Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code
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A. Hedayat
Abstract
In this paper, a 5 MW open pool type research reactor is simulated and analyzed to check and pass a complete and practical safety assessment against various LOFA scenarios. Generally, LOFA may occur in a research reactor when the reactor primary pump or the safety flapper valve fail. The simulation includes the study of a downward core cooling system circulated by the gravity driven force, the performance and failure effects of the safety flapper valve, the flow reversal mode and transition, the natural convection mode, different possible types of LOFA, and impacts of the surrounding empty boxes. First of all, the code nodalization is successfully benchmarked against available experimental data of the reactor operating parameters. Then, all possible DBAs in this field of study are simulated and discussed in detail. Results are completely satisfactory to simulate and analyze a pool-type research reactor in response to LOFA using the RELAP5 code. Furthermore, transients including the natural convection mode and even flow reversal mode are following naturally without any oscillation or source of errors. Finally, TRR is completely safe against the DBA type of LOFA scenarios.
Kurzfassung
In diesem Beitrag werden verschiedene Loss of Flow (LOFA)-Szenarien für einen 5 MW-Forschungsreaktor vom Typ Open Pool analysiert, um hierfür eine Sicherheitsbewertung durchführen zu können. Allgemein treten LOFA in einem Forschungsreaktor auf, wenn die Primärpumpe des Reaktors oder das Sicherheitsklappenventil ausfallen. Die Analyse umfasst die Untersuchung eines Kernkühlsystems, das schwerkraftgetrieben nach unten gerichtet kühlt, die Untersuchung der Auswirkungen des Betriebs und des Ausfalls von Sicherheitsklappenventilen, die Untersuchung des Einflusses der Strömungsumkehr und des Übergangs dazu und des natürlichen Konvektionsmodus sowie des Einflusses umgebender leerer Boxen. Insgesamt werden verschiedene LOFA-Szenarien untersucht. Zunächst wird die Code-Nodalisierung erfolgreich mit den verfügbaren experimentellen Daten der Reaktorbetriebsparameter verglichen. Anschließend werden alle Berechnungen mit dem Programm RELAP 5 der verschiedenen DBAs vorgestellt und ausführlich diskutiert.
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© 2019, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- New research reactor protection system
- Ultimate response guideline strategy evaluation for Maanshan power plant
- Application of two alternative shutdown severe accident management guideline (SSAMG) entry conditions for CPR1000
- Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code
- Numerical study on cut-off diameter of aerosol particle for filtered containment venting system in nuclear power plant
- Reanalysis of environmental qualification radiation parameters for Kuosheng nuclear power plant
- Effect of reactor operator intervention during unprotected LOFA in a typical MTR
Articles in the same Issue
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- New research reactor protection system
- Ultimate response guideline strategy evaluation for Maanshan power plant
- Application of two alternative shutdown severe accident management guideline (SSAMG) entry conditions for CPR1000
- Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code
- Numerical study on cut-off diameter of aerosol particle for filtered containment venting system in nuclear power plant
- Reanalysis of environmental qualification radiation parameters for Kuosheng nuclear power plant
- Effect of reactor operator intervention during unprotected LOFA in a typical MTR