Development of a High Fidelity System Analysis Code for Generation IV Reactors
Traditional nuclear reactor system analysis codes such as RELAP and TRAC employ an operator split methodology. In this approach, each of the physics (fluid flow, heat conduction and neutron diffusion) is solved separately and the coupling terms are done explicitly. This approach limits accuracy (first order in time at best) and makes the codes slow in running since the explicit coupling imposes stability restrictions on the time step size. These codes have been extensively tested and validated for the existing LWRs. However, for GEN IV nuclear reactor designs which tend to have long lasting transients resulting from passive safety systems, the performance is questionable and modern high fidelity simulation tools will be required. The requirement for accurate predictability is the motivation for a large scale overhaul of all of the models and assumptions in transient nuclear reactor safety simulation software. At INL we have launched an effort with the long term goal of developing a high fidelity system analysis code that employs modern physical models, numerical methods, and computer science for transient safety analysis of GEN IV nuclear reactors. Modern parallel solution algorithms will be employed through utilizing the nonlinear solution software package PETSc developed by Argonne National Laboratory. The physical models to be developed will have physically realistic length scales and time scales. The solution algorithm will be based on the physics-based preconditioned Jacobian-free Newton-Krylov solution methods. In this approach all of the physical models are solved implicitly and simultaneously in a single nonlinear system. This includes the coolant flow, nonlinear heat conduction, neutron kinetics, and thermal radiation, etc. Including modern physical models and accurate space and time discretizations will allow the simulation capability to be second order accurate in space and in time. This paper presents the current status of the development efforts as well as some results from analyzing a simplified primary system model of GNEP’s advanced burner test reactor (ABTR) designed by Argonne. Various transient analyses are performed with this simplified ABTR model to study two fundamental issues related to system analysis codes – accuracy of numeric algorithm and efficiency. The accuracy study is carried by comparing the second order method with the first order method. The results show that numerical errors in the first order method are large and it is very difficult to distinguish numerical errors from physical modeling errors. On the other hand, second order method yields small numerical errors and it is very easy to spot physical modeling errors. The efficiency study is carried out by comparing the time steps for the fully implicit solution algorithm versus CFL stability limit methods. The dynamic time steps used in a fully implicit method will adjust the time step to resolve the time scale during the various stages of a long lasting transient. This will make a computer code based on fully implicit methods run more efficiently versus a CFL stability limit method code like RELAP, in which a particle of fluid cannot cross a control volume in a single time step.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- DOE - NE
- DOE Contract Number:
- DE-AC07-99ID-13727
- OSTI ID:
- 935466
- Report Number(s):
- INL/CON-07-13363; TRN: US0804336
- Resource Relation:
- Conference: ICAPP'08,Anaheim, CA, USA,06/08/2008,06/12/2008
- Country of Publication:
- United States
- Language:
- English
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