Reduced Activation Ferritic Steel F82H Research Articles

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.

Oxide layer formation in reduced activation ferritic steel F82H under DEMO reactor blanket condition

Tritium permeation through structure materials in fusion blanket systems is a critical issue from the perspectives of fuel loss and radiological hazard. In the previous studies, detailed hydrogen isotope permeation behaviors in reduced activation ferritic/martensitic steels have been investigated; however, oxidation of the steel surface is expected under an actual DEMO reactor condition, and then the tritium permeation behavior will be changed. In this study, deuterium permeation through the steels heat-treated under simulated environment conditions has been investigated for more precise predictions of tritium loss at DEMO reactor blankets. Reduced activation ferritic/martensitic steel F82H substrates were heat-treated in helium gas flow containing 1 vol% hydrogen at 300, 400 and 500 °C for 100 and 200 h to simulate a solid breeder DEMO reactor blanket condition. After surface observation and analysis for the heat-treated samples, gas-driven deuterium permeation measurements were performed. An iron oxide layer was formed on the sample surface, and the thickness of the layer was 50 nm‒12 μm. The oxide layer on the sample surface heat-treated at 500 °C for 100 h decreased deuterium permeation by a factor of 5. After the permeation tests, dissipation of the oxide layers was confirmed.

Micro-tensile testing of reduced-activation ferritic steel F82H irradiated with Fe and He ions

Micro-tensile tests were conducted on reduced-activation ferritic steel F82H specimens after irradiation with Fe and He ions. The displacement damage levels were 50 and 200 dpa, and helium concentrations were 0, 2000 and 5000 appm. The specimens had dimensions of 8 × 1 × 1 µm for the gage section and they were tension tested at room temperature in a vacuum. The yield strengths of the unirradiated specimens were close to the literature value reported for a large, unirradiated specimen. The increases in yield strength and ultimate tensile strength due to Fe ion irradiation were clearly observed. The loss of work hardening was confirmed for the 200 dpa specimens (0 or 2000 appmHe). Higher yield strength and ultimate tensile strength were confirmed for the 50 dpa/5000 appmHe specimen compared to the no helium specimen (50 dpa/0 appmHe). However, no apparent helium effects were confirmed for the 200 dpa/2000 appmHe specimen by the micro-tensile test at room temperature.

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.

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

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

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.

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.

Neutronics design of the low aspect ratio tokamak reactor, VECTOR

A neutronic study has been carried out for VECTOR, which is a low aspect ratio tokamak with the aspect ratio of 2.3, to achieve simultaneously sufficient shielding for super-conducting toroidal field coils (TFCs) and tritium breeding ratio (TBR) larger than unity, using the three-dimensional Monte Carlo neutron transport code MCNP-4C. The tritium production is mainly made by outboard blankets consisting of the self-cooled lithium lead (LiPb) as tritium breeding and neutron multiplying material, and the reduced activation ferritic steel F82H as structural material. The inboard blanket consisting of vanadium hydrate (VH 2) and F82H as structural material plays the role of radiation shield for the TFCs. Adding Be layers to the LiPb layers effectively improves the TBR. By employing a 0.8 m thick inboard blanket with a 0.13 m thick LiPb layer, sufficient shielding for the inboard TFC and a total TBR of 1.08 are achieved.

Exposure of reduced activation ferritic steel F82H to TEXTOR plasma

Reduced activation ferritic steel, which is a leading structural material candidate for a fusion demonstration reactor, has been installed in TEXTOR and exposed to neutral beam heated plasma. When the heat load was lower than 5 MW/m 2, the insertion of the test limiter did not influence the plasma parameters. Visible spectroscopy near the surface showed that the emission related to constituent element such as iron and chromium is comparable to that from the stainless steel. It means that the sputtering from the ferritic steel is comparable to that from the stainless steel even for plasma exposure in high heat flux condition. When the heat load was ∼10 MW/m 2, the limiter was melted slightly. The strange motion of melted layer was observed by 2D camera. It slowly moved (∼0.5 cm/s) in a spiral motion, forming a small droplet ∼1 mm in diameter which was then released. The released droplet turned toward the limiter and hit the surface. The mechanism of this motion is discussed.

Tritium breeding experiments with blanket mock-ups containing [formula omitted]-enriched lithium titanate and beryllium irradiated with DT neutrons

Experiments have been conducted with breeding blanket mock-ups consisting of layers of candidate materials for JAERI’s pebble bed blanket design: lithium titanate, beryllium and reduced activation ferritic steel F82H. The mock-ups were irradiated with 14 MeV neutrons from the DT neutron source at FNS. The lithium titanate was enriched to 40 and 95% in Li 6 . The local tritium production rate in the 12 mm thick breeding layers was measured by means of pellet detectors of lithium titanate, Li 2CO 3 and liquid scintillation counting techniques. Additional verification of the neutron spectrum was obtained from material activation such as activation foils. The results were analyzed with the Monte Carlo neutron transport code MCNP-4C and several transport cross section libraries such as FENDL/MC-2.0. The calculation/experiment ratio was found to be within the combined error estimate of calculation and experiment but in some cases, the calculations showed an overestimation tendency.

Neutronics experiments for DEMO blanket at JAERI/FNS

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li2TiO3 blocks with a 6Li enrichment of 40% and 95%, and beryllium blocks. Sample pellets of Li2TiO3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross-sections needed to calculate the formation of the radioactive nuclei (56Co, 184Re, 48V, 206Bi, 65Zn and 51Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections increased remarkably while coming closer to polyethylene board, which was a substitute for water. As a result of this present study, it has become clear that the sequential reaction rates are important factors in the accurate evaluation of induced activity in fusion reactor design.

High heat flux test of a HIP-bonded first wall panel of reduced activation ferritic steel F-82H

Reduced activation ferritic steel F-82H is a primary candidate structural material of DEMO fusion reactors. In fabrication technology, development of the DEMO blanket in JAERI, a hot isostatic pressing (HIP) bonding method, especially for the first wall structure with built-in cooling tubes has been proposed. A HIP-bonded F-82H first wall panel was successfully fabricated with selected manufacturing parameters. A high heat flux test of the HIP-bonded F-82H first wall panel has been performed to examine the thermo-mechanical performance of the panel including the integrity of the HIP-bonded interfaces and the fatigue behavior. A maximum heat flux of 2.7 MW/m2 was applied to accelerate the fatigue test up to 5000 cycles in test blanket inserted ITER. The maximum temperature of the panel was ∼450°C under this heat flux. Through this test campaign, no damage such as cracks was observed on the surface of the panel, and no degradation in heat removal performance was observed either from the temperature responses. The thermal fatigue lifetime of the panel was found to be longer than the fatigue data obtained by mechanical testing.

833 ねじり疲労における微小き裂の発生・成長に及ぼす磁場の影響

Reversed tortional fatigue test of sandglass type specimen with a small flat surface notch was conducted at room temperature for a pure iron and a reduced activation ferritic steel F-82H in a solenoid coil cooled by water. Specimen was not constrained lonitudinally to avoid the additional stress induced by magnetostriction. Magnetic flux density in specimen, about 0.7 Tesra, was in rotation magnetization condition. Initiation and propagation behavior of large number of small cracks were observed on the flat surface notch by using image processing technique. For pure iron, crack initiation became earlier and crack growth faster in magnetic field. For F-82H steel influence of magnetic field did not observed.

Effects of hydrogen atmosphere on mechanical properties and surface conditions of a reduced activation ferritic steel F82H

Effects of hydrogen atmosphere on mechanical properties and surface conditions have been studied for a reduced activation ferritic steel F82H (Fe–8Cr–2W–V–Ta). Test specimens of F82H were exposed to hydrogen gas atmosphere simulating the operating conditions of a fusion reactor. Surface analyses were performed by Auger electron spectroscopy in order to evaluate the depth profiles of surface atomic compositions of F82H, and surface oxidation layers with a thickness of up to 56 nm were observed on the surface, mainly due to the high diffusion rate of oxygen in F82H matrix. A quantitative analyses of the specimens exposed to hydrogen atmosphere were performed by a gaschromatograph, and absorbed hydrogen in F82H was found negligibly small, less than 5 ppm. Tensile and Charpy impact tests were carried out on these specimens, and mechanical properties were found comparable to those of as-received materials.

Application of HIP bonding to first wall panel fabrication made from reduced activation ferritic steel F82H

As a course of fabrication technology development of DEMO breeding blankets, fabrication of small-scaled first wall panels made from reduced activation ferritic steel F82H has been attempted by applying HIP (Hot Isostatic Pressing) method. By applying the conditions identified in the previous studies, two flat panels with ten cooling channels each have been fabricated. One of the fabricated panels was destructively tested to examine metallurgically sound HIP bonding and to characterize the change of the micro-structure of the F82H steel due to the HIP process. Destructive examination has confirmed sound bonding for all of the HIP interfaces and satisfactory dimensional tolerance, and applicability of HIP bonding to the complex component made from this steel has been demonstrated.

Diffusion bonding of reduced activation ferritic steel F82H for demo blanket application

A reduced activation ferritic steel, a grade F82H developed by JAERI, is a promising candidate structural material for the blanket and the first wall of DEMO reactors. In the present study, diffusion bonding of F82H has been investigated to develop the fabrication procedures of the blanket box and the first wall panel with cooling channels embedded by F82H. The parameters examined are the bonding temperature (810–1050°C), bonding pressure (2–10 MPa) and roughness of the bonding surface (0.5–12.8 μ R max), and metallurgical examination and mechanical tests of the diffusion bonded joints have been conducted. From the tests, sufficient bonding was obtained under the temperatures of 840–1050°C (compressive stress of 3–12 MPa), and it was found that heat treatment following diffusion bonding is essential to obtain the mechanical properties similar to that of the base metal.