Tensile behaviour of 9Cr martensitic steels irradiated up to 20 dpa in a spallation environment

International audienceTENSILE BEHAVIOUR OF 9Cr MARTENSITIC STEELS IRRADIATED UP TO 20 DPA IN A SPALLATION ENVIRONMENTJ. Henry1), X. Averty1), Y. Dai1), J-P. Pizzanelli1)1) Commissariat a L-Energie Atomique, CEA-Saclay, DEN-DMN 91191 Gif-sur-Yvette, France2) Paul Scherrer Institut, CH-5232 Villigen PSI, SwitzerlandTensile samples of 9Cr 1Mo (EM10) and mod 9Cr 1Mo (T91) martensitic steels were irradiated in the SINQ target (STIP 2 irradiation) to a maximum dose of 20 dpa with the mean irradiation temperature ranging from about 115DC to 350DC. For EM 10, two geometries were used for the tensile samples, the S (gauge section dimensions were 5 mm long, 1 mm wide, 0.4 mm thick) and L (5 mm long, 1.5 mm wide and 0.75 mm thick) geometries while for T91 only S type specimens were irradiated. Tensile tests were conducted at Room Temperature (RT) and at 250DC and for a few samples at 350DC.At RT, both EM10 and T91 samples showed considerable hardening and loss of ductility as classically observed for martensitic steels irradiated below 350DC. Moreover, above about 15 dpa, the samples exhibited a total loss of ductility and some specimens irradiated to 19-20 dpa broke in the elastic regime. Likewise, in the case of the two samples tested at 350DC and which had been irradiated to the maximum dose, almost no plastic deformation was measured. For the specimens tested at 250DC such a drastic embrittlement was not observed, however the maximum exposure of the specimens tested at this temperature was about 16 dpa. In addition, one EM10 and one T91 specimen irradiated to 20 dpa were annealed for 1h at 750DC before being tested at RT. Following the heat treatment both materials regained significant uniform elongation while still exhibiting a large increase in yield stress compared to the unirradiated materials

Bending tests on T91 samples implanted with 0.25 at.% helium: Experiments and mechanical analysis

In order to investigate helium effects on the fracture properties of martensitic mod 9Cr-1Mo (T91) steel, miniature Charpy specimens were implanted at 250 degrees C in the notch region to 0.25 at.% helium using a degraded 34 MeV He-3 ion beam and subsequently submitted to static bending tests at room temperature. For the six implanted specimens, a 'pop-in' phenomenon, which is an arrested unstable crack extension, was systematically recorded during testing. In the implanted zones of the samples, the fracture mode was fully brittle with both intergranular and cleavage fracture, whereas for unimplanted samples tested at -170 degrees C, the fracture mode was found to be 100% cleavage. Finite element simulations of the tests performed on unimplanted and implanted specimens were also carried out to determine stress and strain fields at the onset of crack propagation. Based on these computations, the fracture toughness of implanted T91 was tentatively evaluated using the Beremin model of the local approach to brittle fracture. (c) 2006 Elsevier B.V. All rights reserved

Effect of a high helium content on the flow and fracture properties of a 9Cr martensitic steel

An experimental characterization was conducted of helium effects on the mechanical properties of a 9Cr martensitic steel. Six sub-size Charpy samples were implanted in the notch region at 250 degrees C with 0.25 at.% helium and subsequently tested in 3-point bending at room temperature. Brittle fracture mode (cleavage and intergranular fracture) was systematically observed in the implanted zones of the samples. Finite element calculations of the tests, using as input the tensile properties measured on a helium loaded sample, were performed in order to determine the fracture stress at the onset of brittle crack propagation. Preliminary TEM investigations of the implantation-induced microstructure revealed a high density of small helium bubbles. (c) 2007 Elsevier B.V. All rights reserved

Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel

International audienceThe hardening and the embrittlement under neutron irradiation of a RPV steel considering three different coarse grained microstructures (bainite, bainite-martensite and martensite) have been investigated. Bainite and martensite have a comparable irradiation induced temperature transition shifts but hardening is smaller for martensite than for bainite. Bainite appears to be less sensitive to non-hardening embrittlement (NHE) than martensite. The onset of brittle intergranular fracture after irradiation for the martensitic microstructure is likely the origin of this difference of sensitivity to NHE. However, the bainitic-martensitic microstructure has the largest irradiation induced shift and is the most sensitive to NHE amongst all the tested microstructures. These results show that the microstructure is an important parameter to control the irradiation induced non hardening embrittlement and the enhancement of the brittle intergranular fracture in reactor pressure vessel steels

A model to describe the anisotropic viscoplastic mechanical behavior of fresh and irradiated Zircaloy-4 fuel claddings under RIA loading conditions

International audienceThis paper presents a unified phenomenological model to describe the anisotropic viscoplastic mechanical behavior of cold-worked stress relieved (CWSR) Zircaloy-4 fuel claddings submitted to reactivity initiated accident (RIA) loading conditions. The model relies on a multiplicative viscoplastic formulation and reproduces strain hardening, strain rate sensitivity and plastic anisotropy of the material. It includes temperature, fluence and irradiation conditions dependences within RIA typical ranges. Model parameters have been tuned using axial tensile, hoop tensile and closed-end internal pressurization tests results essentially obtained from the PROMETRA program, dedicated to the study of zirconium alloys under RIA loading conditions. Once calibrated, the model provides a reliable description of the mechanical behavior of the fresh and irradiated (fluence up to View the MathML source or burnup up to 64 GWd/tU) material within large temperature (from 20 °C up to 1100 °C) and strain rate ranges (from View the MathML source up to View the MathML source), representative of the RIA spectrum. Finally, the model is used for the finite element analysis of the hoop tensile tests performed within the PROMETRA program

Mechanical properties of zircaloy-4 PWR fuel cladding with burnup 54-64MWd/kgU and implications for RIA behavior

The PROMETRA material testing program is a support program related to the study of high burnup fuel rod behavior under Reactivity Initiated Accidents (RIA) and to the interpretation of the CABRI REP-Na RIA test results. Hoop and axial tensile tests have been performed on fresh and irradiated Zircaloy-4 cladding alloy first at CEA Grenoble hot labs and now at CEA Saclay in order to assess the cladding mechanical behavior during RIA transients. Efforts have been continuously carried out in order to improve the prototipicallity of the tests for RIA studies involving new specimens and new testing techniques. The corrosion level of irradiated specimens reached up to 130 μm of oxide layer thickness. The influence of in-pile oxide layer spallation has also been addressed. High strain-rate material properties of irradiated Zircaloy-4 and the consequences of hydride embrittlement can be derived from the PROMETRA program. Copyright © 2005 by ASTM International

Irradiation creep of SA 304L and CW 316 stainless steels: Mechanical behaviour and microstructural aspects. Part I: Experimental results

International audienceSolution annealed 304L (SA 304L) and cold work 316 (CW 316) austenitic stainless steel irradiation creep behaviour have been studied thoroughly. Irradiations were carried out in fast breeder reactors BOR-60 (at 330 degrees C, up to 120 dpa) and EBR-II (at 375 degrees C, up to 10.5 dpa), and in the OSIRIS mixed spectrum reactor (at 330 degrees C, up to 9.8 dpa). After an incubation threshold, the irradiation creep of the austenitic stainless steels is linear in stress and in dose. Creep appears to be athermal in this temperature range. A significant difference in the behaviour is measured between the creep of SA 304L and CW 316. In order to study the anisotropy of loop population, which would be the signature of a possible stress induced preferential absorption (SIPA) mechanism for irradiation creep, special attention was given to the measurement of anisotropy of loop distribution between the four families. The anisotropy induced by an applied stress has been shown to be in the range of the statistical scatter in the situation where no stress is applied. TEM microstructural analyses performed on this sample show slight difference between the microstructure of specimens deformed under irradiation and the microstructure of specimens irradiated without stress under the same irradiation conditions. (C) 2011 Elsevier B.V. All rights reserve

Deformation under irradiation of 304L and 316 austenitic stainless steels

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