Qualification of creep, fatigue and fracture design of PFBR components based on tests

In 500 MWe Prototype Fast Breeder Reactor, the critical out-of-core components are main vessel (MV), control plug, inner vessel, intermediate heat exchangers, steam generators (SG) and hot pipelines. The salient structural mechanics features are large size thin walled shell structures, relatively low operating pressure (< 1 MPa, except SG which operates at 17 MPa), high operating temperatures (820 K for hot pool) and large thermal gradients (AT of 150 K between hot and cold pool). These components are designed by analysis (employing numerical techniques such as FEM) to meet the requirements of French Design Code RCC-MR for the design life of 40 y. In order to ensure that the design, analysis, indigenous material and indigenous manufacturing technology comply with the design and construction code rules, tests are carried out on a few important full scale components and mockups having component features such as welds, multiaxiality and stress concentration effects under simulated loading conditions. Particularly in the domain of creep, fatigue and fracture design, a series of tests were conducted in Structural Mechanics Laboratory (SML) with the objectives of qualifying the performance of components in the reactor and the fracture assessment procedure for the FBR application and for demonstrating leak before break (LBB) argument for MV, sodium piping and SG. This paper highlights the summary of theoretical analyses that have been carried out on creep, fatigue and fracture design of critical components. Subsequently, the paper deals with a few of the experimental investigations that have been carried out essentially to qualify the creep-relaxation behaviour of IHX tube to tubesheet joint, creep rupture strength of SG tubes, fatigue and fracture assessment of SG tube bends and LBB justification of a typical full scale Tee of secondary sodium circuit.