R.L. Klueh

Low-activation ferritic and martensitic steels for fusion application

Abstract 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.

Ferritic/martensitic steels - overview of recent results

Considerable research work has been conducted on the ferritic/martensitic steels since the last International Conference on Fusion Reactor Materials in 1999. Since only a limited amount of that work can be reviewed in this paper, four areas will be emphasized: (1) the international collaboration under the auspices of the International Energy Agency (IEA) to address potential problems with ferri tic/marten si tic steels and to prove their feasibility for fusion, (2) the major uncertainty that remains concerning the effect of transmutation helium on mechanical properties of the steels when irradiated in a fusion neutron environment, (3) development of new reduced-activation steels beyond the F82H and JLF-1 steels studied in the IEA collaboration, and (4) work directed at developing oxide dispersion-strengthened steels for operation above 650degreesC

Current status and future R&D for reduced-activation ferritic/martensitic steels

Abstract International research and development programs on reduced-activation ferritic/martensitic steels, the primary candidate-alloys for a DEMO fusion reactor and beyond, are briefly summarized, along with some information on conventional steels. An International Energy Agency (IEA) collaborative test program to determine the feasibility of reduced-activation ferritic/martensitic steels for fusion is in progress and will be completed within this century. Baseline properties including typical irradiation behavior for Fe–(7–9)%Cr reduced-activation ferritic steels are shown. Most of the data are for a heat of modified F82H steel, purchased for the IEA program. Experimental plans to explore possible problems and solutions for fusion devices using ferromagnetic materials are introduced. The preliminary results show that it should be possible to use a ferromagnetic vacuum vessel in tokamak devices.

Progress and critical issues of reduced activation ferritic/martensitic steel development

The inherent properties of reduced activation ferritic/martensitic (RAFM) steels include reduced swelling and high recycling potential, which make them likely candidates for application in commercial fusion power plants. The International Energy Agency (IEA) agreement has been an effective framework for international co-operation in developing RAFM steels. The progress and critical issues observed in this co-operation are reported. The production of RAFM steels on an industrial scale has been demonstrated. Various methods of fusion welding and solid hot isostatic pressing (HIP) are feasible for joining the steels. Manufacturing of complex shapes with the powder HIP method works well for RAFM steels. Major critical issues addressed concern the effects of simultaneous introduction of helium and displacement damage. The availability of a 14 MeV neutron source is identified as an essential tool to determine this effect. Finally, the potential of oxide dispersion strengthening to increase the operating temperature of RAFM steels is considered as an issue that has to be resolved to enlarge the application temperature window of RAFM steels.

Tensile and Creep Properties of an Oxide Dispersion-Strengthened Ferritic Steel

The tensile and creep properties of two oxide dispersion-strengthened (ODS) steels with nominal compositions of Fe–12Cr–0.25Y2O3 (designated 12Y1) and Fe–12Cr–2.5W–0.4Ti–0.25Y2O3 (12YWT) were investigated. Optical microscopy, transmission electron microscopy, and atom probe field ion microscopy studies indicated that the 12YWT contained a high density of extremely fine Y–Ti–O clusters, compared to the much larger oxide particles in the 12Y1. The fine dispersion of particles gave the 12YWT better tensile and creep properties compared to commercial ODS alloys and ferritic/martensitic steels that would be replaced by the new ODS steel.

Development of an extensive database of mechanical and physical properties for reduced-activation martensitic steel F82H

Abstract Tensile, fracture toughness, creep and fatigue properties and microstructural studies of the reduced-activation martensitic steel F82H (8Cr–2W–0.04Ta–0.1C) before and after irradiation are reported. The design concept used for the development of this alloy is also introduced. A large number of collaborative test results including those generated under the International Energy Agency (IEA) implementing agreements are collected and are used to evaluate the feasibility of using reduced-activation martensitic steels for fusion reactor structural materials, with F82H as one of the reference alloys. All the specimens used in these tests were prepared from plates obtained from 5-ton heats of F82H supplied to all participating laboratories by JAERI. Many of the results have been entered into relational databases with emphasis on traceability of records on how the specimens were prepared from plates and ingots.

Status of R&D activities on materials for fusion power reactors

Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, the Peoples Republic of China, Russia and the USA. They relate to the development of new high temperature, radiation resistant materials, the development of coatings that will act as erosion, corrosion, permeation and/or electrical/MHD barriers, characterization of candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

Microstructural analysis of neutron-irradiated martensitic steels☆

Abstract Four martensitic steels for fusion applications were examined by transmission electron microscopy after irradiation in the Fast Flux Test Facility (FFTF) at 420°C to 7.8 X 10 26 n/m 2 ( E > 0.1 MeV), about 35 dpa. There were two commercial steels, 9Cr-IMoVNb and 12Cr-1MoVW, and two experimental reduced-activation steels, 9Cr-2WV and 9Cr-2WVTa. Before irradiation, the tempered martensite microstructures of the four steels contained a high dislocation density, and the major precipitate was M 23 C 6 carbide, with few MC carbides. Irradiation caused minor changes in these precipitates. Voids were found in all irradiated specimens, but swelling remained below 1%, with the 9Cr-1MoVNb having the highest void density. Although the 12Cr-IMoVW steel showed the best swelling resistance, it also contained the highest density of radiation-induced new phases, which were identified as chi-phase and possibly α′. Radiation-induced chi-phase was also observed in the 9Cr-1MoVNb steel. The two reduced-activation steels showed very stable behavior under irradiation: a high density of dislocation loops replaced the original high dislocation density; moderate void swelling occurred, and no new phase formed. The differences in microstructural evolution of the steels can explain some of the mechanical properties observations made in these steels.

Helium effects on void formation in 9Cr-1MoVNb and 12Cr-1MoVW irradiated in HFIR☆

Up to 2 wt% Ni was added to 9Cr-1MoVNb and 12Cr-lMoVW ferritic steels to increase helium production by transmutation during HFIR irradiation. The various steels were irradiated to ∼39 dpa. Voids were found in all the undoped and nickel-doped steels irradiated at 400°C, most of them at 500°C, but not in any of them at 300 or 600°C. Bubble formation, however, was increased at all temperatures in the nickel-doped steels. Maximum void formation was found at 400°C, but swelling remained less than 0.5% even with up to 440 appm He. Irradiation at 300 to 500°C caused dissolution of as-tempered M23C6 precipitates and coarsening of the lath/subgrain structure in the 9-Cr steels, whereas the microstructure generally remained stable in the 12-Cr steels. Irradiation in this temperature range also caused compositional changes in the as-tempered MC phase in all the steels, and produced combinations of fine M6C, G-phase, and M2X precipitates in various steels. The subgrain boundaries appear to be strong sinks that enhance resistance to void formation. Higher helium production during irradiation appears to shorten the incubation period for void formation. The effects of helium on steady state void swelling behavior, however, remain unknown.

Embrittlement of reduced-activation ferritic/martensitic steels irradiated in HFIR at 300°C and 400°C

Abstract Miniature tensile and Charpy specimens of four ferritic/martensitic steels were irradiated at 300°C and 400°C in the high flux isotope reactor (HFIR) to a maximum dose of ≈12 dpa. The steels were standard F82H (F82H-Std), a modified F82H (F82H-Mod), ORNL 9Cr–2WVTa, and 9Cr–2WVTa–2Ni, the 9Cr–2WVTa containing 2% Ni to produce helium by (n,α) reactions with thermal neutrons. More helium was produced in the F82H-Std than the F82H-Mod because of the presence of boron. Irradiation embrittlement in the form of an increase in the ductile–brittle transition temperature (ΔDBTT) and a decrease in the upper-shelf energy (USE) occurred for all the steels. The two F82H steels had similar ΔDBTTs after irradiation at 300°C, but after irradiation at 400°C, the ΔDBTT for F82H-Std was less than for F82H-Mod. Under these irradiation conditions, little effect of the extra helium in the F82H-Std could be discerned. Less embrittlement was observed for 9Cr–2WVTa steel irradiated at 400°C than for the two F82H steels. The 9Cr–2WVTa–2Ni steel with ≈115 appm He had a larger ΔDBTT than the 9Cr–2WVTa with ≈5 appm He, indicating a possible helium effect.