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Abstract: Liquid metal embrittlement (LME) is defined as the brittle fracture (loss of ductility) of usually ductile materials in the presence of a liquid metal. The sensitivity to LME is likely to increase with irradiation hardening as localised stresses can promote the aggressive action of a liquid metal. To investigate the mechanical response of irradiated materials in contact with a liquid metal, an instrumented hot cell has been developed. The testing machine installed inside allows mechanical testing of active materials in liquid lead lithium under well controlled chemistry conditions. Typical mechanical tests that can be carried out are slow strain rate tests (SSRT), constant load and rising load tests at temperatures from 150°C to 500°C. In this paper the first results of the SSRT tests with EUROFER97 in argon and lead–lithium at different temperatures with different strain rates will be presented. The SSRT test method has been chosen due to the accelerated nature of the test, i.e., during straining the oxide layer will be ruptured and wetting of the sample surface by the lead–lithium melt is promoted. The results collected up till now showed no sign of LME. Tests with longer pre-exposure times and tests with irradiated samples will be carried out in the next phase. A longer pre-exposure time can enhance wetting and so the susceptibility to LME. An increase of the yield stress due to irradiation can also enhance the susceptibility to LME. [Copyright &y& Elsevier]

Abstract: The effect of post irradiation annealing on the mechanical properties and the radiation induced defect structure was investigated on stainless steel, of type AISI 304, that was irradiated up to 24dpa in the decommissioned Chooz A reactor. The material was investigated both in the as-irradiated state as well as after post irradiation annealing. In the as-irradiated specimen the typical radiation induced defects were found as well as γ′-precipitates (Ni3Si). Annealing at 400°C had almost no effect on the radiation induced defects, but annealing at 500°C resulted in the immediate unfaulting of the Frank loops. As to the mechanical properties, annealing at 400°C did not strongly affect the yield strength and the ductility of the material, although the fraction of intergranular fracture during slow strain rate tensile tests under pressurised water reactor conditions, was significantly reduced. Annealing at 500°C did reduce the yield strength and restored substantially the ductility and the strain hardening capability of the material. The microstructure investigated by transmission electron microscopy correlates to the mechanical test results. It was found that the observed defect changes after post irradiation annealing provide a reasonable explanation for the observed changes of the mechanical properties. [Copyright &y& Elsevier]

Abstract: The effects of neutron irradiation on the susceptibility to liquid metal embrittlement of two primary selected materials for MYRRHA project an accelerator driven system (ADS), was investigated by means of slow strain rate tests (SSRT). The latter were carried out at 200°C in nitrogen and in liquid Pb–Bi at a strain rate of 5×10−6 s−1. The small tensile specimens were irradiated at the BR-2 reactor in the MISTRAL irradiation rig at 200°C for 3 reactor cycles to reach a dose of about 1.50dpa. The SSR tests were carried out under poor and under dissolved oxygen conditions (∼1.5×10−12 wt% dissolved oxygen) which at this temperature will favour formation of iron and chromium oxides. Although both materials differ in structure (fcc for A316L against bcc for T91), their flow behaviour in contact with liquid lead bismuth eutectic before and after irradiation is very similar. Under these testing conditions none of them was found susceptible to liquid metal embrittlement (LME). [Copyright &y& Elsevier]

Accurate knowledge of the irradiation temperature is a key concern in irradiations of reactor pressure vessel (RPV) steel. We report results of passive temperature monitoring of RPV steel with SiC. Two un-instrumented capsules containing RPV steel blocks were irradiated in Belgian Reactor 2, at SCK CEN in Belgium. Because un-instrumented capsules were used, the irradiation conditions (gamma heating and irradiation temperature) were calculated. To have experimental verification of the irradiation conditions each capsule contained a passive silicon carbide (SiC) temperature monitor. After irradiation the resistivity and the radiation-induced swelling (lattice parameter) is measured using x-ray diffraction (XRD). These measurements were repeated after annealing treatments to determine the peak irradiation temperature. The best method to determine the peak irradiation temperature with the lowest uncertainty in this study was the resistivity technique. Although lattice parameter measurements could be improved by using material without preferential orientation. Tensile tests of the RPV blocks were compared to a large available database of RPV material with the proper sensitivity towards neutron fluence and irradiation temperature to find the irradiation temperature of the RPV steel. The data showed a discrepancy between the temperature obtained from SiC and the temperature obtained from tensile tests. The temperature differences were discussed and rationalized by finite element modelling with respect to uncertainties in the helium gap between the capsule and the RPV steel block. The aluminium holder accounted for the temperature difference and the use of a steel holder is recommended for future similar irradiations to minimize the temperature difference between SiC and tensile specimens. In-depth analysis showed that when the irradiation temperature dropped at the end of the irradiation for a considerable time, the SiC temperature monitor had a memory effect. In this case the post-irradiation resistivity analysis showed two peak irradiation temperatures with a region of constant resistivity between the two. [ABSTRACT FROM AUTHOR]