PARCS-Subchanflow-TRANSURANUS Multiphysics Coupling for High Fidelity PWR Reactor Core Simulation: Preliminary Results

摘要

Traditionally, reactor core simulators use simplified models to predict the fuel temperature and thermal-hydraulic conditions in the core. To achieve better accuracy detailed models should be used to describe all different physical processes (Multiphysics approach). Simplified solvers for the fuel temperature don't capture the material behavior under irradiation such as, swelling, cracking, pellet-clad interaction, etc. These phenomena affect properties such as fuel thermal conductivity, the fuel rod gap conductance which has an impact in the calculation of the fuel temperature. It is known that the gap conductance during reactor lifetime, depends strongly on the irradiation and power history as shown for instance in [Bielen 2015] thermal-hydraulic, and fuel thermo-mechanical behavior of the core components. Typically in current generation reactor physics analysis these three component areas are given separate consideration or are at best loosely coupled. Within this work, a methodology for tightly coupling the core neutronics code PARCS, thermal-hydraulics code PATHS, and fuel rod simulator code FRAPCON was developed. This coupled code package, referred to as FRAPARCS, was applied to two fuel depletion problems: a pin cell and a 5x5 assembly mini-core. The results of the depletion calculations indicate that standalone PARCS does not adequately capture the evolution of fuel rod behavior which influences the Doppler fuel temperature used in cross section evaluation, and as a result significant differences in computed core performance can be seen. In particular, the behavior of the fuel-cladding gap and associated temperature drop was found to be important. FRAPARCS was then applied to the pin cell calculation to evaluate the uncertainty and sensitivity of the nuclear performance of the core due to the influence of fuel thermo-mechanical models available for manipulation in FRAPCON. A sensitivity study was conducted to determine which fuel models were influential on the neutronics outputs; we determined that fuel thermal conductivity, fuel thermal expansion, cladding creep, and fuel swelling had an important influence on the core Doppler temperature and reactivity. Additionally, the heat transfer coefficient was found to be important. Then, FRAPARCS was integrated within the DAKOTA uncertainty package. Two varieties of sampling-based methods (Random and Latin Hypercube Sampling). There are only few publications about mutiphysics simulations in the area of fuel behavior studies [e.g. Magedanz et al. 2015; Hales et al. 2014] and, even less containing studies of reactor core simulations [Holt et al. 2016; Holt et al 2014].