Abstract
The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA.
Highlights
All pressurized nuclear power plant (NPP) designs require thorough evaluation to ensure compatibility with existing safety and regulatory standards
It is required that NPP safety analysis codes should be validated—that is, a true representative set of calculations should be tested against measured or otherwise acceptable data
This study focused on the analysis of the post-accident transient performance behavior of VVER-1000
Summary
All pressurized nuclear power plant (NPP) designs require thorough evaluation to ensure compatibility with existing safety and regulatory standards. Pressurized NPPs should be designed such that, under no credible normal or off-normal situation, can radioactive material be released from the core to the environment [1]. The primary concern in these analyses is that the large fission product inventory produced in the reactor core is not released in any conceivable accident situation [3]. There are several barriers to such fission product release. The primary barrier is the metal clad of the fuel itself, which isolates the fuel pellets from the coolant [4]. The provision of leak tight “barriers” between the radioactive source and the public are generally three: fuel cladding, the primary system pressure boundary, and the containment [5]
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