This article presents the parametric study of pull-out radius of steam generator shell nozzle junction for fast breeder reactor. An efficient finite element modeling for shell nozzle junction has been presented in which shell elements are employed to idealize the whole region. In shell nozzle junction, pull-out region is an important part, so that region is taken and studied with different radius of curvature. The pull-out radius varies from 40 to 80 mm. Five models are taken into consideration and each with different radius of curvature. The optimized stress values for all the models are presented here.
The present power scenario in India is characterized by major contribution from fossil power stations. Under these circumstances, introduction of Fast Breeder Reactors (FBR) on commercial scale is possible only if their economic competitiveness is demonstrated viz-a-viz fossil power stations. For Sodium Fast Reactors, the Steam Generator is known to be a key component in terms of competitiveness and plant availability with safety implications through the sodium/water reaction. One of the critical locations in SG is the shell nozzle junction. This junction is subjected to an end bending moment and internal pressure. Since the shell nozzle junction is the critical location of SG a double-ended guillotine rupture will result in leakage of large quantity of sodium, which is not desirable. Hence safety requirements demand that LBB criteria with assumed initial flaw have to be demonstrated. For all these analysis, the basic requirement is to predict the state of stress precisely in the shell nozzle junction under various loading conditions. An efficient finite element modeling for shell nozzle junction has been presented in which shell elements are employed to idealize the whole region. The stress and deformation values are presented and compare with experimental study. Based on these analysis, the crack is initiated at the intersection of straight vertical shell and the cone ie. at the pullout region. These results are used for the analysis of leak before break concept.
The steam generators (SG) control the capacity factor of sodium cooled fast reactor plants and hence they are designed with high reliability. One of the strategies to enhance the reliability is to demonstrate leak before break (LBB) justification. LBB analysis is reported for 500 MWe sodium cooled fast reactor (SFR) in this paper. The material of construction is modified 9 Cr-1 Mo and the critical location is the shell nozzle junction. The initial surface crack is postulated at the critical locations at the shell nozzle junction. The critical crack length is computed by adopting the philosophy of CEGB-R6 procedure, for which Jintegral is computed by finite element method. For determining the detectable through-wall crack length, the crack opening area is also determined by finite element method. Finally, it is demonstrated that it is possible to justify leak before break argument for SFR SG with adequate margins. The required material properties are extracted from French Design Code RCC-MR-2002 and validated with tests on plates.
The present work presents plastic limit load solutions for branch pipe tee connection under internal pressure and in plane bending moment based on detailed three dimensional finite element limit analysis using elastic -perfectly plastic materials. To assure reliability of the FE limit loads, modeling issues are addressed first, such as the effect of kinematic boundary conditions and branch pipe geometries on the FE limit loads. Several models of branch pipe tee connection are meshed with shell elements and submitted to internal pressure with end in plane bending moment. Results are compared with lower and upper bound analytical solutions and experimental results reported in the literature. Computations with 20 noded elements are proposed to validate this analysis.
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