Deuterium permeation behavior of erbium oxide coating on austenitic,ferritic, and ferritic/martensitic steels

Tritium permeation barrier is required in fusion blanket for reduction of loss of fuel and health hazard. In this study, deuterium permeation experiments have been performed on four kinds of steels and erbium oxide coatings fabricated by a filtered arc deposition method. The permeation flux of uncoated samples shows diffusion-limited regime in the temperature range 573–723 K and the permeability is corresponding to literature data. The coated samples deposited at room temperature have been tested at 773 K. It is found that the coatings suppress the deuterium permeation to a close level in spite of different types of steel substrates. In addition, the exponent of the driving pressure slightly changes compared to the uncoated sample. However, the permeation regime is still near diffusion limited.     

Irradiation-induced precipitation modelling of ferritic steels

In high strength low alloy (HSLA) steels typically used in reactor pressure vessels (RPV), irradiation-induced microstructure changes affect the performance of the components. One such change is precipitation hardening due to the formation of solute clusters and/or precipitates which form as a result of irradiation-enhanced solute diffusion and thermodynamic stability changes. The other is irradiation-enhanced tempering which is a result of carbide coarsening due to irradiation-enhanced carbon diffusion. Both effects have been studied using a recently developed Monte Carlo based precipitation kinetics simulation technique and modelling results are compared with experimental measurements. Good agreements have been achieved.     

Fe-Cr-V ferritic steels for possible nuclear applications

Basic mechanical and metallurgical properties of specific ferritic Fe-Cr-V alloys and steels with 5, 10, and 15 wt.% vanadium were investigated. Vanadium is an effective carbide former and can also form a brittle sigma phase with chromium. Therefore, the microstructural investigations focused on the determination and analysis of possible precipitations. The present study showed that sigma phase precipitates increase significantly in alloys with 10 wt.% Cr and 10 wt.% V. The addition of carbon led to grain refinement due to the stabilizing effect of VC. In this way, precipitation hardening as well as fine grain strengthening could be quantified for this class of material. However, compared to typical martensitic steels, the strength of the considered ternary Fe-Cr-V alloys and steels is still lower.     

Investigation on the precipitation behavior of M3C phase in T91 ferritic steels

As traditional 9-12% Cr heat-resistant ferritic steels, T91 steels have been considered as candidate reduced-activation materials for nuclear engineering applying due to its excellent creep resistance and high resistance to void formation during neutron irradiations at elevated temperature. Needle-like M₃C precipitates are produced during the routine normalizing process before tempering. Differential scanning calorimetry and infrared radiation thermometer have been employed to study the precipitation behavior of the secondary M₃C particles upon subsequent cooling process after austenization. Various austenization conditions (holding time, temperature and the subsequent cooling rate) were carried out to clarify effect of normalizing condition on the formation of the M₃C phase. In spite of various austenization conditions applied, it is found that the precipitation of M₃C phase is depends only on the cooling rate applied. Furthermore, the precipitation of M₃C phase occurs before the onset of the martensite transformation, which is contrary to the previous statement that it takes place during the auto-tempering stage after martensitic transformation. The above observation points out that the precipitation of M₃C would produce an effect on the subsequent martensitic transformation behavior, leading to the formation of wide martensite laths with a low dislocation density.     

Effect of tungsten and tantalum on the low cycle fatigue behavior of reduced activation ferritic/martensitic steels

Reduced activation ferritic/martensitic (RAFM) steels are candidate materials for the test blanket modules of International Thermonuclear Experimental Reactor (ITER). Several degradation mechanisms such as thermal fatigue, low cycle fatigue, creep fatigue interaction, creep, irradiation hardening, swelling and phase instability associated irradiation embrittlement must be understood in order to estimate the component lifetime and issues concerning the structural integrity of components. The current work focuses on the effect of tungsten and tantalum on the low cycle fatigue (LCF) behavior of RAFM steels. Both alloying elements tungsten and tantalum improved the fatigue life. Influence of Ta on increasing fatigue life was an order of magnitude higher than the influence of W on improving the fatigue life. Based on the present study, the W content was optimized at 1.4 wt.%. Softening behavior of RAFM steels showed a strong dependence on W and Ta content in RAFM steels.     

Tensile properties of ferritic/martensitic steels irradiated in STIP-I

Specimens of ferritic/martensitic (FM) steels T91, F82H, Optimax-A and the electron beam weld (EBW) of F82H were irradiated in the Swiss spallation neutron source (SINQ) Target-3 in a temperature range of 90-370 °C to displacement doses between 3 and 12 dpa. Tensile tests were performed at room temperature and the irradiation temperatures. The tensile test results demonstrated that the irradiation hardening increased with dose up to about 10 dpa. Meanwhile, the uniform elongation decreased to less than 1%, while the total elongation remained greater than 5%, except for an F82H specimen of 9.8 dpa tested at room temperature, which failed in elastic deformation regime. At higher doses of 11-12 dpa, the ductility of some specimens recovered, which could be due to the annealing effect of a short period of high temperature excursion. The results do not show significant differences in tensile properties for the different FM steels in the present irradiation conditions.     

Impurity effects on reduced-activation ferritic steels developed for fusion applications

Reduced-activation steels are being developed for fusion applications by restricting alloying elements that produce long-lived radioactive isotopes when irradiated in the fusion neutron environment. Another source of long-lived isotopes is the impurities in the steel. To examine this, three heats of reduced-activation martensitic steel were analyzed by inductively coupled plasma mass spectrometry for low-level impurities that compromise the reduced-activation characteristics: a 5-ton heat of modified F82H (F82H-Mod) for which an effort was made during production to reduce detrimental impurities, a 1-ton heat of JLF-1, and an 18-kg heat of ORNL 9Cr–2WVTa. Specimens from commercial heats of modified 9Cr–1Mo and Sandvik HT9 were also analyzed. The objective was to determine the difference in the impurity levels in the F82H-Mod and steels for which less effort was used to ensure purity. Silver, molybdenum, and niobium were found to be the tramp impurities of most importance. The F82H-Mod had the lowest levels, but in some cases the levels were not much different from the other heats. The impurity levels in the F82H-Mod produced with present technology did not achieve the low-activation limits for either shallow land burial or recycling. The results indicate the progress that has been made and what still must be done before the reduced-activation criteria can be achieved.     

Influence of liquid lead and lead-bismuth eutectic on tensile, fatigue and creep properties of ferritic/martensitic and austenitic steels for transmutation systems

In this paper, the tensile, fatigue and creep properties of the Ferritic/Martensitic (F/M) steel T91 and of the Austenitic Stainless (AS) Steel 316L in lead-bismuth eutectic (LBE) or lead, obtained in the different organizations participating to the EUROTRANS-DEMETRA project are reviewed. The results show a remarkable consistency, referring to the variety of metallurgical and surface state conditions studied. Liquid Metal Embrittlement (LME) effects are shown, remarkable on heat-treated hardened T91 and also on corroded T91 after long-term exposure to low oxygen containing Liquid Metal (LM), but hardly visible on passive or oxidized smooth T91 specimens. For T91, the ductility trough was estimated, starting just above the melting point of the embrittler (T_M,E = 123.5 °C for LBE, 327 °C for lead) with the ductility recovery found at 425 °C. LME effects are weaker on 316L AS steel. Liquid Metal Assisted Creep (LMAC) effects are reported for the T91/LBE system at 550 °C, and for the T91/lead system at 525 °C. Today, if the study of the LME effects on T91 and 316L in LBE or lead can be considered well documented, in contrast, complementary investigations are necessary in order to quantify the LMAC effects in these systems, and determine rigorously the threshold creep conditions.     

Depinning transition of a dislocation line in ferritic oxide strengthened steels

The dynamics of an edge dislocation in a medium with random oxide dispersoid particles acting as pinning centres is analysed. The dislocation line undergoes a depinning transition, where the order parameter is the dislocation line velocity v, which increases from zero for driving external resolved shear stresses τ beyond to a threshold value τ_c, known as the critical resolved shear stress. The critical stress is obtained by means of statistical analysis of the motion of a single dislocation in its glide plane, using overdamped, discrete dislocation dynamics simulations.     

Corrosion of various ferritic steels in an isothermal sodium loop system

Ferritic chromium-molybdenum steels with chromium contents of 1 wt% up to 12 wt% have been exposed for 8370 h to flowing sodium at 550°C. The oxygen content of the sodium was 6–7 ppm by weight. Weight measurements, carbon analyses and metallographic examinations were carried out. The low chromium steels show weight loss and decarburisation. The high chromium steels show weight gain and carburisation. The crossover point is at about 5 wt% Cr. The composition at the utmost surface (<10 μm) of the various steels tends to about 8 wt% chromium, about 2 wt% nickel and 0.02–0.09 wt% carbon. Sodium chromite crystals were present on the steels with a chromium content of 5 wt% or more. At the exposed surface of the $\text{2}\text{1}\text{4}$ wt% chromium steel sodium chromite crystals were found locally.     

Characterization of recrystallization of 12Cr and 15Cr ODS ferritic steels

The recrystallization behavior of 12Cr and 15Cr oxide dispersion-strengthened (ODS) ferritic steels, which are the promising candidate materials for long-life core materials of the advanced fast breeder reactors, was investigated in terms of an intermediate softening heat treatment. It was clarified that keeping recovery structure at the intermediate heat treatment is indispensable for producing recrystallized structure at the final heat treatment. Prevention of repeating recrystallization is owing to the stable {100} 〈110〉 texture formation with less stored strain energy by the cold-rolling of the recrystallized structure. The two-step softening process was proposed to suppress the recrystallization and obtain adequate hardness reduction at the intermediate heat treatment. This process is effective for producing a stable recrystallized structure at the final heat treatment of the manufacturing process of ODS ferritic steel cladding.     

High heat flux testing of 12-14Cr ODS ferritic steels

The thermal performance of Fe-(12-14)Cr-2W-0.3Ti-0.3Y₂O₃ ODS reduced activation ferritic steels, which are considered as candidate first wall materials for the future fusion power reactors and were manufactured by mechanical alloying in hydrogen and hot isostatic pressing, was assessed by high heat flux (HHF) testing with the electron beam JUDITH facility at the Forschungszentrum Jülich (FZJ), Germany. An analysis of the microhardness and microstructure of the specimens was done before and after HHF tests.In general, both materials present a ferritic (α-Fe, bcc) microstructure with a wide range of grain sizes from 100 to 500 nm up to a few micrometers. The coarse grains are almost dislocation-free, while the smaller ones are surrounded by tangles of dislocations. Oxide and carbide impurities (about a few hundreds nm in size) and a high density of Y-Ti-O nano-clusters, with a mean size of about 5 nm, are also present. The microhardness, density and tensile strength of the 14Cr material are slightly larger than those of the 12Cr material.HHF tests revealed that there is no difference in thermal performance, level of degradation and erosion behaviour of 12Cr and 14Cr ODS steels. The onset of melting of the materials occurs for an energy density between 1 and 1.5 MJ/m². Below this value only some kind of thermal etching takes place. This is a significant improvement compared to stainless steel, for which severe plastic deformation at the material surface was observed.     

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Irradiation damage caused by neutron irradiation below 425-450 °C of 9-12% Cr ferritic/martensitic steels produces microstructural defects that cause an increase in yield stress. This irradiation hardening causes embrittlement observed in a Charpy impact test as an increase in the ductile-brittle transition temperature. Little or no change in strength is observed in steels irradiated above 425-450 °C. Therefore, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study, significant embrittlement was observed in F82H steel irradiated at 500 °C to 5 and 20 dpa without any change in strength. Earlier studies on several conventional steels also showed embrittlement effects above the irradiation-hardening temperature regime. Indications are that this embrittlement is caused by irradiation-accelerated or irradiation-induced precipitation. Observations of embrittlement in the absence of irradiation hardening that were previously reported in the literature have been examined and analyzed with computational thermodynamics calculations to illuminate and understand the effect.     

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.     

Void-swelling in irons and ferritic steels: I. Mechanisms of swelling suppression

Point defect and dislocation behaviour in α-iron and ferritic steels relevant to the understanding of their void-swelling resistance during elevated temperature irradiation is summarized. The key role played by both interstitial and substitutional solutes in inducing enhanced mutual recombination of point defects through trapping processes and also in lowering dislocation mobility and bias for preferential self-interstitial capture in these materials, and thereby controlling their void-swelling response, is emphasized.     

Influence of HIP pressure on tensile properties of a 14Cr ODS ferritic steel

An oxide dispersion strengthened ferritic steel with a nominal composition of Fe–14Cr–2W–0.3Ti–0.3Y₂O₃ (in wt.%) was consolidated by hot isostatic pressing at 1150 °C under various pressures in the range of 185–300 MPa for 3 h. The microstructure, microhardness and high temperature tensile properties of the steel were investigated. With increasing compaction pressure the density of specimens also increased, however OM and SEM observations revealed residual porosity in all tested specimens and similar ferritic microstructure with bimodal-like grains and numerous of large oxide particles, located at the grain boundaries. Mechanical testing revealed that compaction pressure has negligible influence on the hardness and tensile strength of the ODS steel, however improves the material ductility.     

Damage region formation in ferritic steels induced by light water reactor neutron spectrum

A new theoretical model for damage region formation is proposed. The model is based on numerical solution of the Boltzmann transport equation for knocked-on atoms. A key point of this model is the selfconsistent determination of subcascade overlapping energy E_over (the threshold energy for distinguished damage region formation). Damage region density and size distributions in ferritic steels (Fe–0.2 wt% Cu and Fe–0.2 wt% Cu–0.3 wt% Si) under neutron irradiation in light water reactor spectrum were calculated.     

Metallography studies and hardness measurements on ferritic/martensitic steels irradiated in STIP

In this work metallography investigations and microhardness measurements have been performed on 15 ferritic/martensitic (FM) steels and 6 weld metals irradiated in the SINQ Target Irradiation Program (STIP). The results demonstrate that all the steels have quite similar martensite lath structures. However, the sizes of the prior austenite grain (PAG) of these steels are quite different and vary from 10 to 86 μm. The microstructure in the fusion zones (FZ) of electron-beam welds (EBWs) of 5 steels (T91, EM10, MANET-II, F82H and Optifer-IX) is similar in respect to the martensite lath structure and PAG size. The FZ of the inert-gas-tungsten weld (TIGW) of the T91 steel shows a duplex structure of large ferrite gains and martensite laths. The microhardness measurements indicate that the normalized and tempered FM steels have rather close hardness values. The unusual high hardness values of the EBW and TIGW of the T91 steel were detected, which suggests that these materials are without proper tempering or post-welding heat treatment.