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.

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

It is crucial to do safety evaluation of different postulated transient scenarios in actual nuclear power plants (NPPs). Some of the common analyzed scenarios are primary coolant tube rupture and station blackout (SBO). In this research, it was supposed that after establishment of a steady-state condition, an instantaneous guillotine large-break loss-of-coolant accident (LB-LOCA) of 850-mm inside diameter in one of the reactor vessel cold legs occurred, accompanied with SBO. The event progression and the variation of different reactor parameters like loop pressures, mass flow rates, fuel and clad temperature, injection rate of accumulators (ACCs), decay, and reactor power were investigated using the RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program. The reactor consequences due to availability and unavailability of passive ACCs were compared. These kinds of analyses assist in estimating the time available to perform operator safety actions. This in turn aids in emergency planning and severe accident management. The results reveal that fuel damage decreased after the introduction of ACCs. Actuation of ACCs at their actuation setpoints provided core cooling by injecting water into the reactor core. However, ACCs alone are inadequate to contain long-term core cooling during a persistent LB-LOCA. The results obtained in the research were compared with MELCOR 2.1 and ASTEC V1.3, and a cohesive agreement was obtained. Therefore, RELAP5/SCDAPSIM/MOD3.5 is capable of modeling a LB-LOCA and SBO in VVER-1000, and it provides a significant analytical capability of safety systems for specialists in the field in NPP safety.