I 27 31 2019 EVOLUTION OF THE MICROSTRUCTURE

ХI конференция по реакторному материаловедению 27– 31 мая 2019 года, г. Димитровград EVOLUTION OF THE MICROSTRUCTURE OF AUSTENITIC STEELS AFTER IRRADIATION AT 300 -350˚С IN REACTORS BOR-60 AND SM-3 © D. E. Markelov, V. S. Neustroev, V. K. Shamardin (SSC “RIAR”) B. Z. Margolin, A. A. Sorokin (NRC “Kurchatov Institute” – CRISM “Prometey”), 2019

Actuality Knowledge of regularities of austenitic steels microstructure changes with the damaging dose becoming higher makes it possible to reliably predict the performance of non-replaceable internals and reactor components as their lifetime grows. To ensure reliable operation and to justify the extension of the designed lifetime, data are used obtained in experiments on samples irradiation and data from studies of real items and components operated in a fast reactor for a long time. The results on microstructural changes in austenitic steels were obtained mainly after irradiating steels at 400– 550°C while irradiation at temperatures below 400°C does not generate enough data. Changes in the microstructure that occurred during low-temperature irradiation (300– 400°C) of austenitic steel Х 18 Н 10 Т [1 -3] and of other steels with similar composition provide the basis for analyzing mechanisms of radiation damage of these steels. The aim of this work is to study the microstructure of austenitic steels AISI 304 L (Х 18 Н 9) and cold worked steels AISI 316 CW (Х 16 Н(11– 14)М 3) similar in composition to Х 18 Н 10 during irradiation in the BOR-60 and SM-3 reactors up to high damage doses and to plot the dose dependences of microstructure parameters. 2

Elemental composition of test materials Test material, steel 08 Х 18 Н 10 Т Mass fraction of elements, % C Si Mn S P Cr Ni Ti ≤ 0, 08 ≤ 0, 8 ≤ 2, 0 ≤ 0, 02 ≤ 0, 035 17, 0 -19, 0 -11, 0 0, 6 Mo - AISI 316 CW 0, 05 0, 68 1, 12 0, 027 16, 6 10, 6 - 2, 2 AISI 304 L 0, 022 0, 36 1, 79 0, 0007 0, 032 18, 61 9, 86 0, 5 - 3

Material and irradiation conditions 4 irradiated "Gagarin" samples were selected for the microstructure studies. № Test material, steel Reactor Damage dose, dpa Irradiation temperature, °С 1 Х 16 Н(11– 14)М 3 (AISI 316 CW) BOR-60 93 320 -350 2 Х 16 Н(11– 14)М 3 (AISI 316 CW) BOR-60 125 320 -350 3 Х 16 Н(11– 14)М 3 (AISI 316 CW) SM-3 4, 5 ~300 4 Х 18 Н 9 (AISI 304 L) SM-3 4, 2 ~300 The "Gagarin" sample is a cylindrical sample for mechanical tensile tests with a working section length of 15 mm and a working section diameter of 3 mm. 4

Microstructure study methods and sample preparation 1. 2. TEM samples were remotely cut off in the hot cell from “Gagarin” samples and their thickness was reduced to 0. 3 mm with a grinding machine. Then the discs with a diameter of 3 mm were transported to a protective glove box where they were brought to the thickness of 0. 08 mm using sand paper, and for finally thinning unit Tenupol 5 was used. The microstructure of the samples was examined using a transmission electron microscope at the accelerating voltage of 200 k. V. 5

Main Results General view of the microstructure of steel AISI 316 CW sample after irradiation in the BOR-60 reactor up to a damage dose of 93 dpa 6

Main Results General view of the microstructure of steel AISI 316 CW sample after irradiation in the SM-3 reactor up to a damage dose of 4, 5 dpa 7

Alpha-phase in steel AISI 316 CW after irradiation in the BOR-60 reactor up to damage doses of 93 and 125 dpa In steel AISI 316 CW samples irradiated in the SM-3 reactor alpha -phase was not found. γ-Fe α-Fe 8

G-phase in steel AISI 316 CW after irradiation in the BOR-60 reactor up to damage doses of 93 and 125 dpa G-phase particles in matrix γ-Fe The average particle size for 93 dpa – 6 nm, for 125 dpa – 8 nm Me 6 Ni 16 Si 7 α-Fe G-phase particles in α-Fe The average particle size for 93 dpa – 4 nm, for 125 dpa – 5 nm The particle size of a sample from the SM-3 reactor is 5 nm. 9

Dislocation structure of AISI 316 CW steel after irradiation in the reactors BOR-60 and SM-3. Frank loops. SM-3 4, 5 dpa BOR-60 93 and 125 dpa The average loop size is ~8 nm 10

Main Results General view of the microstructure of steel AISI 304 L sample irradiated in the SM-3 reactor up to a damage dose of 4, 2 dpa 11

Alpha-phase in AISI 304 L steel after irradiation in the SM-3 reactor up to a dose of 4, 3 dpa The equivalent diameter of marked areas was about 2 μm. The α-Fe has a bcc lattice with the parameter а=0, 2866 nm. Elemental analysis along the scan line showed that the alpha phase is depleted in nickel and enriched in chromium. 12

G-phase and dislocation structure of steel AISI 304 L sample after irradiation in the SM-3 reactor up to a dose of 4, 3 dpa G-phase particles in matrix γ-Fe The average particle size is 5 nm G-phase particles in α-Fe The average particle size is 3 nm Dark-field image of Frank loops The average loop size is ~5 nm Light-field image of the dislocation structure 13

Microstructural parameters of the studied samples Steel Reactor grade AISI Damage dose, dpa Frank loops in austenitic matrix Finely dispersed phase particles Size Lavg, nm ρ, 22 10 m– 3 Davg in α-Fe, nm Ρavg in α-Fe, 1021 m– 3 Davg in -Fe, nm ρavg in -Fe, 1021 m– 3 304 4, 2 4, 9 19 3, 1 25 5, 0 0, 2 316 4, 5 8, 4 8 – – 4, 5 9 316 93 7, 7 7 4, 3 76 5, 8 6 316 125 8, 4 5 5, 4 96 7, 7 5 SM-3 BOR-60 The measurement error of the average diameter and length is 10%, and the concentration of 30% 14

Conclusion BOR-60 Dependence of the average size of dislocation loops and G-phase particles on the damage dose in γ-Fe austenitic matrix of AISI 316 CW steel: - G-phase [1]; - G-phase; - dislocation loops [1]; - dislocation loops. Dependence of the average size of dislocation loops and G-phase particles on the damage dose in γ-Fe austenitic matrix of AISI 304 L steel and Х 18 Н 10 Т: - G-phase [1]; - G-phase (Х 18 Н 10 Т); - G-phase (AISI 304 L) SM-3; - dislocation loops [1]; - dislocation loops (Х 18 Н 10 Т); - dislocation loops (AISI 304 L) SM-3. [1] A. É. Renault, C. Pokor, J. Garnier, J. Malaplate. Microstructure and grain boundary chemistry evolution in austenitic stainless steels irradiated in the BOR-60 reactor up to 120 DPA // 14 th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems, At Virginia Beach (USA), August 2009 15

Conclusion Dependence of dislocation loops concentration on the damage dose in γ-Fe austenitic matrix of AISI 316 CW steel : ― [1]; ― this work; Dependence of dislocation loops concentration on the damage dose in γ-Fe austenitic matrix of AISI 304 L steel and Х 18 Н 10 Т: ― [1] ― steel Х 18 Н 10 Т ― steel 304 L SM-3 [1] A. É. Renault, C. Pokor, J. Garnier, J. Malaplate. Microstructure and grain boundary chemistry evolution in austenitic stainless steels irradiated in the BOR-60 reactor up to 120 DPA // 14 th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems, At Virginia Beach (USA), August 2009 16

Conclusion The results on the quantitative characteristics of the microstructural parameters of steels AISI 316 CW and AISI 304 L samples after irradiation in the SM-3 and BOR-60 reactors were obtained. These data were compared with the data of French researchers who previously studied the same steels after irradiation [1]. 1. Dislocation Frank loops, fine dispersed G-phase particles (in matrix γ-Fe) and α-Fe phase (except for the sample of 316 CW steel irradiated in the SM-3 reactor) were found in all 4 samples. 2. The quantitative characteristics (diameter and concentration) of the dislocation structure and the fine dispersed phase in austenitic matrix of both steels coincide with the results of French studies [1] and complement well the dose dependencies of these characteristics for steels AISI 316 CW, AISI 304 L and Х 18 Н 10 Т, which suggests the authenticity of obtained quantitative estimates. 3. Data on Х 18 Н 10 Т steel were added to the dose dependence of dislocation and phase structure of AISI 304 L, and the subsequent comparison with the result of French researchers suggests that all dependencies of structure parameters have reached certain saturation levels at high damage doses achieved in the BOR-60 reactor. 17

Conclusion 4. Small α-Fe grains were found in the microstructure of steels AISI 316 CW and AISI 304 L and the size and concentration of the found fine dispersed phase were quantified. The concentration of particles was 12– 15 times higher in α-Fe than in matrix after irradiation in the BOR-60 reactor up to a high damage dose. The concentration of particles was 100 times higher in α-Fe than in matrix after irradiation in the SM-3 reactor up to a damage dose of about 4 dpa. Such a high concentration of the fine dispersed phase can significantly affect the mechanical properties of austenitic steels by various hardening mechanisms and, in particular, on austenitic steels containing alpha-phase grains. The results on steels AISI 316 CW and AISI 304 L can be used to study the microstructure evolution mechanisms of austenitic steels irradiated to high damage doses, as well as the mechanism of irradiation-induced stress corrosion of austenitic steel Х 18 Н 10 Т and to develop a model for predicting the lifetime of water-cooled power reactors internals in terms of corrosion cracking. 18

Спасибо за внимание! Thank you for your attention! Junior Researcher D. E. Markelov