T.M. Bulanova

Structural materials for fusion power reactors : the RF R&D activities

Recent progress in the RF low activation structural materials R&D road map towards DEMO via the FBR tests (BOR-60, BN-600, BN-800) and the TBM tests in ITER is overviewed. The properties of the RAFMS RUSFER-EK-181 (Fe?12Cr?2W?Ta?V?B?C) and the V?4Ti?4Cr alloys are presented. The next important steps include further studies on the influence of high dose and high-temperature irradiation on the properties of base structural materials and joints. Activation, transmutation and radiation damage of the materials in BN-600 and DEMO-RF (Kurchatov Institute project) neutron spectra are calculated. The results of the application of the internal friction (ultrasonic) non-destructive method to research the DBTT are in good agreement with the results of the destructive impact method. The important influence of boron on the heat resistance of materials and the He concentration level under irradiation are calculated. The new special regimes of the heat treatments of the alloys are suggested to widen the temperature windows of the applications. The results of the BOR-60 examinations of RUSFER-EK-181 (irradiation temperature 320?340??C and doses up to 15?dpa) are presented. The BN-600 projects for the high dose and high-temperature irradiation tests of manufactured alloys are presented.

Mechanical properties and microstructure of advanced ferritic–martensitic steels used under high dose neutron irradiation

Abstract Some results of the study of mechanical properties and structure of ferritic–martensitic chromium steels with 13% and 9% chromium, irradiated in the BOR-60 reactor up to different damage doses are presented in this report. Results concerning the behaviour of commercial steels, containing to molybdenum, vanadium and niobium, and developed for the use in fusion reactors, are compared to low-activation steels in which W and Ta replaced Mo and Nb. It is shown that after irradiation to the dose of ∼10 dpa at 400°C 0.1C–9Cr–1W, V, Ta steels are prone to lower embrittlement as deduced from fracture surface observations of tensile specimens. Peculiarities of fine structure and fracture mode, composition and precipitation reactions in steels during irradiation are discussed.

Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a δ-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 °C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 °C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

Evolution of the mechanical properties and microstructure of ferritic–martensitic steels irradiated in the BOR-60 reactor

Abstract The effect of neutron irradiation on mechanical properties of low-activation ferritic–martensitic (FM) steels 0.1C–9Cr–1W, V, Ta, B and 0.1C–12Cr–2W, V, Ti, B is studied under tension at temperatures of 330–540 °C and doses of 50 dpa. Steel 0.1C–13Cr–Mo, V, Nb, B was chosen for comparison. At irradiation temperatures of 330–340 °C, the radiation hardening of steel with 9%Cr achieves saturation at a dose of 10 dpa. In this case as compared to steels with 12%Cr, the fracture surface is characterized as ductile without cleavage traces. At irradiation temperatures higher than 420 °C, there is no difference in the behavior of the materials under investigation. The data on radiation creep obtained by direct measurement and from the profilometry data satisfy a model e / σ =B 0 +D S , when B0 and D have the values typical for steels of FM type.

Metallurgical aspects of possibility of 9–12% chromium steel application as a structural material for first wall and blanket of fusion reactors

Abstract Steels containing 9–12% Cr are considered to be candidate structural materials for the first wall and blanket of a fusion reactor at the operation temperature up to 650°C. The optimal structure, phase composition and the specific chemical composition of the steels ensure their high heat resistance, yield strength and ductility as well as adequate thermophysical properties. The susceptibility of chromium steels for low temperature irradiation embrittlement can be influenced by changing their structural state via alloying, heat treatment and method of melting. Steels having a uniform martensite structure are less susceptible to irradiation conditions and have more stable tensile properties as compared to steels having δ-ferrite in their structures.

Change in the properties of FeCrNi and FeCrMn austenitic steels under mixed and fast neutron irradiation

Abstract Detailed investigations are performed on mechanical properties, swelling and structure of different types of FeCrNi and FeCrMn austenitic stainless steels irradiated in the SM-2 high-flux research reactor and BOR-60 fast reactor. Steel irradiation temperatures are ranging from 100 up to 800°C and the maximum achieved level of damage doses is 60 dpa for FeCrMn steel (with 4–5% of Ni) and 30 dpa for steels of the C12Cr20MnWT type. Presented are dose dependencies of swelling and mechanical properties of FeCrNi and FeCrMn steels. It is shown that at temperatures below 530°C the investigated FeCrMn steel systems are less susceptible to swelling as compared to FeCrNi ones. FeCrMn steels showed a lower value of irradiation embrittlement after irradiation in the mixed spectrum at temperatures from 100 up to 400°C and much higher embrittlement after irradiation from 350 up to 400°C in the fast spectrum in comparison with FeCrNi steels. Higher hardening rate of FeCrMn steels after their irradiation in BOR-60 is attributed to the presence of dislocation loops and defects of high density in the structure. The structural change features in FeCrMn steels under irradiation are considered taking into account austenite stabilization in the initial state.

Study of Fe-12Cr-20Mn-W-C austenitic steels irradiated in the SM-2 reactor

Abstract A comparison has been made between the mechanical properties and swelling of austenitic stainless steels EP-838 (Fe-Cr-Mn) and 316SS (Fe-Cr-Ni) irradiated in the mixed-neutron spectrum of the SM-2 reactor in the temperature range 400–800°C (every 100°C) to 16 dpa dose with 1000 and 3000 appm helium generation correspondingly, determined by nickel content. EP-838 exhibited less susceptibility to void swelling and radiation hardening. Fe-12Cr-20Mn-W-0.1C steel without nickel irradiated at 100°C to 21 dpa exhibited significant radiation hardening accompanied by α-phase formation in the steel structure.

Mechanical characteristics and swelling of austenitic Fe-Cr-Mn steels irradiated in the SM-2 and BOR-60 reactors

Abstract Three types of austenitic Fe-Cr-Mn stainless steels were irradiated simultaneously with Fe-Cr-Ni austenitic steel at temperatures from 400 to 800°C in the mixed spectrum of the high flux SM-2 reactor to 10 dpa and 700 appm of He and in the BOR-60 reactor to 60 dpa without He generation. The paper presents the swelling and mechanical properties of steels irradiated in the BOR-60 and SM-2 as a function of the concentration of transmuted He and the value of atomic displacement.

Irradiation effect on mechanical properties and structure of nickel-free ferrum-chromium-manganese steel Fe-12Cr-20Mn-W-C

Abstract In the present paper there will be given the investigation results of mechanical properties, swelling and structure of the home-made steel 0.1C-12Cr-20Mn-1W under austenitization condition after irradiation in the BOR-60 reactor at 350 and 400°C by the doses 8 and 10 dpa, respectively, and American steel 0.25C-12Cr-20Mn-2W-Ti with stable austenite in the state of 20% cold work.