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Large leak sodium-water reaction in the steam generator is one of the important events to be considered for the Liquid Metal cooled Fast Breeder Reactor (LMFBR). For the design of secondary sodium circuit components a design basis water/steam leak has been defined. For PFBR (Prototype Fast Breeder Reactor), under commissioning at Kalpakkam, India, a computer code SWEPT developed for the pressure transient analysis is used for estimating the maximum design pressure of the components. In order to study the possibility of any cliff edge effect, large scale sodium-water reaction analyses beyond the design basis leak have been carried out using SWEPT and CAST3M computer codes and the results are presented in this paper. Based on the studies it is found that simultaneous double ended guillotine failure of 15 tubes can be managed with the available design margin. Possibility of core sub assembly vibration is also studied using the CAST3M computer code. [ABSTRACT FROM AUTHOR]

In a Liquid Metal cooled Fast Breeder Reactor (LMFBR), liquid sodium is used as the reactor coolant to extract the heat energy from the core of the reactor. Liquid sodium has a chemical characteristic that makes it to violently react with water or steam. When large sodium-water reaction occurs due to defect, leakage or breakage in the steam generator tubes, high pressure waves will be generated and get transmitted throughout the secondary sodium circuit and interact with the components like sodium headers on top/bottom of steam generator, surge tank, intermediate heat exchanger etc., which also affects the safety of primary sodium circuit and core subassemblies. In LMFBR, sodium water reactions are included in the Design Basis Events. This dynamic event needs to be analyzed from the perspective of integrity of core subassemblies considering the phenomenon of fluid structure interaction. To evaluate the integrity of core subassemblies, the behavior of high pressure waves at intermediate heat exchanger is considered and the response of the subassemblies to the dynamic pressure load has been studied. The dynamic analysis of the reactor model is carried out using a three-dimensional finite element analysis in CAST3M computer code. In the code, the fluid elements are represented by volume elements and the coupling between fluid and structure by surface elements. Fluid damping is considered in the simulation. Validation of the code has been carried out using an experimental setup. Analysis results shows the maximum displacement of the primary circuit components viz., IHX, Inner vessel and subassemblies are 5.0 mm, 2.0 mm and 0.07 mm respectively for a design basis sodium water reaction event. Due to this displacement there is negligible change in the reactivity of the core. [ABSTRACT FROM AUTHOR]

Abstract: In Prototype Fast Breeder Reactor, the lower stay plate made of SS 316 LN in control plug has cut opening in three locations for the purpose of easiness of assembling the sampling tubes through the lower stay plate. These openings are then covered with appropriate cover plates bolted with the stay plate. The stay plate along with cover plate will be seeing thermal transients during reactor scrams. The thermal transient will cause the membrane stress due to global temperature increase in cover plate and bending stress due to temperature gradient across the thickness. This paper aims the creep and fatigue damages estimation of the stay plate and covers plate. The material of stay plate and cover plate is SS 316 LN. The stay plate and cover plate is analysed using CAST3M FE software developed by CAE France. A small square segment including a bolt and portion of junction between stay plate and cover plate is modeled with 8 noded, CUB8 elements. The fatigue damage for 851 number of thermal transient has been calculated. The creep damage due to the thermal stress has been evaluated. The interaction of creep-fatigue damage has also been evaluated as per RCC-MR code rule and is found within the limit. [Copyright &y& Elsevier]