Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3
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S. H. Abdel-Latif
Abstract
The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.
Abstract
Der Stations-Black-Out (SBO) ist eine der wichtigsten Unfallsequenzen, die im Bereich der schweren Unfälle zu betrachten sind. Um das Verhalten eines Kernkraftwerks im Zusammenhang mit diesem Unfall zu bewerten, werden mit dem Integralcode ASTEC-V2.1.1.3 (Accident Source Term Evaluation Code) SBO-Störfallsequenzen berechnet, die zu einem schweren Unfall führen können. Ziel dieser Arbeit ist es, die Modellierungsprinzipien für das Phänomen des Kernschmelzens und der Schmelzeverlagerung im Reaktorbehälter des WWER-1000 zu diskutieren. Das Szenario des SBO wird mit dem ASTEC-Code und seinen Grundmodulen simuliert. Anschließend wird die Simulation mit erneut durchgeführt, nachdem die Module ISODOP, DOSE, CORIUM und RCSMESH aktiviert wurden, um die Schmelze außerhalb des Reaktorbehälters zu simulieren. Die Ergebnisse der beiden Simulationen werden miteinander verglichen. Infolge des SBO sind die aktiven Sicherheitssysteme nicht verfügbar und konnten ihre Sicherheitsfunktionen, die die Sicherheitsanforderungen für einen sicheren Betrieb des Kernkraftwerks aufrechterhalten, nicht erfüllen. Infolgedessen werden die Sicherheitsanforderungen verletzt, was zu einer Überhitzung des Kerns führt. Außerdem kommt es zu einer möglichen Kernzerstörung. Die vorliegende Studie konzentriert sich auf das Versagen des Reaktordruckbehälters und die Verlagerung des Coriums in den Sicherheitsbehälter. Außerdem werden der Transfer von Spaltprodukten (FP) vom Reaktor in den Sicherheitsbehälter, die Zeit für die Kernaufheizung, die Wasserstoffproduktion und die Menge des Coriums im unteren Plenum des Reaktordruckbehälters ermittelt.
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© 2021 Walter de Gruyter GmbH, Berlin/Boston, Germany
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- Evaluation of human factor engineering influence in nuclear safety using probabilistic safety assessment techniques
- Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm
- Application research on neutron-gamma discrimination based on BC501A liquid scintillator
- Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3
- Study on specific heat capacity and thermal conductivity of uranium nitride
- Review and outlook of the integral test facility PKL III corresponding studies
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- Calendar of events
Articles in the same Issue
- Frontmatter
- ANSYS-CFX simulation of the SRBTL test loop core with nanofluid coolant
- Evaluation of human factor engineering influence in nuclear safety using probabilistic safety assessment techniques
- Optimization of the TiO2 nanofluid as a coolant in the VVER-1000 nuclear reactor based on the thermal reactivity feedback coefficients via the genetic algorithm
- Application research on neutron-gamma discrimination based on BC501A liquid scintillator
- Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3
- Study on specific heat capacity and thermal conductivity of uranium nitride
- Review and outlook of the integral test facility PKL III corresponding studies
- Effects of some level density models and γ-ray strength functions on production cross-section calculations of 16,18O and 24,26Mg radioisotopes
- An application research for near-surface repository of strontium-90 sorption kinetic model on mudrocks
- Calendar of events
