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One of the important operations in nuclear power plants is load-following in which the imbalance in axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load-following operation. On the other hands, precursors produce delayed neutrons which are important with respect to reactor period and control, but xenon concentration and precursors densities cannot be measured directly. In this paper, the non-linear sliding mode observer which has the robust characteristics facing the parameters uncertainties and disturbances is proposed based on the two point nuclear reactor model equations with three groups of the delayed neutrons to estimate the xenon concentration and delayed neutrons precursor densities of the pressurized-water nuclear reactor (PWR) using reactor power measurements. The stability analysis is provided by means of the Lyapunov approach, thus the system is guaranteed to be stable over a wide range. The employed method is easy to implement. This estimation is done taking into account the effects of reactivity feedback due to temperature and xenon concentration. Simulation results clearly show that the sliding mode observer follows the actual system variables accurately and is satisfactory in the presence of the parameter uncertainties and disturbances. [ABSTRACT FROM AUTHOR]

• Direct Lyapunov function in a self-pressurized, natural circulation integral reactor is revealed. • The primary coolant in a self-pressurized, integral reactor is asymptotical global stable. • Neutronic and thermal hydraulic feedbacks can guarantee a stable operation. The non-linear stability of a small modular self-pressurized reactor is assessed here by means of a direct Lyapunov function. The trial and error method is analyzed by applying an appropriate dynamical model for this type of reactors. A function containing all the states of the system which has conditions of a direct Lyapunov function is presented. This continuous function is positive definite and its derivative is negative semi-definite. The results indicate that the primary coolant is asymptotically global stable over its entire power range. Also, we assess the effects of neutronic and thermal hydraulic feedbacks on the system stability. Its been found that in the presence of these feedbacks, the system is stable in the entire power range. [ABSTRACT FROM AUTHOR]

• The system is stable over its entire power range. • The difference between the results of linearized and nonlinear models is negligible. • An increase in chimney height improves system inherent safety. • Higher operational limits can be achieved by applying a heater and a spray. The stability analysis of a natural circulation integrated self-pressurized water reactor is investigated by the Lyapunov approach. The analysis of the pressurized water reactors (PWRs), particularly the integrated self-pressurized water reactors, is essential in keeping the neutronic and thermal-hydraulic parameters of the system stable. An appropriate nonlinear dynamic model is introduced based on conservation of mass, momentum and energy which is then linearized, in a state-space model. The Lyapunov approach and Routh-Hurwitz criterion are applied to assess the stability of this linearized system over its entire power range. The analysis is done for the primary coolant circuit in the RPV by assuming the steam dome pressure as the fixed parameter. It is found that the system remains stable over its entire power range. The influence of different geometrical features is studied at nominal conditions. It has been found that by reducing the chimney height results in a decrease in the coolant flow rate and a downward motion of the onset of flashing while the average core coolant temperature rises. For lower values of friction losses coefficient, the coolant flow rate increases, and the onset of flashing moves upward and the average core coolant temperature decreases. A change in independent parameters, which are effective in generating the natural circulation of the coolant, can influence the inherent safety of the system: An increase in reactor power and chimney height and a decrease in friction losses coefficient improve the system inherent safety. Two input functions as extra reactivity for increasing and decreasing the power from the nominal state are implemented into the dynamic model and the model response is therefore assessed by considering the lack of two phase flow entry to the core restriction. The boundary of the system stability lies in the range 32–107 MWt and using the system outside this range the system pressure needs to be controlled through spray and heater systems. The obtained results are based on the initial information, data and primary design of these types of reactors. Determining a more accurate boundary requires more detailed design and assessment together with experimental test facilities with respect to other restricting parameters of the system. The results presented in this paper can be implemented in further research on this type of reactors, particularly for nonlinear stability analysis and finding nonlinear Lyapunov function. [ABSTRACT FROM AUTHOR]

Dual cooled annular nuclear fuel which is an internally and externally cooled annular fuel has many advantages in heat transfer. One of the most prominent of these advantages is the ability to harvest more of this type of fuel that can be taken advantage to increase the thermal power of nuclear plants. In this paper, the core of a VVER-1000 reactor is designed based on the use of internally and externally cooled annular fuels and thermal hydraulic parameters of the fuel rods in this type of reactor is analysed. In addition, the amount of thermal power uprate in a VVER-1000 reactor using annular fuels is investigated. For this purpose, at the first, using of cell and core neutronics calculations codes the proper pitch length of fuel rods in the core is designed at Clean and Cold conditions. Then, using a simulated fuel rod in hot channel by simulation codes of CFD, thermo-hydraulic calculations are performed, and are compared with a conventional VVER-1000 reactor that does not use this kind of fuel. As one of the most important results of the analysis, annular fuel shows a sufficient margin available on DNB and fuel pellet temperature relative to cylindrical fuel. The margin amount seems accommodating a 129% power-uprate seems viable. [ABSTRACT FROM AUTHOR]

• For the first time, dual cooled annular fuel in CAREM nuclear reactor is studied. • Neutronics and natural circulation parameters are optimized. • Developed Artificial Neural Network predicts CAREM nuclear reactor parameters. • Optimal geometry of fuel is determined using genetic algorithms. The small modular nuclear reactor (SMR) with annular fuel which is cooled internally and externally has the potential to maintain high power density with a sufficient safety margin. The benefit of the dual cooled annular fuel design is that heat can be transferred to the coolant from both the outer and inner surfaces. In this paper, design of the Argentina small modular reactor (CAREM) core with hexagonal assemblies using dual-cooled annular fuels with internal and external cooling is presented. Neutronics and thermal-hydraulic parameters are evaluated and analyzed by changing internal fuel radius. For this purpose, at the first, the dual-cooled annular fuel under clean and cold conditions is modeled and the effective multiplication factor has been calculated for different inner clad diameter. Indeed, the internal and external radius have been changed in such a way to maintain the reactor under moderated.Then, these annular fuels under full power conditions are modeled and power peaking factor has been calculated. Finally, natural circulation parameters are performed for a simulated fuel rod in the hot channel using computational fluid dynamics (CFD) simulation codes. These results are compared with the conventional CAREM reactor. One of the most prominent advantages of the annular fuels is the ability to make softer neutrons in which reduce maximum power peaking factor and improves fuel management and safety parameters in the reactor core. For data fitting, an artificial neural network is trained using the observed data. The input consists of different internal and external radiuses and output consists of fuel rods' pitch, thermal-hydraulic and neutronic parameters. Finally, the optimal geometry of fuels is determined using the neural network by implementing the genetic algorithms.Indeed, developed Artificial Neural Network (ANN) utilizing the obtained data, predicts the thermal-hydraulic and neutronic parameters of the CAREM reactor core with dual-cooled annular fuels. Presented optimization algorithm, which has a significant ability to attain the best solutions, also determines the optimal values of natural circulation parameters (V max /V avg , V out -V in , and pressure drops), heat transfer coefficients, MDNBR, RPPF, and excess reactivity of the reactor. Also, the designed artificial neural network and genetic algorithm has been validated using neutronic and thermal hydraulic calculations. Results indicate that by using annular fuels, better reactor safety and thermal efficiency are achieved. [ABSTRACT FROM AUTHOR]

Abstract: In this study, using porous media approach, the compressible flow within the core of a Pebble Bed Modular Reactor (PBMR) is simulated. This reactor has been composed of coated fuel particles with compressible gas as a coolant and graphite as a moderator and reflector. Containing about 450,000 fuel complexes, the reactor core is considered a porous medium subject to high temperature and high pressure helium flow. The porosity and permeability parameters are calculated and utilized. The coolant compressibility has been introduced as an effective parameter in the thermal–hydraulic analysis. Accordingly, using the ANSYS CFX code, which is capable of simulating compressible flow in porous media, the reactor core is simulated and thermal–hydraulic parameters of the core are obtained through Computational Fluid Dynamic (CFD) approach. The heat flux in the core is first obtained in axial and radial coordinates by MCNP code and is then used in CFD simulation as a semi sine and an algebraic function. The major characteristics of the flow field have been calculated whereby the thermal–hydraulic parameters such as temperature and pressure profiles have been obtained and compared with other data. Comparing the results obtained with other codes and software, the outcomes show that the inclusion of compressibility is reasonable and will lead to a slight difference between the measured and actual temperature, pressure and velocity. In another stage, pressure drop, flow vortices and helium flow lines are explored for two fuel complexes. The empirical formula of pressure drop presented by Kugeler and Schulten is modified and gas density is considered as a function of the core length. Fuel complexes in the reactor core are randomly arranged. However, because Body Center Cubic (BCC) is the closest arrangement to the random distribution, flow parameters are obtained using the BCC arrangement and they are found to deviate very slightly when compared with predictions of other codes. [Copyright &y& Elsevier]

Abstract: In this paper, the effect of perforated plate in horizontal steam generator (SG) has been studied. The injected feed water into the SG is cold and heavy, so it pulls down the fluid around the feed water injection pipeline. The perforated plate has been designed above tube bundles in the SG to relax this asymmetrically void generation. In this work, with consideration of the perforated plate and feed injection effects in tube side, it is illustrated that generated steam will be distributed in the top level of the SG homogeneously. Therefore, the steam collector contains high quality homogeneously distributed dry steam and the perforated plate prevents the water from ascending in the cold side containing cold collector and the steam from descending on the other side. In addition, it can be seen that without the perforated plate, the void fraction distribution becomes heterogeneous in the top level of the SG. In this analysis, the 3D numerical model of a large conventional WWER1000 steam generator in the nuclear industry has been presented. For the computational fluid dynamic (CFD) study of desired steam generator in ANSYS CFX, the SG geometry is prepared with details and interfacial relations of mass, momentum and heat transfer are defined by appropriate functions. In momentum source terms, the interfacial drag forces are defined with Ishii and Zuber model. An Euler–Euler approach is applied to modeling boiling heat transfer and condensation. Porosity model is applied to the primary side in which the tube bundles are not described in detail but they are modeled as sources of enthalpy and pressure loss. The primary side effect is modeled based on a 1D thermal heat source model. Finally, the importance of perforated plate is demonstrated and it shows that the results are in a good agreement with published and experimental data. [Copyright &y& Elsevier]