Presentation on theme: "The International Workshop Influence of atomic displacement rate on radiation-induced ageing of power reactor components: Experimental and modeling October."— Presentation transcript:
The International Workshop Influence of atomic displacement rate on radiation-induced ageing of power reactor components: Experimental and modeling October 3 – 7, 2005, Ulyanovsk Microstructure and mechanical properties of austenitic stainless steel 12Х18Н9Т neutron irradiated at extremely low dose rates S.I. Porollo, A.M. Dvoriashin, Y.V. Konobeev, A.A. Ivanov, S.V. Shulepin State Scientific Center of Russian Federation, The Institute of Physics and Power Engineering, Obninsk, Russia F.A. Garner Pacific Northwest National Laboratory, Richland, USA
Introduction Internals of Russian power reactors (WWER-440, WWER-1000, BN-600) are made of type X18H9 or X18H9T (18Cr-9Ni or 18Cr-10Ni-Тi) austenitic stainless steels. In Western PWRs the AISI 304 steel (with the chemical composition similar to 18Cr-9Ni steel) is used for this purpose. Now the problem of reactor life-time prolongation over design is very important for the majority of Russian and Western reactors of this type. For substantiation of reactor life-time prolongation it is important to have a reliable information on how properties of structural materials of internals change with increasing neutron dose. In practice this question is more often solved by using surveillance samples, which are being located at the fast reactor core periphery and, hence, are irradiated with more high neutron fluxes. Investigations of such samples allow to judge about change of material properties of internals. Recently it became clear, that data on surveillance samples are insufficient, mainly due to influence of neutron flux on many irradiation characteristics of steels (intensity of irradiation). The number of experiments carried out on the study of influence of dose rate on swelling and microstructure evolution in austenitic stainless steels is small. This is related to impossibility to obtain data for sufficiently wide range of dose rates at approximately the same dose only for reactor core assemblies. For this purpose it is necessary to examine assemblies irradiated at the core periphery or another components located even more far from the core. In the present paper results of investigating swelling, microstructure and mechanical properties of Russian 12Х18Н9Т (0.12C-18Cr-9Ni-Ti) austenitic stainless steel irradiated as a structural material of the BР-10 fast reactor first vessel with dose of 0.64 dpa at extremely low displacement rate of 1.9 10 -9 dpa/s are presented.
Materials and techniques of investigation. Samples for investigation of microstructure and mechanical properties were cut out from the first vessel of BR-10 reactor, which was replaced by a new vessel in 1979. The first vessel was made from half-finished product chiseled on various diameters with the maximum outside diameter of 535 mm and the total length above 4 m. At the location of fuel assemblies the vessel has the outside diameter of 366 mm and wall thickness of 7 mm. The vessel material is 12X18H9T austenitic stainless steel in solution treated condition. The nominal chemical composition of the steel is (wt. %): С 0.12; Si 0.8; Mn 2.0; Cr - 17 20; Ni - 8 11; Ti 0.8. The first vessel was in operation during 20 years since July, 1959 till October, 1979 with three cycle runs, two of them with PuO 2 fuel and one with UC fuel. The total time of the reactor operation on capacity equals 3930 days or 2562.6 effective days. The total neutron fluence accumulated by the vessel at the core midplane is equal to 8.44 10 26 n/m 2 that corresponds to the dose of 33.1 dpa (NRT). At its inner side the vessel was in contact with the sodium coolant flowing from reactor bottom to top, but at the outer side it was in contact with air in the gap between the vessel and an insurance jacket.
Irradiation conditions for samples - templates cut out from the BR-10 reactor vessel
Materials and techniques of investigation. (continuance) Using a remote milling machine, strips with the cross section 10 mm 2 mm or 7 mm 2 mm were cut out from these templates in an axial direction. Then from the strips half-finished products of TEM-specimens and samples for measurements of short-term mechanical properties of the vessel structural material were prepared. Mechanical properties were measured for flat samples having the gauge length of 12 mm and cross section of 2 mm 2 mm. The tests were carried out at the temperatures of 25 and 350 С. The test temperature of 350 С equals the inlet coolant temperature in the core for the most part of reactor operation time and was approximately equal to the reactor vessel temperature at the basket bottom level. The initial strain rate was 1.4 10 -3 s -1. At each temperature 3 4 samples were tested with averaging of obtained results. TEM-specimens in the form of disks of 3 mm in diameter with a perforated central hole were prepared by a standard technique using two- jet-polishing STRUERS device. Microstructural investigations were carried out at an accelerating voltage of 100 kV using a JEM-100CX electron microscope.
Experimental results Microstructure of the unirradiated steel 12Х18Н9Т from the template cut out from the upper flange of the BR-10 reactor first vessel. 1 m
Experimental results (continuance) Dislocations and TiC- precipitates in unirradiated steel 12Х18Н9Т (cross section of the BR-10 reactor vessel at the level of the upper flange) 0.5 m
Experimental results (continuance) Dislocation loops in neutron irradiated 12Х18Н9Т steel (cross section of the BR-10 reactor vessel at a level of the basket bottom): left-hand - general view, right-hand - a dislocation loop cluster along sub-grain boundaries 0.5 m 100 nm
Experimental results (continuance) Voids in neutron irradiated 12Х18Н9Т steel (cross section of theBR-10 reactor vessel at a level of the basket bottom): left-hand - large voids on sub-grain boundaries, right-hand - spatial distribution of smaller voids. 50 nm100 nm
Experimental results (continuance) Results of mechanical tests of flat samples from steel 12Х18Н9Т, cut out from the first vessel of the BR-10 reactor.
Discussion The cross section of the BR-10 vessel at the basket bottom level is most remote from the reactor core. The dose of 0.64 dpa in this cross section has been accumulated in the vessel steel for 2563 eff. days or for 3930 days of reactor operation. Hence, maximum dose rate in this cross section was equal to 0.64 dpa/2.2 10 8 s=2,9 10 -9 dpa/s with the average dose rate of 1.9 10 -9 dpa/s. For comparison, the dose rate at the center of BR-10 core equals 3.5 10 -7 dpa/s. In the BN-600 fast reactor core this rate even higher and is equal to 1.8 10 -6 dpa/s. Internals of power reactors (BN-600, WWER-440, WWER-1000) operate at considerably lower dose rates.
Discussion (continuance) Doses accumulated in various internals during 30 years of operation and dose rates
Discussion (continuance) Dependence of steel 12Х18Н9Т swelling on dose. Light circles - wrappers of fuel assemblies and fuel pin claddings of BR-10 reactor, black circle – reactor first vessel.
The International Workshop Influence of atomic displacement rate on radiation-induced ageing of power reactor components: Experimental and modeling October 3 – 7, 2005, Ulyanovsk Conclusions 1. Neutron irradiation under conditions investigated resulted in a significant reduction of swelling incubation dose up to < 1 dpa as compared with 4-7 dpa incubation dose of swelling in cladding and wrapper materials of BR-10 reactor (typical dose rate of 1.3 10 -7 dpa/s). 2. The spatial distribution of dislocation loops and voids in the irradiated steel 12Х18Н9Т is non-uniform and is caused by initial non-uniformity of dislocation structure. 3. Irradiation resulted in an essential hardening (the yield strength measured at room temperature increased by 359 MPa) accompanied with a ductility loss.