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The swelling behavior of ferritic/martensitic steels (FC92–B/-N, HT9, and Gr.92) was investigated following 3.5 MeV Fe++ ion irradiation. Tested alloys were helium-pre-implanted up to the peak contents of 120 and 240 appm with He/dpa ratio of 1 appm/dpa at room temperature and then exposed to self-ion beam to the peak damage conditions of 120 and 240 dpa at 475 °C. Field-emission transmission electron microscopy was used to characterize the cavity evolution. FC92–B exhibited the highest resistance to swelling among the irradiated alloys. The final volumetric swelling of FC92–B reached 1.3 % at 70 dpa and 2.9 % at 140 dpa. On the other hand, HT9 exhibited the highest swelling, reaching 7.4 % at 140 dpa. Comparing the present swelling results at 140 dpa/140 appm He with swelling data at 280 dpa/280 appm He from our previous study, it was observed that Gr.92 and FC92–N swelled more at 140 dpa/140 appm He than at 280 dpa/280 appm He. This negative correlation between swelling and dose in Gr.92 and FC92–N is primarily attributed to the helium-associated swelling suppression at higher helium concentration of 280 appm. A bimodal cavity size distribution appeared only in Gr.92 and FC92–N at 280 dpa/280 appm. This result demonstrates that the excess amount of helium over 200 appm promoted early-stabilization of new-born cavities, resulting in preferentially enhanced cavity nucleation, while impeding the growth of nucleated cavities. An inhibition in cavity growth possibly led to an extended duration of nucleation-dominant stages, finally suppressing swelling in ion-irradiated Gr.92 and FC92–N alloys.

This paper presents the results of studying the evolution of the structure and mechanical properties of 12Cr–2W ferritic/martensitic steel after cold rolling and additional heat treatment. X-ray diffraction analysis, transmission electron microscopy, and electron backscattering diffraction were used to study the structure. A significant increase of tensile strength up to a tensile strength of 1380 MPa was observed after cold rolling to 50% reduction and subsequent re-quenching from the austenite region. Compared to samples subjected to standard treatment, the use of this combined treatment led to the increase of impact toughness by 22 times up to 550 kJ/m2. The principles of achieving enhanced strength and toughness by reducing the grain size and increasing the fraction of carbides and the share of coincident site lattice (CSL) boundaries are discussed. [ABSTRACT FROM AUTHOR]

This article provides a comprehensive review of the advancement and material development for advanced ultra-supercritical thermal power plant technology applications. The development of these alloys is of high interest to the power generation industries. Adopting supercritical and ultra-supercritical power plants with increased steam parameters significantly improve efficiency, reducing fuel consumption and the emission of environmentally damaging gases. Materials development for the A-USC power plant with steam temperatures of 700 °C and above has been performed in order to achieve high efficiency and low CO2 emissions. With the idea of boiler components materials selection for Ultra supercritical steam parameters are in focus. The history of material advancement for thermal power generation has been summarized. Selection and design criteria for heating surfaces area in boiler components have been outlined. The relationship between creep strength at different operating temperatures has been illustrated. The literature shows insight into the material development for boiler pressure parts components and the ability to withstand the ultra-supercritical parameters during their service life. The paper also indicates microstructure stability for an alloy for long-term service life in conditions like high temperature and pressure. Recent and influential papers have been summarized, with the key findings outlined; also the type of alloy/s being used in boiler design has been headlined for ease of navigation, out of available alloys which are under development and to be ready to fabricate the boiler components.

The oxides with Al2O3‐containing and/or Y‐containing are introduced into 9Cr ferritic/martensitic steel fabricated by vacuum casting to modify its microstructures and improve its mechanical properties. The results show that the micron nonmetallic particles are refined, and the number of submicron particles is increased after the introduction of Al2O3‐containing oxides, while the number of submicron particles is decreased when both Al2O3‐and Y‐containing oxides are introduced. The average grain size, the width of martensite, and the M23C6 phases in the tempered steel are optimized when the Al2O3‐containing oxides are introduced. However, with the further introduction of Y‐containing oxides, these tempered microstructures do not change much, except for the precipitation of the δ‐Fe phase. Due to the grain refinement and second‐phase particle strengthening after the introduction of Al2O3‐containing particles, the tensile strength is improved not only at room temperature but also at 550 °C, but the impact energy drops significantly, which is caused by a large number of TiN particles promoted by Al2O3 particles through heterogeneous nucleation. At the same time, the further introduction of Y‐containing oxides reduces the tensile strength and increases the elongation at room temperature, which is attributed to the precipitation of soft δ‐Fe promoted by the addition of Y. [ABSTRACT FROM AUTHOR]

The high‐temperature oxidation resistance of 12Cr ferritic/martensitic steels with Zr contents in the range of 0–1.3607 wt% is investigated at 650 and 800 °C in air. The results show that the oxidation resistance of steels is improved by adding Zr. The oxide layer on the surface of steels after oxidation is mainly composed of MnCr2O4 and Cr2O3, where traces of Mn2O3 and ZrO2 oxides can also be detected there. The oxide layer of steels consists of two layers, that is, the outer Mn‐rich oxides (MnCr2O4, Mn2O3) and the inner Cr‐rich oxides (Cr2O3). For none‐Zr steels oxidized at 650 °C, Cr2O3 oxides are also formed in the outer layer. The addition of Zr promotes the outer oxides to change from Cr2O3 oxides to MnCr2O4 oxides and reduces the growth rate of MnCr2O4 oxides. The effect of Zr on the high‐temperature oxidation resistance of steels can be attributed to its promoting effect on the formation of outer Mn‐rich oxides, which can refine the size of outer Mn‐rich oxides and form a dense outer oxide layer. The dense outer oxide layer can inhibit the inward diffusion of oxygen and improve the oxidation resistance of steel. [ABSTRACT FROM AUTHOR]

Because of their superior radiation damage tolerance, Ferritic/Martensitic (F/M) steels are candidate materials for fuel cladding tubes in nuclear reactors. This study applies extreme gradient boosting (Xgboost) and artificial neural network (ANN) algorithms to predict the outer-diametral strains of several types of F/M steel cladding tubes when subjected to the synergistic effects of neutron irradiation and creep. Key input variables were identified using the SHapley Additive exPlanations (SHAP) algorithm and a Genetic Algorithm-based feature selection method. After deploying these machine learning (ML) algorithms, they were trained and tested across their respective hyperparameter ranges. Thereafter, the ML models' performance was assessed on an unseen/validation dataset. The outcomes revealed that the fine-tuned Xgboost model exhibited superior predictive capabilities for the validation dataset compared to the ANN model. Moreover, the trained Xgboost model surpassed the predictions made by the empirical model that couples irradiation creep with void swelling. Additionally, a more in-depth analysis of the Xgboost model was conducted to determine its ability to capture intricate relationships between input variables like irradiation dose and hoop stress and the output variable—namely, outer-diametral strain. Through synthetic experiments, it was revealed that the Xgboost model could indeed comprehend these complex interconnections effectively. Despite the challenges, such as smaller and sparser datasets with uncertainties, ML models successfully overcame the complexities and effectively predicted the outer-diametral strains in F/M steel tubes during neutron-irradiation creep. [Display omitted] [ABSTRACT FROM AUTHOR]

Dislocation-free ferritic/martensitic steel: F82H was prepared by heat-treatment in the appropriate condition in order to investigate the sink strength of grain boundaries for point defects. The size and the areal density of irradiation-induced secondary defects, such as dislocation loops and voids, were investigated in electron-irradiated F82H. The in-situ irradiation experiment resulted in the formation of a defect free zone and the gradient distribution of loop and cavity as a function of distance from boundaries, especially around prior austenite grain boundaries. The atomic density ratio of grain boundary planes is a key parameter in determining the width of defect free zone; a grain boundary with a higher atomic density ratio typically exhibits a wider defect free zone compared to one with a lower ratio. These results clearly indicate that the distribution of irradiation-induced secondary defects in ferritic/martensitic steels is not uniform and is likely controlled by the sink strength, as characterized by the atomic density ratio of grain boundary planes. • Electron irradiation experiment revealed no relationship between the defect free zone (DFZ) widths and misorientation angles of grain boundaries (GBs). • Width of DFZ would be wider at a higher irradiation temperature. • GB character was defined by the atomic density ratio (ADR) of GB planes. And the atomic density of GB plane was defined as (Number of atoms of plane1)/(Unit-area). • GB ADR>>1 showed a wider defect free zone formation compared with GB ADR∼1. • The GB ADR>>1 would be non-equilibrium and has a higher sink strength and a larger strain field compared to the quasi-equilibrium GB ADR∼1. [ABSTRACT FROM AUTHOR]

Ferritic/martensitic steel was chosen as a primary candidate structural material for China Initiative Accelerator Driven System (CiADS) which started in 2015. A new kind of tempered martensitic steel, Steel of Institute of Modern Physics (SIMP), was designed to meet the requirements of the special operating environment. Structural materials suffer from liquid Pb‐Bi corrosion and liquid metal wetting at 450°C. Liquid metal wetting can seriously affect the mechanical properties of structural materials due to the decrease in surface energy and transition from martensitic laths to ferritic grains. Creep‐to‐rupture of SIMP steel was explored in stagnant liquid Pb‐Bi eutectic at 450°C. The possible reasons for creep‐to‐rupture are discussed. The results of the present study provide a new insight into challenges related to the application of SIMP steel in CiADS. [ABSTRACT FROM AUTHOR]

Small-angle neutron scattering (SANS) has been utilized for micro-structural investigation on laboratory heats of oxide dispersion strengthened (ODS) 13.5 Cr wt % ferritic steel, with 0.3 wt% Y2O3 and with variable Ti and W contents. The results show that increasing the Ti content from 0.2 to 0.4 wt% a distribution of nano-clusters develops, tentatively identified as Y2Ti2O7, with average radii as small as 6.5 Å and volume fractions increasing from 0.021 to 0.032. The measured SANS cross-sections show also the growth of much larger defects, possibly Cr oxides. Furthermore, the ratio of magnetic to nuclear SANS components shows that the defect composition varies both with their size and with the Ti and the W content. These results are in qualitative agreement with transmission electron microscopy (TEM) observations, showing a striking influence of Ti addition on particle size refinement. However, while TEM is limited in statistics and minimum observable size of the Ti-rich nano-clusters, the defect distributions obtained by these SANS measurements provide complementary information useful for morphological characterization of the micro-structure in the investigated material.

A ferritic/martensitic steel, HT9, was irradiated in the BOR-60 reactor to ∼17.1 and ∼35.1 displacements per atom (dpa) at 650 ± 23 K (377 ± 23 °C). Irradiated samples were comprehensively characterized using analytical scanning/transmission electron microscopy and atom probe tomography, with emphasis on the role of irradiation dose on the microstructure evolution. Radiation-induced Mn/Ni/Si-rich (G-phase) and radiation-enhanced Cr-rich (αʹ) precipitates were observed within the martensitic laths at all doses. In addition, the G-phase was also observed to precipitate heterogeneously at various defect sinks. The number density for these second-phase precipitates decreases while the size increases with increasing dose, resulting in an increase of the volume fraction. Both a <100> and a /2 <111> type loops were observed with the a <100> type being the predominant type at both doses. The proportion of a <100> loops is consistent with that previously observed in HT9 ion irradiated to similar doses at ∼693–743 K (∼420-470 °C). Only small cavities (diameter < 2 nm) were observed at ∼17.1 dpa whereas both small and large cavities were observed at ∼35.1 dpa, resulting in a bi-modal cavity size distribution at this dose. Alloying elements, Ni and Si, were observed to segregate to the cavity surface, forming Ni/Si-rich shells around the cavities. The swelling at ∼17.1 dpa is evaluated at ∼0.02% while the swelling at ∼35.1 dpa is found to be ∼0.07% with variations from grain to grain. attributed to the spatial variation of the density of large cavities (in different grains). The swelling data obtained in this study was compared with the neutron data of F/M steels available in the literature. [ABSTRACT FROM AUTHOR]

• Irradiation-hardness of F82H, T91, SIMP and Fe-15%Cr based ODS steels was investigated. • Irradiation causes a significant hardness at 300℃, but not at 450℃. • The hardness change is the lowest in ODS, while is the highest in T91. To access irradiation conditions of candidate materials used in advanced nuclear energy systems, simultaneous helium and iron-ion irradiation have been widely used to study on the synergistic effects of helium bubble nucleation and displacement damage. In present study, nano-indentation was used to study the irradiation hardening of F82H, T91, SIMP and ODS steels under dual Fe3+/He+-ion irradiation. Two different irradiation temperatures (300℃ and 450℃), three different helium injection ratios (60 appm He/dpa, 200 appm He/dpa and 600 appm He/dpa, displacements per atom: dpa), the same irradiation damage (5 dpa), were chosen. The effect of helium concentration on irradiation hardening was investigated. The irradiation hardness of these four kinds of common structural materials was compared. The research results are useful for the development of irradiation hardening resistant Fe-Cr alloy steels in future. [ABSTRACT FROM AUTHOR]

Abstract This article presents a novel method combining ion irradiation, in-situ transmission electron microscopy (TEM), and microchemistry analysis before/after irradiation, which allows to examine same microstructural areas throughout ion irradiation. A 12 wt% Cr Ferritic/Martensitic steel (HT9) was irradiated in the TEM to 1.17 × 1020 ions·m−2 at 440 °C using 1 MeV Kr2+ ions, and the in-situ characterization focused on radiation-induced precipitation and segregation. Results of in-situ experiments were compared with those obtained from ex-situ experiments, to showcase how this method helps to better understand precipitation kinetics in the irradiated material examined ex-situ, for which only snapshots are available at limited doses. Graphical abstract Unlabelled Image [ABSTRACT FROM AUTHOR]

This contribution presents a comparative microstructural investigation, carried out by Small-Angle Neutron Scattering (SANS), of ferritic/martensitic steel Eurofer97 (0.12 C, 9 Cr, 0.2V, 1.08Wwt%) neutron irradiated at two different neutron sources, the HFR-Petten (SPICE experiment) and the BOR60 reactor (ARBOR experiment). The investigated “SPICE” sample had been irradiated to 16dpa at 250°C, the investigated “ARBOR” one had been irradiated to 32dpa at 330°C. The SANS measurements were carried under a 1 T magnetic field to separate nuclear and magnetic SANS components; a reference, un-irradiated Eurofer sample was also measured to evaluate as accurately as possible the genuine effect of the irradiation on the microstructure. The detected increase in the respective SANS cross-sections of these two samples under irradiation is attributed primarily to the presence of micro-voids, for neutron contrast reasons; it is quite similar in the two samples, despite the higher irradiation dose and temperature of the “ARBOR” sample with respect to the “SPICE” one. This is tentatively correlated with the higher helium content produced under HFR irradiation, playing an important role to stabilize the micro-voids under irradiation. In fact, the size distributions obtained by transformation of the SANS data yield a micro-void volume fraction of 1.3% for the “SPICE” sample and of 0.6% for the “ARBOR” one.

Abstract In this contribution, small-angle neutron scattering (SANS) studies of the micro-structural radiation damage in technical steels, developed for application in future fusion reactors, are presented. The effect of neutron irradiation at 250 °C–450 °C, up to dose levels of 16 dpa (displacement per atom), has been investigated in the European reference ferritic/martensitic steel Eurofer97 (0.12 C, 9 Cr, 0.2 V, 1.08 W wt%), both in its standard composition and mechanically alloyed with B (up to 1000 appm) to enhance the helium production, via transmutations, and reproduce fusion relevant helium/dpa ratios. The obtained SANS results suggest that in the irradiated standard Eurofer97 micro-voids distributions are present, with very small average radii (a few Å) and volume fractions (around 0.01); such low values, close to the SANS resolution limit, are consistent with the good resistance of this steel to radiation damage, for low helium concentrations. In the irradiated B-alloyed Eurofer97, more complex changes are observed: the helium bubble distributions appear strongly dependent on the irradiation temperature and on the helium concentration; furthermore, micro-structural effects possibly related to modifications in the steel matrix are detected. [ABSTRACT FROM AUTHOR]

Abstract 9Cr ferritic/martensitic steels are promising materials for in-core components in advanced Gen-IV reactors. In these applications, their long-term microstructural stability under thermal exposure and resistance to neutron irradiation are essential. Tempered (unaged) and long-term thermal aged T92 samples were used to evaluate the effects of thermal aging and ion irradiation on the microstructure and micromechanical properties of the steel. Both the tempered and aged samples were irradiated with 3 MeV Fe11+ ions to 0.25, 0.50, 1.00 and 5.00 dpa at room temperature. Using the nanoindentation technique, the irradiation hardening behaviors of T92 steel were investigated. The irradiation hardening effect was observed in both the tempered and aged T92 samples. To eliminate the soft substrate effect, the critical indentation depth was determined using the ratio of the average hardness of irradiated and unirradiated samples at the same depth. Under the same irradiation conditions, the macroscopic hardness values of the aged T92 samples after irradiation were lower than those of the tempered samples. The irradiation hardening effect was more significant in the aged T92 due to the decreased dislocation density and the coarsened martensitic lath after long-term thermal aging. [ABSTRACT FROM AUTHOR]

Ferritic/martensitic steels show exceptional void-swelling resistance during neutron irradiation compared to conventional austenitic stainless steels, such as Types 316 and 304. Explanations for the difference have been proposed based on the different crystal structures and the different microstructures of the steels. In this paper, swelling behaviour of ferritic/martensitic steels HT9 and modified 9Cr-Mo are analysed to demonstrate how the complicated tempered martensite microstructure determines their excellent swelling resistance. The variation in steady-state swelling rate observed for different heats and different heat treatments is explained in terms of how chemical composition and heat treatment affect the microstructure. Microstructures of these steels are characteristically non-uniform, containing precipitate-rich and precipitate-deficient regions. Swelling resistance increases as number and size of precipitate-deficient regions are reduced by proper heat treatment. [ABSTRACT FROM AUTHOR]

The paper presents a comparative study of different nanoindentation methods as applied to the ferritic/martensitic steels T91 and Eurofer97, here investigated in the non-irradiated reference state, but envisaged as structural materials for nuclear fission and fusion applications, respectively. Depth-controlled single cycle measurements at various indentation depths, force-controlled single cycle, force-controlled progressive multi-cycle measurements, and continuous stiffness measurements (CSM) using a Berkovich tip at room temperature have been combined to determine the indentation hardness and the elastic modulus, and to assess the robustness of the different methods in capturing the indentation size effects (ISE) of those steels. The Nix−Gao model is found inappropriate because it does not account for the breakdown of the scaling regime at small indentation depths that is linked to the extremely high density of dislocations associated with martensitic lath boundary misorientation. A generalization of the Nix–Gao model is therefore developed which allows the prediction of the dislocation densities in the lath structure in accordance with neutron diffraction results. Amplitude and frequency of the CSM oscillations influence the ISE observed. Differences of the microstructure-based parameters describing the ISE of quasi-static and dynamic measurements on T91 and Eurofer97 may reflect differences in the associated deformation mechanisms and histories. [ABSTRACT FROM AUTHOR]

Characterization of microstructures and creep fracture characteristics of dissimilar welds between Ni-based Inconel 740H superalloy and ferritic/martensitic P92 steel has been investigated. The welds were produced by shielded metal arc (SMA) welding process with the AWS A5.11 Class ENiCrFe-3 filler metal, commonly known as Inconel 182 superalloy. Postweld heat treatment (PWHT) at 760 °C for 4 h was conducted to temper the martensite in heat-affected zone (HAZ) of P92 and to form γ′ strengthener in Inconel 740H. The deposited Inconel 182 austenitic weld metal had a columnar microstructure, and grew epitaxially from the Inconel 740H base metal. The weld interface between P92 and weld metal was characterized by the discrete line with a minimal inter-diffusion, where the fresh untempered martensite was found to be formed during PWHT due to higher Ni contents diffused from Inconel 182 into P92. The P92 base metal and Inconel 182 weld metal displayed low hardness values (~240 Hv), and the sharp hardness increase was detected at coarse-grained HAZ of P92 while the hardness minimum occurred at the fine-grained HAZ (FGHAZ). Fracture location after creep was found to shift from the P92-sided fusion line to the FGHAZ of P92 with increasing creep temperature. The carbon migration from the P92 into the weld metal was not significant at lower creep temperature (600 °C), where the crack initiation at the P92-sided fusion line appeared to be related with a significant strain incompatibility across the fusion line. At both 650 °C and 700 °C where diffusion-controlled creep was more pronounced, the type IV cracking was observed. This may be attributed to the strain localization at the weak grain boundaries of FGHAZ due to the smaller grain size as well as the lack of grain boundary strengthening by carbides. [ABSTRACT FROM AUTHOR]

The microstructural effects of mixed spectrum neutron irradiation at 250 °C and 300 °C, for 2.7 dpa, 8.4 dpa, and 16.3 dpa doses, have been investigated in standard Eurofer97 (0.12 C, 9 Cr, 0.48 Mn, 0.2 V, 1.08 W, 0.14 Ta wt%) by means of small-angle neutron scattering (SANS) compared with un-irradiated Eurofer97. The observed SANS effects are attributed to the development of micro-voids, also detected by electron microscopy. The micro-voids distributions have been obtained by an improved transformation method of the SANS cross-sections providing consistent results both before and after subtraction of the un-irradiated reference. Mono-disperse micro-voids distributions are found, with average radii increasing with the dose, namely 4.4 Å for irradiation to 2.7 dpa at 300 °C, 6.6 Å for 8.4 dpa at 300 °C, and 12.9 Å for 16.4 dpa at 250 °C; the corresponding volume fractions are 0.001, 0.006, and 0.004, respectively. The differences in such distributions might reflect different damage evolution mechanisms for the different irradiation conditions, as also suggested by the comparison with a Eurofer97 sample irradiated under fast spectrum. A good resistance of Eurofer97 to micro-structural radiation damage, at least under these irradiation conditions, is suggested by the analysis of these experimental results.

Ferritic/Martensitic HT9 steel was irradiated at 432 °C to 16.6 displacements per atom (dpa) (at the depth of 600 nm) using a defocused beam of 5 MeV Fe ++ ions, while co-implanted with 3.22 appm He at the same depth. The helium concentration profile was designed so to follow the damage curve with a 0.22 appm He/dpa ratio at the depth from 300 to 1000 nm in the material. The depth-dependence of the cavity size and number density were characterized by both transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) imaging methods. A comparison between the two techniques was done showing good agreement. Cavity number density and the resulting swelling were found to be suppressed by the injected interstitial effect in the vicinity of the ion induced damage peak. The region between 300 and 750 nm depth which excludes the injected interstitial effect was thus proposed for improved cavity swelling analysis. The swelling ratio in this region was found to be ~(0.86–1.02) × 10 − 2 %/dpa. In addition, ChemiSTEM characterization revealed radiation-induced segregation occurring throughout the irradiated region and precipitation of G-phase particles. Segregation of Ni to cavity surfaces was also observed and its possible synergistic influence on swelling was discussed. [ABSTRACT FROM AUTHOR]

Creep behaviors of CNS-2 steel have been investigated under different conditions at 600–650 °C and 105–180 MPa, and the microstructures and precipitations of crept and aged specimens have been characterized. It is showed that the CNS-2 steels have remarkable creep resistance at 600°C/105MPa until at least 1500 h, and the accelerated creep occurs at the elevated temperature of 650 °C and applied stress of 180MPa. The CNS-2 steels are toughened after creep without hardening and embrittlement. Their grain microstructures are stable but the sub-structures change during the creep. The subgrains coarsen with creep strain and the density of dislocations decreases due to the recovery at elevated temperature. Three kinds of precipitations including M 23 C 6 , MX-type carbonitride and Laves phase are identified in the crept steels by TEM, EDS and SEAD analysis. It is also found that the applied stress promotes the precipitations formed at the grain or subgrain boundary. [ABSTRACT FROM AUTHOR]

A paper describes the most popular alloys for power plant application as well as the most promising alloys for future application in that technology. The components in power plants operate in severe conditions (high temperatures and pressures) and they are expected reliable service for 30 years and more. The correct choice of the material is, thus, of a very importance. The paper describes the development as well as advantages and disadvantages of convenient ferritic/martensitic steels, ferritic/bainitic steels, austenitic stainless steels and the new alloys for the application at temperatures of 650°C and more.

• Ion irradiation effects on a reduced activation ferritic/martensitic steel CLF-1 have been investigated by TEM analysis and nanoindentation tests. • It is confirmed the increase of <100>-type dislocation loops with the increasing irradiation temperature. These lattice defects induced irradiation hardening. • Lattice swelling after Fe ion irradiation at 500°C was observed, which was induced by nano-sized cavities around the dislocation loops and along dislocation lines. A kind of reduced-activation ferritic-martensitic steel, CLF-1, was developed by the Southwestern Institute of Physics, China, as a candidate structural material for fusion and fourth-generation fission reactors. This work reports the results from the nanoindentation and transmission electron microscopy study performed on this steel after irradiation with 2.75 MeV Fe11+ ions up to 1 and 10 dpa at temperatures of 300 - 500 °C. An increase in hardness was observed in the damaged layer of samples irradiated at 400 °C, while the samples irradiated at 500 °C exhibited a softening of the damaged layer. The hardening can be primarily, although not exclusively, attributed to the increase of <100>-type dislocation loops observed at the increasing irradiation temperature. Nano-sized cavities around the dislocation loops and along dislocation lines were also observed, explaining the lattice swelling after irradiation at 500°C. The research results provide a valuable reference for the development of irradiation-hardening-resistant Fe-Cr alloys for advanced nuclear energy systems. [ABSTRACT FROM AUTHOR]

Abstract: Successful development of fusion energy will require the design of high-performance structural materials that exhibit dimensional stability and good resistance to fusion neutron degradation of mechanical and physical properties. The high levels of gaseous (H, He) transmutation products associated with deuterium–tritium (D–T) fusion neutron transmutation reactions, along with displacement damage dose requirements up to 50–200displacements per atom (dpa) for a fusion demonstration reactor (DEMO), pose an extraordinary challenge. One or more intense neutron source(s) are needed to address two complementary missions: (1) scientific investigations of radiation degradation phenomena and microstructural evolution under fusion-relevant irradiation conditions (to provide the foundation for designing improved radiation resistant materials), and (2) engineering database development for design and licensing of next-step fusion energy machines such as a fusion DEMO. A wide variety of irradiation facilities have been proposed to investigate materials science phenomena and to test and qualify materials for a DEMO reactor. Some of the key technical considerations for selecting the most appropriate fusion materials irradiation source are summarized. Currently available and proposed facilities include fission reactors (including isotopic and spectral tailoring techniques to modify the rate of H and He production per dpa), dual- and triple-ion accelerator irradiation facilities that enable greatly accelerated irradiation studies with fusion-relevant H and He production rates per dpa within microscopic volumes, D–Li stripping reaction and spallation neutron sources, and plasma-based sources. The advantages and limitations of the main proposed fusion materials irradiation facility options are reviewed. Evaluation parameters include irradiation volume, potential for performing accelerated irradiation studies, capital and operating costs, similarity of neutron irradiation spectrum to fusion reactor conditions, temperature and irradiation flux stability/control, ability to perform multiple-effect tests (e.g., irradiation in the presence of a flowing coolant, or in the presence of complex applied stress fields), and technical maturity/risk of the concept. Ultimately, it is anticipated that heavy utilization of ion beam and fission neutron irradiation facilities along with sophisticated materials models, in addition to a dedicated fusion-relevant neutron irradiation facility, will be necessary to provide a comprehensive and cost-effective understanding of anticipated materials evolution in a fusion DEMO and to therefore provide a timely and robust materials database. [Copyright &y& Elsevier]

Abstract: The irradiation-induced behaviour of iron–chromium model alloys and high chromium ferritic/martensitic steels have been studied using positron annihilation measurements. Both the positron lifetime and coincidence Doppler broadening techniques were used in a complementary way. It is shown that the presence of chromium in iron-based alloys significantly reduces the concentration of vacancies to the extent that the formation of clusters is hindered. The observed vacancy behaviour as a function of the chromium content is fully consistent with the findings of void swelling suppression in chromium-rich alloys and can be rationalized based on existing models. Finally, the information obtained for the ferritic/martensitic steels shows good agreement with the results for binary model alloys. Compared with the vacancy behaviour in pure iron, this confirms the major influence of chromium on the nanostructural evolution of these steels under irradiation. [Copyright &y& Elsevier]

A paper describes the most popular alloys for power plant application as well as the most promising alloys for future application in that technology. The components in power plants operate in severe conditions (high temperatures and pressures) and they are expected reliable service for 30 years and more. The correct choice of the material is, thus, of a very importance. The paper describes the development as well as advantages and disadvantages of convenient ferritic/martensitic steels, ferritic/bainitic steels, austenitic stainless steels and the new alloys for the application at temperatures of 650 °C and more. [ABSTRACT FROM AUTHOR]

The quenched and tempered reduced-activation ferritic/martensitic steel EUROFER 97 subjected to cycling at room temperaure has been studied. Under Low-Cycle Fatigue (LCF) test this steel shows, after the first few cycles, a pronounced cylic softening accompanied by microstructural changes such as the decrease of the dislocation density inside the subgrain. During LCF tests, the softening seems to be governed by a mechanism independent of the plastic strain range imposed to the specimen. From the analysis of the peak tensile stress of the hysteresis loops and its respective correlation with the transmission electron microscopy observations can be concluded that the cyclic softening observed at room temperature could be attriuted to the progressive annihilation of dislocactions located in the interior of the subgrains. [ABSTRACT FROM AUTHOR]

Design concepts for the next generation of nuclear power reactors include water-cooled, gas-cooled, and liquid-metal-cooled reactors. Reactor conditions for several designs offer challenges for engineers and designers concerning which structural and cladding materials to use. Depending on operating conditions, some designs favor elevated-temperature ferritic/martensitic steels for in-core and out-of core applications. Such steels have been investigated in previous work on international fast reactor and fusion reactor research programs. Steels from these fission and fusion programs will provide reference materials for future fission applications. In addition, new elevated-temperature steels have been developed in recent years for conventional power systems that also need to be considered. [ABSTRACT FROM AUTHOR]

Abstract: High radiation resistant structural materials for fusion and fission nuclear power plants are a key issue for the development of both types of reactors. Selection criteria, elements of metallurgy of the selected materials, and the major issues as they are revealed by the results of the present development programmes, are presented. At low temperature (∼300 °C) ferritic/martensitic steels are suffering from He-embrittlement, associated with possible hardening due to unmixing. The kinetics of hardening and embrittlement versus dose, especially saturation with dose, are still open key issues, difficult to settle on the basis of a purely experimental programme. Important progress is still to be made in mastering the initial microstructure, inclusion cleanness and joining techniques of oxide dispersion strengthened steels for higher heat resistance. Physics modeling as presented in this issue should promote guidance to the understanding of the mechanisms involved, provide solutions to master the initial microstructure and phase stability, and mitigate the in-service property degradation. To cite this article: J.-L. Boutard et al., C. R. Physique 9 (2008). [Copyright &y& Elsevier]

Abstract: Creep rupture strength of tempered martensitic steel Eurofer’97 has been investigated. Different products form (plate and bar) have been tested in the temperature range from 450°C to 650°C at different loads. No significant differences in the creep rupture properties have been found between the studied product forms. The Eurofer’97 has shown adequate creep rupture strength levels at short creep rupture tests, similar to those of the F-82H mod. steel. However, for long testing times (>9000h) the results available up to now at 500°C and 550°C seem to indicate a change in the creep degradation mechanism. [Copyright &y& Elsevier]

In the 1970s, high chromium (9–12%Cr) ferritic/martensitic steels became candidates for elevated temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, a nominally Fe–12Cr–1Mo–0·5W–0·5Ni–0·25V–0·2C steel (composition in wt-%), were considered in the USA, Europe and Japan. Now, a new generation of fission reactors is in the planning stage, and ferritic, bainitic and martensitic steels are again candidates for in-core and out-of-core applications. Since the 1970s, advances have been made in developing steels with 2–12%Cr for conventional power plants that are significant improvements over steels originally considered. The present study will review the development of the new steels to illustrate the advantages they offer for the new reactor concepts. Elevated temperature mechanical properties will be emphasised. Effects of alloying additions on long-time thermal exposure with and without stress (creep) will be examined. Information on neutron radiation effects will be discussed as it applies to ferritic and martensitic steels. [ABSTRACT FROM AUTHOR]

Reduced activation ferritic/martensitic steels are leading candidates for blanket/first-wall structures of the D-T fusion reactors. In fusion application, structural materials will suffer cyclic stresses caused by repeated changes of temperature and electromagnetic forces according to reactor operation scenarios. Therefore, creep–fatigue behaviors are extremely important to qualify reduced activation steels as fusion structural materials. In this work, microstructural stability of reduced activation ferritic/martensitic steels under various external stresses, such as constant stress cyclic stress, was studied. The materials used are JLF-1 steel (9Cr–2W–V,Ta) and JLS-2 steel (9Cr–3W–V,Ta). The microstructure inspection by means of transmission electron microscopy (TEM) and scanning electron microscopy (SEM) was performed following creep rupture tests, fatigue and creep–fatigue tests at elevated temperatures. In order to examine precipitation morphology in detail, the improved extracted residue and extracted replica methods were applied. From the microstructural observation of creep rupture-tested specimen, intergranular precipitates such as M23C6 and Laves phase coarsened by applying the static stress. [Copyright &y& Elsevier]

Ferritic/martensitic (F/M) steels whose matrix is Fe-Cr are important candidate materials for fuel cladding of fast reactors, and they have excellent irradiation-swelling resistance. However, the mechanism of irradiation-swelling of F/M steels is still unclear. We use a first-principles method to reveal the influence of irradiation defects, i.e., Frenkel pair including atomic vacancy and self-interstitial atom, on the change of lattice volume of Fe-13Cr lattice. It is found that vacancy causes lattice contraction, while a self-interstitial atom causes lattice expansion. The overall effect of a Frenkel pair on the change of lattice volume is lattice expansion, leading to swelling of the alloy. Furthermore, the diffusion properties of point defects in Fe-13Cr are investigated. Based on the diffusion barriers of the vacancies and interstitial atoms, we find that the defects in Fe-13Cr drain out to surfaces/grain boundaries more efficiently than those in pure α-Fe do. Therefore, the faster diffusion of defects in Fe-13Cr is one of important factors for good swelling resistance of Fe-13Cr compared to pure α-Fe. [ABSTRACT FROM AUTHOR]

• Void swelling was studied in self-ion irradiated HT.9, Gr.92, FC92-B, and FC92-N. • FC92-N showed the greatest swelling resistance of 1.76% at 240 dpa and 2.17% 318 dpa. • M 2 X evolved only in FC92 series, in which bimodal swelling profiles were observed. • RIP and outward Cr sinking to the surface synergistically formed a low-alloyed zone. • RIP modified swelling-depth profiles, and in turn, determined double peak swelling. The radiation responses of two newly developed ferritic/martensitic steels, FC92-B and -N, were tested in comparison to reference alloys HT9 and Gr.92. Ion irradiations on the steels were performed up to 480 dpa at 475°C using 3.5-MeV Fe++ ions with a helium pre-implantation of 1 appm/dpa. Void swelling and M 2 X precipitation were characterized using FE-TEM and EDS. Swelling resistance was the greatest in FC92-N, which showed suppressed void nucleation and growth. The swelling rate in FC92-N was determined as 0.005 %/dpa, indicating that FC92-N did not reach the steady-state swelling regime with void nucleation behavior. The least swelling-resistant alloy was HT9 with a swelling rate of 0.048 %/dpa. Cr-rich carbide, M 2 X, was observed in only 9Cr-FC92 series; however, its formation did not depend on radiation damage. This exceptional M 2 X evolution in FC92 series may be attributed to B and N alloying, which resulted in suppressed M 23 C 6 carbide formation during metallurgical production and sequentially high C contents in the alloy solution of FC92 series. A narrower range (800 nm) of M 2 X evolution compared to that of cavity formation (1,000 nm) indicates that radiation-induced precipitation (RIP) is more sensitive to the injected interstitial effect. Precipitation-induced Cr depletion and preferential interstitial outward sinking to the free surface synergistically modified local chemical composition before void evolution and led to double-peak swelling by locally forming a low-alloyed zone. This study provides the first experimental evidence that RIP modifies the swelling–depth profiles and in turn, determines double-peak swelling in ion-irradiated steels. [Display omitted] [ABSTRACT FROM AUTHOR]

• Small punch tests were conducted on irradiated Optifer-IX and Eurofer 97 ferritic/martensitic steels and MA956 ODS ferritic steel. • The ductile-to-brittle transition temperatures of irradiated steels were evaluated with small punch tests. • The irradiation-induced embrittlement effect in ferritic/martensitic steels is enhanced by helium at high concentration. In order to analyze the irradiation-induced changes of ductile-to-brittle transition temperature (DBTT) of ferritic/martensitic (FM) steels, small punch (SP) tests were conducted on the specimens of Optifer-IX, Eurofer 97 and MA956 steels after irradiation in a target of the Swiss spallation neutron source (SINQ) at doses between 8.5 and 18.8 dpa and temperatures between 165 and 365°C. The SP test results demonstrate that the DBTT of the three steels increases with irradiation dose, which is in agreement with the previous results of SP and Charpy impact tests on FM steels irradiated at SINQ. Both SP and SEM observation results indicate that, among the three steels, Eurofer 97 possesses the highest resistance to irradiation-induced embrittlement. On the other hand, the ferritic ODS steel MA956 exhibits the poorest radiation resistance, manifested by the higher DBTT in both unirradiated and irradiated conditions and higher DBTT shift after irradiation as compared to Optifer-IX and Eurofer 97 steels. [ABSTRACT FROM AUTHOR]

Ferritic/martensitic (F/M) steels have been proposed as important candidates for structural materials of nuclear reactors due to their good mechanical properties and radiation resistance. Compared with traditional normalizing and tempering (N&T) or quenching and tempering (Q&T) processes, recent studies have shown that a quenching and partitioning (Q&P) process can significantly improve the strength of steels due to the multi-phase microstructure. In this study, we developed a Q&P process that significantly improved the ultimate tensile strength/yield strength of a modified 9Cr–2WVTa steel from ~752 MPa/~644 MPa to ~1210 MPa/~927 MPa at room temperature, maintaining a total elongation of ~21.4%. Additionally, the strength of the steel at elevated temperatures (25–600 °C) was significantly improved by the Q&P process. The strengthening effect is attributed to the combined effects of the different microstructure characteristics, i.e., more solute carbon atoms within the matrix, higher dislocation density in the martensite, refined structural units (martensite packets/ultra-fine martensite laths) and finer precipitates (M 23 C 6 , MX, and θ-carbide). The relationship between the microstructures resulted from the two processes and the corresponding mechanical properties has also been comparatively discussed, which provides guidelines for tailoring the microstructure and mechanical properties of F/M steels by a Q&P process. [ABSTRACT FROM AUTHOR]