Verification of a TRACE EPR™ model on the basis of a scaling calculation of an SBLOCA ROSA test

▶ Verification of a TRACE input deck for the EPR™ generation III PWR. ▶ Scaling simulation of an SBLOCA experiment of the integral test facility ROSA/LSTF. ▶ The EPR™ model was compared with the TRACE results of the ROSA/LSTF model. In cooperation with the Finnish Radiation and Nuclear Safety Author...

Full description

Saved in:
Bibliographic Details
Published in:Nuclear engineering and design 2011-03, Vol.241 (3), p.888-896
Main Authors: Freixa, J., Manera, A.
Format: Article
Language:English
Subjects:
Applied sciences
Breaking
Controled nuclear fusion plants
Design engineering
Energy
Energy. Thermal use of fuels
Exact sciences and technology
Fission nuclear power plants
Fuels
Installations for energy generation and conversion: thermal and electrical energy
Mathematical models
Nuclear engineering
Nuclear fuels
Nuclear power generation
Nuclear reactor components
Nuclear reactors
Nuclear safety
Pressurized water reactors
Citations: Items that this one cites
Items that cite this one
Online Access:Get full text
Tags: Add Tag
No Tags, Be the first to tag this record!
Description
Summary:▶ Verification of a TRACE input deck for the EPR™ generation III PWR. ▶ Scaling simulation of an SBLOCA experiment of the integral test facility ROSA/LSTF. ▶ The EPR™ model was compared with the TRACE results of the ROSA/LSTF model. In cooperation with the Finnish Radiation and Nuclear Safety Authority (STUK), a project has been launched at the Paul Scherrer Institute (PSI) aimed at performing safety evaluations of the Olkiluoto-3 nuclear power plant (NPP), the first EPR™, a generation III pressurizer water reactor (PWR); with particular emphasis on small-and large-break loss-of-coolant-accidents (SB/LB-LOCAs) and main steam-line breaks. As a first step of this work, the best estimate system code TRACE has been used to develop a model of Olkiluoto-3. In order to test the nodalization, a scaling calculation from the rig of safety assessment (ROSA) test facility has been performed. The ROSA large scale test facility (LSTF) was built to simulate Westinghouse design pressurized water reactors (PWR) with a four-loop configuration. Even though there are differences between the EPR™ and the Westinghouse designs, the number of similarities is large enough to carry out scaling calculations on SBLOCA and LOCA cases from the ROSA facility; as a matter of fact, the main differences are located in the secondary side. Test 6-1 of the ROSA 1 programme, an SBLOCA with the break situated in the upper head of the reactor pressure vessel (RPV), was of special interest since a very good agreement with the experiment was obtained with a TRACE input deck. In order to perform such scaling calculation, the set-points of the secondary relief and safety valves in the EPR™ nodalization had to be changed to those used in the ROSA facility, the break size and the core power had to be scaled by a factor of 60 (according to the core power and core volume) and the pumps coast down had to be adapted to the ones of the test. The calculation showed very similar results as the experiment and the ROSA-TRACE calculation. The only significant difference observed was a faster primary depressurization after the break flow turned to single-vapor flow. This difference could be explained on the basis of geometrical differences between the EPR™ and ROSA/Westinghouse RPV's designs.
ISSN:0029-5493
1872-759X
DOI:10.1016/j.nucengdes.2010.12.016