Reduced Activation Martensitic Steel Research Articles

Effect of dislocation on helium irradiation defects in low-activation martensitic steel

Reduced-activation martensitic steel is a main candidate structural material for key components of advanced nuclear energy systems because of its good mechanical properties at room temperature, low neutron activation characteristics and satisfactory radiation resistance. In this work, we study the interaction between dislocations and helium-irradiation-induced defects in the steel and the effect of dislocations on the behavior of helium migration and desorption. The well-annealed samples are pre-deformed to 10% and 20% reductions in thickness by using cold rolling mill. After pre-deformation, samples are heat-treated to remove vacancy defects and retain dislocation defects. Then, the samples with reserved dislocations are irradiated with helium at room temperature (50 keV, 1×10<sup>17</sup> He/cm<sup>2</sup>). After irradiation, the samples are characterized by synchrotron radiation grazing incident X-Ray diffraction, positron annihilation Doppler broadening spectroscopy, and thermal desorption spectroscopy. The results show that dislocations hinder the diffusion of helium and helium-vacancy complexes, and reduce the accumulation of radiation damage. Such an effect becomes more significant with the increase of dislocation density. The BCC → FCC phase transition of low activation martensitic steel occurs at 1179 K. The increase of dislocation density will lead to the forward shift of helium desorption peak induced by phase transition. The retention of helium in the undeformed sample, 10% deformed sample and 20% deformed sample is 10.3%, 15.7% and 17.9%, respectively, indicating that high density dislocations promote the retention of helium.

Ion-irradiation hardening and microstructural evolution in F82H and ferritic alloys

A reduced activation martensitic steel (F82H), Fe−8Cr ferrite model steel (8CrFS) and pure iron (Fe) were irradiated with 6.4 MeV Fe3+ ions at 300 °C to a displacement damage of 30 dpa at a nominal depth of 600 nm. Nano-indentation tests were performed at room temperature in continuous stiffness measurement (CSM) mode before and after the irradiation. Nix-Gao model was used to extract the nano-hardness of specimens and irradiation hardening (ΔH) was measured. Irradiation hardening was observed in all materials with F82H lowest (ΔH = 0.85 GPa) and pure iron highest (ΔH = 4.03 GPa). Transmission electron microscopy (TEM) observations indicated that dislocation loops dominated the damage microstructure after ion-irradiation. Both the average diameter and number density of dislocation loops were obtained. The correlation between microstructure and irradiation hardening was established based on dispersed barrier hardening (DBH) model. The obstacle strength factors of <100> and 1/2 <111> loops were assigned to be 0.54 and 0.17 with an assumption of linear rule.

Evolution of Dislocation Loops Induced by Different Hydrogen Irradiation Conditions in Reduced-Activation Martensitic Steel.

Hydrogen can be induced in various ways into reduced-activation ferritic/martensitic (RAFM) steels when they are used as structural materials for advanced nuclear systems. However, because of the fast diffusion of hydrogen in metals, the effect of hydrogen on the evolution of irradiation-induced defects was almost neglected. In the present work, the effect of hydrogen on the evolution of dislocation loops was investigated using a transmission electron microscope. Specimens of reduced-activation ferritic/martensitic (RAFM) steels were irradiated with hydrogen ions to 5 × 1020 H+ • m−2 at 523–823 K, and to 1 × 1020 H+ • m−2 − 5 × 1020 H+ • m−2 at 723 K. The experimental results reveal that there is an optimum temperature for dislocation loop growth, which is ~723 K, and it is greater than the reported values for neutron irradiations. Surprisingly, the sizes of the loops produced by hydrogen ions, namely, 93 nm and 286 nm for the mean and maximum value, respectively, at the peak dose of 0.16 dpa under 723 K, are much larger than that produced by neutrons and heavy ions at the same damage level and temperature. The results indicate that hydrogen could enhance the growth of loops. Moreover, 47.3% a0 <111> and 52.7% a0 <100> loops were observed at 523 K, but a0 <111> loops disappeared and only a0 <100> loops existed above 623 K. Compared with the neutron and ion irradiations, the presence of hydrogen promoted the formation of a0 <100> loops.

Joint inhomogeneity in dissimilar friction stir welded martensitic and nanostructured ferritic steels

ABSTRACTDissimilar welding between oxide dispersion strengthened ferritic (ODS) steel and reduced activation martensitic steel would be required for constructing the advanced blanket of progressive fusion reactors. In this study, we achieved dissimilar joints by friction stir welding, and aimed to characterise and ameliorate joint inhomogeneity. Main results reveal that the joint inhomogeneity is generated from discrepant microstructural evolutions within the martensitic and ODS ferritic steels. The ODS steel achieves evolution by the dynamic recrystallisation, while the martensitic steel undergoes phase transformation that drastically hardens the stir zone. By a proper post-weld heat treatment, the joint inhomogeneity can be effectively ameliorated due to carbide reprecipitation and stress relief in the joint.

Characterization of Microstructure and Stability of Precipitation in SIMP Steel Irradiated with Energetic Fe Ions * *Supported by the Young Scientists Fund of the National Natural Science Foundation of China under Grant No 11505246, and the Major Research Plan of the National Natural Science Foundation of China under Grant No 91426301.

A type of home-made reduced activation martensitic steel, high silicon (SIMP) steel, is homogeneously irradiated with energetic Fe ions to the doses of 0.1, 0.25 and 1 displacement per atom (dpa), respectively, at 300°C and 1 dpa, at 400°C. Microstructural changes are investigated in detail by transmission electron microscopy with cross-section technique. Interstitial defects and defect clusters induced by Fe-ion irradiation are observed in all the specimens under different conditions. It is found that with increasing irradiation temperature, size of defect clusters increases while the density drops quickly. The results of element chemical mapping from the STEM images indicate that the Si element enrichment and Ta element depletion occur inside the precipitates in the matrix of SIMP steel irradiated to a dose of 1 dpa at 300°C. Correlations between the microstructure and irradiation conditions are briefly discussed.

TEM observation of dislocation loops induced by single and sequential helium/hydrogen irradiations in reduced-activation martensitic steel

Dislocation loops induced by single (He or H) ion beam and sequential (H/He or He/H) ion beam irradiation in reduced activation martensitic steels were investigated with transmission electron microscopy (TEM). Specimens were irradiated with 30keV He+ and 18keV H+ respectively under three different doses at 450°C. Dislocation loops with mean size of 41.3nm were observed under single He ion beam irradiated to 0.1dpa. Sequential ion irradiation could induce larger dislocation loops than the case of single ion irradiation. It was interesting that mean size of dislocation loops induced by He/H sequential ion irradiation was larger than that induced by H/He sequential ion irradiation at the same dose. Possible explanations were presented and synergistic effects of He and H on dislocation loops were discussed.

Synergistic effects on dislocation loops in reduced-activation martensitic steel investigated by single and sequential hydrogen/helium ion irradiation

Single-beam and sequential-beam irradiations were performed to investigate the H/He synergistic effect on dislocation loops in reduced-activation ferritic/martensitic (RAFM) steels. The irradiations were carried out with 10 keV H+, 18 keV He+ and 160 keV Ar+, alone and in combination at 723 K. He+ single-beam irradiation induced much larger dislocation loops than that induced by both H+ and Ar+ single-beam irradiation. H+ post-irradiation after He+ irradiation further increased the size of dislocation loops, whilst He+ post-irradiation or Ar+ post-irradiation following H+ irradiation only slightly increased the size of dislocation loops. The experiment results indicate that pre-implanted H+ can drastically inhibit the growth while post-implanted H+ can significantly enhance the growth of dislocation loops induced by He+ irradiation. The mechanisms behind the complex synergistic phenomena between H and He and the different roles that H and He played in the growth of dislocation loops are discussed.

Low-temperature mechanical and magnetic properties of the reduced activation martensitic steel

Mechanical and magnetic properties as well as their relationship in the reduced activation martensitic (RAM) steel were investigated in the temperature range from −90°C to 20°C. Charpy impact tests show that the ductile-to-brittle transition temperature (DBTT) of the RAM steel is about −60°C. Low-temperature tensile tests show that the yield strength, ultimate tensile strength and total elongation values increase as temperature decreases, indicating that the strength and plasticity below the DBTT are higher than those above the DBTT. The coercive field (H C) in the scale of logarithm decreases linearly with the increasing temperature and the absolute value of the slope of lnH C versus temperature above the DBTT is obviously larger than that below the DBTT, also confirmed in the T91 steel. The results indicate that the non-destructive magnetic measurement is a promising candidate method for the DBTT detection of ferromagnetic steels.

Modifications of grain-boundary structure by friction stir welding in the joint of nano-structured oxide dispersion strengthened ferritic steel and reduced activation martensitic steel

The joint of oxide dispersion strengthened (ODS) ferritic steel and martensitic steel was produced by friction stir welding. The thermo-mechanical process during welding causes recrystallization in the ODS steel and phase transformation in the martensitic steel. In stir zones of both steels, low-angle boundaries significantly convert to high-angle boundaries. The recrystallization in ODS steel is characterized by the increase in the Σ3 twin boundary with an almost constant fraction of Σ5–Σ29 boundaries, while the martensitic steel shows an opposite trend.

Synergistic Effects in Reduced-Activation Martensitic Steel Under Single and Sequential Helium/Hydrogen Ion Irradiation

Single-beam (He+ or H+) and sequential-beam (He+/H+ or H+/He+) irradiation of reduced-activation martensitic steels at 450°C was carried out to investigate helium/hydrogen synergistic effects on the microstructure of the steels. After helium implantation, a very low density of helium bubbles was observed. No void was observed after hydrogen implantation. For the He+/H+ sequential-beam irradiation, the number density of bubbles increased rapidly at low hydrogen dose, and the average size of bubbles increased at higher hydrogen dose. Helium bubbles were also found in the H+/He+ sequential-beam irradiated specimen, but its swelling was smaller than that for the He+/H+ sequential conditions. It was concluded that the effect of hydrogen atoms on the nucleation and growth of bubbles may enhance the diffusion of helium-atom/helium-vacancy clusters. Hydrogen irradiation after helium irradiation can increase the nucleation and growth of helium bubbles, while helium bubbles may enhance the retention of hydrogen atoms to form He-H-vacancy complexes.

Positron beam Doppler broadening spectra and nano-hardness study on helium and hydrogen irradiated RAFM steel

The reduced activation martensitic steel was irradiated with 250 keV He2+ to 3000 appm and 130 keV H+ to 15,000 appm alone and in sequence at 450 degrees C. The S-W curve obtained by the positron beam Doppler broadening spectra reveals that (He, H)-V complexes were induced by irradiation. The S parameters show that the H irradiation can decrease defects by forming H-V complexes. The He irradiation can form He-V complexes and the sequentially implanted H can accelerate the formation of (He, H)-V complexes. Hardening induced by He+H sequential irradiation is greater than that by He or H ion irradiation alone, suggesting synergistic effects of He and H on defects and hardening. (C) 2014 Elsevier Ltd. All rights reserved.

Transport of hydrogen and deuterium in the reduced activation martensitic steel ARAA

Advanced reduced activation alloy (ARAA) is a reduced activation ferritic/martensitic (RAFM) steel under development at the Korea Atomic Energy Research Institute. The transport of hydrogen and deuterium in ARAA was investigated in an elevated temperature range of 250–600°C. A continuous-flow method, a time-dependent gas-phase technique, was used for the measurements. Complete sets of transport parameters (permeability, diffusivity, solubility, trap site density, and trapping energy) of hydrogen and deuterium in ARAA were successfully obtained. We show that appreciable trapping effects are observed only at low temperatures (250–350°C) and that the isotope effect ratio for the diffusivity differs from the classical prediction. However, the measured values of permeability, effective diffusivity, and effective solubility of ARRA were within the range of results reported for other RAFM steels.

Damage behavior in helium-irradiated reduced-activation martensitic steels at elevated temperatures

Dislocation loops induced by helium irradiation at elevated temperatures in reduced-activation martensitic steels were investigated using transmission electron microscopy. Steels were irradiated with 100keV helium ions to 0.8dpa between 300K and 723K. At irradiation temperatures Tirr⩽573K, small defects with both Burger vectors b=1/2〈111〉 and b=〈100〉 were observed, while at Tirr⩾623K, the microstructure was dominated by large convoluted interstitial dislocation loops with b=〈100〉. Only small cavities were found in the steels irradiated at 723K.

Convoluted dislocation loops induced by helium irradiation in reduced-activation martensitic steel and their impact on mechanical properties

Helium irradiation induced dislocation loops in reduced-activation martensitic steels were investigated using transmission electron microscopy. The specimens were irradiated with 100keV helium ions to 0.8dpa at 350°C. Unexpectedly, very large dislocation loops were found, significantly larger than that induced by other types of irradiations under the same dose. Moreover, the large loops were convoluted and formed interesting flower-like shape. The large loops were determined as interstitial type. Loops with the Burgers vectors of b=〈100〉 were only observed. Furthermore, irradiation induced hardening caused by these large loops was observed using the nano-indentation technique.

Effects of Ti element on the microstructural stability of 9Cr–WVTiN reduced activation martensitic steel under ion irradiation

Microstructure of 9Cr–WVTiN reduced-activation martensitic steels with two different Ti concentrations irradiated with Fe+, He+ and H+ at 300°C was studied with transmission electron microscopy. Small dislocation loops were observed in the irradiated steels. The mean size and number density of dislocation loops decreased with the increase of Ti concentration. The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in the low Ti-concentration 9Cr–WVTiN steel.

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li2TiO3 pebbles were the breeder material, EUROFER was irradiated up to 1.3dpa at 340–580°C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5dpa. The irradiation temperature was 350°C (LIBRETTO-4/1), 550°C (LIBRETTO-4/2), and 300–500°C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.

Tritium permeation and desorption in reduced activation martensitic steels studied in EXOTIC-9/1 irradiation experiment

The EXOTIC-9/1 has been irradiated in the High Flux Reactor in Petten within the European framework for the development of the helium cooled pebble bed breeder concept. The EXOTIC-9/1 irradiation assembly consists of a pebble bed with Li2TiO3 pebbles enclosed in a EUROFER97 containment. The irradiation of EXOTIC-9/1 proceeded for 301 full power days reaching 1.3dpa in steel. Irradiation temperatures varied between 613 and 853K.Tritium permeation in a EUROFER97 containment wall was studied, in-pile, during temperature and purge gas chemistry transients. Under reference purge gas conditions (He+0.1%H2) permeation proceeded in the diffusion-limited regime. Tritium permeability in EUROFER97 was obtained from analysis of the in-pile data.During the post irradiation examination program a number of EUROFER97 containment wall segments were ramp annealed in a temperature programmed desorption setup. This study determines the amount of tritium retained in the samples and presents analysis of the desorption spectra.

Microstructural evolution of reduced-activation martensitic steel under single and sequential ion irradiations

Microstructural evolution of super-clean reduced-activation martensitic steels irradiated with single-beam (Fe+) and sequential-beam (Fe+ plus He+) at 350°C and 550°C was studied. Sequential-beam irradiation induced smaller size and larger number density of precipitates compared to single-beam irradiation at 350°C. The largest size of cavities was observed after sequential-beam irradiation at 550°C. The segregation of Cr and W and depletion of Fe in carbides were observed, and the maximum depletion of Fe and enrichment of Cr occurred under irradiation at 350°C.

Eurofer Steel, Development to Full Code Qualification

Abstract European Union fusion materials community has been developing a reduced activation martensitic steel called Eurofer for almost 2 decades. This steel has now reached maturity and is being proposed for addition to RCC-MRx. At first, its code qualification is sought for ITER Test Blanket Modules, where the maximum irradiation dose is low, 70 dpa and continuous operation are anticipated. In this paper, after a brief recall of the Eurofer steel development, types of materials and joints used during its characterization are presented. This is followed by the procedures employed for establishment of its materials properties databases and determination of its design allowable values, with emphasis on creep, fatigue and creep-fatigue properties.

Effect of Carbon Reduction on the Toughness of 9CrWVTaN Steels

Nitride-strengthened, reduced activation, martensitic steel is anticipated to have higher creep strength because of the remarkable thermal stability of nitrides. Two nitride-strengthened, reduced activation martensitic steels with different carbon contents were prepared to investigate the microstructure and mechanical property changes with decreasing carbon content. It has been found that both steels had the microstructure of full martensite with fine nitrides dispersed homogeneously in the matrix and displayed extremely high strength but poor toughness. Compared with the steel with low carbon content (0.005 pct in wt pct), the steel with high carbon content (0.012 pct in wt pct) had not only the higher strength but also the higher impact toughness and grain coarsening temperature, which was related to the carbon content. On the one hand, carbon reduction led to Ta-rich inclusions; on the other hand, the grain grew larger when normalized at high temperature because of the absence of Ta carbonitrides, which would decrease impact toughness. The complicated Al2O3 inclusions in the two steels have been revealed to be responsible for the initiated cleavage fracture by acting as the critical cracks.