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A full scope thermal-neutronic couple has been developed to analyze WWER-1000 nuclear reactor core by using PARCS v2.7, COBRA-EN and WIMSD-5B. First, WIMSD-5B has been used to calculate macroscopic cross sections for each fuel assembly. Neutronic calculations have been done by PARCS v2.7 code to obtain multiplication factor and PPF distributions of all fuel assemblies in the core. COBRA-EN calculates thermal-hydraulic parameters for all the fuel assemblies. This couple can be used for steady-state and transient calculations. Neutronic accidents such as Rod Ejection Accident that its initial event is a neutronic event (control rod movement that yields to reactivity changes) and thermal-hydraulic accidents such as LOFA that its initial event is a thermal-hydraulic event (coolant flow rate decrease) can be analyzed by this couple. In this study, steady-state calculations and LOFA have been simulated by this couple in the reactor core. A new “steady-state convergence” method is introduced. Results have been compared to Bushehr Nuclear Power Plant FSAR. The results show great similarities with FSAR results; and confirm the ability of this package to simulate accident in the core; especially thermal-hydraulic accident that is out of the ability of PARCS code individually.

Nuclear safety has been one of the major issues to be studied since the inception of the nuclear industry. Establishing and maintaining core cooling and ensuring containment integrity are the two main goals that nuclear safety must guarantee. The specific type of large LOCA is DECL (Double Ended Cold Leg) break which means a total guillotine type of break in cold leg pipe and is one of the most dangerous accidents in the reactor containment. In this paper, thermal-hydraulic parameters (Temperature and Pressure) of WWER-1000 (Bushehr Nuclear Power Plant) containment in a DECL accident have been simulated by CONTAIN2.0 code and a single cell model. The Containment has been divided to 23 cells in CONTAIN code but for simplicity only one cell has been considered in modeling. The model has been programmed by MATLAB. The accident has been simulated for Short-term (0 < time

The consequences of sever reactor accident depend greatly on containment safety features and containment performance in retaining radioactive material. The specific type of large LOCA is DECL (Double Ended Cold Leg) break which means a total guillotine type of break in cold leg pipe and is one of the most dangerous accidents in the reactor containment. In this paper, thermal–hydraulic parameters (temperature and pressure) of WWER-1000 (Bushehr Nuclear Power Plant) containment in a DECL accident have been simulated by CONTAIN 2.0 code and a single cell model. The containment has been divided to 23 cells in CONTAIN code but for simplicity only one cell has been considered in modeling. The model has been programmed by MATLAB. The accident has been simulated for a short time (initial 200 s) and all of the results have been compared with Bushehr’s Nuclear Power Plant FSAR.

The rod ejection accident is defined as the postulated rupture of a control rod drive mechanism housing that results in the complete ejection of a rod cluster control assembly from the reactor core. The consequences of the mechanical failure are a rapid positive reactivity insertion and an increase in the local power peaking with high local energy deposition in the fuel assembly, accompanied by an initial pressure increase in the reactor cooling system. In this study, the REA has been simulated in a WWER-1000 reactor by using WIMSD-5B and PARCS v2.7 codes. First, macroscopic cross-sections have been calculated for various types of fuel assemblies using WIMSD-5B. Results have been fed as input to PARCS v2.7 code. Steady-state, transient and specially thermal–hydraulic feedback blocks of PARCS code have been handled in this simulation. Finally, results have been compared with Final Safety Analysis Report of WWER-1000 reactor. The results show a great similarity and confirm the ability of PARCS code in simulation of transient accidents.

In this paper a whole-core transport technique using response matrix and collision probability (CP) methods is presented for large-scale, highly-heterogeneous reactors. Integral transport method has been used to provide sufficient accuracy for response matrix formation of pin cell sized node with considerably less computational expense. Ray tracing is efficiently applied using macro-bands. For practical application, double P 2 (DP 2 ) Legendre polynomial expansion is applied to approximate interface angular flux that is used to couple nodes. The proposed method is based on a sound mathematical foundation and leads to dramatically reduced memory requirements in contrast to the conventional transport method. This method is also applied to several problems such as C5G7, containing mixed oxide (MOX) and UO 2 fuel assemblies, to show the effectiveness of the proposed method. The results clearly indicate that the method is quite promising and acceptable.

In this paper, a robust water level control system for the horizontal steam generator (SG) using the quantitative feedback theory (QFT) method is presented. To design a robust QFT controller for the nonlinear uncertain SG, control oriented linear models are identified. Then, the nonlinear system is modeled as an uncertain linear time invariant (LTI) system. The robust designed controller is applied to the nonlinear plant model. This nonlinear model is based on a locally linear neuro-fuzzy (LLNF) model. This model is trained using the locally linear model tree (LOLIMOT) algorithm. Finally, simulation results are employed to show the effectiveness of the designed QFT level controller. It is shown that it will ensure the entire designer’s water level closed loop specifications.