Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)
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A. Hamedani
Abstract
Steam generators are one of the most important components of pressurized-water reactors. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper, steady-state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, the subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&W) 19- tube once through steam generator experimental data. Thermal- hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach have been proved.
© 2020 by Walter de Gruyter Berlin/Boston
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- CALENDAR OF EVENTS · VERANSTALTUNGSKALENDER
- Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
- Appraisal of new density coefficient on integrated-nanoparticles concrete in nuclear protection
- Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation
- Developing a futuristic multi-objective optimization of the fuel management problems for the nuclear research reactors
- Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask
- Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)
- Note · Imprint
Articles in the same Issue
- CONTENTS
- CALENDAR OF EVENTS · VERANSTALTUNGSKALENDER
- Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
- Appraisal of new density coefficient on integrated-nanoparticles concrete in nuclear protection
- Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation
- Developing a futuristic multi-objective optimization of the fuel management problems for the nuclear research reactors
- Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask
- Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)
- Note · Imprint
