The Irradiation Performance and Microstructural Evolution in 9Cr-2W Steel Under Ion Irradiation

Grade 92 steel (9Cr-2W) is a ferritic-martensitic steel with good mechanical and thermal properties. It is being considered for structural applications in Generation IV reactors. Still, the irradiation performance of this alloy needs more investigation as a result of the limited available data. The irradiation performance investigation of Grade 92 steel would contribute to the understanding of engineering aspects including feasibility of application, economy, and maintenance. In this study, Grade 92 steel was irradiated by iron ion beam to 10, 50, and 100 dpa at 30 and 500 °C. In general, the samples exhibited good radiation damage resistance at these testing parameters. The radiation-induced hardening was higher at 30 °C with higher dislocation density; however, the dislocation density was less pronounced at higher temperature. Moreover, the irradiated samples at 30 °C had defect clusters and their density increased at higher doses. On the other hand, dislocation loops were found in the irradiated sample at 50 dpa and 500 °C. Further, the irradiated samples did not show any bubble or void.     

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.     

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.     

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.     

Structural and phase states and irradiation-induced defects in Ni-Cr alloys

The structural and phase states in alloys of the Ni-Cr-Mo system which were induced by both heat aging and electron irradiation at elevated temperatures have been studied by the methods of measurement of residual resistivity and positron annihilation. Migration of irradiation-induced defects during irradiation at 300°C is shown to initiate processes of ordering or phase separation depending on the initial alloy microstructure and chromium content. It has been established that in the alloy with 32 wt % Cr the concentration of accumulated vacancy defects in the state of short-range ordering after irradiation with 5-MeV electrons to a dose of ～1.5 × 10^?4 dpa at 200°C is half as high as that in the state of long-range ordering with a homogeneous distribution of domains (to 10 nm in size) of the ordered Ni₂Cr phase in the matrix.     

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.     

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.     

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.     

Cyclic deformation behaviors of Ti-46Al-2Cr-2Nb-0.15B alloy during thermo-mechanical fatigue tests

Thermo-mechanical fatigue tests were carried out on the gamma-TiAl alloy in the temperature range of 500-800 °C under mechanical strain control in order to evaluate its cyclic deformation behaviors at elevated temperature.Cyclic deformation curves,stress-strain hysteresis loops under different temperature-strain cycles were analyzed and dislocation configurations were also observed by TEM.The mechanisms of cyclic hardening or softening during thermo-mechanical fatigue(TMF) tests were also discussed.Results showed that thermo-mechanical fatigue lives largely depended on the applied mechanical strain amplitudes,applied types of strain and temperature.On the hysteresis loops appeared two apparent asymmetries:one was zero asymmetry and the other was tensile and compressive asymmetry.Dislocations configuration and slip behaviors were contributed to cyclic hardening or cyclic softening.     

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.     

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.     

Strain Hardening During Hot Compression Through Planar Dislocation and Twin-Like Structure in a Low-Density High-Mn Steel

In this work, we study the flow hardening behavior and microstructural evolution in a low-density Fe-21Mn-0.1C-2.0Al-2.5Si (wt.%) transformation-twinning-induced plasticity steel during hot compression. The substructures were examined by transmission electron microscopy and electron backscatter diffraction methods. The alloy exhibits a combination of strain hardening (at low and medium true compression strains, <0.2) followed by a strain softening at 800 °C. We explain the flow hardening behavior in terms of substructure refinement due to planar dislocation, as well as twin-like structure. The transmission electron microscopy results suggest that short-range ordering triggers slip planarity after straining to 0.05 at 800 °C. Twin-like structure with misorientation angle higher than 20° is formed during further plastic deformation.     

The Kinetics of Phase Transformations During Tempering in Laser Melted High Chromium Cast Steel

The precipitation of secondary carbides in the laser melted high chromium cast steels during tempering at 300-650?°C for 2?h in air furnace was characterized and the present phases was identified, by using transmission electron microscopy. Laser melted high chromium cast steel consists of austenitic dendrites and interdendritic M₂₃C₆ carbides. The austenite has such a strong tempering stability that it remains unchanged at temperature below 400?°C and the secondary hardening phenomenon starts from 450?°C to the maximum value of 672 HV at 560?°C. After tempering at 450?°C fine M₂₃C₆ carbides precipitate from the supersaturated austenite preferentially. In addition, the dislocation lines and slip bands still exist inside the austenite. While tempering at temperature below 560?°C, the secondary hardening simultaneously results from the martensite phase transformation and the precipitation of carbides as well as dislocation strengthening within a refined microstructure. Moreover, the formation of the ferrite matrix and large quality of coarse lamellar M₃C carbides when the samples were tempered at 650?°C contributes to the decrease of hardness.     

Embrittlement of a Cr-Mo Low-Alloy Steel due to Low-Temperature Neutron Irradiation

Embrittlement of a 2.25CrlMo steel stemming from neutron irradiation at 270℃ is studied by virtue of small punch testing in conjunction with scanning electron microscopy. The ductile-brittle transition temperature determined by the small punch test is much lower than that determined by the standard Charpy test. There is some irradiation-induced embrittlement effect after the steel is irradiated for 46 days with a neutron dose rate of 1.05×10-3-adpa/s(displacement per atom per second).     

Evolution of radiation defects in nickel under low-temperature neutron irradiation

Neutron irradiation of pure nickel samples in an IBB-2M research reactor has been performed at a temperature of 305 K to damaging doses of 0.0015 and 0.15 dpa. Radiation defects formed in the material under irradiation have been investigated using transmission electron microscopy. It has been established that the main types of defects are vacancy clusters and interstitial dislocation loops. Sizes of vacancy clusters have been measured, and histograms of the cluster-size distribution have been constructed. It has been shown that, after irradiation with a dose of 0.15 dpa, the average cluster size is nearly half of that for samples irradiated with a dose of 0.0015 dpa. In the framework of the model of the migration of point defects, their evolution under irradiation has been analyzed. It has been shown that, at a temperature of 305 K, vacancies in nickel are immobile and migrating interstitials falling into clusters recombine with vacancies in them, which results in the exhaustion of clusters. The average life span of clusters has been calculated, and average concentrations of vacancies and interstitials under irradiation have been estimated.     

Dissolution corrosion of 4H‐SiC in lead‐bismuth eutectic at 550°C

The corrosion behavior of 4H‐SiC in lead‐bismuth (Pb‐Bi) eutectic (LBE) at 550°C is investigated. To clarify the effect of irradiation damage on corrosion, samples with and without Si⁵⁺ ion irradiation are contrastively evaluated. The main results show that dissolution corrosion occurs in both unirradiated and irradiated samples, while the irradiation damage can accelerate the corrosion rate. The corroded surface is characterized by the loss of C element and the formation of amorphous layers with a slight enrichment of Si atoms. The possible reasons are discussed.     

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.     

Effect of Mn addition on one-dimensional migration of dislocation loops in body-centered cubic Fe

Elucidation of the one-dimensional (1-D) motion of dislocation loops is important for describing the microstructural development of materials under irradiation. In this study, the effect of Mn on radiation-induced microstructure evolution in body-centered cubic Fe was experimentally investigated by focusing on the migration of dislocation loops. Pure Fe and Fe–1.4Mn alloy were irradiated with Fe³⁺ ions to introduce dislocation loops. In pure Fe, inhomogeneous distribution of loops in the vicinity of the residual dislocation was observed. However, in Fe–1.4Mn, isolated dislocation loops were homogeneously distributed in a high number density. In situ transmission electron microscopy during annealing revealed that 1-D motion of dislocation loops occurred in pure Fe at 623 K, while 1-D motion of dislocation loops occurred minimally in Fe–1.4Mn annealed at temperatures below 773 K. These results indicate that 1-D motion of dislocation loops play a key role in producing the differences in the microstructures between pure Fe and Fe–1.4Mn. In pure Fe, dislocation loops were mobile and trapped in the strain field of a dislocation, leading to the formation of loop decoration of dislocations. However, in Fe–1.4Mn, dislocation loops were less mobile and dislocation loops were homogeneously formed in high density in the matrix. The migration of dislocation loops by Mn solute is strongly suggested as one of the key mechanisms of microstructure development in irradiated Fe–Mn alloy.     

Characteristics of radiation porosity and structural phase state of reactor austenitic 07C–16Cr–19Ni–2Mo–2Mn–Ti–Si–V–P–B Steel after neutron irradiation at a temperature of 440–600°C to damaging doses of 36–94 dpa

The phase composition and the characteristics of vacancy voids in cold-worked steel 07C–16Cr–19Ni–2Mo–2Mn–Ti–Si–V–P–B (CW EK164-ID) after neutron irradiation at damaging doses of 36–94 dpa and temperatures of 440–600°C are investigated. In the entire range of damaging doses and temperatures, voids with different sizes are observed in the material. The maximum void size increases with irradiation temperature up to ~550°C, whereas their concentration decreases. At higher irradiation temperatures, almost no coarse voids are observed. The concentration of fine voids (to 10 nm in size) sharply increases with temperature from 440 to 480°C. Further increases in the temperature do not result in the noticeable concentration growth. In the irradiation temperature range of 440–515°C, second phases precipitate (G phase, γ’ phase, and complex fcc carbides). At higher irradiation temperatures, there are Laves-phase particles, fine second carbides of the MC type, and needle shape precipitates identified as phosphides in the material.