skip to main content

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

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 orders of numerical methods. It is an important part of modern verification and validation process. The core channel component in RELAP-7 is designed to simulate coolant flow as well as the conjugated 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 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 define numerical errors. These cases prove that the mass and energy are well conserved and 2ndmore » order convergence rates for both time and space are achieved in the core channel model.« less
Authors:
; ; ;
Publication Date:
OSTI Identifier:
1165502
Report Number(s):
INL/CON-13-30982
DOE Contract Number:
DE-AC07-05ID14517
Resource Type:
Conference
Resource Relation:
Conference: International Topical Meeting on Advances in Thermal Hydraulics - 2014 (ATH '14),Reno, NV,06/15/2014,06/19/2014
Research Org:
Idaho National Laboratory (INL)
Sponsoring Org:
DOE - NE
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS convergence study; RELAP-7; verification