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.

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.

Tehran Research Reactor (T.R.R.) is a pool-type, 5 MW thermal research reactor. One probable event is that if some external objects or debris fall down into the reactor core and cause obstruction of the coolant flow through one of the fuel assemblies, decreasing the surface flow area, ceases the coolant flow, and also raises the fuel and sheaths temperature. Thermal hydraulic analysis of this event has been studied using RELAP5 system code. This report is related to the partial and total obstruction of a single Fuel Element (F.E.) and cooling channel of 27 F.E. equilibrium core of the T.R.R. Such event may lead to severe accident for such type of research reactors, since it may cause a local dry out and eventually loss of the F.E. integrity. Two scenarios are analysed in order to emphasize the severity of the mentioned accident. The first is a partial blockage of hot F.E. which is considered for four different obstruction levels of the nominal flow area: 25%, 50%, 75% and 93%. The second is related to an extreme case which consists of the total blockage of the same F.E. The reactor power is derived through the kinetic point calculation in the RELAP5 code. The point kinetic feedbacks including the fuel temperature (Doppler coefficient) and the coolant density coefficient have been considered through the applied model. The main results obtained from the RELAP5 calculations are as follows: 1. In case when the flow blockage is under 93% of the nominal flow area of an average F.E., only the increase of the coolant and clad temperatures are observed with no integrity of the F.E. consequences. The mass flow rate remains sufficient enough and cools the clad safely 2. In the case of a total obstruction in the nominal flow area, it is seen that the severe accident is due to dryout conditions and reaches promptly, while melting of the cladding occurs.