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Ragusa, J.

Publications and source records attributed to Ragusa, J..

RELAP-7 Numerical Stabilization: Entropy Viscosity Method

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL's modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5's capability and extends the analysis capability for all reactor system simulation scenarios. RELAP-7 utilizes a single phase and a novel seven-equation two-phase flow models as described in the RELAP-7 Theory Manual (INL/EXT-14-31366). The basic equation systems are hyperbolic, which generally require some type of stabilization (or artificial viscosity) to capture nonlinear discontinuities and to suppress advection-caused oscillations. This report documents one of the available options for this stabilization in RELAP-7 -- a new and novel approach known as the entropy viscosity method. Because the code is an ongoing development effort in which the physical sub models, numerics, and coding are evolving, so too must the specific details of the entropy viscosity stabilization method. Here the fundamentals of the method in their current state are presented.

42 ENGINEERING↗

Multiphysics simulations for LWR analysis

Accurate prediction of the neutron and temperature distributions within an operating nuclear reactor requires the solution of multiple coupled physics equations. In a light water reactor (LWR), there is a very strong coupling between the power distribution (described by the radiation transport equation) and the temperature and density distributions (described by a thermal diffusion equation in combination with a fluid flow model). This study aims to begin to quantify the impact of such feedback mechanisms as well as identify numerical difficulties associated with such multiphysics problems. A description of the multiphysics model and current solution strategy within the Exnihilo code package for coupling between 3-D radiation transport and 3-D heat transfer is given. Numerical results detailing the effects of varying the nature of the coupling and the impact of mesh refinement for a representative 3x3 pressurized water reactor (PWR) 'mini-assembly' are presented. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗