Creep Response Of Fuel Channels In Candu Nuclear Reactors Computer Code Cdepth Version 2

Author : Ahmet Durmayaz
Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (85 download)

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Download or read book A Loss-of-coolant Accident Analysis for Near Stagnation Flow in the Horizontal Fuel Channel of Candu Reactor written by Ahmet Durmayaz and published by . This book was released on 1995 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: In this thesis, a two-phase thermalhydraulic model has been developed inorder to investigate the near stagnation flow phenomena in the horizontal fuelchannel of a CANDU nuclear reactor following a postulated small break _loss-_of-çoolant-accident (LOCA) in the inlet feeder or inlet header. In this model, the unequal velocity unequal temperature (UVUT) modelis used in which the velocities and enthalpies (hence temperatures) for both liquidand vapor phases, void fraction and a common dynamic pressure are computed. Stratified smooth and dispersed bubbly flows are considered together for the near stagnation flow and dispersed bubbly flow is considered for the normal operating conditions of a CANDU reactor as the flow regimes. Geometrical configuration of a typically CANDU reactor fuel channel is considered. Thermodynamic andtransport properties of heavy water are also considered as functions of pressure and temperature. In the derivations of the conservation equations, all of the terms are area and time (ensemble) averaged. A fully implicit discretization method and a staggered grid control volume approach are applied in order to obtain an algebraic equation set from the governing differential conservation equations. Hvbrid scheme. in which central difference and upwind schemes are considered by turns depending on the value of Peclet number, is used while deriving the discretization equations. To solve the near stagnation flow problem by using this model, the IPSA algorithm (Inter-Phase Slip Algorithm) which was developed to solve the multi-phase flows originally is modified to give a solution for a horizontal channel which is full of single phase fluid in some part whereas it is full of two- phase fluid in the other part. A computer code that is called as the ANESTA ( Analysis for NEar STAgnation flow in the fuel channel of a CANDU reactor ) code has been developed. The model mentioned above is applied to both normal operating conditions and the near stagnation flow (LOCA) conditions by this code. A set of approximate functions for the fast calculation of the thermodynamic properties of heavy water (D20 ) has also been derived. For this aim, the data given in the steam tables for heavy water thermodynamic properties have been accurately and successfully fitted to the curves as functions of pressure and temperature in order to obtain some simple functions by usingthe least-square method.