Low Activation Ferritic Steels Research Articles

Dynamics of MHD instabilities near a ferromagnetic wall

Prospective fusion component testing and DEMO power reactor concepts are expected to employ low-activation ferritic steels because of their ability to withstand the high neutron flux of the reactor environment. However, theory suggests that ferromagnetic material may amplify certain external MHD instabilities. Using its ferromagnetic-resistive wall mode (FRWM) upgrade, the High Beta Tokamak-Extended Pulse (HBT-EP) experiment has observed approximately doubled growth rates when operating with a close-fitting ferromagnetic first wall, compared to operation with a stainless steel first wall. The presence of a ferromagnetic wall correlates with earlier disruptions, and FRWM growth rates increase with decreasing mode rotation, as expected due to the increased skin depth allowing greater mode interaction with the bulk ferromagnetic material. It is also seen that introducing low-n asymmetries into the toroidal distribution of ferromagnetic material, similar to the ITER test blanket module toroidal asymmetry, changes the phase preference of rotating modes; meanwhile, a similar change in purely conducting material does not significantly change the mode’s phase preference.

Design and installation of a ferromagnetic wall in tokamak geometry.

Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak-Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics and overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability.

Active and passive kink mode studies in a tokamak with a movable ferromagnetic walla)

High-resolution active and passive kink mode studies are conducted in a tokamak with an adjustable ferromagnetic wall near the plasma surface. Ferritic tiles made from 5.6 mm thick Hiperco® 50 alloy have been mounted on the plasma-facing side of half of the in-vessel movable wall segments in the High Beta Tokamak-Extended Pulse device [D. A. Maurer et al., Plasma Phys. Controlled Fusion 53, 074016 (2011)] in order to explore ferritic resistive wall mode stability. Low-activation ferritic steels are a candidate for structural components of a fusion reactor, and these experiments examine MHD stability of plasmas with nearby ferromagnetic material. Plasma-wall separation for alternating ferritic and non-ferritic wall segments is adjusted between discharges without opening the vacuum vessel. Amplification of applied resonant magnetic perturbations and plasma disruptivity are observed to increase when the ferromagnetic wall is close to plasma surface instead of the standard stainless steel wall. Rapidly rotating m/n=3/1 external kink modes have higher growth rates with the nearby ferritic wall. Feedback suppression of kinks is still as effective as before the installation of ferritic material in vessel, in spite of increased mode growth rates.

Compatibility of Low Activation Ferritic Steels with Liquid Lithium

Fe-Cr-W based low activation ferritic steels are widely regarded as promising blanket structural materials for fusion reactors, while liquid lithium breeder/coolant concept provides an attractive option for high efficiency and simplicity of blanket system. However, past compatibility tests of ferritic steels with liquid lithium were almost limited to conventional Fe-Cr-Mo steels. In this study, the corrosion behavior of the candidate reduced activation ferritic steel, JLF-1(Fe-9Cr-2W-0.1C) in lithium was investigated. Static immersion tests were carried out using coupon specimens (16 × 4 × 0.25 mm) at 873K and 973K for 100hr. At 973K, the phase transformation from martensite to ferrite resulted in decrease in hardness from 250 to 140Hv. This seemed to be caused by depletion of C. Examinations of binary Fe-Cr and pure iron were also carried out for comparison with JLF-1.

Corrosion characteristics of low activation ferritic steels in liquid lithium

Corrosion characteristics of low activation ferritic steels in liquid lithium

Microstructural examination of low activation ferritic steels following irradiation in ORR at 330 and 400 °C to ∼10 dpa

Microstructural examinations are reported for a series of low activation steels containing Mn following irradiation in the Oak Ridge Reactor at 330 and 400 °C to ∼10 dpa. Alloy compositions included 2% Cr, 9% Cr and 12% Cr steels with V to 1.5% and W to 1.0%. Results include compositional changes in precipitates and microstructural changes as a function of composition and irradiation temperature. It is concluded that temperatures in ORR are on the order of 50 °C higher than anticipated.

Evaluation on fracture toughness of low activation ferritic steel (JLF-1)

Low activation ferritic steels have been developed for a candidate of structural materials in nuclear fusion reactor applications. This paper dealt with the effects of rolling direction and side groove on the fracture toughness of low activation ferritic steel, JLF-1 Steel (9Cr-2W-V-Ta). The fracture toughness of JLF-1 steel showed the anisotropy in the rolling direction. In the case of J-R curve, JQ and slope of the longitudinal JLF-1 steel (BML) for the rolling direction were a slight larger than those of transverse JLF-1 Steel (BMT) at room temperature. In order to obtain the valid fracture toughness of JLF-1 Steel through the UC method, the CT specimen with a side groove of 40% was recommended. As one of low activation ferritic steels for the IEA test program, the JLF-1 (9Cr-2W-V-Ta) steel has been developed and investigated by the active cooperation of Japanese Universities (1,3,4). The low activation ferritic steels mounted in a high magnetic field such as fusion reactor components may be affected by the operating cycle and the mechanical vibration that is associated with the magnetic damping. For the sound design of fusion reactor, it is necessary to investigate the effect of magnetic damping on the mechanical properties of JLF-1 steel. Prior to its widespread, practical use, however, several fracture properties of the JLF-1 steel including tensile property, fatigue property and fracture toughness must be evaluated, preferentially. Especially, the investigation of fracture toughness and J-R curve is essential for practical applications of the JLF-1 steel into fusion areas, since it suffered from the elastic-plastic fracture behavior caused by the history of operation temperature. When the JLF-1 steel is used as the blanket structural material of fusion reactors, the experimental data for tensile properties and fracture toughness are very important to estimate safety life span of the blanket by critical crack length with the irradiation. The tensile properties and fracture toughness characteristics of unirradiated materials are required before the estimation of irradiated materials (2,5). Finally, the correlativity between magnetic field and fracture properties of JLF-1 steel should be established for its active application. Due to the limited amount of materials, the unloading compliance method with single-specimen method (ASTM E813-89), which was the alternative of multi-specimen method, was recommended for the examination of fracture toughness (6,9).

Tensile response of low activation ferritic steels irradiated in ORR at temperatures in the range 60–400 °C

Post-irradiation tensile test results are reported for a series of low activation steels containing manganese following irradiation in the Oak Ridge Reactor at 60, 200, 330 and 400 °C to ∼10 dpa. Alloy compositions included 2Cr, 9Cr and 12Cr steels with V to 1.5% and W to 1.0%. Strengths are higher in all alloys for irradiation conditions below 400 °C, with peak hardening occurring following irradiation at 200 °C. The 9Cr alloy class exhibited the smallest increases in hardening. Test results were consistent with previous results obtained on fast flux test facility-irradiated specimens. Manganese does not appear to play a role in the hardening observed at these low irradiation temperatures.

Demonstration of ripple reduction by ferritic steel board insertion in JFT-2M

In the JFT-2M tokamak, the application of low activation ferritic steels to plasmas has been investigated (the so-called Advanced Material Tokamak Experiment (AMTEX) programme). In the first stage, toroidal field ripple reduction was examined by ferritic steel boards (FBs) inserted between the toroidal field coils and the vacuum vessel. It is demonstrated that FB insertion reduced toroidal field ripple and the losses of fast ions produced by tangential co-NBI. By optimizing the FB thickness, such that fundamental mode ripple is minimized to 0.07% at the shoulder of the inside wall, ripple trapped loss is reduced to an almost negligible level. It is determined that the reductions of the fundamental mode ripple and the ripple banana diffusion coefficients at the shoulder are most effective in reducing ripple ion losses. Ripple loss reduction by FBs is also confirmed with perpendicular beam injection. The insertion of FBs causes no undesirable effects on plasma production and control.

Low-temperature irradiation effects on tensile and Charpy properties of low-activation ferritic steels

Tensile and Charpy properties of low-activation ferritic steel, F82H irradiated up to 0.8 dpa at low temperature below 300°C were investigated. The helium effect on these properties was also investigated using the boron isotope doping method. Neutron irradiation increased yield stress accompanied with ductility loss, and it also shifted the ductile-to-brittle transition temperature (DBTT) from −50°C to 0°C. Boron-doped F82H showed larger degradation in DBTT and ductility than boron-free F82H, while they had the same yield stress before and after irradiation.

Mechanical Properties of a High-Purity Fe–9Cr–2W–0.1C Model Alloy for Low-Activation Ferritic Steels for Fusion Reactors

Reduced-activation ferritic steels, such as JLF-1 (Fe-9Cr-2W-0.1C--V-Ta-N) steel, are candidate structural materials for use in the next fusion device. It is essential to develop low-activation ferritic steels which have been purified in order to reduce the amount of radioactive impurity elements. It is important to clarify whether or not purification causes any significant changes in the mechanical properties of the ferritic steels, as they will be used as structural materials. In this study a high-purity Fe-9Cr-2W-0.1C model alloy was prepared from an ultra-high-purity iron. A significantly lower tensile strength was observed in the high-purity ferritic steel at room temperature and 573 K, as compared to that of JLF-1 steel. The difference in yield strength at room temperature is considered to be mainly due to a lower content of nitrogen and soluble impurities, a larger packet diameter and a larger lath length in the case of the high-purity steel. The hardness of the as-quenched high-purity steel and the recovery behavior below 823 K are not significantly different from those of conventional ferritic steel.

Production of low activation steel; JLF-1, large heats – Current status and future plan

Based on the excellent basic properties prior to neutron irradiation and under neutron irradiation up to 100 dpa, the 9Cr–2W type low activation ferritic steel (LAF). JLF-1, had been selected as one of the reference materials to the IEA low activation ferritic steels R&D activity. This paper provides the general information about the second large heat of JLF-1 steel. The basic property of thick plates (24 and 12 mm in thickness) and their welded joints together with the information about microstructure are also provided.

The possibility of the commercial production of low-activation structural steels for fusion energy in the Russian Federation

Laboratory meltings of low activation austenitic and ferritic steels (Fe0.1C12Cr20Mn1W0.2V and Fe0.1C9Cr1W0.15Ta) were done in an inductive furnace. The panoramic analysis of the charge materials and steels were made by means of spark mass-spectrometry. Series of calculations of activation parameters for steels composing from hypothetically pure, isotopically enriched macrocomponents and for real materials were conducted. The possible metallurgical technologies of industrial production of the low activation steels in the Russian Federation including processes of initial charge materials purification are considered.

The development of ferritic steels for DEMO blanket

The development of low-activation ferritic/martensitic steels is a key to the achievement of nuclear fusion as a safe, environmentally attractive and economically competitive energy source. The Japanese and the European Fusion Materials programs have put low-activation ferritic and martensitic steels R&D at the highest priority for a demonstration reactor (DEMO) and the beyond. An international collaborative test program on low-activation ferritic/martensitic steels for fusion is in progress as an activity of the International Energy Agency (IEA) fusion materials working group to verify the feasibility of using ferritic/martensitic steels for fusion by an extensive test program covering the most relevant technical issues for the qualification of a material for a nuclear application. The development of a comprehensive data base on the representative industrially processed reduced-activation steels of type 8-9Cr-2WVTa is underway for providing designers a preliminary set of material data for the mechanical design of components, e.g. for DEMO relevant blanket modules. The current design status of FFHR and SSTR utilizing low-activation ferritic steels is reviewed and future prospects are defined.

Low-activation ferritic and martensitic steels for fusion application

This paper reviews the history and the present status of the development of low-activation ferritic/martensitic steels for fusion applications, followed by a summary of the status of the International Energy Agency fusion materials working group activities, where an international collaborative test program on low-activation ferritic/martensitic steels for fusion is in progress. The objective of the test program is to verify the feasibility of using ferritic/martensitic steels for fusion by an extensive test program covering the most relevant technical issues for the qualification of a material for nuclear application. The development of a comprehensive data base on the representative industrially processed reduced-activation steels of type 89Cr2WVTa will provide designers a preliminary set of material data within about 3 years for the mechanical design of components, e.g., for demo relevant blanket modules to be tested in ITER. Knowledge on the current limitations of low-activation ferritic steels for application in advanced fusion systems is reviewed and future prospects are defined.

R&D of low activation ferritic steels for fusion in japanese universities

Following the brief review of the R&D of low activation ferritic steels in Japanese universities, the status of 9Cr-2W type ferritic steels development is presented. The main emphasis is on mechanical property changes by fast neutron irradiation in FFTF. Bend test, tensile test, CVN test and in-reactor creep results are provided including some data about low activation ferritic steels with Cr variation from 2.25 to 12%. The 9Cr-2W ferritic steel, denoted as JLF-1, showed excellent mechanical properties under fast neutron irradiation as high as 60 dpa. As potential materials for DEMO and beyond, innovative oxide dispersion strengthened (ODS) quasi-amorphous low activation ferritic steels are introduced. The baseline properties, microstructural evolution under ion irradiation and the recent progress of new processes are provided.

Effects of neutron irradiation on microstructural evolution in candidate low activation ferritic steels

Fe-(2.25-12)Cr-2W-V, Ta low activation ferritic steels (JLF series steels) were developed in the fusion materials development program of Japanese universities. Microstructural observations, including precipitation response, were performed after neutron irradiation in the FFTF/MOTA. The preirradiation microstructure was stable after irradiation at low temperature (< 683 K). Recovery of martensitic lath structure and coarsening of precipitates took place above 733 K. Precipitates observed after irradiation were the same as those in unirradiated materials in 7–9Cr steels, and no irradiation induced phase was identified. The irradiation induced shift in DBTT in the 9Cr-2W steel proved to be very small which is a reflection of stable precipitation response in these steels. A high density of fine α' precipitates was observed in the 12Cr steel which might be responsible for the large irradiation hardening found in the 12Cr steel. Void formation was observed in 7–9Cr steels irradiated at 683 K, but the amount of void swelling was very small.

Irradiation creep of low-activation ferritic steels in FFTF/MOTA

Irradiation creep behavior of low-activation steels, developed as structural materials for fusion reactors (JLF series steels), was investigated to obtain a fundamental understanding of these alloys under fast neutron irradiation in FFTF. (2.25–8)Cr(1–2)W bainitic steels and 12Cr-2W ferritic steels showed superior creep resistance to type-316 stainless steels under fast neutron irradiation up to 520°C. At temperatures below 460°C the creep strain increased with increasing Cr content up to 7 Cr, and further increments of Cr content up to 12% reduced the creep strain. At temperatures between 460 and 600°C, 7–8 Cr ferritic steels showed the largest creep strain. Swelling-enhanced creep, near the peak swelling temperature of 410°C, was also observed. The 9Cr-2W ferritic steel JLF-1 presented excellent properties, suggesting it as a leading candidate alloy for structural components of fusion reactors.

Small specimen test techniques for the evaluation of toughness degradation

The small punch (SP) test was applied to neutron irradiated TEM disks to obtain fracture related mechanical properties (DBTT, J IC) of low activation ferritic steels of 8Cr-2WV and 8Cr-2WVTa and conventional 12Cr-1MoVW steel. The SP test using 3 mm diameter TEM disks is confirmed to be useful, especially for J IC evaluation. DBTT determination is also thought to be possible with a sufficient number of specimens.

Recent Results on Fusion Reactor Materials from FFTF/MOTA Program. Radiation Responses of Low Activation Ferritic Steels.

Reduced activation is realized to be an important issue from reactor safety, maintenance, waste management and other view points.Low activation ferritic steels are potential options to this issue and heavy radiation effects on these materials have been an important part of FFTF/MOTA program.This paper provides effects of fast neutron irradiation in FFTF on mechanical property changes and microstructural evolution of some low activation ferritic steels developed for fusion.