نشریه فناوری و انرژی هسته ای
پیاپی 4 (زمستان 1401)

Design and simulation of research test facilities have great importance among countries nuclear technology. The operation of the pressure control system plays an essential role in the safe operation of these research test loops. In this paper, high-pressure and high-temperature circuit are designed and simulated using Aspen HYSYS engineering software. A pressure control system is connected to the primary circuit to control the simulated circuit pressure. This system keeps the circuit pressure constant at a certain level (reference limit) by injecting and removing nitrogen in the upper head of the pressurizer. The simulation includes a PID control system designed within the Simulink MATLAB, in coupling with Aspen HYSYS engineering software. To ensure the correct operation of the pressure control system in transients, various scenarios have been performed to evaluate the pressure control system.

The main control room is one of the parts of a Nuclear Power Plant that is often identified as a part of the high fire risk in fire risk assessments. Due to the importance of fire in the main control room of Bushehr NPP-Unit 1 (BNPP-1), in this study, by using the CFAST code which is one of the deterministic safety analysis code, the fire in the main board of BNPP-1 is simulated and the results are evaluated. Required data has been extracted from safety analysis reports, and operating documents of BNPP-1 and NUREG6850 report to simulate fire. The highest recorded temperature was measured in the upper layer of the standing cabinet, about 1013C, which was result of flames and heat transfer through smoke (convection). Another highest recorded temperature was at the lower wall of the standing cabinet, measured about 700C; this temperature was due to flames of fire and also because of radiation heat transfer. The first ventilation system was blocked in 350 seconds after the start of the fire, and the second ventilation system was blocked in 860 seconds after the start of the fire.

The use of dual-cooled annular fuel is one of the proposed methods to increase the safety of nuclear reactors. This type of fuel allows the cooling fluid to pass through the internal channel and as a result more heat is removed from the fuel surface. For this purpose, in this study, a dual-cooled annular fuel assembly with 11×11 array in a small modular reactor has been investigated by neutronic and thermal-hydraulic simulations. In the results of neutronic calculations, designed dual-cooled annular fuel assembly were compared with the reference reactor, it was found that these two designs have similar neutronic properties. But in the thermohydraulic results, a significant effect of the use of dual-cooled annular fuel was extracted. The use of this type of fuel reduced the maximum fuel temperature, the pressure drop and increased minimum departure from nucleate boiling ratio.

In order to enhance applicability of Tehran Research Reactor (TRR) for irradiation test of nuclear fuel and materials and considering TRR potential to provide required neutronic and thermal hydraulic conditions for irradiation tests on domestic fuel samples of research reactors, TRR fuel test loop with 10 bar nominal pressure and 20m3/h nominal flow rate is designed and fabricated to simulate thermal hydraulic conditions of the desired research reactor. In this paper, consequences of loss of flow in the loop due to primary pump breakdown are investigated and the functions of safety systems in response to this accident to provide continuous cooling of the fuels under test and mitigate any undesirable consequence are evaluated. In this regard, the probable scenarios are simulated using a RELAP5 model and the ability of the safety features of the loop to prevent any damage to the fuels under test is evaluated.

The design and use of passive safety systems as one of the effective ways to increase reliability and safety level of nuclear power plants in new generation reactors has received much attention. In this paper, using RELAP5 computational code, the thermo-hydraulic analysis of the Passive Heat Removal System (PHRS) in the VVER-1000 reactor is investigated to ensure the capability of its performance under various conditions. For this purpose, the system nodalization is first extracted according to the available documentation and then the RELAP5 code input is developed. The results show that for 6.27 MPa pressure in steam generators and air temperature of 36 ° C, the system is able to remove 95.52 MW heat.

In recent years SMRs have received a lot of attention due to their small size and portability, as well as some nuclear propulsion. In this research, VVER-1000 fuel assembly has been used as a basis and has been compacted and reduced by special methods and arguments. In this paper, a hexagonal fuel assembly without soluble boron and rely on suitable burnable absorbers has been designed and proposed for use in the SMRs. The objectives of this research are elimination of boric acid from the coolant, determine the number of fuel rods and guide channels inside the fuel assembly, determine the optimal arrangement of fuel rods and guide channels inside the fuel assembly, keeping the fuel to moderator ratio and determine the appropriate dimensions Boron-free hexagonal fuel assembly for use in SMR reactors. Also, one of the challenges of this design was to keeping the ratio of fuel volume to moderator of the VVER-1000 reactor fuel assemblies for SMR reactor, which was solved with neutronics and safety considerations. Due to the maintenance of the fuel volume to moderator ratio, pitch of fuel rods of the VVER-1000 reactor is also intended for the boron-free fuel assembly. Also the desired cycle length for the designed fuel assembly is about 40 month, for which the fuel enrichment has been set at 6.6%. According to increase in fuel enrichment in the SMR cores, the reactivity at the beginning of the cycle is very high, therefore, this reactivity needs to be compensated. One of the most important challenges and concerns in this research is the control of reactivity at the beginning of the cycle, during the cycle and the reduction of the radial power distribution In the absence of boric acid, the burnable absorber of gadolinium was solved problem with proper design. The designed boron-free fuel assembly has 156 fuel rods, 12 guide channels and 1 central tube. It should be noted that MCNP and WIMS codes have been used to simulate fuel assembly and also to ensure the correct design process.