Ferritic Martensitic Research Articles

Thermomechanical Processing for Improved Mechanical Properties of HT9 Steels.

Thermomechanical processing (TMP) of ferritic-martensitic (FM) steels, such as HT9 (Fe-12Cr-1MoWV) steels, involves normalizing, quenching, and tempering to create a microstructure of fine ferritic/martensitic laths with carbide precipitates. HT9 steels are used in fast reactor core components due to their high-temperature strength and resistance to irradiation damage. However, traditional TMP methods for these steels often result in performance limitations under irradiation, including embrittlement at low temperatures (<~430 °C), insufficient strength and toughness at higher temperatures (>500 °C), and void swelling after high-dose irradiation (>200 dpa). This research aimed to enhance both fracture toughness and strength at high temperatures by creating a quenched and tempered martensitic structure with ultrafine laths and precipitates through rapid quenching and unconventional tempering. Mechanical testing revealed significant variations in strength and fracture toughness depending on the processing route, particularly the tempering conditions. Tailored TMP approaches, combining rapid quenching with limited tempering, elevated strength to levels comparable to nano-oxide strengthened ferritic alloys while preserving fracture toughness. For optimal properties in high-Cr steels for future reactor applications, this study recommends a modified tempering treatment, i.e., post-quench annealing at 500 °C or 600 °C for 1 h, possibly followed by a brief tempering at a slightly higher temperature.

Comparative assessment of a continuum damage mechanics-based creep damage models for India-specific RAFM steel

ABSTRACT In the present investigation, a comparative assessment of Kachanov-Rabotnov (KR) and Liu-Murakami (LM) Creep Damage Models is presented for India-Specific Reduced Activation Ferritic Martensitic (IN-RAFM) steel, in the framework of Continuum Damage Mechanics (CDM) approach using ABAQUS finite element analysis software. The assessment is made based on the prediction of entire creep curve and creep rupture life of IN-RAFM steel at 823 K under uniaxial tensile loading at the stress levels of 200–260 MPa. Results disclosed that LM model accurately predicts the rupture life and creep deformation in comparison to KR model and the corresponding R-square values are 0.999 (LM model) and 0.993 (KR model), respectively. The material constant q of LM model controlled the damage evolution in tertiary creep stage. In contrast, the high value of constant ∅ in damage rate equation of KR model, led to over prediction of creep strain owing to high damage rate evolution.

Effects of Zr content on microstructure and properties of 9Cr ferritic/martensitic steel

The effect of the Zr content range in 0.03–0.63 wt% on the microstructure, mechanical properties and high-temperature oxidation behaviors of 9Cr ferritic/martensitic (FM) steels was investigated. The microstructural evolution of steels was studied using Thermo-Calc software, SEM, EBSD and TEM technologies. A full martensite structure was obtained in steels with various Zr contents. The small-sized M23C6 decrease and the larger-sized Zr(C, N) increase with the increasing Zr content, which leads to the increase of martensitic lath width and martensitic block size. The proper Zr content can enhance the oxidation resistance of steel, but Zr addition will reduce the mechanical properties. The decrease in strength and hardness can be ascribed to the reduction of precipitation strengthening and gain size strengthening. The increase in Zr content promotes the formation of Zr oxides, which reduces the formation of MnCr2O4 oxides and Cr2O3 oxides, and finally improves the oxidation resistance of steels.

Correlation between microstructure and hardening of 9Cr ferritic/martensitic steels irradiated in spallation neutron source

Significant hardening can be caused in ferritic/martensitic (FM) steels because of intensive displacement damage and transmutation helium/hydrogen when irradiated with high-energy protons and spallation neutrons. It is important but difficult to correlate microstructure and corresponding hardening of the FM steels irradiated in the Swiss spallation neutron source (SINQ). Based on the results of the China low activation martensitic (CLAM) steel and comparison with other 9Cr FM steels irradiated in SINQ up to 20.8 dpa/328 °C, total hardening of 9Cr FM steels was considered to be made up of helium bubbles, defect clusters/loops and α′ phase. Barrier strength of helium bubbles and clusters/loops in the CLAM specimens were obtained and investigated. Results show that the hardening of 9Cr FM specimens irradiated in SINQ increased with dose up to 20.8 dpa. After annealing at 600 °C, the hardening of FM specimens approximately showed a linear relationship with dose. Annealing made the clusters/loops in FM steels almost completely disappear under all irradiation conditions. In contrast, annealing changed the bubble size slightly and changed the bubble density to some degree. Barrier strength of helium bubbles in the CLAM specimens approximately increased linearly with size below 1.79 nm. Barrier strength of clusters/loops in the CLAM specimens also increased from 0.23 to 0.48 when their size increased from 3.5 nm to 19.1 nm. The obtained barrier strength of helium bubbles and loops in the CLAM steel is useful to explain the composition of irradiation hardening of other FM specimens irradiated in SINQ.

Effect of Impurities on the Compatibility of Steels in Supercritical CO2 at 450°–650°C

Abstract Direct-fired supercritical CO2 (sCO2) power cycles are a pathway to low-CO2 fossil energy but contain O2 and H2O in the sCO2 from combustion. The effect of impurities on structural steels was investigated at 450°–650 °C in 30 MPa sCO2. The test matrix included 9 and 12%Cr ferritic-martensitic (FM) steels and conventional and advanced austenitic steels exposed for 1000–2000 h with and without additions of 1%O2 and 0.1%H2O to simulate the cycle after water removal. For FM steels, the mass gains and scale thicknesses were similar with and without impurities with the formation of thick, duplex Fe-rich scales in all cases including the observation that Fe2O3 only formed with 1%O2. For the austenitic steels, higher mass gains were observed at all temperatures with increased formation of Fe-rich oxides when impurities were added. Carbon ingress was assessed by bulk combustion analysis, glow discharge optical emission spectroscopy (GDOES) and measuring postexposure room temperature tensile properties. Bulk C content was strongly increased at 650 °C but not at 450° or 550 °C.

Synchrotron based investigation of anisotropy and microstructure of wire arc additive manufactured Grade 91 steel

Additive manufacturing (AM) rapidly produces complex shapes crucial for energy technologies and engineering designs. In this work, the microstructure and mechanical properties of additive manufactured Grade 91 (modified 9Cr–1Mo) ferritic/martensitic (FM) steel were investigated. Computational thermodynamics, synchrotron X-ray diffraction, scanning electron microscopy, and microhardness testing were utilized to quantify microstructure (crystallographic phases, microstrain), lattice parameters, and mechanical properties as functions of the build directions and post-build heat-treatment. The as-built microstructures were found to contain a large fraction of microstrain due to two-dimensional defects, and a low-carbon martensite phase (maximum value 38.4 vol%), as quantified from the X-ray diffraction analysis. It was shown that a post-build heat-treatment can effectively remove this martensitic phase and promote the formation of desired microstructures with dispersed M23C6 carbides which was consistent with Thermo-Calc predictions. From the X-ray diffraction analysis, an ∼82.5% reduction in microstrain, a 61.5% increase in the coherent grain size and complete elimination of body-centered tetragonal (BCT) phase demonstrated that an appropriate heat-treatment of AM built FM steel yields microstructures and hardness comparable to conventionally processed steel. However, based on lattice parameter and hardness map statistics, heat-treatment on the as-manufactured steel did not eliminate anisotropy.

Strength-ductility synergy through tailoring heterostructures of hot-rolled ferritic-martensitic steels containing varying Si contents

By utilizing the feature that adding Si is able to effectively increase Ac1 and Ac3 temperatures of ferritic-martensitic (FM) steels, four FM steel sheets with different Si additions (0–1.0 wt%) were prepared by hot rolling at the same intercritical temperature to obtain various heterostructures. Their mechanical properties were evaluated and well correlated with detailed microstructural characteristics revealed by multiple characterization methods. Results show that the 0 Si specimen has a nearly complete martensitic structure; the 0.4 Si specimen consists of ∼23.5 vol% island-like ferrites embedded in martensites exhibiting a typical bi-modal structure (BMS); the 0.7 Si specimen consists of ∼46.8 vol% lamellar ferrites alternating with lamellar martensitic structures corresponding to a typical heterogeneous lamella structure (HLS); the 1.0 Si specimen is mostly comprised of ferrites no longer belonging to typical heterostructures. With increasing Si content, sizes of martensitic laths gradually decrease while the ferritic grains are coarsened. The 0.4 Si and 0.7 Si specimens exhibit high values (∼17 GPa%) of the product of ultimate tensile strength (UTS) and total elongation (TE), demonstrating a good strength-ductility synergy. Comprehensive analyses reveal that the 0.4 Si specimen with BMS exhibits a considerable hetero-deformation induced (HDI) strengthening effect, facilitating the production of the highest UTS (1476.3 MPa), while the 0.7 Si specimen with HLS shows a stronger HDI strain hardening effect leading to a higher TE (12.5%). This study could offer an important reference for designing and processing of FM steels with excellent strength-ductility synergy.

Innovative Isostatic Processing Technologies for Defence and other Strategic Applications

Powder Metallurgy (P/M) involving Cold Isostatic Pressing (CIP) and Hot Isostatic Pressing (HIP) techniques is the best suited processing route for manufacture of critical components from ceramics and highly complex metallic materials. The CIP followed by sintering route was adopted for the development of fused silica radomes for application in high-speed target seeking missiles. On the other hand, the HIP, a single-step processing technique has been used for the development of First-Wall (FW) of Test Blanket Modules (TBMs) from India Specific Reduced Activation Ferritic Martensitic (INRAFM) steel. The FW is an important component required for International Thermo-nuclear Experimental Reactor (ITER) application.The details of innovative processes which have been established to overcome the specific technical challenges for the development of CIP and HIP technologies for manufacture of silica radomes and INRAFM steel first-wall respectively for intended applications are highlighted in this paper.

Influence of thermal ageing on tensile-plastic flow and work hardening parameters of Indian reduced activated Ferritic Martensitic steel

The present study investigates the influence of thermal ageing (873K/5000h) on the microstructure, tensile-plastic flow behavior, and work hardening parameters of normalized and tempered (N&amp;T) 1.4W-0.06Ta Indian Reduced Activated Ferritic Martensitic (INRAFM) steel. To comprehensively understand the tensile-plastic flow response of the INRAFM steel across a broad temperature spectrum of 300-873 K, the Hollomon, Ludwigson and Voce equations were employed. The results reveal an augmentation in the strain hardening exponent following the ageing process, attributed to the heightened work hardening capability of the aged steel, while the strain hardening coefficient exhibited a decline post ageing. To elucidate the dislocation debris structure?s formation and movement within both the N&amp;T and thermally aged specimens, transmission electron microscopy (TEM) specimens were procured in close proximity to the tensile tested specimens at varying temperatures of 300, 573 and 873 K. The tensile plastic flow behavior at diverse temperatures was aptly described by Hollomon, Ludwigson, and Voce equations. The fitting accuracy of these equations was determined using the goodness of fit, as indicated by the lowest ?2 values. The constitutive Voce equation successfully captured the yield strength (YS) and ultimate tensile strength (UTS) with the initial stress and saturation stress acting as fitting parameters. Distinct patterns of initial stress and saturation stress variations were observed concerning both N&amp;T and aged steel in relation to temperature. Furthermore, absolute value of the Voce strain component (nv) demonstrated a reduction consequent to ageing, manifesting a two-stage behavior corresponding to temperature elevation. Notably, the deceleration of the recovery process during high temperature conditions was more pronounced in the thermally aged steel when contrasted with the N&amp;T steel. Conclusively, the Voce relation proved efficacious in predicting the yield stress (YS) and ultimate tensile strength (UTS) of both the thermally aged and N&amp;T INRAFM steel at varying temperatures.

Radiation-induced precipitates in ferritic-martensitic steels and their radiation resistance effects

Designing high-performance radiation-tolerant materials and understanding the atomistic mechanism of radiation resistance are important topics in nuclear energy structural materials research. In this study, we investigated the atomistic mechanism of the formation of rod-like precipitates in helium-irradiated ferritic-martensitic steels (F-M steels) by positron annihilation spectroscopy (PAS) and transmission electron microscopy with energy dispersive X-ray spectroscopy (TEM/EDX). The results indicated that vacancies tended to accumulate near dislocation lines to form complexes in the reduced activation ferritic-martensitic (RAFM) steel. Moreover, Fe and Cr atoms diffused toward these complexes and eventually depleted therein, which was not conducive to the formation of rod-like precipitates. Conversely, impurity atoms such as C atoms were found to segregate near dislocations in the Y-bearing oxide dispersion strengthened (Y-ODS) steel. When Fe and Cr diffused toward dislocations, they exchanged positions with C atoms until Fe and Cr exceeded their saturation solubilities and precipitated on the glide plane. We also analyzed the distribution of vacancy defects and bubbles/voids in the four irradiated materials and discussed the radiation resistance mechanism of the precipitates. The results of this study are significant in demonstrating a microscopic mechanism of radiation-induced precipitates to swelling resistance in materials. The raw/processed data required to reproduce these findings cannot be shared online at this time due to technical limitations, but they are available upon contacting the corresponding author. Materials exposed to extreme environments have received great attention recently in the context of advanced energy systems, defense, and aerospace technologies. Specifically, radiation-damage-tolerant materials are in great demand for the safe operation and advancement of nuclear energy and accelerator systems. Designing high-performance radiation-tolerant materials and understanding the atomistic mechanism of radiation resistance are important materials science topics. In this study, we reported the contribution of a microscopic mechanism of radiation-induced precipitates to swelling resistance in helium-irradiated ferritic-martensitic (F-M) steels/alloys. Significant findings included the formation of large areas of rod-like precipitates, which exhibited swelling resistance effects in steels subjected to helium irradiation and a possible microscopic mechanism for their formation. • Dense rod-like precipitates were observed in samples irradiated with high-dose helium ions. • The Cr and C contents were important factors affecting the formation of rod-like precipitates. • We clarified the atomistic mechanism of the formation of rod-like precipitates in the steels. • High-density rod-like precipitates as "radiation-damage antibodies" reduced void swelling.

APT characterization of irradiation effects on MX phase in reduced-activation ferritic/martensitic steels

Reduced activation ferritic/martensitic steels (RAFMs), such as Eurofer 97 and F82H, are designed to enhance the high-temperature strength by forming the fine dispersive particles of the MX phase. However, the stability of MX phase is still not fully understood. The objective of this study is to determine irradiation effects on the MX phase of Eurofer 97 and F82H. The irradiation experiments were performed at the Swiss Spallation Neutron Source (SINQ) at a temperature of ∼300 °C to a dose of ∼20 dpa. The current work reveals the irradiation effect on the MX phase on an atomic scale by atom probe tomography (APT). The irradiation on the one hand can affect the composition of the MX phase, on the other hand, induces interface segregation. Compared with the nitride in the unirradiated Eurofer 97, the nitride in the irradiated specimen contains less V and Ta, while more Cr, Fe, Mn, and W. The irradiation-induced segregation shows consistency among different interfaces between the MX nanoprecipitate, the matrix and the M23C6 precipitate that the undersized elements, Mn, Si, and Cu are enriched, oversized elements W, V and Ta are depleted, while the spallation transmutants Ca, Sc, and K, being largely oversized elements, enriched at the phase boundary as well. The irradiation-induced changes of the MX phase indicates that in the current irradiation condition the MX phase endures an initial transformation process to Z-phase and in the meanwhile a dissolution process at the edge. Ta in the RAFM steels plays a similar role as Nb in the conventional ferritic/martensitic (FM) steels during the MX phase to the Z-phase transformation process where Nb diffuses out of the MX phase.

Microstructure Stability and Mechanical Properties of Reduced Activated Ferritic Martensitic Steel during Thermal Aging at 550 °C for 5000 h

Microstructure Stability and Mechanical Properties of Reduced Activated Ferritic Martensitic Steel during Thermal Aging at 550 °C for 5000 h

Creep behavior of an additively manufactured 9Cr steel in the as-built condition

Limited studies have evaluated the creep behavior of additively manufactured (AM) ferritic-martensitic (FM) steels. This work investigated the creep behavior of a 9Cr FM steel fabricated by powder blown directed energy deposition (DED) technique. The creep testing at 550–650 °C and 150 MPa for the specimens along the deposition direction in the as-built condition, together with corresponding microstructural characterization, revealed a threshold temperature between 600 and 625 °C, below which the steel has creep resistance comparable with Grade 91 cross-welds and noticeably greater than 9Cr-1Mo steel. The threshold temperature distinguishes the creep behavior in two regimes differentiated in creep activation energy, creep deformation, and failure mechanism. Unlike the creep rupture surface ∼45° from the loading direction when tested above the threshold temperature, the creep rupture for testing below the threshold temperature resembles type IV failure in the cross-welds of ferritic steels. The DED-induced layer structure in the as-built steel played a significant role on the change of creep behavior.

The effect of helium on cavity swelling in dual-ion irradiated Fe and Fe-10Cr ferritic alloys

Ferritic-martensitic (FM) steels for in-vessel components in proposed fusion reactors are expected to suffer from high levels of displacement damage and helium (He) generation by neutron transmutation. However, a thorough understanding of the He synergistic effects on the cavity swelling in FM steels is still not well established. To gain fundamental insights into the He effect on cavity swelling, high purity Fe and Fe-10 wt.% Cr ferritic model alloys were irradiated with 8 MeV Ni ions and co-implanted He ions at 400 to 550 °C up to 30 displacements per atom (dpa) with He implantation rates of 0.1, 10 and 50 appm He/dpa. The current study focuses on the 50 appm He/dpa, 500 °C behavior vs. 0.1 and 10 appm He/dpa. Irradiation-induced defects, including cavities, dislocation loops, and dislocation networks were characterized using transmission electron microscopy (TEM). In the grain interior, a bimodal cavity size distribution was observed in the 10 and 50 appm He/dpa samples, but not for 0.1 appm He/dpa. Cavity swelling was maximized at intermediate He implantation rates of ∼10 appm He/dpa for both ion-irradiated Fe and Fe-10Cr alloys. The cavity swelling behavior as a function of He implantation rate appears to be controlled by the He/dpa-dependent variation of cavity sink strengths. Treating small bubbles as biased sinks for interstitial absorption can significantly increase the ratio of biased to unbiased sink strengths (Q) and results in maximized cavity swelling for a Q ratio close to one.

Radiation induced amorphization of carbides in additively manufactured and conventional ferritic-martensitic steels: In-situ experiments on extraction replicas

In-situ irradiations using 1 MeV Kr2+ ions in a transmission electron microscope were performed on extraction replica samples containing M23C6 carbides and MX carbonitrides from two ferritic-martensitic (FM) steels: (i) 9%Cr-1%W-TaV based Eurofer97 and (ii) 9%Cr-1%Mo-VNb (all in wt.%) based additively manufactured (AM) Grade91 steel. The irradiations were performed between 100 and 773 K up to a maximum dose of ∼2.4 displacements per atom (dpa). The M23C6 carbides are highly susceptible to radiation induced amorphization (RIA), while the MX type nanoprecipitates are highly amorphization resistant across the entire irradiation temperature range. Between 100 and 423 K, RIA of M23C6 carbides occurs very rapidly with critical amorphization doses ranging between ∼0.35 and 0.9 dpa, increasing to higher doses at 573 K. Complete amorphization of the M23C6 carbides up to doses of ∼2.4 dpa is not possible at 773 K under the present irradiation conditions. The critical amorphization dose of M23C6 carbides increases nearly exponentially with irradiation temperature. The critical temperature for the crystalline-to-amorphous phase transformation (Tc→a) of M23C6 carbides irradiated as replica samples, i.e. without the surrounding metal matrix, was estimated to be ∼812 K. Comparing the present results with neutron irradiation data on bulk samples reveals a decrease in the critical amorphization dose for M23C6 and an increase of Tc→a, highlighting the effect of dose rate on the amorphization behavior that is qualitatively consistent with literature on other non-metals. Changes in the minor chemistry of M23C6, such as presence or absence of W, V, Nb, Mo, seems to have little effect on the amorphization behavior. In-situ irradiations on extraction replica samples provide a novel pathway to explore radiation tolerance of nanoprecipitates in nuclear structural materials

Effect of grain boundary planes on radiation-induced segregation (RIS) at near Σ3 grain boundaries in Fe-Cr alloy under ion irradiation

Due to their good mechanical properties and corrosion resistance, Ferritic Martensitic and ODS steels are widely studied for applications in the next generation of nuclear reactor. Under irradiation, migration of point defects leads to different mechanisms including radiation-induced segregation (RIS). This type of segregation affects defect sinks such as grain boundaries (GB). Many parameters influence the amplitude of the RIS such as temperature, dose and chemical composition of the alloy. This paper presents the quantification of GB plane variation on RIS. Two different near Σ3 (60° 〈111〉) GB were investigated using EBSD and TKD for structural characterization and APT and STEM/EDXS for chemical quantification. This work revealed W-shaped profiles of Cr across GBs and a heterogeneous precipitation of Cr C rich particles in GBs planes after irradiation. Our work shows that not only the amplitude but also the sign of the Gibbsian Excess may be affected by the GB habit plane. • After irradiation segregation of Cr and C was obtained on near ∑3 in Fe-13%Cr. • Difference in orientation of habit plan has an effect on segregation profile. • Heterogeneous precipitation of Cr C rich phase was observed on the habit plan of near ∑3. • W shape profile was observed with Cr across each grain boundary after irradiation.

Core Materials for Sodium-Cooled Fast Reactors: Past to Present and Future Prospects

Abstract The Sodium-Cooled Fast Reactor (SFR) is the most promising and technologically evolved among the six nuclear reactor concepts selected by the Generation IV International Forum. Although it is compatible with a closed fuel cycle for sustainable energy production, its competitiveness depends on being able to achieve high fuel burn-up up to 200 GWd/t. This would enable efficient fuel utilization to minimize fuel cycle costs. Hence, development of optimized core materials, especially for fuel cladding tubes that are subjected to extreme conditions of intense fast-spectrum neutron irradiation, high temperatures, and mechanical and chemical fuel–cladding interactions, continues to be a priority worldwide. Resistance to void swelling, irradiation creep, and embrittlement are required to be enhanced to minimize dimensional changes and loss of ductility in core components, which limit achievable burn-up. Current SFR core materials comprise austenitic stainless steels (SS) and ferritic-martensitic (FM) steels. Cold-worked austenitic SS grades 304, 316, 316Ti, niobium-stabilized grades FV548, EI-847, EP-172, and Ti-modified 15Cr-15Ni SS, such as D9, 1.4970, 15/15Ti SS, and ChS-68, have been used for fuel cladding and ducts in SFRs built to date, and they withstand neutron irradiation damage up to 80–100 dpa. Subsequent improvements have been made by optimizing minor elements (titanium, silicon, and phosphorous) and multi-stabilization to develop advanced grades such as Si-modified 15-15Ti, Indian Fast Reactor Advanced Clad (IFAC-1), PNC-316, EK-164, and HT-UPS, which could support burn-up of 150 GWd/t. In the case of FM steels, several commercial grades are found suitable in view of their inherent void swelling resistance to 200 dpa, and are being developed for improved creep strength by oxide dispersion strengthening to realize burn-up of 200–250 GWd/t. This article presents the development of structural materials for SFR fuel pin cladding and ducts and associated enhancement of permissible fuel burn-up. Indian and international experience resulting from extensive in-pile and out-of-pile mechanical testing and fuel–cladding interactions has been covered.

Varied enhancements in mechanical properties and sodium compatibility of Grade 92 by thermomechanical treatments

As a candidate material for advanced nuclear reactors, a ferritic-martensitic (FM) steel Grade 92 in the standard normalized and tempered (NT) condition was subjected to two types of thermomechanical treatments (TMTs) with one having a lower level of deformation (T5A) and the other having a higher level of deformation (T5B). The testing results of the three conditions indicated that the TMTs significantly improved yield strength, creep resistance, fracture toughness, and sodium corrosion resistance compared with the NT condition. T5A excelled at creep resistance while T5B at sodium corrosion resistance. Microstructural characterization of the three conditions supports the varied enhancements in the properties. This study suggests that an appropriate TMT can be developed and applied to FM steels to better meet specific application requirements.

Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9

Ferritic Martensitic (F/M) steel HT9 specimens were irradiated in the Advanced Test Reactor up to 4.16 dpa in three temperature ranges (roughly from 300 to 600 °C). The post-irradiation microstructure, including dislocation structure , precipitation and radiation-induced segregation (RIS) was characterized using analytical scanning / transmission electron microscopy (S/TEM), and atom probe tomography (APT). Irradiation hardening was measured using nanoindentation . The results reveal a distinctive pattern of dislocation and precipitate evolution at high temperature, around 600 °C, where various defects and precipitates formed in the low dose regime followed by a recovering process with increasing dose. Dislocation loops formed in all temperature ranges, and the growth of dislocation loops is unconstrained above certain critical temperature, contributing to the increasing dislocation density even prior to doses of 0.5 dpa at 600 °C. Ni/Mn/Si clusters were identified in all temperature ranges and the compositions of these clusters converged to G phase stoichiometrically. Significant coarsening of G phase particles was observed at 600 °C, accompanied by the formation of G phase on grain boundaries. α ’ precipitates were only found in the medium and low temperature ranges (below 500 °C). The number density and volume fraction were higher in the low temperature specimens, while larger particles were observed in the medium temperature range. RIS of Cr, Ni, Mn, Si, P was identified at dislocation lines , grain boundaries and phase boundaries, and the temperature dependence is consistent with previous studies. The RIS of Cr to the existing VN particles was confirmed by APT and may accelerate the transition of VN to Cr-rich nitrides . The irradiation hardening contribution from dislocation loops, dislocation lines, G phase and α ’ phase was parsed based on a linear dispersed barrier hardening model. The results suggest that most irradiation hardening at high temperature is due to increasing dislocation density with dose.

A Joining Process between Beryllium and Reduced-Activation Ferritic-Martensitic Steel by Plasma Sintering.

To investigate the growth kinetics of the reaction layer and mechanical strength of joined materials, we joined beryllium and reduced-activation ferritic–martensitic steel (F82H) by plasma sintering under various conditions and characterized the joined region. Scanning electron microscopy revealed that the thickness of the reaction layer increased with an increase in the joining time and temperature. Line analyses and elemental mapping using an electron microprobe analyser showed that the reaction layer consists of Be–Fe intermetallic compounds, including Be12Fe, Be5Fe, and Be2Fe, with small amounts of chromium and tungsten. Owing to the time and temperature dependence of the reaction-layer thickness, the layer growth of Be–Fe intermetallic compounds obeys the parabolic law, and the activation energy for the reaction-layer growth was 116.2 kJ/mol. The bonding strengths of the joined materials varied inversely with the thickness of the reaction layer.