Published October 2017 | Version v1
Journal article

An advanced new fully implicit numerical method for two-phase flow subchannel analysis based on the Drift Flux Model

  • 1. Research School of Reactor, Nuclear Science and Technology Research Institute, P.O. Box 11365-3486, Tehran, Islamic Republic of (Iran, Islamic Republic of)
  • 2. Research School of Radiation Applications, Nuclear Science and Technology Research Institute, P.O. Box 11365-3486, Tehran, Islamic Republic of (Iran, Islamic Republic of)
  • 3. Department of Energy Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran, Islamic Republic of (Iran, Islamic Republic of)

Description

Highlights: • A new fully-implicit numerical scheme is presented to simulate two phase flow in rod bundle subchannels. • The Drift Flux Model (DFM) discretized equations are solved based on the Newton method. • A set of equations is derived to calculate the exact Jacobian matrix. • The BFBT-BWR is simulated at steady-state and transient for benchmarking case studies. • Transients are simulated at different time scales without any stability limit. - Abstract: In this paper, a new and efficient fully implicit numerical model is introduced using the Drift Flux Model (DFM) for simulation of the reactor core thermal-hydraulics using subchannel analysis. Although several fully implicit models have been developed to simulate two-phase flow in a single channel, but development of an efficient model for more realistic conditions, i.e. reactor core, would be useful. Based on the benefits of the Newton method, a procedure for the accurate approximation of the inverse of the Jacobian matrix and a fully implicit numerical scheme is developed. To benchmark the present model, a well-scaled 8 × 8 rod bundle was simulated and fifteen steady-state test series and two transient cases were selected to analyses the subchannel grade void distributions in NUPEC (Nuclear Power Engineering Corporation) 8 × 8 rod bundle test facility. The steady-state void distributions predicted by the model are in agreement with the measured data for a wide range of thermal-hydraulic parameters investigated. Transient calculations were also performed with different time scales and it is concluded that the model is not subject to any time step restrictions.

Availability note (English)

Available from http://dx.doi.org/10.1016/j.anucene.2017.04.046

Additional details

Identifiers

DOI
10.1016/j.anucene.2017.04.046;
PII
S0306454916309161;

Publishing Information

Journal Title
Annals of Nuclear Energy (Oxford)
Journal Volume
108
Journal Page Range
p. 351-365
ISSN
0306-4549
CODEN
ANENDJ

Optional Information

Copyright
Copyright (c) 2017 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.