NUMERICAL VERIFICATION OF THE RELAP-7 CORE CHANNEL SINGLE-PHASE MODEL

摘要

The RELAP-7 code is the next generation of nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). All the physics in RELAP-7 are fully coupled and the errors resulted from the traditional operator-splitting approach are eliminated. By using 2nd order methods in both time and space and eliminating operator-splitting errors, the numerical error of RELAP-7 can be minimized. Numerical verification is the process to verify the convergence orders of numerical methods. It is an essential part of the modern verification and validation process. The core channel component in RELAP-7 is designed to simulate coolant flow as well as the conjugate heat transfer between coolant flow and the fuel rod. A special treatment at fuel centerline to avoid numerical singularity for the cylindrical heat conduction in the continuous finite element mesh is discussed. One steady state test case and one fast power up transient test case are utilized for the numerical verification of the core channel model with single-phase flow. Analytical solution for the fuel pin temperature and figures of merit such as peak clad temperature and peak fuel temperature are used to quantify numerical errors. These cases prove that the mass and energy are well conserved and 2nd order convergence rates for both time and space are achieved in the core channel model.