Irradiation-induced hardening and softening of CLAM steel under Fe ion irradiation

The irradiation-induced hardening and softening of CLAM steel irradiated with 3.5 MeV Fe¹³⁺ ions at temperatures of 300 °C and 550 °C were investigated by nanoindentation tests in combination with microstructures. Irradiation- induced hardening occurred in the steel irradiated at 300 °C to doses of 0.46 dpa, 0.94 dpa, and 2.79 dpa. The hardening occurred at 300 °C is mainly attributed to the formation of irradiation-produced dislocation loops and a network of tangled dislocations in the irradiated steel samples. Significant hardening was found in the steel irradiated at 550 °C to 0.38 dpa. On the contrary, irradiation-induced softening occurred in the steel irradiated at 550 °C to both 0.76 dpa and 2.75 dpa. Irradiation-produced dislocation loops are not dominant effect on the irradiation hardening of the steel samples irradiated at 550 °C. The hardening and softening of the irradiated steel were explained in terms of the irradiation-produced defects and recovery process occurred during the irradiation.     

Radiation effects on the microstructure of a 9Cr-ODS alloy

Oxide dispersion strengthened (ODS) steels are prime candidates for high-temperature, high-dose cladding in advanced nuclear reactors. When a 9Cr-ODS alloy was irradiated with 5 MeV nickel ions at temperatures of 500–700°C to doses up to 150 dpa, there was no significant change in the dislocation arrangement. For oxide particles, there is a small shrinkage in size and increase in density with increasing irradiation dose. This work confirms that oxide particles and the microstructure of the 9Cr-ODS show minimal changes under irradiation at temperatures up to 700°C and doses up to 150 dpa.     

Effects of MeV Fe Ions Irradiation on the Microstructure and Property of Nuclear Grade 304 Stainless Steel:Characterized by Positron Annihilation Spectroscopy,Transmission Electron Microscope and Nanoindentation

Nuclear grade 304 stainless steel was irradiated by 3.5 MeV Fe ions,with fluxes of 3.05E+ 15 ions/cm2 and 1.55E+ 16 ions/cm2.Irradiation effects were studied by positron annihilation spectroscopy (PAS),transmission electron microscope (TEM)and nanoindentation techniques.PAS results showed that different types of defects were produced after irradiation and that there was significant variance in defects formed when the samples were subjected to different irradiation doses.TEM char-acterization showed that the irradiation-induced dislocation loops enlarged in average size,but decreased in number density at higher irradiation doses.Nanoindentation test showed obvious irradiation hardening phenomenon,which was in good agreement with the PAS and TEM results.Irradiation hardening effect increased with an increase in irradiation dose and saturation occurred with an increase in irradiation dose from 3.2 to 16 dpa.Further statistical analysis showed that barrier strength of the Frank loop depends on the loop size and density produced by the ion irradiation.     

Defect characterization,mechanical and thermal property evaluation in CVD-W after low-dose neutron irradiation

Ultra-high purity (> 99.9999 wt%) chemical vapour deposited tungsten (CVDW) samples were neutron irradiated in the BR2 reactor (Belgium) at T_irr < 210 °C to ~ 0.15 dpa, followed by isochronal annealing at 500, 800 and 1100 °C. Defect characterization showed that dislocation loops dominated the as-irradiated damage microstructure and were mostly ≤5 nm. Void formation was observed after post-irradiation annealing at 1100 °C. The mechanical and thermal properties of CVD-W were evaluated based on tensile tests, Vickers hardness and temperature wave analysis. Fractography study suggested that a transition from intergranular fracture to cleavage fracture took place in the material after neutron irradiation. Hardening was found ~ 23% after irradiation. Subsequent annealing below 800 °C saw further increase in hardness, featuring a maximum value of about Hv = 487. Softening occurred at 1100 °C. Thermal diffusivity dropped by ~ 65% after irradiation and ~ 40% of this degradation recovered at 1100 °C.     

Effect of dose and temperature parameters of neutron irradiation to maximum damaging dose of 77 dpa on characteristics of porosity formed in steel 0.07C-16Cr-19Ni-2Mo-2Mn-Ti-Si-V-P-B

The microstructure of samples of cladding tubes made of steel 0.07C-16Cr-19Ni-2Mo-2Mn-Ti-Si-V-P-B (EK164) irradiated to different damaging doses (up to 77 dpa) in the BN-600 reactor at temperatures from 440 to 600°C has been investigated. Characteristics of radiation porosity formed during irradiation in different temperature intervals have been determined. The dependences of the porosity characteristics on the rate of generation of atomic displacements and temperature of neutron irradiation have been established.     

Effect of tensile and compressive stresses on the radiation swelling and creep strain in austenitic steel Kh18N10T

Effect of tensile and compressive stresses on the radiation swelling, microstructure, and creep strain in austenitic steel Kh18N10T is considered. The gas-filled samples of a complex shape prepared from steel Kh18N10T were irradiated in a BOR-60 reactor for 2 years to a damaging dose of 15 dpa at a temperature of 420–450°C. In the shells of the irradiated samples, compressive and tensile stresses were created. Samples were also irradiated, in which these stresses practically were absent.     

Evolution of magnetic properties of cladding austenitic steel under irradiation in a reactor

Magnetic properties of samples of austenitic steel ChS-68 cut from the cladding of a fuel element, which was irradiated in a BN-600 fast-neutron reactor to a maximal damage dose of ～80 displacements per atom (dpa) at temperatures of 370–587°C, have been investigated. It has been established that irradiation with fast neutrons leads to the formation of ferromagnetic microregions, the effective sizes and concentration of which depend on the damage dose. It has been shown that, at damage doses higher than ～55 dpa, small spontaneous magnetization and magnetization hysteresis, which are characteristic of the ferromagnetic state, appear in the samples. It is assumed that the ferromagnetic microregions are the nuclei of the α′ phase and the radiation-induced segregation microregions, in which the spacing between the nearest iron atoms exceeds the critical distance that determines the change in the sign of exchange interaction. Arguments in favor of this assumption are presented.     

Electrochemical Evaluation of Radiation-Induced Segregation in Austenitic Stainless Steels with Oversize Solute Addition

Effect of different levels of oversize element, cerium, on radiation-induced segregation (RIS) in type 316 stainless steel was investigated. The effect of prior cold-work on RIS was also investigated. Samples with 0.00, 0.01, and 0.04 wt.% cerium were irradiated to 0.70 dpa using 4.8 MeV protons at 300 °C. Characterization of proton-irradiated specimens was carried out using electrochemical potentiokinetic reactivation (EPR) test followed by atomic force microscopic examination. The specimen with prior cold-work (without cerium addition) showed the lowest EPR values indicating the lowest chromium depletion in this material. The specimen with 0.04 wt.% cerium showed the lower EPR value as compared to the specimen with 0.01 wt.% Ce. The irradiated specimen with prior cold-work showed linear features after the EPR tests and such features were attributed to decoration of dislocations, generated due to prior cold-work, by point defects produced during irradiation. The resistance to RIS offered by cold-work (linear features) has been more effective as compared to that by the addition of oversize solute addition.     

Evolution of secondary-phase precipitates during annealing of the 12Kh18N9T steel irradiated with neutrons to a dose of 5 DPA

The formation and evolution of thermally-induced secondary precipitates in an austenitic stainless steel 12Kh18N9T irradiated in the core of a laboratory reactor VVR-K to a dose of 5 dpa and subjected to post-radiation isochronous annealings for 1 h in a temperature range from 450 to 1050°C have been studied using transmission electron microscopy (TEM) and microhardness measurements. It has been shown that the formation of stitch (secondary) titanium carbides and M ₂₃C₆ carbides at grain and twin boundaries after annealing at 1050°C is preceded by a complex evolution of fineparticles of secondary phases (titanium carbides and nitrides) precipitated at dislocation loops and dislocations during annealing at temperatures above 750°C.     

Ar ion irradiation hardening of high-Cr ferritic/martensitic steels at 700 °C

High-Cr ferritic/martensitic (FM) steels are being considered for applications as fuel cladding or core structures for Generation-IV reactors. Because high temperatures approaching 923-973 K (650-700 °C) are envisioned in the designs of Generation IV reactors, irradiation response of high-Cr FM steels at the high temperatures requires investigations. Response of two high-Cr FM steels P92 and 11Cr to irradiation at 973 K (700 °C) was investigated through Ar ion irradiation in combination with damage simulations, nanoindentation measurements and microstructure analyses. Irradiation hardening occurred in both steels after Ar ion irradiation at 973 K (700 °C) to 10 dpa, providing the first evidence that irradiation hardening can occur at a high irradiation temperature of 973 K (700 °C) in high-Cr FM steels. Argon bubbles with a very high number density and an average diameter of about 2.6-3 nm formed in the two steels after the irradiation. The irradiation hardening occurred in the two steels is attributed to the formation of these high-number-density fine argon bubbles produced by the irradiation homogeneously distributed in the matrix. Difference in the magnitude of irradiation hardening between the two steels was also discussed.     

Effects of ion irradiation on Zr52.5Cu17.9Ni14.6Al10Ti5 (BAM-11) bulk metallic glass

Bulk metallic glasses are intriguing candidates for nuclear applications due to their inherent amorphous structure, but their radiation response is largely unknown due to the relatively recent nature of innovations in bulk metallic glass fabrication. Here, microstructural and mechanical property evaluations have been performed on a Zr_52.5Cu_17.9Ni_14.6Al₁₀Ti₅ bulk metallic glass (BAM-11) irradiated with 3 MeV Ni⁺ ions to 0.1 and 1.0 dpa at room temperature and 200 °C. Nanoindentation hardness and Young's modulus both decreased by 6–20% in samples irradiated at room temperature, with the sample irradiated to 1.0 dpa experiencing the greatest change in mechanical properties. However, no significant changes in properties were observed in the samples irradiated at 200 °C, and transmission electron microscopy showed no visible evidence of radiation damage or crystallization following ion irradiation at any of the tested conditions. These results suggest that BAM-11 bulk metallic glass may be useful for certain applications in nuclear environments.     

Atom probe tomography of nanoscaled features of oxide-dispersion-strengthened ODS Eurofer steel in the initial state and after neutron irradiation

Atom probe tomography has been used to investigate nanoscale features in the yttrium oxide dispersion strengthened steel ODS Eurofer, which is a perspective structural material for the reactor cores. In the initial material, a large number (∼2 × 10²⁴ m⁻³) of ultrafine (∼2.5 nm in diameter) clusters enriched in yttrium, oxygen, nitrogen, and vanadium have been revealed. The investigation of the ODS Eurofer steel irradiated at 330°C to 32 dpa in the BOR-60 fast reactor has also revealed a large number of ultrafine (1–3 nm in diameter) nanoclusters significantly enriched in yttrium, oxygen, manganese, and chromium. In the irradiated material, an increase in the concentration of clusters and changes in the chemical composition of the clusters and matrix have been noted. The irradiation by fast neutrons leads to a partial transition of vanadium from the clusters into the surrounding matrix and to a general increase in the concentrations of yttrium and oxygen in the volumes under investigation.     

Irradiation Effects in Ni–17Mo–7Cr Alloy Bombarded with MeV Au Ions

Irradiation effects in Ni–17Mo–7Cr alloy have been systematically investigated by using 3 Me V Au ions at different fluences ranging from 8 9 1013cm-2to 2.3 9 1015cm-2,corresponding to doses of 1–30 dpa.The results indicated that sample microstrain increased gradually from 0.14 to 0.22% as dose increased from 0 to 30 dpa.Besides,the nanohardness of Ni–17Mo–7Cr alloy increased with irradiation dose until 10 dpa,and then,softening effect became dominant while further increasing dose to 30 dpa.After being irradiated at room temperature,the swelling rate of Ni–17Mo–7Cr alloy was found to be around 0.04% per dpa.These data are helpful in estimating the irradiation resistance of this newly developed Ni–17Mo–7Cr alloy in nuclear energy systems.     

Investigation of the Microstructural Changes and Hardness Variations of Sub-Zero Treated Cr-V Ledeburitic Tool Steel Due to the Tempering Treatment

The microstructure and tempering response of Cr-V ledeburitic steel Vanadis 6 subjected to sub-zero treatment at ??196 °C for 4 h have been examined with reference to the same steel after conventional heat treatment. The obtained experimental results infer that sub-zero treatment significantly reduces the retained austenite amount, makes an overall refinement of microstructure, and induces a significant increase in the number and population density of small globular carbides with a size 100-500 nm. At low tempering temperatures, the transient M₃C-carbides precipitated, whereas their number was enhanced by sub-zero treatment. The presence of chromium-based M₇C₃ precipitates was evidenced after tempering at the temperature of normal secondary hardening; this phase was detected along with the M₃C. Tempering above 470 °C converts almost all the retained austenite in conventionally quenched specimens while the transformation of retained austenite is rather accelerated in sub-zero treated material. As a result of tempering, a decrease in the population density of small globular carbides was recorded; however, the number of these particles retained much higher in sub-zero treated steel. Elevated hardness of sub-zero treated steel can be referred to more completed martensitic transformation and enhanced number of small globular carbides; this state is retained up to a tempering temperature of around 500 °C in certain extent. Correspondingly, lower as-tempered hardness of sub-zero treated steel tempered above 500 °C is referred to much lower contribution of the transformation of retained austenite, and to an expectedly lower amount of precipitated alloy carbides.     

Influence of Plasma Nitriding on the Microstructure, Wear, and Corrosion Properties of Quenched 30CrMnSiA Steel

The oil-quenched 30CrMnSiA steel specimens have been pulse plasma-nitrided for 4 h using a constant 25% N₂-75% H₂ gaseous mixture. Different nitriding temperatures varying from 400 to 560 °C have been used to investigate the effects of treatment temperature on the microstructure, microhardness, wear, and corrosion resistances of the surface layers of the nitrided specimens. The results show that significant surface-hardened layer consisting of compound and diffusion layers can be obtained when the oil-quenched steel (α′-Fe) are plasma-nitrided at these experimental conditions, and the compound layer mainly consists of ε-Fe_2-3N and γ′-Fe₄N phases. Lower temperature (400-500 °C) nitriding favors the formation of ε-Fe_2-3N phase in surface layer, while a monophase γ′-Fe₄N layer can be obtained when the nitriding is carried out at a higher temperature (560 °C). With increasing nitriding temperature, the compound layer thickness increases firstly from 2-3 μm (400 °C) to 8 μm (500 °C) and then decreases to 4.5 μm (560 °C). The surface roughness increases remarkably, and both the surface and inner microhardness of the nitrided samples decrease as increasing the temperature. The compact compound layers with more ε-Fe_2-3N phase can be obtained at lower temperature and have much higher wear and corrosion resistances than those compound layers formed employing 500-560 °C plasma nitriding.     

Characterization of Cold-Sprayed IN625 and NiCr Coatings

Ni-based coatings IN625^® and Ni20%Cr were cold sprayed on a low-alloy steel (AISI 4130) substrate, using Helium as the process gas. Dense coatings up to 3-mm thickness were deposited, having a hardness of 500-550 HV. The coatings showed a hardness maximum, with heat treatment, before dropping to a lower value. The coating microstructure revealed two distinct types of regions, comprising grains with a high dislocation density and elongated shear bands having twins. Heat treatment led to 30-50 nm grains in the IN625 coating, and >1-2 μm grains for NiCr. Both coatings showed a compressive residual stress in the as-sprayed condition, which relaxed to a zero residual stress, at 650 °C. The NiCr coatings showed a much higher compressibility, as compared to IN625. The IN625 coatings induced a much larger deformation on the 4130 steel. Overall, while both types of Ni-based alloy coatings showed similarities in terms of hardness and microstructure, they revealed distinct differences in their deformability, thermal stability, and substrate deformation, indicating a different behavior between a binary solid solution (NiCr) as compared to a multielement solid solution (IN625), as elucidated via a detailed characterization of these coatings.     

Microstructure and mechanical property of irradiated Zr−2.5Nb pressure tube in Wolsong unit-1

With the aim of assessing the degradation of Zr−2.5Nb pressure tubes operating in the Wolsong unit-1 nuclear power plant, characterization tests are being conducted on irradiated Zr−2.5Nb tubes removed after 10-year operation. The examined tube had been exposed to temperatures ranging from 264 to 306°C and a neutron fluence of 8.9×10²¹ n/cm² (E>1 MeV) at the maximum. Tensile tests were carried out at temperatures ranging from RT to 300°C. The density of a-type and c-type dislocations was examined on the irradiated Zr-2.5Nb tube using a transmission electron microscope. Neutron irradiation up to 8.9×10²¹ n/cm² (E>1 MeV) yielded an increase in a-type dislocation density of the Zr−2.5Nb pressure tube to 7.5×10¹⁴ m⁻², which was highest at the inlet of the tube exposed to the low temperature of 275°C. In contranst, the c-component dislocation density did not change with irradiation, keeping an initial dislocation density of 0.8×10¹⁴ m⁻² over the whole length of the tube. As expected, the neutron irradiation increased mechanical strength by about 17–26% in the transverse direction and by 34–39% in the longitudinal direction compared to that of the unirradiated tube at 300°C. The change in the mechanical properties with irradiation is discussed in association with the microstructural change as a function of temperature and neutron fluence.     

Oxidation Behavior of Carbon Steel: Effect of Formation Temperature and pH of the Environment

The nature of surface oxide formed on carbon steel piping used in nuclear power plants affects flow-accelerated corrosion. In this investigation, carbon steel specimens were oxidized in an autoclave using demineralized water at various temperatures (150-300 °C) and at pH levels (neutral, 9.5). At low temperatures (< 240 °C), weight loss of specimens due to dissolution of iron in water occurred to a greater extent than weight gain due to oxide formation. With the increase in temperature, the extent of iron dissolution reduced and weight gain due to oxide formation increased. A similar trend was observed with the increase in pH as was observed with the increase in temperature. XRD and Raman spectroscopy confirmed the formation of magnetite. The oxide film formed by precipitation process was negligible at temperatures from 150 to 240 °C compared to that at higher temperatures (> 240 °C) as confirmed by scanning electron microscopy. Electrochemical impedance measurement followed by Mott–Schottky analysis indicated an increase in defect density with exposure duration at 150 °C at neutral pH but a low and stable defect density in alkaline environment. The defect density of the oxide formed at neutral pH at 150-300 °C was always higher than that formed in alkaline environment as reported in the literature.     

Effect of tensile stresses on the evolution of vacancy porosity in the Fe-18% Cr-10% Ni-Ti steel irradiated in BOR-60 reactor

The results of the studies of the microstructure of the 08Kh18N10T austenitic steel irradiated in a BOR-60 reactor to a damaging dose of 36 dpa at a temperature of about 420°C have been presented. The results have been compared with previously published data on the effect of stresses on the development of vacancy porosity in austenitic steels irradiated in the BOR-60 reactor.     

Gamma Ray Irradiation Effects on the Ferroelectric and Piezoelectric Properties of Barium Titanate Ceramics

The effect of heavy dose gamma ray irradiation on the ferroelectric and piezoelectric properties of barium titanate (BaTiO₃) ceramics has been investigated. It is found that on irradiation the ferroelectric property decreases and polarization behavior shows double loop hysteresis. The piezoelectric properties including piezoelectric charge constant (d ₃₃), electromechanical coupling coefficient (K _p), and electrostrictive strain also decreases. The most probable reason for decreased ferroelectric and piezoelectric properties may be the occurrence of random local strain upon irradiation. The phase transition temperature from ferroelectric to paraelectric decreases and degree of diffuseness increases on irradiation. The thermoluminescence (TL) glow curve showed a peak at 226 °C showing that irradiated BaTiO₃ has TL properties. Presence of TL clearly indicates that gamma ray irradiation causes trapped holes and electrons and these trapped charges are released at temperature higher than 226 °C. The creation of trapped holes and electrons effected the microstrain of BaTiO₃ ceramic leading to change in the ferroelectric and piezoelectric properties of BaTiO₃ ceramic.