Hardening and microstructural evolution in A533B steels under neutron irradiation and a direct comparison with electron irradiation

A533B steels irradiated at 290 °C up to 10 mdpa in the Kyoto University Reactor were examined by hardness, positron annihilation and atom probe measurements. Dose dependent irradiation hardening and formation of Cu-rich clusters were confirmed in medium Cu (0.12% and 0.16%Cu) steels whereas neither hardening nor cluster formation was detected in low Cu (0.03%Cu) steel. No microvoids were formed in any of the steels. Post-irradiation annealing in medium Cu steels revealed that the hardening recovery at temperatures above 350-400 °C could be attributed to compositional changes and dissociation of the Cu-rich clusters. Compared to electron irradiation at almost the same dose and dose rate, KUR irradiation caused almost the same hardening and produced Cu-rich clusters, more solute-enriched with larger size and lower density. Considering lower production of freely-migrating vacancies in neutron irradiation, the results suggested that cascades enhance the formation of Cu-rich clusters.     

Effects of dose rate change under irradiation on hardening and microstructural evolution in A533B steel

Iron-ion irradiations were carried out for 0.09wt%Cu A533B steel specimens at 290°C to investigate effects of dose rate change during irradiation; the irradiations consisted of the base irradiation (with an unchanged dose rate) and an additional one with changed dose rates from 1 to 50 times that of the base one. Nano-indentation hardness measurements showed that the increase in hardness was higher for lower dose rate of the base irradiation. A similar trend was identified during the additional irradiation. Transmission electron microscope (TEM) and three-dimensional atom probe (3DAP) analyses were carried out for the quantitative characterization of defect features. Mn/Ni/Si/Cu-enriched clusters and dislocation loops were observed in all specimens. The increase in hardness mainly depended on the formation of the solute atom clusters. The square root of the volume fraction of the solute atom clusters provided a good correlation with the increase in hardness. The effects of dose rate and dose rate change during irradiation were explained by the formation of solute atom clusters.     

Effects of stress on radiation hardening and microstructural evolution in A533B steel

Bent specimens of A533B steel (0.16 wt% Cu) were irradiated at 290 °C to 1 dpa with 6.4 MeV Fe³⁺ ions. Calculated tensile stresses at the irradiated surface were set to 0, 250, 500 and 750 MPa. The specimens were subjected to hardness measurements, transmission electron microscopy (TEM) observations and three-dimensional atom probe (3DAP) analysis. The radiation-induced hardening decreased with increasing stress to 500 MPa which was near the yield strength. TEM and 3DAP results showed that well-defined dislocation loops and solute clusters were formed. The diameter of dislocation loops increased and the number density decreased when the stress was applied, whereas the diameter and number density of solute clusters decreased. The hardening was mainly attributed to solute cluster formation. Application of tensile stress would control hardening by suppressing the solute cluster nucleation and growth.     

Dislocation loop evolution under ion irradiation in austenitic stainless steels

A solution annealed 304 and a cold worked 316 austenitic stainless steels were irradiated from 0.36 to 5 dpa at 350 °C using 160 keV Fe ions. Irradiated microstructures were characterized by transmission electron microscopy (TEM). Observations after irradiation revealed the presence of a high number density of Frank loops. Size and number density of Frank loops have been measured. Results are in good agreement with those observed in the literature and show that ion irradiation is able to simulate dislocation loop microstructure obtained after neutron irradiation.Experimental results and data from literature were compared with predictions from the cluster dynamic model, MFVIC (Mean Field Vacancy and Interstitial Clustering). It is able to reproduce dislocation loop population for neutron irradiation. Effects of dose rate and temperature on the loop number density are simulated by the model. Calculations for ion irradiations show that simulation results are consistent with experimental observations. However, results also show the model limitations due to the lack of accurate parameters.     

Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.     

Sustained load crack growth in A533B-1 steel under neutron irradiation in a water environment

A study of the combined effects of radiation, water and temperature on sustained load crack growth behavior of reactor pressure vessel steel A533B-1 is reported. To complete this study wedge opening loading (WOL) T-type fracture toughness specimens were prepared from a sample of A533B-1 steel which had a copper content of 0.13%. The crack length change was measured after 939 hr of irradiation in a water environment. An electrical potential method was successfully used to measure the crack length of rusted radioactive specimens. Sustained load crack growth occurred at initial stress intensity factor K_Ii as low as . The value of stress corrosion cracking threshold factor K_Iscc after neutron irradiation in a water environment appears to be in the range of . The results of neutron irradiation in a water environment are to apparently increase the susceptibility of A533B-1 steel to stress corrosion cracking and hydrogen embrittlement.     

Understanding silicon-rich phase precipitation under irradiation in austenitic stainless steels

Atom probe samples have been Fe⁺ ion irradiated at different doses (from 0.5 to 10 dpa) and different temperatures (between 300 and 400 °C) in order to understand the mechanism of formation, under irradiation, of Si-rich phases in austenitic stainless steels. Atom probe results show the presence of Si-enriched clusters which can also be enriched in Ni and depleted in Cr. Number densities of solute clusters can be linked to number densities of dislocation loops already observed by transmission electron microscopy in a previous work. This suggests that solute clusters are formed by heterogeneous precipitation on dislocation loops. Furthermore, the evolution of the composition of solute clusters as a function of the irradiation temperature is consistent with a radiation-induced mechanism. Results are also compared with previous results obtained after neutron irradiation at lower dose rate (in term of dpa s⁻¹). The comparison is, here again, consistent with the radiation-induced mechanism. Thus, Si-rich clusters may be formed by radiation-induced segregation to dislocation loops. Results also show that Si is probably dragged to sinks via the interstitial mechanism.     

Influence of irradiation temperature and dose gradients on the microstructural evolution in neutron-irradiated 316SS

A cold worked 316SS baffle bolt was extracted from the Tihange pressurized water reactor and sectioned at three different positions. The temperature and dose at the 1-mm bolt head position were 593 K and 19.5 dpa respectively, whereas at two shank positions the temperature and dose was 616 K and 12.2 dpa at the 25-mm position and 606 K and 7.5 dpa at the 55-mm position. Microstructural characterization revealed that small faulted dislocation loops and cavities were visible at each position, but the cavities were most prominent at the two shank positions. Measurable swelling exists in the shank portions of this particular bolt, and accompanying this swelling is the retention of very high levels of hydrogen absorbed from the environment. The observation of cavities in the CW 316SS at temperatures and doses relevant to LWR conditions has important implications for pressurized water reactors since SA 304SS plates surround the bolts, a steel that usually swells earlier due to its lower incubation period for swelling.     

Defect and void evolution in oxide dispersion strengthened ferritic steels under 3.2 MeV Fe ion irradiation with simultaneous helium injection

In an attempt to explore the potential of oxide dispersion strengthened (ODS) ferritic steels for fission and fusion structural materials applications, a set of ODS steels with varying oxide particle dispersion were irradiated at 650°C, using 3.2 MeV Fe⁺ and 330 keV He⁺ ions simultaneously. The void formation mechanisms in these ODS steels were studied by juxtaposing the response of a 9Cr–2WVTa ferritic/martensitic steel and solution annealed AISI 316LN austenitic stainless steel under the same irradiation conditions. The results showed that void formation was suppressed progressively by introducing and retaining a higher dislocation density and finer precipitate particles. Theoretical analyses suggest that the delayed onset of void formation in ODS steels stems from the enhanced point defect recombination in the high density dislocation microstructure, lower dislocation bias due to oxide particle pinning, and a very fine dispersion of helium bubbles caused by trapping helium atoms at the particle–matrix interfaces.     

Solute segregation along non-migrated and migrated grain boundaries during electron irradiation in austenitic stainless steels

Solute distribution and precipitation in the vicinity of the grain boundary in Type 316 steels were studied during electron irradiation up to about 50 dpa at temperatures from room temperature to 873 K. Undersized solute atoms, such as nickel, silicon and phosphorus, segregate toward the grain boundary, and oversized solutes, chromium and molybdenum, segregate away from the grain boundary during irradiation in the temperature range between 623 and 873 K. Enrichment of silicon and phosphorus along the grain boundary occurs after the irradiation at room temperature. The segregation of solute atoms increases with irradiation temperature except for silicon and phosphorus; the concentration of silicon and phosphorus along the grain boundary exhibits a maximum at 773 K. Remarkable depletion of chromium with enrichment of nickel, silicon and phosphorus occurs in the area swept by the migrating grain boundary. Massive M₂₃C₆ type carbide precipitates in front of the migrating grain boundary during irradiation in the temperature range from 723 to 873 K in the steels.     

Sputtering and ion-induced electron emission of graphite under high-dose nitrogen bombardment

The dependence of the sputtering yield Y and the electron emission coefficient γ of isotropic graphites (POCO-AXF-5Q and Russian MPG-LT) on ion fluence and ion incidence angle θ at near room temperatures and the dependence of γ on target temperature under high dose 30 keV N₂⁺ ion irradiation were measured. It was found that Y and γ are stabilized at fluences F?1×10¹⁹ N/cm². A specific target surface topography develops. At steady-state conditions, the N concentration in MPG-LT is 19 at.% and in POCO16 at.%. In the angular range θ=0-80°, Y and γ increase and the angular dependence of Y is slightly stronger than that of γ. Sputtering yields of POCO are 1.5 times higher than those of MPG-LT. The reasons of the difference between the experimental and calculated sputtering yields using the TRIM.SP code are discussed. The dependence of γ on the target temperature manifests a step-like increase at ?250 °C which may be due to radiation induced structure transformation in the modified surface layer.     

Nano-scale quasi-melting of alkali-borosilicate glasses under electron irradiation

Quasi-melting of micro- and nano-samples during transmission electron microscope irradiation of glassy materials is analysed. Overheating and true melting by the electron beam is shown not to be an explanation due to the ultra-sharp boundary between transformed and intact material. We propose that the observed fluidisation (quasi-melting) of glasses can be caused by effective bond breaking processes induced by the energetic electrons in the electron beam. The bond breaking processes modify the effective viscosity of glasses to a low activation energy regime. The higher the electron flux density the lower is the viscosity. Quasi-melting of glasses at high enough electron flux densities can result in shape modification of nano-sized particles including formation of perfect beads due to surface tension. Accompanying effects, such as bubble formation and foil bending are revisited in the light of the new interpretation.     

Non-equilibrium intragrain concentration redistribution of the alloying elements in austenitic steels under irradiation

Radiation-induced and thermally activated decomposition of austenitic 16Cr15Ni3Mo1Ti, 32Ni, 34Ni steels at high temperature (500–650°C) were examined. High doses (up to 10–200 dpa ) in 16Cr15Ni3Mo1Ti stainless steel with 1.5 MeV Kr ions and 450 keV Fe ions at 550–650°C lead to the appearance of relatively large regions (up to 200–400 nm) of concentration-oscillations with ‘mosaic’ dark–white diffraction contrast in TEM images. The radiation-induced redistribution of alloying elements takes place thanks to inverse Kirkendall effect and, in particular, removal of Cr to cell volume from boundaries of coarse cellular structure. The competing formation of ultrafine subgrains and grains 10–50 nm in size structure in Ti-free steel (16Cr15Ni3Mo) impedes the development of large ingrain segregations. The Mössbauer investigations showed that the 32Ni and 34Ni steels with purposefully produced concentration-oscillations were fully homogenised in that high-temperature region. This fact indicated the absence of the thermal decomposition dome in the Fe–Ni equilibrium diagram.     

Time dependent ductile crack extension in A533B class I steel

Four wide plate specimens manufactured in A533B Class I, 90 mm thick by 500 mm wide containing through-thickness or semi-elliptical surface fatigue cracks were tested at +70°C. These specimens were subjected to a series of increasing applied loads, each of 100 h duration, until failure. Testing was performed using a computer interactive 40 MN load controlled tensile testing rig. Values of the fracture toughness parameters J and crack tip opening displacement (CTOD) were derived from the recorded values of applied load, plate extension and crack mouth opening displacement.The influence of loading rate, degree of yield containment and crack orientation on the time dependent behaviour is assessed and compared with data obtained from wide plate and bend tests under monotonic loading and from bend tests conducted with a variable loading rate, with hold periods, under crack mouth opening control. Interpretation of the results provides a clearer understanding of low temperature time dependent ductile crack extension and enables the identification of the conditions under which this phenomenon is apparent, to allow the necessary adjustments to failure assessments.     

Swelling of austenitic stainless steels under fast neutron irradiation at elevated temperatures

A discussion of studies on volume increase phenomena observed in highly irradiated austenitic stainless steels is given. Volume increases are related to the production of polyhedral cavities. The likely source of swelling is a combination of the effects of two reactions: the helium-producing (n, α) reaction, and the vacancy-producing displacement reaction. It is believed that helium atoms so generated stabilize the cavity nuclei.     

电子束辐照钻石热传导过程的数值模拟

应用Matlab程序，对钻石内部和边界采用不同的差分迭代格式，计算模拟了其中的热传导过程。数值计算结果，所用方法与PDE工具箱的计算精度相当，但计算时间明显缩短。同时讨论了辐照时间、吸收剂量率、环境温度等因素对于钻石内部温度分布的影响，这对于钻石的辐照工艺具有重要意义。     

Microstructural investigation of the stability under irradiation of oxide dispersion strengthened ferritic steels

Some fuel pin cladding made from a ferritic steel reinforced by titanium and yttrium oxides were irradiated in the French experimental reactor Phénix. Microstructural examination of this alloy indicates that oxides undergo dissolution under irradiation. This irradiation shows the influence of dose and, in a smaller part, of temperature. In order to better understand the mechanisms of dissolution, three ferritic steels reinforced by Y₂O₃ or MgO were irradiated with different charged particles. Inelastic interactions induced by 1 MeV He ion irradiation do not lead to any modification, neither in their chemical composition, nor in their spatial and size distribution. In contrast, isolated Frenkel pairs created by electron irradiation lead to significant oxide dissolution with a radius decrease proportional to the dose. Moreover, the comparison between irradiation with ions (displacements cascades) and electrons (Frenkel pairs only) shows the importance of free point defects in the dissolution phenomena.     

Heterogeneity of ion irradiation-induced hardening in A533B reactor pressure vessel model alloys

Reactor pressure vessels comprise bainitic steel structures, and are heterogeneous on the mesoscale. Nanoindentation techniques were used to evaluate the hardness of these structures on the micrometer scale, and to evaluate the heterogeneity in a specimen using the distribution of the hardness. Three A533B model alloys were irradiated by 2.8 MeV Fe²⁺ ions at 563 K, and the effects of ion fluence, ion flux, and chemical composition on the change in the hardness distribution were examined. Heterogeneity of the hardening is observed in high-copper specimens irradiated up to (2–10) × 10¹⁴ ions/cm², where the average hardness increases the most. In these specimens, the hardness distribution broadens, and demonstrates that the hardening in certain positions (possibly where the initial hardness is high) is greater than in other positions. Variation in initial chemical composition (especially copper and carbon) or sink strength may cause a difference in the curing behavior.     

Neural-network analysis of irradiation hardening in low-activation steels

An artificial neural network has been used to model the irradiation hardening of low-activation ferritic/martensitic steels. The data used to create the model span a range of displacement damage of 0-90 dpa, within a temperature range of 273-973 K and contain 1800 points. The trained model has been able to capture the non-linear dependence of yield strength on the chemical composition and irradiation parameters. The ability of the model to generalise on unseen data has been tested and regions within the input domain that are sparsely populated have been identified. These are the regions where future experiments could be focused. It is shown that this method of analysis, because of its ability to capture complex relationships between the many variables, could help in the design of maximally informative experiments on materials in future irradiation test facilities. This will accelerate the acquisition of the key missing knowledge to assist the materials choices in a future fusion power plant.     

Identifying defect energy levels using DLTS under different electron irradiation conditions

Electron beams of 0.5, 1.5, 2.0, and 5.0 MeV were used to irradiate n-Si diodes to fluences of5.5×10~(13), 1.7×10~(14), and 3.3×1014 e cm~(-2). The forward voltage drop, minority carrier lifetime, and deep level transient spectroscopy(DLTS) characteristics of silicon p–n junction diodes before and after irradiation were compared. At the fluence of 3.3×10~(14) e cm~(-2), the forward voltage drop increased from 1.25 V at 0.5 MeV to 7.96μs at 5.0 MeV, while the minority carrier lifetime decreased significantly from 7.09 ls at 0.5 MeV to 0.06μs at 5.0 MeV. Six types of changes in the energy levels in DLTS spectra were analyzed and discussed.