Currently, passive safety systems are critical for enhancing nuclear reactor safety and dependability. To limit the chance of the core being uncovered in pool-type research reactors, a siphon pipe with a penetration in the pool wall higher than the core level can be used as the pool outlet pipe. Using a siphon breaker as a passive safety system is vital. The hydraulic study of the siphon breaker line passive safety system for a pool-type research reactor is carried out using the RELAP5 code. The hydraulic analysis and modeling are carried out on a 16-inch coolant outlet siphon pipe, taking into account 16-inch and 8-inch break diameters, as well as siphon breaker line diameters of 2, 2.5, 3, and 4 inches. As a consequence, the undershooting height for a 16-inch break and a 4-inch siphon breaker line is -36.7 cm. The undershooting height is -51.4 cm when using an 8-inch break and a 2-inch siphon breaker line. Compared with the findings to the reference experimental data, the largest difference is -3.1 cm and the smallest difference is -0.1 cm. The findings obtained indicate a substantial agreement between the simulated and experimental results.

Following the Fukushima Daiichi accident, the simulation of accidents related to the Spent Fuel Pool (SFP) became more important due to the high content of long-lived radionuclides, and lack of the protection by the pressure vessel despite its low decay heat. Therefore, the loss-of-cooling accident in the SFP of the Bushehr NPP was first simulated in this paper. The RELAP5 (as the Best Estimate code) and MELCOR (as a Severe Accident code) codes were used for simulation of the loss-of-cooling accident. The decay heat power calculation was performed by the ORIGEN code. The nodalization of SFP was done by using the Final Safety Analysis Report (FSAR) of Bushehr NPP. Different phenomena such as increasing water temperature in the pool, water boiling and decreasing of pool water level, spent fuel uncovering, increasing fuel temperature and the onset of fuel melting, hydrogen production, and release of radio-nuclides were observed and investigated. The steady-state results were validated by Bushehr NPP operating data. Verification of transient and accident results was performed by code-to-code (RELAP5 & MELCOR) comparison approach and Bushehr NPP data, the results showed that a good agreement together.

As, the Russian designed VVER-1000 reactor of the Bushehr Nuclear Power Plant (BNPP) by taking into account the change from German technology to that of Russian technology, and with the design of elevation change in the cold legs has been developed; therefore safety assessment of these systems for loss of coolant accident (LOCA) in elevation change in the cold legs and comparison results for non change elevation in the cold legs for a typical reactor (normal design of nuclear reactors) is the main important factor to be considered for the safe operation. In this article, the main objective is the simulation of the loss of coolant accident scenario by the ELAP5/MOD 3.2 code in two different cases; first, the elevation change in the cold legs, and the second, non change in it. After comparing and analyzing these two code calculations the results have been generalized for a new design feature of Bushehr reactor. The design and simulation of the elevation change in the cold legs process with RELAP5/MOD3.2 code for PWR reactor is performed for the first time in the country, where it is introducing several important results in this respect.

Small and medium break LOCA accidents at low pressure and under low velocity conditions have been studied in the TTL-2 Thermo-hydraulic Test Loop, experimentally. TTL-2 is a thermal hydraulic test facility which is designed and constructed in NSTRI to study thermal hydraulic parameters under normal operational and accident conditions of nuclear research reactors. A nodalization has been developed for the TTL-2 and experimental results have been compared with RELAP5/MOD3.2 results. The considered accidents are a 25% and 50% cold leg break without emergency core cooling systems. Results show good agreement between experiments and RELAP5/MOD3.2 results. This research provides experimental data for evaluation of thermo hydraulic codes for nuclear research reactors, and verifies that RELAP5/MOD3.2 has a good capability to estimate the thermal hydraulic behavior of low pressure and low velocity thermal hydraulic systems, such as research reactors under steady state and transient conditions.