A simulation of a steam generator tube rupture in a VVER-1000 plant

• Published in: Energy conversion and management Vol. 49; no. 7; pp. 1972 - 1980
• Main Authors: Nematollahi, M.R., Zare, A.
• Format: Journal Article Conference Proceeding
• Language: English
• Published: Oxford Elsevier Ltd 01.07.2008; Elsevier
• Subjects: Applied sciences; Energy; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Installations for energy generation and conversion: thermal and electrical energy; Steam generator; Tube rupture and thermo-hydraulics; VVER-1000; INW, Russia; ISW, Iran; Iran, Bushehr; VVER-1000; Tube rupture and thermo-hydraulics; Steam generator; Tube; Accident; Simulation; Rupture; Thermohydraulics; Pressurized water reactor; Nuclear reactor
• ISSN: 0196-8904; 1879-2227
• DOI: 10.1016/j.enconman.2007.12.018

In the operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions. One of the main accidents is steam generator tube rupture (SGTR), in the field of nuclear safety. In this research the variation of thermo-hydraulics parameters in a primary loop under SGTR accidents at Bushehr VVER-1000 Nuclear Power Plant (NPP) is analyzed by RELAP5/MOD 3.2 thermo-hydraulics code. A model of Bushehr unit 1 has been developed based on RELAP5/MOD 3.2 at Shiraz University in Iran. The results obtained by this code are compared with the Preliminary Safety Assessment Report (PSAR) data of Bushehr NPP. The DINAMICA-97 code was used to evaluate the SGTR accident in Bushehr PSAR by Federal Agency on Nuclear Energy of Russia. In simulation of this accident, it was supposed that after the establishment of a steady-state condition in the system, primary-to-secondary coolant leakage is as a result of an instantaneous break happened with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes in one of the fourth steam generators (SG#2). Loss of off-site power and two diesel generators of in-site power in loops 1and 2 of the primary cycles follow this event. The results show that after initiating the accident, the pressure above the core decreases rapidly from 16 MPa to 8 MPa during 130 s and finally, it will stabilize to about 7.1 MPa and it becomes practically equal to the secondary pressure and it leads to stop leakage completely at the end of transient. The coolant temperature at the reactor inlet decreases from 566 K to 473 K during the transient time and the cooldown of the coolant of the primary loop is performed. The primary-to-secondary coolant leakage reduces from 800 kg/s to approximately 0.0 kg/s during the first 150 s of the accident. As seen from the results, the thermo-hydraulics code correctly predicts the behavior of the main plant parameters in comparison with the reported PSAR data.