Design Modification of Feed Water Ring Header Assembly of Nuclear Steam Generator by CFD And Thermo-Mechanical Analysis

Abstract

Steam generator is a shell and tube heat exchanger used for steam production in nuclear power plant to generate electricity. Nowadays, pressurized heavy water reactor steam generators are used to generate steam at high temperature and high pressure which is further used to run turbine. L & T heavy engineering, Hazira, is designing and manufacturing four Nuclear steam generators of 700 MWe for Gorakhpur Haryana Anu Vidyut Pariyojana (GHAVP) for Nuclear Power Corporation of India Limited (NPCIL). Each reactor is connected with four numbers of 700 MWe steam generator. The customer has suggested few design modifications in the existing design of feed water ring header viz. change in dimensions of ring header, elimination of expansion bellow and modification in connecting pipe. Feed water is supplied to steam generator through feed water ring header located in the annular gap of shroud and shell. Feed water ring header is provided with 92 numbers of inverted spray tubes, which ensured that in every condition feed water is fully filled with water. The objective of the present study is to analyse the effects of these modification using numerical approach. The 3D model of newly designed feed water ring header assembly with cylindrical and conical configuration was modelled in solid works and different grids were taken for analysis. Quarter symmetric model of feed water ring header assembly was taken for analysis for saving computational time and resources. As boundary conditions, mass flow inlet and pressure outlet were considered. Grid independence test was carried out to study the effects of variation in grid size. Standard k-𝜀 turbulence model was used to consider the turbulent effects. The average mass flow rate of different spray tubes was found to be 2.92 kg/s. Configuration which result out within 10% variation in mass flow rate over spray tube was considered to be the better model. The selected model has been evaluated for Thermo-Mechanical Stress individually and combined over the applied transient cycle as specified by NPCIL. The induced stress values were analysed and verified by the ASME Section III Div I NB 3200. The induced stress values were found within the limit as specified in Code.