The recent availability of powerful computer and computational techniques has enlarged the possibilities to perform best estimate simulations of complex scenarios in nuclear power plants. Nowadays, the coupled codes method, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used. It is particularly suited for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Within this framework the Peach Bottom BWR 2 Turbine Trip test was selected since it involves a rapid pressw"e induced positive reactivity addition into the core. It is also characterized by a self-limiting power course due to compensated inherent reactivity mechanisms. To perform a numerical simulation of the turbine trip a reference case was nm for the coupled thermal-hydraulic system code RELAP5/mod3.3 and 3D neutron kinetic PARCS/2.3 code. The overall data comparison shows good agreements between the calculations and most of the significant global aspects observed experimentally. However, the test is very sensitive to the feedback modeling and requires a tightly accurate simulation of the thermal-hydraulic and the cross sections parameters. For this purpose, sensitivity studies have been carried out in order to identify the most influential parameters that govern the transient behavior. The considered cases showed that the self-limiting power amplitude as predicted by the coupled code calculation is mainly due to delayed feedback mechanisms whereas the experimental data shows that the power quenching before the Scram is governed by prompt feedback effects.
Sensitivity analysis of the Peach Bottom Turbine Trip 2 experiment
D'AURIA, FRANCESCO SAVERIO;
2004-01-01
Abstract
The recent availability of powerful computer and computational techniques has enlarged the possibilities to perform best estimate simulations of complex scenarios in nuclear power plants. Nowadays, the coupled codes method, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used. It is particularly suited for simulating transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Within this framework the Peach Bottom BWR 2 Turbine Trip test was selected since it involves a rapid pressw"e induced positive reactivity addition into the core. It is also characterized by a self-limiting power course due to compensated inherent reactivity mechanisms. To perform a numerical simulation of the turbine trip a reference case was nm for the coupled thermal-hydraulic system code RELAP5/mod3.3 and 3D neutron kinetic PARCS/2.3 code. The overall data comparison shows good agreements between the calculations and most of the significant global aspects observed experimentally. However, the test is very sensitive to the feedback modeling and requires a tightly accurate simulation of the thermal-hydraulic and the cross sections parameters. For this purpose, sensitivity studies have been carried out in order to identify the most influential parameters that govern the transient behavior. The considered cases showed that the self-limiting power amplitude as predicted by the coupled code calculation is mainly due to delayed feedback mechanisms whereas the experimental data shows that the power quenching before the Scram is governed by prompt feedback effects.I documenti in IRIS sono protetti da copyright e tutti i diritti sono riservati, salvo diversa indicazione.
