Reduced Activation Ferritic Steel Research Articles

Multiphysics modeling of the FW/Blanket of the U.S. fusion nuclear science facility (FNSF)

The dual coolant lead-lithium (DCLL) blanket concept, which is utilized in the Fusion Nuclear Science Facility (FNSF) conceptual design, is based on a helium-cooled first wall and blanket structure with RAFS (Reduced Activation Ferritic Steel) and a self-cooled LiPb breeding zone. The objective of this work is to develop a multiphysics modeling process in order to optimize the design and achieve long lifetime, maintainability, and high reliability. 3D finite element multiphysics modeling of the DCLL first wall and blanket (midplane of one sector) has been performed using COMSOL 5.2. The multiphysics aspect of the design is demonstrated via coupling of Computational Fluid Dynamics (CFD), conjugate heat transfer and solid mechanics. Both normal and off-normal loading conditions have been analyzed. The results of velocity, pressure, and temperature distributions of helium flow, as well as the primary and thermal stress of the structure were obtained. This was followed by determination of the factors of safety along three critical paths based on the ITER Structural Design Criteria for In-vessel Components (ISDC-IC). We show here that the structural design meets the ITER-ISDC design rules under both normal and off-normal operating conditions, though the safety factors under off-normal condition with 8MPa helium pressure are marginal. Thus simple design optimization was conducted based on a parametric study on first wall dimensions to improve the design.

Bi-directional hydrogen isotopes permeation through a reduced activation ferritic steel F82H

The first wall of a magnetic fusion reactor serves to separate the edge plasma from breeding blankets, which will be subjected to bi-directional hydrogen isotopes permeation: in one direction by plasma-driven permeation, and in the other direction by gas-driven permeation. In this work, hydrogen isotopes transport through a reduced activation ferritic steel F82H has been investigated in the temperature range of 150–500°C. The transport parameters including permeability, diffusivity, solubility and surface recombination coefficient have been measured and the isotopic mass effect has been discussed. Bi-directional hydrogen isotopes gas- and plasma-driven permeation has been demonstrated for the first time under controlled experimental conditions.

MHD effects and heat transfer analysis in magneto-thermo-fluid-structure coupled field in DCLL blanket

The study on DCLL blanket in ITER is related to magneto-thermo-fluid-structure coupled field issue. A key element in the DCLL concept is the flow channel insert (FCI) which serves as an electrical insulator to reduce the magnetohydrodynamic (MHD) pressure drop, and as a thermal insulator to decouple the high temperature PbLi from the reduced activation ferritic steel (RAFS) structure. In the present work, 16 geometrical models of flow ducts are introduced to study the MHD flow and heat transfer in DCLL blanket in magneto-thermo-fluid-structure coupled physical field. The PISO method on unstructured collocated meshes is applied to simulate the metal liquid flow and heat transfer in the blanket. The consistent and conservative scheme is employed to solve the incompressible Navier-Stokes equations with the Lorentz force included based on the electrical potential formula. The finite element method is used to study thermal mechanical behaviors of FCI. The velocity distribution, MHD pressure drop, electric current stream lines and temperature distribution in liquid blanket, thermal deformations of FCI in various geometrical models under external strong magnetic field are investigated. The pressure drop reduction factor is defined to analyze the influence of FCI structure on the MHD effects in the liquid metal blanket. The nonlinear coupling effects among magnetic field, heat transfer and FCI structure are revealed. The results show that thicker FCI and wider gap would increase the MHD pressure drops in bulk flow; the thicker FCI has smaller displacement but its strain and stress change non-monotonously; the wider gap can enhance the heat transfer performance but lead to larger stress in FCI. The optimal design is essential for the structural safety and high heat efficiency of the system.

Deuterium permeation behavior in iron-irradiated erbium oxide coating

Tritium permeation barrier has been developed for mitigating fuel loss and radiological concern at a fuel breeding/recovery system in a D-T fusion reactor. Recent research effort has been dedicated to erbium oxide coatings, and various hydrogen permeation behaviors except for irradiation effects have been elucidated. In this study, irradiation effects on deuterium permeation through erbium oxide coatings have been investigated by iron-ion irradiation at elevated temperature followed by deuterium gas-driven permeation experiments. The coatings deposited on reduced activation ferritic steel substrates with displacement damages of 0.01–1 dpa showed one or two orders of magnitude different permeabilities at 300–500°C; however, the permeabilities became comparable and lower than that of unirradiated at 550–700°C, indicating the grain growth and the formation of grain boundaries with a lower permeability. Cross-sectional transmission electron microscopy with selected-area electron diffraction for the coatings before and after the permeation experiments indicated the formation of a defect-accumulated region. The stability of the region strongly depends on the irradiation condition: damage concentration and annealing time, resulting in the difference of the permeability and diffusivity in the lower temperature range.

Hydrogen Isotopes Plasma-Driven Permeation through Tungsten Coated Reduced Activation Ferritic Steel F82H

Hydrogen Isotopes Plasma-Driven Permeation through Tungsten Coated Reduced Activation Ferritic Steel F82H

Optimization of Welding Process Parameters for Reduced Activation Ferritic-Martensitic (RAFM) Steel

Reduced-activation ferritic and martensitic (RAFM) steels are considered as struc tural materials for fusion reactor applications as they can loose their induced radioactivity quite early. In RAFM steel weld joints produced by A-TIG welding process, weld bead width (BW), depth of penetration and heat affected zone (HAZ) width play an important role in determining the mechanical properties and also the performance of the weld joints during service. To obtain the desired depth of penetration, and HAZ width, it becomes necessary to set up the welding process parameters. In the present work, intelligent modeling using genetic algorithm (GA) was used for optimization of the welding process parameters. Then GA suggested welding process parameters were validated for achieving the desired depth of penetration and HAZ width during A-TIG welding of RAFM steels. There was good agreement between the predicted and the measured values of depth of penetration and the HAZ width obtained.

低放射化フェライト鋼 F82H 鋼の高温水素ガス中における疲労き裂進展特性と SSRT 特性

低放射化フェライト鋼 F82H 鋼の高温水素ガス中における疲労き裂進展特性と SSRT 特性

Effect of recrystallization on ion-irradiation hardening and microstructural changes in 15Cr-ODS steel

The effects of recrystallization on ion-irradiation hardening and microstructural changes were investigated for a 15Cr-ODS ferritic steel. Dual ion-irradiation experiments were performed at 470°C using 6.4MeV Fe3+ ions simultaneously with energy-degraded 1MeV He+ ions. The displacement of damage at 600nm depth from the specimen surface was 30dpa. Nano-indentation test with Berkovich type indentation tip was measured by constant stiffness measurement (CSM) technique. Results from nano-indentation tests indicate irradiation hardening in ODS steels even at 470°C, while it wasn’t observed in reduced activation ferritic steel. Recrystallized ODS steel shows a larger irradiation hardening, which is considered to be due to the reduction of grain boundaries and interfaces of matrix/oxide particles. In 20% cold rolled ODS steel after recrystallization, both the hardening and bubble number density were lower than those of recrystallized ODS steel, suggesting that dislocations generated by cold rolling suppress bubble formation. Based on the estimation of irradiation hardening from TEM observation results, it is considered that the bubbles are not the main factor controlling ion-irradiation hardening.

Dynamic tensile properties of reduced-activation ferritic steel F82H

The dynamic tensile properties of reduced-activation ferritic steel F82H was characterized by high-speed tensile tests over a strain-rate range of 0.01–1400s−1 at temperatures of 296 and 423K. As a result, the tensile strength of the F82H steel appears to be sensitive to strain-rate. This is explained by the activation volume of gliding dislocation. The uniform elongation is also sensitive to strain-rate but true fracture strain is considered to be less sensitive. Zerilli–Armstrong bcc model is found adequate to describe the dependence of yield stress of F82H on temperature and strain-rate tested in the present study.

ARIES-ACT2 DCLL Power Core Design and Engineering

ARIES-ACT2 is a conventional tokamak power plant conceptual design that uses a dual-coolant lead-lithium (DCLL) blanket concept with a RAFS (reduced-activation ferritic steel) first-wall (FW) and blanket structure. The design concept is the first fully integrated study of the DCLL blanket in a tokamak power plant. The major engineering efforts were to develop a credible configuration that can meet aggressive maintenance goals and achieve high availability and maintainability; to design a DCLL blanket that can meet tritium breeding requirements with reasonable helium and Pb-17Li cooling schemes to remove the surface and volumetric thermal power in the blanket while keeping the helium pressure drop, magnetohydrodynamic (MHD) pressure drop, and total pumping power low, and material temperatures and stresses at an acceptable level; to design manifolding and access pipes to connect/disconnect the inboard and outboard blanket sectors to the ring headers located underneath the reactor without affecting maintenance operations and creating major MHD effects when feeding all the Pb-17Li/He mass flow. Detailed three-dimensional finite element analysis of the DCLL blankets together with design iterations have been performed to finalize and optimize the major design parameters of the FW and blanket structure. The helium-cooled W plate-type divertor concept was adopted and integrated into the ACT2 DCLL power core to accommodate the peak surface heat flux of ∼10 MW/m2 predicted by edge plasma physics.

The ARIES Advanced and Conservative Tokamak Power Plant Study

Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5,a βtotalN of 5.75, an H98 of 1.65, an n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m2. The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced-activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βtotalN of 2.5, an H98 of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.

Gas- and plasma-driven hydrogen permeation through a reduced activation ferritic steel alloy F82H

The first wall of a magnetic fusion power reactor will be subjected to hydrogen isotope permeation by the two mechanisms: one is gas-driven and the other is plasma-driven. Hydrogen transport through a reduced activation ferritic steel alloy F82H has been investigated using a steady-state laboratory-scale plasma device. Permeation parameters including permeability, solubility and diffusivity have been measured in the temperature range from 150 to 520°C. The surface recombination coefficient for hydrogen has also been estimated by a one-dimensional steady-state permeation model with the input data taken from experiments. Using these parameters, the hydrogen plasma-driven permeation flux and inventory for a 0.5cm thick first wall around 500°C are estimated to be ∼1.0×1013atomcm−2s−1 and ∼2×1016atomcm−3, respectively. Also, the implications of all these data on reactor operation are discussed.

Corrosion behavior of F82H exposed to high temperature pressurized water with a rotating apparatus

The present study reports the corrosion behavior of a reduced-activation ferritic martensitic steel F82H exposed to high temperature pressurized water for 28 and 100h using a rotating disk apparatus at rotation speeds of 500 and 1000rpm at a temperature of 573K under a water pressure of 15MPa with <20ppb dissolved oxygen. Static tests without rotation increased the weight of sample with time, whereas rotating tests decreased the weight of sample with the exposure time and rotating speed. Oxide particles in the size of several micron meters on the surface of static samples were not observed on the rotated samples which became covered with a thin Cr-rich oxide layer. These results suggest an occurrence of erosion–corrosion and/or flow-accelerated corrosion of F82H under the rotating condition.

In-situ Fe+ Ion Irradiation of an Oxide Dispersion Strengthened Steel

Oxide Dispersion Strengthened (ODS) reduced activation ferritic steels are promising candidate materials for structural components of both nuclear fission and fusion reactors. However, when irradiated with energetic particles, they may suffer changes on their microstructures that degrade their mechanical performance. In-situ transmission electron microscopy studies on ion-irradiated ODS steels can give remarkable insights into fundamental aspects of radiation damage allowing dynamic observations of defect formation, mobilities, and interactions during irradiation. In this investigation, a commercially available PM2000 ODS steel was in-situ irradiated with 150 KeV Fe+ at room temperature and 700°C. These experiments showed that the oxide nanoparticles in these steels remain stable up to the higher irradiation dose (~ 1.5 dpa), and that these particles seem to be effective sinks for irradiation induced defects.

Depth-dependent nanoindentation hardness of reduced-activation ferritic steels after MeV Fe-ion irradiation

The irradiation hardening behavior of F82H reduced-activation ferritic steels after MeV Fe-ion beam irradiation experiments was investigated with a nanoindentation test. Two sets of ion-irradiation experiments were conducted at 270°C with 10.5MeV Fe3+ ions up to 5dpa at a 1000nm depth at TIARA facility and at 290°C with 6.4MeV Fe3+ ions up to 3dpa at a 600nm depth at DuET facility, respectively. The measured nanoindentation hardness was converted to the depth-dependent bulk-equivalent hardness based on a combination of the Nix–Gao model to explain the indentation size effect and the film/substrate model to explain the damage gradient effect in the ion-irradiated region and softer substrate effect of the non-irradiated region beyond the irradiated depth range.

Compatibility of Ni and F82H with liquid Pb–Li under rotating flow

The present study reports the compatibility of a reduced-activation ferritic steel F82H and Ni with liquid Pb–Li under rotating flow conditions at 600°C. Cross-sectional observation of Ni using field emission electron probe micro-analyzer (FE-EPMA) after exposure for 100h revealed that severe grain boundary penetration of Pb into Ni occurred up to approximately a depth of 700μm, causing liquid metal embrittlement (LME). In contrast, the results for F82H after exposure for 500h showed the formation of pitting holes and a Cr-depleted layer at the surface with an approximate maximum depth of 10μm. Oxide particles were also found in the Pb–Li region in the F82H specimen after exposure. The radio-frequency glow discharge spectrometers successfully detected Li and indicated Li oxide formation at the surface.

Underwater explosive welding of tungsten to reduced-activation ferritic steel F82H

The present study reports the underwater explosive welding of commercially pure tungsten onto the surface of a reduced-activation ferritic steel F82H plate. Cross-sectional observation revealed the formation of a wave-like interface, consisting of a thin mixed layer of W and F82H. The results of nanoindentation hardness testing identified a gradual progressive change in the interface, with no hardened or brittle layer being observed. Small punch tests on the welded specimens resulted in cracking at the center of the tungsten, followed by crack propagation toward both the tungsten surface and the tungsten/steel interface.

Stability of the strengthening nanoprecipitates in reduced activation ferritic steels under Fe2+ ion irradiation

The stability of MX-type precipitates is critical to retain mechanical properties of both reduced activation ferritic–martensitic (RAFM) and conventional FM steels at elevated temperatures. Radiation resistance of TaC, TaN, and VN nanoprecipitates irradiated up to ∼49dpa at 500°C using Fe2+ is investigated in this work. Transmission electron microscopy (TEM) utilized in standard and scanning mode (STEM) reveals the non-stoichiometric nature of the nanoprecipitates. Irradiation did not alter their crystalline nature. The radiation resistance of these precipitates, in an order of reduced resistance, is TaC, VN, and TaN. Particle dissolution, growth, and reprecipitation were the modes of irradiation-induced instability. Irradiation also facilitated formation of Fe2W type Laves phase limited to the VN and TaN bearing alloys. This result suggests that nitrogen level should be controlled to a minimal level in alloys to gain greater radiation resistance of the MX-type precipitates at similar temperatures as well as postpone the formation and subsequent coarsening of Laves phase.

Investigation on different oxides as candidates for nano-sized ODS particles in reduced-activation ferritic (RAF) steels

Future generation reactor concepts are based on materials that can stand higher temperatures and higher neutron doses in corrosive environments. Oxide dispersion strengthened steels with chromium contents ranging from 9 to 14wt.% – produced by mechanical alloying – are typical candidate materials for future structural materials in fission and fusion power plants.Y2O3 has proven to be a good candidate for addition to ferritic steels during mechanical alloying to form nano-sized dispersion oxide particles during compacting of the material. These oxide particles have many positive effects on the material such as improved high-temperature properties and higher corrosion resistance. However, there is potential for improvements by choosing different oxides.In this present work, four different oxides (MgO, La2O3, Ce2O3 and ZrO2) are selected by looking at their thermal stabilities and Gibbs free enthalpies of various chemical compositions. These oxides are mixed and mechanically alloyed with ferritic steel powder (Fe13Cr1W0.3Ti) and compared to a reference material produced with Y2O3 (Fe13Cr1W0.3Ti+Y2O3). The materials were characterized in terms of their mechanical properties and detailed microstructural investigations by transmission electron microscopy and electron backscatter diffraction.All further results of the mechanical testing and microstructural characterizations are analyzed, compared, and discussed in this paper.

Mechanical properties and thermal stability of nanostructured ODS RAF steels

The main goal of the research reported in this paper was to evaluate the microstructure and mechanical properties of an oxide dispersion strengthened (ODS) reduced activation ferritic (RAF) steel after a long-term and high temperature annealing. The mechanically alloyed ODS RAF steel powder with a composition of Fe–14Cr–2W–0.3Ti–0.3Y2O3 was consolidated by hot isostatic pressing at 1150°C for 3h followed by a high speed hydrostatic extrusion at 900°C. The samples were subsequently exposed to various heat treatments, such as annealing at 850°C–1350°C for 1h in argon and to a long-term isothermal annealing at 750°C up to 10000h in air. The results indicate that due to the Zener pinning of the nanoparticles at the grain boundaries, the ODS RAF steel is thermally stable up to high annealing temperature of 1250°C. However, annealing at 1050°C applied after the hydrostatic extrusion of the steel significantly improves its ductility. Further increase in the temperature up to 1350°C resulted in nanoparticles coarsening (size ∼8.0nm, number density 2.5×023m−3) accompanied by an increase in the volume fraction of the coarse grains and a reduction of microhardness from 460 to 330HV0.1. Long-term annealing at 750°C up to 10000h brought about a systematic loss of both strength and ductility. However, the microhardness increases up to 500HV0.1. TEM observations revealed that the grain size after ageing remains almost unchanged. On the other hand, the number of dislocation free larger grains and of larger precipitates with a size of ∼25nm located at grain boundaries significantly increase.