Assessment of the nuclear power plant behaviour during a small break LOCA combined with a SGTR assuming a station black out

• Simulation of multiple failure scenario based on a “SB LOCA combined with a SGTR simultaneously with a SBO”. • Comparison of calculated RELAP5 and TRACE results for assessment of VVER1000 plant behavior. • Investigation of 3D TRACE code representation of the reactor pressure vessel compared to the 1D in RELAP5 code. This paper presents the comparison of the results achieved by performing calculations using integral computer codes RELAP5/mod3.3 and TRACEv5.05. The scenario selected for this purpose is a multiple failure transient based on a “Small break loss of coolant accident (SB LOCA) combined with a steam generator tube rupture (SGTR) simultaneously with a Station blackout (SBO)”. In the investigated scenario is assumed that the break with equivalent diameter of 30 mm is located on the main coolant loop #1 between the main coolant pump and the inlet of the reactor vessel. In addition, a double ended guillotine tube rupture in SG #4 with the equivalent diameter of 13 mm is simulated simultaneously with a SBO. The main objective of this work is focused on evaluation of VVER1000 plant response during multiple failure events and the ability of system thermal hydraulic computer codes to simulate the key phenomena and processes taking place during the evolution of the accidental scenario. The computer codes used to perform the study are based on different approach for the modeling of the reactor pressure vessel (RPV) and core physics. The RELAP5 code is based on 1D thermal hydraulics modelling of RPV and point kinetics for reactor physics modelling, while in the TRACE code used 3D thermal hydraulics of RPV modelling and 3D neutron kinetics model for reactor physics modelling. The calculation is performed for the end of cycle (EOC) eight, when the reactor is operated at nominal power. The investigations were performed in the frame of the EU H2020 CAMIVVER project. The comparison of the calculated results has shown that the codes predict the VVER1000 plant behavior under SB LOCA with SGTR simultaneously with a SBO in a good manner.

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Publication Details

Journal
Annals of Nuclear Energy
Published
2025-12-04
DOI
https://doi.org/10.1016/j.anucene.2025.112038
Primary Topic
Nuclear Engineering Thermal-Hydraulics
Type
article
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article

Assessment of the nuclear power plant behaviour during a small break LOCA combined with a SGTR assuming a station black out

Pavlin Groudev, Víctor Hugo Sánchez-Espinoza, A. Стефанова, Petya Vryashkova
Annals of Nuclear Energy
Nuclear Engineering Thermal-Hydraulics
article

Assessment of the nuclear power plant behaviour during a small break LOCA combined with a SGTR assuming a station black out

Pavlin Groudev, Víctor Hugo Sánchez-Espinoza, A. Стефанова, Petya Vryashkova
article en

Abstract

• Simulation of multiple failure scenario based on a “SB LOCA combined with a SGTR simultaneously with a SBO”. • Comparison of calculated RELAP5 and TRACE results for assessment of VVER1000 plant behavior. • Investigation of 3D TRACE code representation of the reactor pressure vessel compared to the 1D in RELAP5 code. This paper presents the comparison of the results achieved by performing calculations using integral computer codes RELAP5/mod3.3 and TRACEv5.05. The scenario selected for this purpose is a multiple failure transient based on a “Small break loss of coolant accident (SB LOCA) combined with a steam generator tube rupture (SGTR) simultaneously with a Station blackout (SBO)”. In the investigated scenario is assumed that the break with equivalent diameter of 30 mm is located on the main coolant loop #1 between the main coolant pump and the inlet of the reactor vessel. In addition, a double ended guillotine tube rupture in SG #4 with the equivalent diameter of 13 mm is simulated simultaneously with a SBO. The main objective of this work is focused on evaluation of VVER1000 plant response during multiple failure events and the ability of system thermal hydraulic computer codes to simulate the key phenomena and processes taking place during the evolution of the accidental scenario. The computer codes used to perform the study are based on different approach for the modeling of the reactor pressure vessel (RPV) and core physics. The RELAP5 code is based on 1D thermal hydraulics modelling of RPV and point kinetics for reactor physics modelling, while in the TRACE code used 3D thermal hydraulics of RPV modelling and 3D neutron kinetics model for reactor physics modelling. The calculation is performed for the end of cycle (EOC) eight, when the reactor is operated at nominal power. The investigations were performed in the frame of the EU H2020 CAMIVVER project. The comparison of the calculated results has shown that the codes predict the VVER1000 plant behavior under SB LOCA with SGTR simultaneously with a SBO in a good manner.

Annals of Nuclear EnergyVol. 228
Karlsruhe Institute of Technology (DE), Institute for Nuclear Research and Nuclear Energy (BG)
Openalex Percentile: Top 50%
Nuclear Engineering Thermal-Hydraulics
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