Reduced Activation Ferritic/martensitic Research Articles

Calibration of RAFM Micromechanical Model for Creep Using Bayesian Optimization for Functional Output

Abstract A Bayesian optimization procedure is presented for calibrating a multimechanism micromechanical model for creep to experimental data of F82H steel. Reduced activation ferritic martensitic (RAFM) steels based on Fe(8–9)%Cr are the most promising candidates for some fusion reactor structures. Although there are indications that RAFM steel could be viable for fusion applications at temperatures up to 600∘C, the maximum operating temperature will be determined by the creep properties of the structural material and the breeder material compatibility with the structural material. Due to the relative paucity of available creep data on F82H steel compared to other alloys such as Grade 91 steel, micromechanical models are sought for simulating creep based on relevant deformation mechanisms. As a point of departure, this work recalibrates a model form that was previously proposed for Grade 91 steel to match creep curves for F82H steel. Due to the large number of parameters (9) and cost of the nonlinear simulations, an automated approach for tuning the parameters is pursued using a recently developed Bayesian optimization for functional output (BOFO) framework (Huang et al., 2021, “Bayesian optimization of functional output in inverse problems,” Optim. Eng., 22, pp. 2553–2574). Incorporating extensions such as batch sequencing and weighted experimental load cases into BOFO, a reasonably small error between experimental and simulated creep curves at two load levels is achieved in a reasonable number of iterations. Validation with an additional creep curve provides confidence in the fitted parameters obtained from the automated calibration procedure to describe the creep behavior of F82H steel.

Assessment of the impact behavior of CLF-1 steel with Charpy V-notch testing and miniature Charpy V-notch testing

China Low-activation Ferrite steel (CLF-1), as one of the structural materials developed in China for fusion reactors, is a Reduced activation Ferritic/Martensitic (RAFM) steel modeled after the traditional T91 steel (9Cr-1Mo-0.2V-0.08Nb). For fusion reactor materials, the Charpy impact testing is an important method to assess their notch sensitivity. In this study, Charpy impact tests were conducted on CLF-1 steel using Charpy V-notch specimens (CVN) and Miniature Charpy V-notch specimens (Kleinstprobe, KLST) to assess the impact behavior of CLF-1 steel, and data normalization was performed on the KLST specimen. The results show that as the specimen size of CLF-1 steel decreased, the curves of impact absorbed energy, lateral expansion, and shear fracture appearance shifted towards lower temperatures, and the impact absorbed energy significantly decreased. Both the macroscopic and microscopic characteristics of the specimens obtained from the experiments indicated that KLST specimens could achieve the same fracture characteristics as CVN specimens. Additionally, at the same experimental temperature, KLST specimens exhibited a higher proportion of ductile fracture regions.

Effects of Carbon Deposition on Deuterium Plasma-Driven Permeation Through Tungsten-Coated RAFM Steel

In steady-state operation of fusion reactors, eroded materials and contaminations, especially carbon (C), may deposit on the surface of plasma-facing components. In this work, the effects of C deposition on hydrogen isotope permeation behavior through tungsten (W)–coated reduced activation ferritic/martensitic (RAFM) steel were systematically investigated by plasma-driven permeation (PDP) measurements in the temperature range of 633 to 893 K. A C deposition layer with thickness of ~200 nm was prepared by magnetron sputtering to simulate the formation of C impurities in the first-wall area of tokamaks. The implantation depth of incident deuterium (D) ions was estimated to be <10 nm at incident energy of 114 eV. Deuterium effective diffusion coefficients (Deff’s) for W-coated RAFM steel with/without a C layer were obtained. It was found that the C layer tended to increase Deff in the low-temperature region of ~675 to 820 K. At high temperature, however, Deff was measured be lower than that without a C layer. Nevertheless, the addition of a C layer had no significant effect on Deff compared to the W coating alone with respect to bare RAFM steels. For steady-state D-PDP flux, it was found that the C layer significantly decreased D permeation flux at low temperature. But, the permeation flux difference between the samples with/without a C layer became smaller with increasing temperature, indicating that the influence of C deposition on D permeation was negligible at high temperature. Similar D-PDP behavior was detected as increasing the incident ion flux by means of increasing plasma discharge power. Surface reemission of absorbed D as well as the D concentration gradient throughout the sample was found to be influenced by C deposition; therefore, D permeation flux changed correspondingly.

Pre-conceptual design and proof of principle assessments of self-cooled Toroidally symmetric lead-lithium (TSLL) blanket

A new self-cooled liquid metal blanket concept called TSLL (Toroidally Symmetric Lead-Lithium) blanket is proposed and assessed, including analysis for magnetohydrodynamic (MHD) flows, structural analysis, and heat transfer and neutronics assessments using the ARC reactor with demountable magnets designed by the Commonwealth Fusion Systems (CFS) as a testbed. The proposed blanket utilizes lead-lithium (PbLi) alloy as breeder/coolant and reduced activation ferritic/martensitic (RAFM) steel as structural material. A special feature of the new concept is the toroidally symmetric flow in the blanket integrated first wall and the breeding zone to reduce the MHD pressure drop, while using anchor links to strengthen the first wall construction. Provided analysis suggests acceptable MHD pressure drop, required mechanical integrity and high tritium breeding ratio of ∼ 1.64. As a result of these assessments, the new blanket concept can be recommended for more detailed studies as a promising blanket candidate for implementation in future fusion devices.

Simulations of radiation damage accumulation in Fe-9Cr under pulsed irradiation conditions representative of inertial fusion energy

Structural materials for laser-based inertial fusion energy (IFE) reactor concepts are expected to operate under pulsed irradiation conditions, with cycles consisting of microsecond-long neutron bursts followed by inter-pulse periods of up to one second in duration. During each laser shot, irradiation damage is introduced at dose rates that are up to six orders of magnitude higher than those in their magnetic fusion energy (MFE) counterparts. Under certain conditions, the inter-pulse periods may last an amount of time sufficient to anneal much of the damage introduced during each shot. This phenomenon is highly temperature dependent, with pulsed heating directly linked to the pulsed damage, large surface temperature spikes may also occur. As such, this intermittent mode of operation has the potential to lead to fundamental differences in how irradiation damage accumulates in structural reactor materials. However, damage to structural materials under IFE conditions has received comparatively much less attention than in MFE, as pulsed conditions add yet an extra dimension to the already extremely challenging problem of microstructural evolution under fusion neutron irradiation in structural materials. In this work we use the stochastic cluster dynamics (SCD) method to simulate the evolution with time of defect cluster concentrations under IFE conditions. We consider the Laser Inertial Fusion Energy (LIFE) reactor concept as the representative IFE design for our study, for which detailed spectral information is available, including gas transmutant production. We simulate several pulse frequencies and three different temperatures, and compare the results with continuous irradiation cases under identical average dose rates. The simulations are run in Fe-9Cr system as a model alloy for reduced-activation ferritic/martensitic (RAFM) steels, which are the leading structural material candidates for first-wall structures in MFE and IFE devices. We find that, in practically all scenarios, pulsed irradiation restricts the formation of helium-vacancy clusters relative to the levels seen under equivalent steady irradiation conditions. As well, although self-interstitial atom clusters do accumulate under pulsed operation, their number densities remain up to an order of magnitude lower than in continuous irradiation conditions. Based on the SCD results, we provide a temperature-pulse rate map to identify regions where pulsed irradiation may lead to larger defect accumulation than under continuous irradiation.

Tensile and creep properties of small specimens of reduced-activation ferritic steel F82H, and the correlation to standard specimen data

Background A reduced-activation ferritic/martensitic (RAFM) steel, F82H steel, is the primary candidate structural material for fusion reactor blanket. Small specimen test technique is essential to develop the blanket materials using limited irradiation volume in high flux neutron field. An international collaboration activity “Towards the Standardization of Small Specimen Test Techniques for Fusion Applications” has been initiated under the framework of the International Atomic Energy Agency Coordinated Research Project for Phase I from 2017 to 2021, and Phase II from 2022 to 2026. The present paper reports the preliminary results on tensile and creep tests as a summary of the above Phase I activity. Methods Tensile and creep tests were conducted at 550 and 650°C, using flat-plate SSJ type small specimens with various gauge thickness ranged from 0.14 to 1.2 mm, while gauge length and width are 5 and 1.2 mm, respectively. In addition, round bar type standard specimens with a gauge geometry of 6 mm in diameter and 30 mm in length was also tested for comparison. Results Tensile yield stress, ultimate tensile strength and uniform elongation were independent of the gauge thickness of SSJ specimens, and agreed with the data from the standard size specimens. On the other hand, total elongation was decreased with decreasing the thickness. In creep tests, rupture time was decreased with decreasing the gauge thickness of SSJ specimens. Standard size specimens exhibited shorter rupture time than the SSJ specimens. Conclusions The SSJ type specimens provided similar tensile parameters to those from the standard specimen, except total elongation. Creep rupture time of the SSJ specimens were different from the standard specimen, and decreased with decreasing the gauge thickness.

Numerical study and experimental validation of heat transfer capacity of flat-type divertor for fusion reactor

The divertor must simultaneously withstand unprecedented high heat fluxes of up to 20 MW/m2 and high-energy neutron irradiation of 14-MeV during fusion reactor operation. Accordingly, it is necessary to simultaneously meet the excellent heat dissipation capacity of the divertor and maintain good material performance in harsh neutron irradiation environments. The flat-type divertor demonstrates better heat transfer performance compared to monoblock divertor. Nevertheless, under the condition of a heat flux of 20 MW/m2, the heat transfer and thermal fatigue performance of flat-type divertor using advanced materials are still unidentified. In this study, we conducted a thorough analysis of the heat transfer capabilities and thermal fatigue characteristics of potassium-doped tungsten (KW)/Cu/Oxide dispersion strengthened copper (ODS-Cu)/reduced activation ferritic/martensitic (RAFM) divertor mockup adopts hypervaportron (HV) structure by numerical simulations combined with experiments. The numerical results indicate that the flat-type KW/Cu/ODS-Cu/RAFM divertor mockup expresses excellent heat transfer capacity. The contact area between the edge of the fins and the bottom surface of the ODS-Cu heat sink induces a significant increase in water flow velocity to ∼15 m/s. The peak temperature of loaded KW surface is only ∼985 °C under the flow rate of 5 t/h and inlet temperature of 20 °C. During the high heat flux tests, the prepared flat-type divertor mockup successful endured 1000 cycles of 20 MW/m2 with the peak temperature of 883 °C, and the surface temperature experienced a fluctuation of 2.4 % during the thermal fatigue tests. This study can provide a strong data reference and technical support for the development of fusion reactors, and is of great significance in advancing the commercialization of fusion energy.

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.

Influence of Oxide Inclusions on the Creep Properties of Electron Beam Welds in Ti-added Reduced Activation Ferritic/Martensitic Steel

In this study, the effects of reducing oxide inclusions on the microstructure, tensile properties at 550 ℃, and creep resistance of electron beam welded (EBW) joints in Ti-added reduced activation ferritic/martensitic (RAFM) steel were investigated. Two RAFM steels, designated as Steel A and Steel B, were prepared by adjusting the Al and Si contents to contain different fractions of oxide inclusions. The microstructures of the base metal and heat-affected zone (HAZ) were analyzed, and the mechanical properties of the welds were evaluated. Reducing the Al and Si content in Steel B resulted in a decrease in both the fraction and size of oxide inclusions compared to Steel A. Both steels exhibited tempered martensitic microstructures with M23C6 and MX precipitates. Hardness tests revealed that the over-tempered HAZ (OTHAZ) was the weakest region. However, both tensile test and creep tests at 550 ℃ showed fractures occurring in the base metal, indicating that the low heat input of EBW effectively minimized the weak HAZ regions. Steel B demonstrated improved tensile properties and creep resistance due to the reduced oxide inclusions. It was found that Steel B had a significantly longer creep life than Steel A at 550 ℃. SEM analysis revealed ductile fracture characterized by dimples, where oxide inclusions containing Al, Si, Ta, and Ti were predominant. The presence of these inclusions reduced the effectiveness of solid solution and precipitation strengthening by consuming Ta and Ti. Additionally, microvoids could easily nucleate at the interface between the hard oxide inclusions and the soft metal matrix during creep. Therefore, the reduction of oxide inclusions improved deformation resistance and high-temperature stability, resulting in better creep resistance in Steel B. In conclusion, the reduction of oxide inclusions in Ti-added RAFM steel enhances the high-temperature mechanical properties and creep resistance of EBW joints, underscoring the importance of oxide control in the development of RAFM steels for fusion reactor applications.

Preparation of high-density and excellent bending strength pure tungsten target by hot oscillatory pressing sintering and its magnetron sputtering coating

The high-strength, high-density pure tungsten (W) targets were fabricated using hot oscillatory pressing (HOP) sintering. This study examined the influence of sintering temperature, oscillation frequency, amplitude, and median pressure on the relative density, grain size, and bending strength of the tungsten targets. A tungsten target exhibiting a relative density of 99.51%, a grain size of 2.95 μm, a bending strength of 1200.1 MPa, and a purity of 99.95% was synthesized at a sintering temperature of 1600 °C, an oscillation pressure of 60 ± 15 MPa, and a frequency of 1 Hz. Grain rearrangement and plastic deformation (dislocation movement and grain boundary slip) were identified as the primary mechanisms for grain refinement and densification, respectively. The key strengthening mechanism was sub-grain boundary strengthening. The optimal tungsten target was utilized for magnetron sputtering on a reduced activation ferritic/martensitic (RAFM) steel substrate. The effects of sputtering time, power, deposition pressure, and substrate temperature on the crystallinity, roughness, thickness, and bonding force of the tungsten films were investigated. The optimal magnetron sputtering process was achieved by sputtering for 45 min at a power of 60 W, a deposition pressure of 15 Pa, and a substrate temperature of 100 °C. This resulted in dense W films with excellent comprehensive properties, including a roughness of Ra = 8.73 nm, Rq = 3.34 nm, a thickness of 999.05 nm, and a bonding force of 2.57 N.

Deuterium permeation performance of V4Cr4Ti/RAFM steel joint with in-situ formed carbide interlayer during diffusion bonding

Vanadium alloy (V4Cr4Ti) is a structural material used in future fusion reactors owing to its inherent advantages of high-temperature strength and neutron irradiation resistance. Oxygen pickup and high hydrogen isotope permeability are the two major concerns for their application in various types of blankets. To solve these issues, a bimetallic joint was designed with a CLF-1 layer of reduced-activation ferritic/martensitic (RAFM) steel to resist corrosion and a CLF-1/V4Cr4Ti bonding interface to suppress hydrogen isotope permeation. Samples were prepared from joints bonded via solid-state diffusion under hot isostatic pressing (HIP). Deuterium (D) gas permeation tests were performed at various temperatures in the range 400–600 °C at 100 kPa. After deuterium permeation tests, the thermal desorption spectra (TDS) of the samples were analyzed. And the permeation behavior of deuterium atom across the joints was schematically discussed. The results showed that bonding with steel could be an effective solution for resisting hydrogen isotope permeation through the V4Cr4Ti alloy. Thermal desorption spectrum (TDS) analysis of the tested samples revealed a desorption peak temperature equivalent to that of TiC. The formation of a 1 µm-wide B2 ordered α'- [Fe, (V, Ti)] phase layer, together with a large number of dispersed carbides around the interface, played a significant role in the deuterium permeation resistance. In addition, microstructural changes in the steel base material after the 900 °C bonding process were beneficial for suppressing deuterium diffusion.

Irradiation Effects on Tensile Properties of Reduced Activation Ferritic/Martensitic Steel: A Micromechanical-Damage-Model-Based Numerical Investigation

The tensile properties of reduced activation ferritic/martensitic (RAFM) steels are significantly influenced by neutron irradiation. Here, a mechanism-based model taking account of the typical ductile damage process of void nucleation, growth, and coalescence was used to study the temperature and irradiation effects. The elastic–plastic response of RAFM steels irradiated up to 20 dpa was investigated by applying the GTN model coupled with different work hardening models. Through a numerical study of tensile curves, the GTN parameters were identified reasonably and satisfying simulation results were obtained. A combination of Swift law and Voce law was used to define the flow behavior of irradiated RAFM steels. The deformation localization could be adjusted effectively via setting the nucleation parameter εn close to the strain where necking occurs. Because εn changed with uniform elongation, εn decreased with the testing temperature and rose with an irradiation temperature above 300 °C. The nucleation parameter fn increased with the testing temperature for RAFM steels before irradiation. For irradiated RAFM steels, fn barely changed when the irradiation temperature was below 300 °C and then it rose at a higher irradiation temperature. Meanwhile, the ultimate strength of the simulated and experimental curves showed good agreement, indicating that this method can be applied to engineering design.

Hot compression bonding of a 9Cr oxide dispersion strengthened alloy and a 9Cr reduced-activation ferritic/martensitic alloy

An innovative method of hot compression bonding is proposed in this work for the joining of 9Cr oxide dispersion strengthened (ODS) alloy and 9Cr reduced-activation ferritic/martensitic (RAFM) alloy. The microstructural evolution of the bonding interface was investigated by scanning electron microscopy (SEM), electron back-scattered diffraction (EBSD), and transmission electron microscopy (TEM). The results verify that the pinning effect of nano-oxides particles (NPs) in 9Cr ODS alloy significantly enhances its dynamic recrystallization (DRX) temperature and deformation resistance. Continuous DRX (CDRX) first occurred on the 9Cr RAFM alloy side, and the areas near the bonding interface were composed of recrystallized grains. With increasing strain, CDRX also showed up on the 9Cr ODS alloy side. Inevitable slight oxidation occurred at the bonding interface during the hot compression bonding (HCB) process, and the interfacial oxides transformed from initial coarse CrO to TiO and finally to Y-Ti-O nanoparticles with sizes comparable to pre-existing NPs dispersed in the 9Cr ODS alloy matrix. It is believed that interfacial oxide transformation and grain structure consistency contributed to the excellent interface healing of the two dissimilar alloy pieces. The effectiveness of the bonding was tested by tensile tests and fractography analysis, revealing that ideal metallurgical bonding could be achieved under a controlled strain level of 10 % at 800 °C followed by soaking at 1000 °C for 4 h.

Low cycle fatigue of EUROFER97 investigated for test blanket modules of ITER: An interlaboratory study

Several reduced-activation ferritic-martensitic (RAFM) steel grades have been proposed for structural applications in fusion reactors since the 1990s by several countries. The European reference RAFM steel EUROFER97 has been produced in four batches since 1998. It is intended for fabrication of Test Blanket Modules (TBM) for ITER as well as the blanket first wall and divertor cassettes of DEMO.RCC-MRx is the design code developed for high-temperature, research and fusion reactors by AFCEN. Its latest edition contains a provisional section dedicated to EUROFER97, aggregating properties of the first two batches, whereas a full qualification and final inclusion of a material requires three batches minimum. The completion of the code is required for the design assessment of TBM units in the frame of the nuclear licensing process.The EUROfusion project coordinates efforts to broaden the knowledge of EUROFER97 properties, which will be used for closing the database gaps in RCC-MRx. Low cycle fatigue (LCF) tests are a significant part of the EUROfusion test programme, since the material fatigue properties are indispensable for the component assessment owing to the cyclic nature of a tokamak operation.In this work we report the results of interlaboratory LCF tests on EUROFER97 batch 3 performed in a temperature range relevant for its application in ITER (RT–550 °C), wide selection of total strain ranges (0.3–1.5 %) and two values of strain ratio Rε (0 and –1). Along with the estimated fatigue life and cyclic behaviour, a fractographic analysis done with scanning electron microscopy (SEM) is presented.

Characteristics of oxide-dispersion strengthened alloys produced by high-temperature severe deformation

This study is to explore an economically attractive and technically feasible processing method for oxide-nanoparticle strengthened alloys for fusion reactor application. Despite many scientific merits of the advanced oxide-dispersion strengthened (ODS) alloys, such as the nanostructured ferritic alloy (NFA) 14YWT, the only viable production path for a high-quality NFA is the high-power mechanical alloying process. This process is often a multi-day high-speed ball milling of alloy powder with a small quantity of yttria (Y2O3) powder, followed by the milled-powder consolidation using extrusion or other methods and additional thermomechanical processing (TMP) for property control. This complex production path, including the low-temperature mechanical alloying in particular, has limited technical advancement toward the cost-effective and scale-up production of ODS alloy components. To overcome such a practical limitation, we proposed to explore alternative low-cost processing routes using traditional thermomechanical processing (TMP) method only. A series of continuous TMP cycles, which were designed to impose high-temperature severe plastic deformation (HT-SPD) conditions to the consolidated powder mixtures, were applied to achieve the effective distribution of oxide particles in nanograin structure and thus desirable mechanical properties. Since the reduced-activation ferritic-martensitic (RAFM) alloy powders (Fe-10Cr and Fe-14Cr alloys with various Y contents) are available in our inventory, we focused to utilize the new solid-state synthesis approach for controlling oxide (oxygen source) dissolution and nanoscale clustering in nanograin structure in those alloys. A combination of powder consolidation at 900 °C and continuous thermomechanical activation at 600 °C yielded two essential ODS alloy microstructure contents–nanograin structure and nanoparticle distribution–and thus demonstrated a good combination of strength and ductility.

Superior radiation resistance of ODS-RAFM steels revealed by Positron annihilation spectroscopy study

Oxide-dispersion-strengthened (ODS) reduced-activation ferritic/martensitic (RAFM) steels are considered as most promising structural materials for fusion reactors and other advanced nuclear systems due to its excellent irradiation resistance under high dose irradiations. However, the extreme complex synergistic effect between transmuted helium/hydrogen and displacement damage which most seriously affects the structure and mechanical properties of ODS-RAFM steels is very difficult to understand, especially the role of hydrogen still remains unclear. Taking advantage of very sensitive to small vacancy clusters of Positron annihilation spectroscopy (PAS) method, we conducted irradiation defect study on ODS-RAFM, RAFM and its corresponding model alloy Fe–9Cr and α-Fe to investigate their anti-swelling properties at the early stage of irradiation with hydrogen-helium synergism and the mechanism. Three distinct irradiation schemes including (1) single Fe ion beam, (2) simultaneous Fe and He (named Fe + He) ion beam, (3) subsequent H ion injection after Fe + He (named Fe + He/H) ion beam irradiation were performed. Hydrogen-helium synergistic effects on vacancy evolution within these steels were explored combining the first-principles calculation. The implantation of He was observed to significantly increase defect concentration among all four steels, while H exhibited interestingly complexity on affecting the evolution of helium bubbles. The H post-injection increased defect concentration obviously in the materials with abundant sinks, but improved poorly in ones with rare sinks. Through the systematic study of the interactions between H, He, and vacancies, it was revealed that even at the early irradiation stage, ODS-RAFM steels has already exhibited excellent irradiation resistance compared to other alloys across three irradiation schemes.

Hydrogen isotope permeation and retention behavior in the RAFM steel manufactured by laser powder bed fusion

Laser powder bed fusion (LPBF) has emerged as a promising additive manufacturing technique for producing complex components across various industries. This method is particularly instrumental in fabricating reduced activation ferritic/martensitic (RAFM) steel, a candidate structural material for the tritium breeding blanket. Hydrogen transport behaviors in the conventional RAFM steel have been largely investigated; however, they are not well understood in the LPBF-RAFM steel. In this work, we studied the hydrogen isotope permeation and retention behavior in the RAFM steel fabricated by LPBF, utilizing the deuterium gas driven permeation and thermal desorption spectroscopy experiments. The results show that the hydrogen isotope permeability of the LPBF-RAFM steel is close to the conventionally fabricated RAFM steel, but the diffusion coefficient is greatly lower and the retention is markedly higher in the LPBF-RAFM steel compared to the conventional RAFM steel. The large number density of microvoids and TiC precipitates in the LPBF-RAFM steel could be the predominant factors that result in the decrease in hydrogen isotope diffusion coefficient and the increase in the total amount of deuterium retention.

Microstructure and thermal stability of a structurally graded tungsten and reduced activation ferritic/martensitic steel joint

The leading design for plasma-facing high heat flux components in proposed fusion reactors involves joining plasma-facing tungsten tiles to underlying reduced activation ferritic-martensitic (RAFM) steel structures. Due to significant differences in the physical properties between W and steel, an effective method for joining them while preserving the mechanical and microstructural properties under service thermo-mechanical load is yet to be demonstrated. A transitional multilayer consisting of three layers (VCrTi, VCrAl, and FeCrAl) between W and steel is designed with the help of thermodynamic simulation and diffusion kinetics in an attempt to form solid solution bonding without forming brittle intermetallic phases. The solid solution bonding is desirable to mitigate the drastically different thermal expansion between W and steels. The multilayer structure was fabricated using spark plasma sintering (SPS). The microstructure of the bonded transition layers was analyzed after long-term annealing at 620 °C up to 1000h, through scanning electron microscope, energy dispersive spectroscopy and electron backscattering diffraction. Interdiffusion kinetics between layers and reliability of mobility database were also analyzed. This work provides a good understanding of thermal stability as well as the efficacy of multilayer functionally graded transition layer for joining W to ferritic/martensitic steel.

Enhanced disperse Ta(C, N) precipitation in reduced activation ferritic/martensitic steels by optimized two-step austenitizing

High-density MX precipitates block the movement of dislocations during creep at high temperature and increase sink strength to point defects under irradiation. Therefore, the precipitation control of MX phase is crucial for the advancement of reduced activation ferritic/martensitic (RAFM) steels. In this paper, an optimized two-step austenitizing strategy was proposed to enhance Ta(C, N) precipitation in RAFM steels, and the temperature effects of the first and second austenitizing on the Ta(C, N) precipitation enhancement were studied. During the first austenitizing, Ta(C, N) particles with a small radius (∼12 nm) and a high number density (∼8.5 × 1019 m−3) precipitated by heterogeneous nucleation on the M23C6 phase boundaries retained by the comparatively low temperature (900 °C). Then in the second austenitizing, enhanced Ta(C, N) precipitates kept good size and thermal stability below 1050 °C. Compared to RAFM steels treated by conventional normalizing and tempering, the strength and creep performance were significantly improved by the pinning effect of the enhanced dispersed Ta(C, N) particles. Furthermore, the optimized two-step austenitizing strategy was validated in the other two RAFM steels, showing broad application prospects in RAFM and heat-resistant steels.

Impact of trace silicon on irradiation hardening and embrittlement of RAFM steel subjected to neutron irradiation

The content of silicon (Si) element in reduced activation ferritic/martensitic (RAFM) steels varies significantly among different candidates of fusion reactor structural materials. Si is not intentionally added to the RAFM steel but rather remains as a residue during the deoxidation process in steel fabrication. To control and reduce the Si content, removing processes and special treatments are required. The additional refining processes significantly increase the cost of steel fabrication. Currently, the role of residual or trace Si in neutron irradiation of steel remains unclear. In the presented study, two 9Cr-2 W RAFM steels with Si contents of 0.1 wt.% and 0.01 wt.% were fabricated and subjected to a neutron irradiation at 563 K with a neutron fluence of 4.8 × 1023m−2. The main results show that 0.1 wt.% Si can contribute to the suppression of irradiation hardening and embrittlement, as evidenced by the reduction in irradiation-increased hardness and yield strength, while maintaining preferable elongation. Microscopic analysis suggests that an appropriate content of Si helps refine the grains in RAFM steel, increases the ductile fracture region of the fracture surface, and effectively reduces the density of dislocation loops induced by neutron irradiation. The findings evidence that retaining a certain amount of trace Si should be an informed and rational choice in the design of RAFM steel.