contrast enhanced μ-CT. Volumes were determined by image processing software and algorithms, with the porosity volume subsequently compared to the ASTM E562 method. The 3-D porosity of a small region (~ 8 x 10 5 µm 3 ) located in the intermediate fuel redistribution zone that was removed via a focused ion beam was determined to be 7.2% via image processing. A detailed 3-D analysis of the porosity with respect to volume and morphology was provided. Five growth stages of pore evolution were observed including: nucleation, growth, coalescence, interconnected porosity, and extended/interconnected porosity. Four distinct phase regions were identified, including the pores, via both absorption contrast and propagation-based phase contrast enhanced μ-CT. A quantitative comparison of fuel swelling between this localized microscopic U-10Zr fuel region, previously irradiated in the Fast Flux Test Facility, and historic U-10Zr fuels irradiated in the Experimental Breeder Reactor-II (EBR-II) was conducted. It was determined that for similar burnups, the localized swelling in this microscopic fuel region was 7.7%, which was approximately one-half of the value of the macroscopic swelling reported in EBR-II irradiated fuel. With the advent of new specimen preparation tools for irradiated materials, as well as advances in characterization techniques, researchers can now obtain previously inaccessible microstructural insight into irradiated fuels. This possibility naturally lends itself to improved fuel performance models for the nuclear community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

In-situ ion irradiation study of alloy 709 stainless steels with different processing histories

Alloy 709 stainless steel (A709 SS) has gained an increased interest in nuclear applications due to its advanced properties at high temperatures over the widely used Type 316 SS. However, how this alloy behaves under irradiation with concurrent environmental factors, such as mechanical deformation and/or thermal annealing, has not been studied. In this work, we used in-situ ion irradiation inside a transmission electron microscope (TEM) to investigate the effects of irradiation on the microstructures of A709 SS samples with different prior processing histories, including as-received, annealed, and creep-deformed. It was found that, compared to room temperature irradiation, 600 degrees C irradiation led to larger dislocation loop sizes and lower loop densities. Furthermore, with 600 degrees C irradiation, the creep-deformed sample, which had the highest precipitate density and the highest dislocation density, had the smallest loop size and the lowest loop density compared to others. This observation has been rationalized with quantitative evaluation of the sink strength factors in each sample. This study shows that in-service deformation can significantly affect the irradiation performance in structural materials at reactor operating conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2023

High temperature annealing of irradiated nuclear grade graphite

Previous work has shown that the material properties of nuclear grade graphite are substantially affected by the atomic and microstructural changes that occur during neutron-irradiation. The parameters that play a role in these changes are irradiation dose, temperature, graphite composition and the initial microstructure of the graphite. Understanding the details of how these material changes occur and exactly what the changes consists of will enable the prediction of the material property changes as a function of these variables. Other work has shown these atomic level and microstructural changes can be healed or annealed out by raising the irradiated graphite above its irradiation temperature. Here in this work, experiments were carried out to investigate how the properties of irradiated graphite recover when heated. By showing property recovery as a function of annealing temperature or energy, insight is provided into the type of damage that occurred during neutron-irradiation. The data presented here shows recovery of thermal diffusivity, coefficient of thermal expansion, Young's modulus and electrical resistivity between annealing temperatures of 500 °C and 2380 °C. Graphite grades NBG-18, IG-110 and PCEA are considered that were both stressed and unstressed during irradiation.

36 MATERIALS SCIENCE↗