Irradiated Steel Research Articles

EFFECT OF POST-RADIATION ANNEALING ON SWELLING, MAGNETIZATION AND CORROSION RESISTANCE OF AUSTENITIC STEEL 12Cr18Ni10Ti, IRRADIATED IN THE REACTOR BN-350

The comparative study has been performed on the post-operation changes in the structural-phase state, magnetization, swelling and corrosion properties of the samples of structural reactor steel 12Cr18Ni10Ti, cut from the walls of the hexagonal ducts of the spent fuel assemblies from various elevations from the core center of the nuclear reactor BN-350. The studies of the aging processes in the highly irradiated corrosion-resistant austenitic steel showed that as the result of isochronous annealing in the region of 400÷450 °C we can observe the effect of magnetization increase, initiation of the processes of nucleation of defects of various morphology, the changes in their density and size, which determines the degree of swelling. The relationship has been established between radiation swelling with evolution of the defective structure, density, magnetization and corrosion resistance of austenitic steel.

Behaviors of bubble-loop complexes in He-irradiated CLAM steels at elevated temperatures

The behavior of bubble-loop complexes formed in reduced activation ferritic/martensitic (RAFM) steels under He+ irradiation at elevated temperatures was investigated by transmission electron microscope (TEM), and corresponding simulations on the interaction between He bubbles and dislocation loops were also carried out by molecular dynamics (MD). China Low Activation Martensitic (CLAM) steels were irradiated under18 keV He+ with the fluence of 1 × 1020 ion/m2 at room temperature, 250 °C, 350 °C, 400 °C, 450 °C and 550 °C, respectively. Bubble-loop complexes were observed in all specimens when irradiated at 250 °C and above. Almost all of He bubbles were located inside the dislocation loops when the irradiation temperature reached up to 550 °C. Moreover, the formation of the complexes was independent of the type of dislocation loops since both a0 <100> and a0/2 <111> loops could form bubble-loop complexes. He bubbles were situated to be more stable inside the dislocation loops than that in the matrix according to the experimental results. The MD simulations showed that the dislocation loop can be locked by the He bubble, which could elucidate the mechanism of the observed bubble-loop complexes formed in He irradiated RAFM steels. The mean size of dislocation loops decreased when irradiation temperature increased to 550 °C, which can be ascribed to the high emission rate of point defects from dislocation loops at high temperature.

Full-field polycrystal plasticity simulations of neutron-irradiated austenitic stainless steel: A comparison between FE and FFT-based approaches

We compare two full-field approaches – a crystal plasticity finite element method (CP-FEM) and crystal plasticity fast Fourier transform-based (CP-FFT) method – for a specific crystal plasticity law introduced for neutron-irradiated austenitic stainless steel SA304L currently used in nuclear reactor vessel internals. This particular law is employed to identify and quantify possible advantages and drawbacks of the two approaches when used in the large-scale simulations to predict the effect of irradiation damage (e.g., crack initiation) in stainless steel microstructures. A comparison is performed in a polycrystalline context for different periodic Voronoi microstructures deformed under tension. Special emphasis is put on studying the performance of the two approaches in terms of mesh convergence analysis using aggregate models with different spatial discretizations. In the CP-FEM approach, the performance of linear as well as quadratic tetrahedral meshes is investigated. A similar performance between the CP-FEM and CP-FFT methods is demonstrated in a smaller 2-grain aggregate. However, a slower mesh convergence is observed for the CP-FEM method when comparing tensile responses of a larger 100-grain polycrystal. A drop in the convergence rate is much more pronounced on linear than on quadratic tetrahedral meshes. As a consequence, largest (average) grain boundary stresses are shown to be overestimated with the linear mesh CP-FEM approach, thus raising a concern of possible over-conservatism employed in the CP-FEM prediction of crack initiation in irradiated stainless steels. On the contrary, the mesh convergence of the CP-FFT approach is found to be practically independent of the applied macroscopic strain and also aggregate size. Therefore, for such steels, the CP-FFT approach seems to be better justified.

Comparative study of deuterium retention in irradiated Eurofer and Fe–Cr from a new ion implantation materials facility

A new facility to study the interaction of hydrogen isotopes with nuclear fusion-relevant first wall materials, and their retention and release, has been produced. The new facility allows for implanting a range of gases into samples, including tritium. An accurate study of isotope effects, such as the isotopic exchange in damaged microstructure, has previously been difficult due to a background signal of light hydrogen. This new capability will allow virtually background free measurements using tritium and deuterium. The design and build of this facility are described and commissioning results are presented. Within the UKAEA-led tritium retention in controlled and evolving microstructure (TRiCEM) project, this facility is used for the comparative study of deuterium retention in self-ion irradiated Eurofer steel and Fe–Cr alloy. Self-ion bombardment with energies of 0.5 MeV is used to mimic the defects created by neutrons in fusion power plants and the created traps are then filled with deuterium in the new facility. Implanted samples are analysed using thermal desorption spectrometry (TDS), secondary ion mass spectrometry (SIMS), and transmission electron microscopy. Results on the total deuterium content as a function of time, TDS spectra and SIMS analysis are presented. A comparison of the results for Eurofer and Fe–Cr revealed several differences. While some of them may be due to experimental details like different time delays between exposure and analysis, others, such as deuterium retention as function of dose, might be genuine and require further studies.

Role of cavities on deformation-induced martensitic transformation pathways in a laser-welded, neutron irradiated austenitic stainless steel

The role of cavities on deformation-induced martensitic phase transformations is studied in a laser-welded and neutron irradiated austenitic stainless steel. Orientation dependent nanoindentation experiments are performed in the base metal and the weld heat affected zone (HAZ) at room temperature. Transmission electron microscopy study of deformed microstructures indicates indentation-induced α’-martensite forms in the base metal, whereas α’- and ε-martensite arise in the HAZ. The different pathway of martensite phase transformation is attributed to the laser weld-induced annealing of cavities. Our results suggest that deformation-induced martensitic phase transformation of austenitic stainless steel is correlated to neutron irradiated cavity structures.

Effect of laser welding on deformation mechanisms in irradiated austenitic stainless steel

Deformation mechanism of a laser weld on neutron irradiated AISI 304L stainless steel was studied by in-situ microcompression test at room temperature. The deformation-induced austenite-to-martensite phase transformation occurs in {101}-oriented grains in the irradiated base metal, while deformation twinning prevails in {101}-oriented grains in the weld heat affected zone (HAZ). A high number density of irradiation-induced voids in the base metal provide sufficient nucleation sites for the austenite-to-martensite phase transformation under compression at room temperature. A deformation map is established to predict critical twinning stress for face-centered cubic (fcc) metals and alloys. Our study suggests that irradiation can tailor the deformation-induced phase transformation in austenitic stainless steels.

Complex study of grain boundary segregation in long-term irradiated reactor pressure vessel steels

Grain boundary (GB) embrittlement of reactor pressure vessel steel occurs at the long-term operation through the radiation-enhanced segregation of impurities. All the known techniques of studying the GB segregation have their advantages and disadvantages, and complete information can be obtained only through a comprehensive analysis by different methods.In this paper the level of GB segregation in VVER-1000 RPV steels with high nickel content was assessed by combined methods of Auger electron spectroscopy (by the P monolayer coverage and atomic concentration of other elements); atom probe tomography (by the Gibbsian interfacial excess) and fractographic analysis (by the maximum fraction of brittle intergranular fracture).The results obtained by different methods are in good agreement and show that phosphorus GB segregation increases with the increase of fast neutron fluence, GB concentration of Ni and Mn, the bulk concentration of Ni and Mn, and correlates with the fraction of brittle intergranular fracture.

Analysis of mechanisms inducing corrosion cracking of irradiated austenitic steels and development of a model for prediction of crack initiation

The austenitic steels used for RPV internals of light water reactors (LWR) are considered. Mechanisms having a potential effect on irradiation assisted stress corrosion cracking (IASCC) of austenitic steels in the LWR environment have been analyzed. Based on the analysis and generalization of reference and original data on IASCC, an IASCC initiation criterion has been formulated. The nature of low-temperature creep of irradiated austenitic steels has been considered, constitutive equations have been derived. Relying on the formulated criterion of grain-boundary microcrack nucleation and the derived creep equations, an IASCC initiation model has been developed. The model allows one to predict the dependence of the threshold stress σthIASCC on neutron dose and also to calculate the IASCC initiation time with stresses exceeding σthIASCC.

MARTENSITE POINTS FOR AUSTENITIC STEELS IRRADIATED WITH NEUTRONS

The effect of neutron irradiation on the Md martensitic point in austenitic chromium-nickel steels X18H9T and 12X18N10T – structural materials of nuclear reactors is considered. It was established that up to a fluence of 2·1020 n/cm2 (E&gt;0.1 MeV), the maximum possible temperature for the manifestation of the phase y→α transition for steel 12X18H10T varies in the range of 100–75 °C.Based on the analysis of the experimental data obtained, it is proposed to consider another characteristic point Mp – a stretching temperature at which the values of the intensities of the direct martensitic y→α transformation are compared for deformable non-irradiated and irradiated steel.

RADIATION RELAXATION OF STRESSES IN REACTOR STEELS IRRADIATED BY NEUTRONS OR PULSE ELECTRONS

Experimental results on the relaxation of internal stresses in structural stainless steels Х18Н9Т, 12Х18Н10Т and AISI 316 after neutron irradiation, effects of stationary or pulsed electrons of various powers are presented. It has been established that in steels irradiated with low neutron fluxes (1016 ÷ 1018 n/cm2), the deformation by extension as a result of relaxation phenomena is the martensitic γ → α transition begins at lower values of critical stresses than after austenization. At the same time, there is an anomalous effect of reducing the yield strength and reducing the amount of mechanical workexpended on the destruction of the steel sample. The features of the relaxation of internal stresses and the formation of the α-phase in stainless steels under irradiation with electrons of various energies are revealed. It is shown that the radiation relaxation of internal stresses caused by the action of neutrons, pulsed or stationary electrons, changing the conditions for the formation of the α-phase, determines the level of mechanical and corrosion properties of irradiated stainless steel.

Modelling the influence of strain fields around precipitates on defect equilibria and kinetics under irradiation in ODS steels: A multi scale approach

Resistance to radiation damage is a crucial property for structural materials in fission or fusion reactors. The particle-matrix interface in oxide dispersion strengthened (ODS) steels is considered a sink for point defects created during irradiation. In this work we address the question, if elastic strain fields around oxide particles cause a long-range interaction between the precipitates and point defects. We use kinetic Monte Carlo simulations to simulate the diffusion of point defects under the influence of elastic strain fields caused by Y2O3 and Y2Ti2O7 precipitates. We show that there is essentially no interaction between vacancies and the strain fields, while the sink strength of precipitates for interstitials increases with misfit strain between precipitate and matrix. Consequently, the interstitial concentration decreases with misfit strain, while due to reduced interstitial-vacancy recombination the vacancy concentration increases with misfit strain. The total change of point defect concentration with misfit strain is, however, rather limited.

Understanding the importance of the energetics of Mn, Ni, Cu, Si and vacancy triplet clusters in bcc Fe

Numerous experimental studies have found the presence of (Cu)-Ni-Mn-Si clusters in neutron irradiated reactor pressure vessel steels, prompting concerns that these clusters could lead to larger than expected increases in hardening, especially at high fluences late in life. The mechanics governing clustering for the Fe-Mn-Ni-Si system are not well-known; state-of-the-art methods use kinetic Monte Carlo (KMC) parameterized by density functional theory (DFT) and thermodynamic data to model the time evolution of clusters. However, DFT-based KMC studies have so far been limited to only pairwise interactions due to lack of DFT data. Here, we explicitly calculate the binding energy of triplet clusters of Mn, Ni, Cu, Si, and vacancies in bcc Fe using DFT to show that the presence of vacancies, Si, or Cu stabilizes cluster formation, as clusters containing exclusively Mn and/or Ni are not energetically stable in the absence of interstitials. We further identify which clusters may be reasonably approximated as a sum of pairwise interactions and which instead require an explicit treatment of the three-body interaction, showing that the three-body term can account for as much as 0.3 eV, especially for clusters containing vacancies.

Precipitation and hardening in irradiated low alloy steels with a wide range of Ni and Mn compositions

Mn-Ni-Si intermetallic precipitates (MNSPs) that are observed in some Fe-based alloys following thermal aging and irradiation are of considerable scientific and technical interest. For example, large volume fractions (f) of MNSPs form in reactor pressure vessel low alloy steels irradiated to high fluence, resulting in severe hardening induced embrittlement. Nine compositionally-tailored small heats of low Cu RPV-type steels, with an unusually wide range of dissolved Mn (0.06–1.34 at.%) and Ni (0.19–3.50 at.%) contents, were irradiated at ≈ 290 °C to ≈ 1.4 × 1020 n/cm2 at an accelerated test reactor flux of ≈3.6 × 1012 n/cm2-s (E > 1 MeV). Atom probe tomography shows Mn-Ni interactions play the dominant role in determining the MNSP f, which correlates well with irradiation hardening. The wide range of alloy compositions results in corresponding variations in precipitates chemistries that are reasonably similar to various phases in the Mn-Ni-Si projection of the Fe based quaternary. Notably, f scales with ≈ Ni1.6Mn0.8. Thus f is modest even in advanced high 3.5 at.% Ni steels at very low Mn (Mn starvation); in this case Ni-silicide phase type compositions are observed.

Insights into the sources of irradiation hardening in a neutron irradiated 304L stainless steel following post-irradiation annealing

The 304L SS was irradiated in Barsebäck 1 BWR reactor to 5.9 dpa followed by post-irradiation annealing at 500 °C and 550 °C up to 20 h. Correlations of irradiation hardening with the main hardening features of dislocation loops and Ni/Si-rich clusters were made using the least squares regression. The dispersed-barrier hardening model tended to overestimate the contribution of Ni/Si-rich clusters in the as-irradiated condition. Contribution of Ni/Si-rich clusters to irradiation hardening appeared to be more appropriately accounted for by the short-range order strengthening mechanism and was dependent on the Ni and Si concentrations in the clusters. Irradiation hardening in the as-irradiated condition was predominantly from dislocation loops with minor contribution from the Ni/Si-rich clusters. This is consistent with recent molecular dynamics simulations which showed that Ni/Si-rich clusters were weak obstacles to dislocation glide. As dislocation loops were annihilated at a faster pace than the Ni/Si-rich solute clusters, the Ni/Si-rich clusters played a more important role in the remaining hardening after thermal annealing. Contribution of Ni/Si-rich clusters to remaining hardening outweighed that of dislocation loops after thermal annealing at 550 °C for 5 h and became predominant after annealing for 20 h.

Model of corrosion cracking of irradiated austenitic steels. Part 1. Analysis of damage mechanisms and formulation of the defining

Mechanisms having a potential effect on irradiation assisted stress corrosion cracking (IASCC) of austenitic steels in the LWR environment have been analyzed. Based on the analysis and generalization of reference and original data on IASCC, an IASCC initiation criterion has been formulated. Conditions for grainboundary microcrack propagation by IASCC mechanism have been formulated. The nature of low-temperature creep of irradiated austenitic steels has been considered, constitutive equations have been derived. Relying on the formulated criterion of grain-boundary microcrack nucleation and the derived creep equations, an IASCC initiation model has been developed. The model allows one to predict the dependence of the threshold stress σIASCC th on neutron dose and also to calculate the IASCC initiation time with stresses exceeding σIASCCth .

Model of corrosion cracking of irradiated austenitic steels. Part 2. Determination of model parameters and its verification

Based on the analysis and generalization of reference and original experimental data unknown coefficients, parameters and functions in constitutive equations of the model developed in the first part of article have been defined. The model has been verified. The model allows one to predict the dependence σIASCCth (below which SCC does not occur at any time of tests) on neutron dose and to calculate the time of SCC initiation at stresses higher than σIASCCth.

Automated Quantitative Analysis of Extended Irradiation Defects - Dislocations, Voids and Precipitates in Neutron Irradiated HT-9 Steel

Automated Quantitative Analysis of Extended Irradiation Defects - Dislocations, Voids and Precipitates in Neutron Irradiated HT-9 Steel

Evaluation of mechanical properties distribution for irradiated stainless steels simulated by utilizing the gradient of the cold working ratio

It is known that irradiation damage can cause a reduction in fracture toughness and an increase in the strength of austenitic stainless steels used for the internal structures of nuclear power plant reactors. A change in the mechanical properties of material, such as fracture toughness and strength, affects the fracture behavior of the components. Actual reactor internal structures, such as the top guides and core shrouds in BWR plants, are irradiated from the core in a single direction. This causes the attenuation of neutron flux, which results in the distribution of neutron fluence in the components, meaning that the distribution of mechanical properties is in the through-wall direction of the components. It is therefore possible for the fracture mode of the components to change in the through-wall direction. It would be valuable for more accurate assessment of the structural integrity of the nuclear components to make the effect of the mechanical properties distribution on the fracture mode clear. However, it is difficult to prepare large-scale test specimens with the mechanical properties distribution induced by uni-directional neutron irradiation. In this study, the mechanical properties distribution was simulated by applying the gradient of the cold working ratio using Type 316L stainless steel plate with an uneven thickness. The targeted range of the cold working ratio was designed by assuming BWR reactor internal structures, such as the top guides. The mechanical properties distribution was evaluated based on the results of hardness, tensile and fracture toughness tests. It was found that the macroscopic mechanical properties of irradiated stainless steels can be simulated by using the cold working and the macroscopic distribution of the mechanical properties of irradiated stainless steels can also be simulated by using the cold working ratio with a gradient. Furthermore, fracture tests were performed using test plates with the mechanical properties distribution assuming irradiated components with a crack. In some cases, the maximum loads obtained by the tests were almost the same value in spite of the different cold working ratio in the ligament. Based on the results of the fracture tests, it was considered that the dominant mechanical properties for the fracture load were those at the tip of the slit, not those in the ligament.

Observation of radiation enhanced phase formation and subsequent dissolution at very high doses in Ar9+ ion irradiated Ti-modified stainless steel

Samples of Ti-modified stainless steel alloy have been irradiated using 315 KeV Ar9+ ion to varying doses resulting in high dpa levels from 100 to 400. The variation of the secondary phase present in the sample as a function of irradiation dose has been characterised using Grazing Incidence X-ray Diffraction measurements. Energy Dispersive X-ray mapping of the irradiated volume has also been carried out to see the variation in the spatial distribution of the major alloying elements in the un-irradiated and irradiated samples as a function of dose. The study reveals that there is enhancement of secondary phases due to irradiation, which at very high doses re-dissolves into the matrix.

Evolution of microstructures and hardening property of initial irradiated, post-irradiation annealed and re-irradiated Chinese-type low-Cu reactor pressure vessel steel

Investigation of microstructures and hardening property of initial irradiated, post-irradiation annealed, and re-irradiated Chinese-type low-Cu (0.01 wt%) reactor pressure vessel (RPV) steel were conducted by slow positron beam (SPB), TEM, and nanoindentation. Results of the SPB measurements indicated that a large number density of open volume defects such as vacancies, vacancy clusters, vacancy-solute/H complexes, and dislocation loops were introduced in both initial-irradiated and re-irradiated specimens. The TEM results indicated that interstitial-type dislocation loops with a number density of ∼1022 m−3 were formed and the mean size (∼3 nm) of the dislocation loops was almost unchanged in both initial-irradiated and re-irradiated specimens. The open volume defects such as vacancy-type defects and dislocation loops were almost annealed out and some stable defects still existed in the post-irradiation annealed specimen. The nanoindentation results identified that the obvious hardening phenomena were found in the initial irradiated, post-irradiation annealed and re-irradiated specimens. The quantitative analysis suggested that irradiation hardening of Chinese-type low-Cu RPV steel were attributed to matrix defects including large-size vacancy clusters, vacancy-solute/H complexes, and dislocation loops produced during initial irradiation and re-irradiation. On the other hand, the existed stable defects such as solute-atom (Mn, Ni, and Si) complexes/precipitates, which cannot be distinguished by SPB or TEM, might be responsible for the rest fraction of irradiation hardening of highly initial irradiated and re-irradiated Chinese-type low-Cu RPV steel.