Basis for Calculating Boron Dilution Scenarios in PWR by 3D Neutron Kinetics
NURETH-14 - 2011 September 25-30


Presented at:
NURETH-14
2011 September 25-30
Location:
Toronto, Canada
Session Title:
B3-4 Core Thermalhydraulics and Subchannel Analysis

Authors:
Patricia Pla (GRNSPG-UNIPI/UPC)
Carlo Parisi (GRNSPG-UNIPI)
Regina Galetti (CNEN)
Francesco S. D'Auria (University of Pisa, GRNSPG, Italy)
Giorgio Galassi (GRNSPG-UNIPI)
Francesc Reventós (Universitat Politécnica de Catalunya)
  

Abstract

The origin of the performed study was the analysis of 20 cm2 small break LOCA in the lower

plenum in a four-loop PWR nuclear reactor by Relap5 code stand-alone (0DNK) in which boron

 dilution was observed in more than one loop seal. In order to have a more precise result of the

boron dilution NK feedback effect, the original nodalization was refined axially in the core area to couple

with PARCS v.2.7 code (3DNK). The neutron macroscopic XSec database was generated by the

lattice transport code HELIOS.

Before using the new model to predict boron dilution transients, a necessary activity is the

qualification of the model (the boron feedback calculated by the Neutronic Cross Sections) against

boron changes, so a group of sensitivity calculations injecting more or less borated water in the cold

leg were performed either with Relap5 code stand-alone (0DNK) and with Relap5 coupled with

PARCS v.2.7 (3DNK) code in order to analyze the reactor power response to the boron injection and

the differences using a 0DNK or a coupled 3DNK nodalization.

To complete the study a benchmark calculation was performed considering a 20 cm2 break in the

lower plenum, in which the reactor trip by control rods has been disabled and boron injection was

simulated in the cold leg. This calculation utilized the Relap5 code stand-alone (0DNK) and the

Relap5 coupled with PARCS v.2.7 (3DNK) code, in order to see the differences using a 0DNK or a

coupled 3DNK model.

Non negligible differences have been found in all cases in the comparison of 0DNK and coupled

3DNK results analyzed, in relation to the core power. These results challenge the evaluation of the

uncertainties in case of coupled thermalhydraulic-3DNK calculations. A comprehensive evaluation

of the relevant uncertainties of the 3D NK TH coupled calculations is needed.

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