V. K. Shamardin

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.

Effect of ITER components manufacturing cycle on the irradiation behaviour of 316L(N)-IG steel

The main options for the manufacturing of high heat flux (HHF) components is hot isostatic pressing (HIP) using either solid pieces or powder. There was no database on the radiation behaviour of these materials, and in particular stainless steel (SS) 316L(N)-IG with ITER components manufacturing thermal cycle. Irradiation of wrought steel, powder-HIP, solid-HIP and HIPed joints has been performed within the framework of an ITER task. Specimens cut from 316L(N)-IG plate, HIP products, and solid-HIP joints were irradiated in the SM-3 reactor in Dimitrovgrad up to 4 and 10 dpa at 175°C and 265°C. The paper describes the results of post-irradiation tensile and fracture toughness tests.

Investigations of BOR-60 Structural Materials and Prospects for Further Work

Since its startup in December in 1969, the BOR-60 reactor has been used effectively for irradiation of structural and fuel materials in a wide range of dose–temperature parameters. Analysis of the actual computational-experimental parameters (irradiation temperature, damage rates) shows that the irradiation conditions are highly reproducible and can be maintained accurately.The investigations made it possible to study phenomena which are important for building reactors using domestic structural materials and to choose the optimal composition and heat treatment of the materials.New directions are indicated for scientific-research work, for improving and increasing the service life of VVÉR type reactors, and for developing new-generation structural materials for fusion reactors being designed.

Influence of combined thermomechanical treatment on impurity segregation in ferritic–martensitic and austenitic stainless steels

Abstract In this study 13Cr2MoVNb ferritic–martensitic steel (FMS) and 16Cr15Ni3MoNb austenitic stainless steel (ASS) tensile specimens were subjected to standard heat treatments and divided into two groups. Specimens in group 1 (FMS only) were aged at 400°C in a stress free and in an elastically stressed state with a tensile load (100 MPa) then doped with hydrogen in an electrolytic cell. Specimens in group 2 were subjected to cold work (up to 10%) and exposed to short-time heating at 500° for 0.5 h. All specimens were fractured at room temperature in an Auger spectrometer and Auger analysis of the fracture surfaces was performed in situ after fracturing. A noticeable increase of N and P segregation levels and a widening of the depth distribution on the grain boundary facets were observed in the FMS after aging in the stressed state. Cold-worked FMS and ASS showed a ductile dimple mode of fracture, but relatively high levels of S, P and N were observed on the dimple surfaces. We consider the origin of such effects in terms of the stressed state and plastic-deformation-enhanced segregation.

Swelling, Mechanical Properties and Structure of Austenitic High-Nickel Alloy Irradiated in a Fast Reactor

Specimens from fuel assembly wrappers and control rod cases of 0.07C-15Cr-35Ni-3Mo-B-Zr-Y alloy were investigated after irradiation in the BOR-60 and BN-600 reactors up to a maximum damage dose of about 110 dpa at temperatures between 340 to 550°C. Maximum swelling occurs at 420 to 430°C with 6% at a dose of about 80 dpa and 13.5% at a dose of 108 dpa. The maximum change in strength properties and corresponding decrease of ductility occur at an irradiation temperature of 385°C. With increasing testing temperature the ductility increases from 3-6% up to 22% at 450°C for the material irradiated at 60-80 dpa. Frank dislocation loops with an average diameter of 25 nm (concentration is 2x10 1 6 cm - 3 ), voids having diameter between 20 to 23 nm (concentration is (6-7)x10 1 4 cm - 3 ) and semicoherent needle-like precipitates with an average length of 21 nm, oriented to several directions, were observed in the alloy structure irradiated to 60 dpa at 430°C. Analysis of the experimental data was performed and the service life time of the investigated alloy was estimated for use as a material for control rod cases of the BOR-60 reactor.