A. Hishinuma

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.

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.

Development of potential low activation ferritic and austenitic steels

Abstract New experimental steels were designed and tested as the reduced activation alternatives to the current austenitic and ferritic prime candidate alloys for applications to the first wall structural components of Tokamak reactors. The design was based on the experience and the knowledge implemented in the recent nuclear and non-nuclear material fields, in which optimization was attempted on the basic properties, keeping their levels equivalent to or better than those of the conventional candidates. Four proto-type alloys were melted for testing with the following compositional specifications: 0.1 C-8 Cr-2 W-0.2 V-Fe, 0.1 C-8 Cr-2 W-0.2 V-0.04 Ta-Fe, 0.25 C-25 Mn-14 Cr-0.2 Ti-Fe and 0.02 C-18 Ni-18 Cr-0.3 Ti-Fe. In the first phase of the evaluation program, tests were made to determine properties that included tensile, creep-rupture, impact before and after thermal aging, wet and dry corrosion, and weldability properties. These results are compaed with those of the conventionally known materials of the equivalent classes. Among the materials tested the ferritic alloy with 0.04 Ta showed particularly attractive features.

Phase stability of reduced activation ferritic steel: 8%Cr-2%W-0.2%V-0.04%Ta-Fe

Abstract A reduced activation ferritic steel, 8Cr-2W-0.2V-0.04Ta-Fe, was aged at 673–923 K. for up to 36 Ms, and Charpy impact tests and metallurgical investigation were conducted on the aged specimens. The results were compared with those for 12Cr-1Mo-V-W, 9Cr-1Mo-V-Nb, and Ta-free 8Cr-2W-0.2V steels. The fracture appearance transition temperature (FATT hereafter) in most cases increased with increase in aging time and aging temperature due to the precipitation of Laves phase and the growth of M23C6 carbides for all the steels tested. Though the change in morphology of the precipitates in 12Cr-1Mo-V-W-Fe was not as large as in the other steels tested, the initial value of the FATT was rather high. The Laves phase (Fe2Mo) precipitation occurred at rather a low temperature, 823–873 K in 9Cr-1Mo-V-Nb-Fe. The FATT of 8Cr-2W-V-Ta-Fe was lower than that of the 9 to 12%Cr steels even after aging at 873 K for 36 Ms because of its slow rate of Laves phase (Fe2W) precipitation.

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

Abstract 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.

Accumulation of engineering data for practical use of reduced activation ferritic steel: 8%Cr2%W0.2%V0.04%TaFe

Abstract A reduced activation ferritic steel, 8Cr2W0.2V0.04TaFe, is one of the candidates for the first wall material of SSTR (Steady State Tokamak Reactor). There is a growing demand for extensive properties data for designing the structure of the first wall. In a program of expanding of the data base, tensile tests at room, and elevated temperatures, creep rupture tests, Charpy impact tests and measurements of the physical properties were performed on the mill production plates as well as some laboratory heats. Some of the tests were conducted on specimens that were thermally aged up to 36Ms. Recent data were summarized and compared with those for the well known ferritic steels, HT9 and modified 9Cr.

Irradiation response on mechanical properties of neutron irradiated F82H

Abstract Tensile and Charpy impact properties of neutron irradiated F82H (Fe8Cr2WVTa) with and without boron have been investigated to obtain the basic irradiation response on mechanical properties in low damage regime less than 1 dpa at the temperature ranging from 300° to 590°C. Boron-doped steel was used for the helium effect due to (n, α) reaction. Typical irradiation hardening was observed at 300°C. The irradiation above 520°C did not reveal increase in yield stress, but the specimen irradiated at 590°C showed some reduction in elongation in room temperature tensile testing. Slight difference in the tensile properties between boron-doped and boron-free were observed at 590°C. No changes in ductile brittle transition temperature (DBTT) occurred at a temperature between 335° and 460°C by Charpy impact testing.

Microstructure of welded and thermal-aged low activation steel F82H IEA heat

Abstract F82H(8Cr–2WVTa steel) IEA heat was used to prepare tungsten-inert-gas (TIG) and electron-beam (EB) weld joints, followed by heat treatment at 720°C for 1 h. Hardening in the weld metal and softening in the heat-affected zone (HAZ) were detected in TIG weld joints. In EB weld joints, hardening in the weld metal was more clearly observed but HAZ softening was hardly observed. Hardness of TIG weld metal was reduced after 550°C thermal-aging, but softening of the base metal was only observed after 650°C thermal-aging. M 23 C 6 phase was the major precipitate in aged base metal and weld joints. The amount of precipitates in aged weld metal was lower than that of normalized and tempered base metal. W-rich Laves phase was also detected in aged weld metal, HAZ and base metal.

Microstructural development of austenitic stainless steels irradiated in HFIR

Abstract Microstructural developments of neutron irradiated JPCA and USPCA, which are Ti-modified austenitic stainless steels and candidate structural material for fusion reactor first wall, have been examined. The irradiation has been performed in High Flux Isotope Reactor (HFIR) at temperatures ranging between 300 and 600°C to a peak neutron fluence corresponding to approximately 34 dpa and 2500 appm helium. Microstructure of PCAs after irradiation at temperatures of 400°C and below suggests that the mutual stability of the radiation enhanced MC precipitation, and fine bubbles associated with such precipitation, has contributed to the extension of the transient regime of swelling to fluences above 34 dpa. At higher irradiation temperatures of 500°C and above, however, the conversion of some of the helium bubbles to voids has occurred at 34 dpa irradiation. MC precipitation on the radiation-induced dislocation lines is reduced at 500°C and above. This reduces the effective sink strength and increase the number of sites for helium bubble formation, which may lead to a severe reduction in the incubation regime of swelling at these temperatures. The factors controlling the stability of the MC precipitates in PCAs is discussed.

Microstructural development of PCAs irradiated in HFIR at 300 to 400° C☆

Abstract Microstructural developments were determined on solution-annealed (SA) and cold-worked (CW) JPCA and U.S. PCAs irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400°C. Irradiation produced damage levels of about 10 and 34 dpa and helium concentrations of around 580 and 2500 appm respectively. High concentrations of fine bubbles and MC precipitates, as well as Frank faulted loops, were observed in both SA and CW PCAs. Mutual stability of the MC particles and associated fine bubbles contributed to the extension of the transient regime of swelling to higher fluence. The irradiation responses of JPCA and U.S.-PCA were similar in the HFIR, despite minor compositional differences (P, B) between the two materials. Useful fusion applications of SA-PCA as well as CW-PCA in the 300 to 400° C temperature range are suggested from these data.