Nuclear Engineering and Design (00295493)432
In order to improve the safety of commercial nuclear reactors, this paper presents neutronic design and analysis of the accident-tolerant fuels (ATF) application to VVER-1000 nuclear reactor. The study commences by simulating a fuel assembly containing 2.4% enriched uranium. The simulation of a VVER-1000 reactor core loaded with standard uranium dioxide (UO2) fuel was then conducted using the MCNPX 2.6 code. This simulation yielded various neutronic and dynamic parameters, including the multiplication factor, excess reactivity, delayed neutron fraction, fuel and coolant temperature reactivity feedbacks, power peaking factor, and fuel burn-up. These values were subsequently compared with existing reference data. Finally, the reactor core was simulated using accident-tolerant fuel. The resulting data was analyzed to identify the optimal combination of fuel and cladding. Throughout the study, the MCNPX 2.6 software was employed for neutronic core analysis. Comparative analysis revealed that uranium carbide (UC) offers superior safety margins in both neutronic and dynamic aspects, outperforming uranium dioxide (UO2) and uranium mononitride (UN). Simulations exploring the use of silicon carbide (SiC) and FeCrAl as alternative cladding materials within the reactor core demonstrated potential advantages over traditional Zirconium (Zr). Results indicate that application of ATF can improve cycle length, temperature reactivity feedbacks and power peaking factor compare to the reference core (UO2 + Zr), significantly. © 2024 Elsevier B.V.
Nuclear Engineering and Design (00295493)432
The core of an operational power plant can include hundreds of fuel assemblies. In cases where more than one single fuel assembly design is present in a core—whether through a change in fuel manufacturer (vendor), offering a better design, or other reasons—the core is described as a mixed core. Ensuring that different types of fuel assemblies do not interact in a harmful manner that causes problems in the reactor core is essential to ensuring nuclear safety. This article thoroughly explores the neutronic modeling of the core of a VVER-1000 reactor using TVS-2 M fuels and the investigation of the mixed core created by loading a UTVS fuel assembly with the optimal fuel composition and location in the core of this reactor. The research is conducted thoroughly to ensure the desired UTVS assembly is placed in the reactor's core with minimal tension regarding neutronic parameters while maintaining the core's performance. The core of the VVER-1000 reactor was initially modeled using DRAGON and PARCS codes. Then, the new UTVS fuel assembly was loaded into each fuel assembly station with varying UTVS fuel enrichments. Neutronic parameters were obtained for each of the created states using computational codes. These results were used to create an artificial neural network in MATLAB. By connecting this neural network to a genetic algorithm, optimization was performed to determine the optimal fuel location and enrichment for loading the UTVS fuel assembly in the desired reactor core. © 2025 Elsevier B.V.
Nuclear Engineering and Design (00295493)438
This study investigates the design of a novel Small Modular Reactor (SMR) concept utilizing Dual-Cooled Accident Tolerant Fuel (DC-ATF). The DC-ATF incorporates U3Si2 fuel pellets clad in FeCrAl, enhancing safety and accident tolerance. A systematic approach was employed, beginning with the evaluation of 4000 unique fuel assembly configurations varying the number and arrangement of Integrated Burnable Absorbers (IBAs). Fifty configurations in each category were rigorously simulated using the MCNP code, and the results were used to train Artificial Neural Networks (ANNs) to predict the performance of the remaining assemblies. This approach facilitated the identification of suitable fuel assembly designs for each IBA category. Subsequently, these assemblies were integrated into 55 distinct reactor core configurations, varying the distribution of IBA-containing assemblies within a 37-assembly core arranged in a square lattice. Neutronic simulations were performed to evaluate core criticality, power distribution, burnup characteristics, and temperature coefficients. The results demonstrate that the proposed DC-ATF SMR exhibits favorable safety margins, including negative temperature coefficients (−2 pcm/K for fuel and −33.89 pcm/K for coolant) and acceptable power peaking factors (1.58 at beginning of the cycle). Burnup calculations indicate a first core cycle length exceeding 1800 effective full power days (EFPD), a significant increase compared to conventional UO2-fueled SMRs of similar size and output power, which typically achieve around 730–1330 EFPD. This improvement is primarily attributed to the higher uranium density of U3Si2 fuel, enabling increased fissile material loading. © 2025 Elsevier B.V.
Results in Engineering (25901230)26
Small Modular Reactors (SMRs), such as the NuScale Power Module (NPM), have gained considerable attention in recent years. However, significant efforts are needed to enhance safety. The utilization of hexagonal Fuel Assemblies (FAs) offers numerous benefits from both neutronic and thermal-hydraulic perspectives. By leveraging these advantages, reactors performance can be improved without compromising safety regulations. Additionally, there has been consideration given to the use of burnable neutron absorbers, such as gadolinia, in novel nuclear fuels. This paper focuses on the redesign and evaluation of the NPM core with hexagonal FAs. The study investigates the impact of altering the arrangement pattern of FAs and the quantity of gadolinia-containing rods within these assemblies. Various models featuring hexagonal FAs, containing different mixed fuels, were tested through a series of simulations. Calculations were repeated under both Clean-Cold and Hot Full Power (HFP) conditions for each model. To examine different aspects of reactor performance, the concentration of the gadolinia absorber was varied, and its impact on parameters such as Temperature Coefficients (TCs), Power Peaking Factors (PPFs), and excess reactivity was investigated. The investigation involved simulations using the WIMS and CITATION neutronic calculation codes. The gadolinia concentration was effectively optimized using the Particle Swarm Optimization (PSO) algorithm. The best model, based on TCs and PPFs, was selected. The analysis findings revealed significant improvements in safety parameters such as TCs and PPFs for optimal core loading pattern with optimized gadolinia (Gd2O3) Concentration. © 2025 The Author(s)
Nuclear Engineering and Design (00295493)433
An investigation into the influence of innovative dual-cooled fuel geometric structural characteristics on critical reactor core parameters, such as temperature reactivity coefficients and convective heat transfer coefficient, is essential for accurately assessing its neutronic and thermohydraulic performance and estimating safety margins. This study calculated the fuel and coolant/moderator temperature reactivity coefficients in the NuScale-type reactor using WIMS-CITATION codes. Additionally, the impact of increasing the size of dual-cooled fuel rods on these coefficients was analyzed. The study also calculated the hot rod convective heat transfer coefficient for proposed fuel rods using a Computational Fluid Dynamics (CFD) and sub-channel model method. The results demonstrated that increasing the internal radius of the fuel decreases the temperature reactivity feedback coefficient of fuels, while the feedback coefficient of coolant exhibits a parabolic trend. All proposed fuel rods exhibited negative coolant and fuel temperature reactivity coefficients, and increasing the internal radius resulted in a reduction in the convective heat transfer coefficient. Furthermore, the study explored the use of Artificial Neural Network (ANN) and Gene Expression Programming (GEP) models to develop an optimal method with low computational cost. Based on statistical indicators, the ANN was found to outperform GEP. Finally, the designed ANN, coupled with the Henry Gas Solubility Optimization (HGSO) algorithm, was employed to determine the optimal dual-cooled fuel based on desired parameters. © 2025 Elsevier B.V.
Nuclear Science and Techniques (22103147)35(1)
Nuclear power plants exhibit non-linear and time-variable dynamics. Therefore, designing a control system that sets the reactor power and forces it to follow the desired load is complicated. A supercritical water reactor (SCWR) is a fourth-generation conceptual reactor. In an SCWR, the non-linear dynamics of the reactor require a controller capable of controlling the nonlinearities. In this study, a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode, and the reactor outgoing steam temperature and pressure were controlled simultaneously. In an SCWR, the temperature, pressure, and power must be maintained at a setpoint (desired value) during power maneuvering. Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation. Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers specified in previous studies. The controlled parameters were reactor power, steam temperature, and pressure. Notably, for these parameters, the PI controller had certain instabilities in the presence of disturbances. The classic sliding mode controller had a higher accuracy and stability; however its main drawback was the chattering phenomenon. HOSMC was highly accurate and stable and had a small computational cost. In reality, it followed the desired values without oscillations and chattering. © 2024, The Author(s), under exclusive licence to China Science Publishing & Media Ltd. (Science Press), Shanghai Institute of Applied Physics, the Chinese Academy of Sciences, Chinese Nuclear Society.
Kerntechnik (21958580)89(3)pp. 292-300
The Steam Generator (SG) is a crucial component of a nuclear power plant. Proper water level control in a nuclear steam generator is of great importance to ensure a sufficient cooling source for the nuclear reactor and to prevent damage to turbine blades. The water level control problem of steam generators has been a leading cause of unexpected shutdowns in nuclear power plants, which must be addressed for plant safety and availability. The control problem is challenging, particularly at low power levels due to shrink and swell phenomena and flow measurement errors. Furthermore, the dynamics of the steam generator vary as the power level changes. Therefore, there is a need to enhance the water level control system of the SG. In this paper, a Gain Scheduled Internal Model Control (IMC) based on the Dynamic Sliding Mode (DSMC) method is developed for the level control problem. The proposed method exhibits the desired dynamic properties throughout the entire output tracking process, independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. The simulation results confirm the improvement in transient response achieved by using the proposed controller. © 2024 Walter de Gruyter GmbH, Berlin/Boston.
Nuclear Engineering and Design (00295493)421
Although nuclear accidents have occurred in the past, advancements in technology and safety measures have made nuclear energy a viable and competitive option for energy generation. However, within nuclear reactors, the issue of spatial oscillations in neutron flux distribution caused by reactivity feedback of xenon must be addressed. Uncontrolled oscillations in power distribution may lead to exceeding power density and power change limits in certain regions of the reactor core, thereby increasing the likelihood of fuel failure. It is imperative to detect the presence of Xenon Oscillations (XO) and devise an appropriate control strategy to regulate spatial power distribution during the design phase of any Pressurized Water Reactor (PWR). This paper introduces a novel approach for maintaining bounded oscillations of axial power distribution within acceptable limits during load-following operation. The proposed method is an Adaptive Robust Feedback-Linearization Control (ARFLC) that utilizes the multi-point kinetics reactor model, incorporating both neutronic and thermal–hydraulic aspects. Furthermore, the paper provides a stability analysis using the Lyapunov approach, ensuring that the system remains stable over a wide range. This paper presents dynamic simulations utilizing a nodal core model of a PWR to illustrate the effectiveness and robustness of the presented method. The simulations aim to demonstrate that the parameters obtained accurately represent the system and provide satisfactory results, notwithstanding the uncertainties and disturbances present in these parameters. The outcomes conclusively exhibit the exceptional accomplishment of implementing ARFLC for robust control across diverse applications, effectively managing system behavior in the face of challenging factors such as model uncertainties and noise disturbances. Moreover, a comprehensive evaluation was conducted, comparing the performance of the newly developed controller with that of existing conventional controllers like Dynamic Sliding Mode Control (DSMC), Feedback-Linearization Control (FLC), and Sliding Mode Control (SMC). Remarkably, the evaluation unveiled a substantial enhancement in load-following capabilities for the proposed controller system, solidifying its superiority and the potential for broader implementation. © 2024 Elsevier B.V.
This paper proposed a new method for maximum power point tracking in photovoltaic power generation systems by combining super-twisting sliding mode control and active disturbance rejection method. An incremental guidance method is used to find the point of maximum power. The non-linear extended state observer is applied to estimate the unmodeled dynamics and external disturbance. The ADRC based on a super-twisting sliding mode is designed to bring the state variables to the desired state. In the next step, the stability of NESO and ADRC are theoretically proved. Finally, the simulation results have been compared with the results of the PI controller, classical sliding mode control, and terminal sliding mode control (TSMC) presented in other articles. The results show the effectiveness and superiority of the proposed method. Also, to check the performance of the proposal method in real-time, real-time results have been compared with non-real-time results. The results obtained from the real-time and non-real-time simulations exhibited a minimal difference. This fact indicates the high accuracy of the modeling and simulations performed. Indeed, the mathematical models and non-real-time simulations have been able to accurately mimic the actual behavior of the photovoltaic system under various operating conditions. © 2024 The Authors
International Journal of Energy Research (1099114X)46(7)pp. 8838-8871
The reactor design includes optimizing parameters such as fuel composition. In this research, the issue of fuel composition optimization in a Small Modular Nuclear Reactor (SMR) is investigated. Indeed, considering the importance of fuel management optimization at the core of a nuclear reactor as a fundamental issue, this study analyzes gadolinium concentration effects on the neutronic and thermal hydraulic parameters in the NuScale reactor. In addition, optimizing the gadolinium concentration in fuel composition has been done via a genetic algorithm (GA) as a Machine Learning method. At first, the core of the NuScale reactor was modeled using neutronic codes (WIMS & CITATION). Then, the core of this reactor was simulated in different concentrations of natural gadolinium with 141 different concentration combinations in fuel assemblies through the mentioned neutronic codes. Furthermore, the neutronic parameters, including excess reactivity, radial and axial power peaking factors (PPFradial, PPFaxial) related to each design, were calculated. Thermal-hydraulic modeling of the hot channel for each design created was performed using ANSYS FLUENT code, and thermal-hydraulic parameters including heat transfer coefficient, MDNBR, pressure drop, Vout-Vin, and Vmax/Vavg, have been obtained. An artificial neural network (ANN) was trained using the obtained data. Finally, optimal gadolinium concentrations in fuels were determined using the ANN by implementing the GA. In the core of the conventional NuScale reactor, there are assemblies with 2%, 6%, and 8% gadolinium concentrations. Via optimization algorithm, in this paper, two sets of optimal gadolinium concentrations have been presented using different appropriate cost functions. First cost function proposed the optimal concentrations as 1.4256%, 4.2606%, and 5.4968%, while, based on the second one, 1.4502%, 4.2552%, and 5.5296% are optimal concentrations. Finally, the reactor core containing the optimal fuel composition has been redesigned and compared with the conventional NuScale reactor core. Most of the neutron and thermal-hydraulic parameters significantly improved over the NuScale reactor's optimally designed core. This optimization leads to better fuel management and safety in the optimized core than in the core of the NuScale reactor and can also economically increase power plant efficiency. © 2022 John Wiley & Sons Ltd.
International Journal of Energy Research (1099114X)46(7)pp. 9456-9485
A nuclear reactor could be a suitable option in space needs such as planetary and space exploration. A low-enriched uranium fuel (LEU) could be utilized by member states of the Non-Proliferation Treaty (NPT) instead of Highly Enriched Uranium (HEU) fuel for security reasons. This paper deals with a conceptual design and optimization of low-enriched uranium (20 wt% enriched uranium in UO2 fuel and 100 kWe) integral space nuclear reactor core according to the neutronic and dynamic analysis. The existing HEU-fueled reactor is considered for comparison. For neutronic evaluation, different fuel mixtures and reflector thicknesses are considered to derive reactivity of excess, power peaking coefficient in radial direction, reactivity factor of fuel temperature, and reactivity factor of coolant temperature. An Adaptive Neuro-Fuzzy Inference System (ANFIS) is implemented for data estimation. This system is a machine learning method that combines the features of neural networks and fuzzy systems. The initial ANFIS parameters are optimized using optimization algorithms. The optimized ANFIS is coupled with the HGAPSO algorithm to derive an optimally designed fuel mixture and reflector thickness considering a cost function. The simulation results show that the coupled ANFIS-HGAPSO method has better performance and superiority (less error) compared with the ANFIS-GA and ANFIS-PSO methods. In addition, the optimized fuel mixture and core geometry have better performance compared to the present design. In each optimization process (fuel mixture and reflector thickness) the optimally designed model is validated utilizing neutronic computations. The overall optimal design of this reactor can provide 7000 days of full-power operation which is significantly more than the core cycle length of the existing HEU-fueled space nuclear reactor. In addition, this scheme has more reactor safety and stability because of a more negative reactivity factor of fuel temperature. Highlights: This paper, for the first time, deals with a conceptual design and optimization of low-enriched uranium kilowatt-class (20 wt% enriched uranium in UO2 fuel and 100 kWe) integrated space nuclear reactor core based on the neutronic and dynamic analysis. neutronic analysis is performed considering different fuel compositions and reflector thicknesses to obtain excess reactivity, radial power peaking factor, fuel reactivity temperature coefficient, and coolant reactivity temperature coefficient The optimized ANFIS is coupled with the HGAPSO algorithm to find an optimal design of the fuel composition and reflector thickness based on an objective function. The simulation result shows that the optimized fuel composition has better performance in comparison with the existing design. The overall optimal design of this reactor can provide 7000 days of full-power operation which is more than core cycle length of existing HEU-fueled space nuclear reactor, significantly. The optimal design is validated using neutronic calculations. © 2022 John Wiley & Sons Ltd.
Annals of Nuclear Energy (03064549)169
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 (Vmax/Vavg, Vout-Vin, 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. © 2021 Elsevier Ltd
Progress in Nuclear Energy (01491970)153
In this paper, the effect of hybrid Nanofluids on heat transfer of flow around a vertical rod with cosine heat flux is investigated. TiO2 and Al2O3 Nanoparticles are added to deionized water to produce hybrid Nanofluids. Experiments are conducted for both turbulence and laminar flow regimes. The test rig is an annulus in which a pitch-variable helical element applies a cosine heat flux to its inner tube. Results reveal that the increase in concentrations would reduce maximum rod temperature significantly. This is near l20% for 1.5%Al2O3 compared to water. The effect of pressure was also studied. The data obtained at pressures of 1, 2, 5, and 10 bar shows that pressure does not affect heat transfer performance. The thermal enhancement of Nanofluid decreases as the Reynolds number increases. © 2022 Elsevier Ltd
International Journal of Energy Research (1099114X)46(15)pp. 22314-22335
The main purpose of this paper was to review nanofluid applications as the main coolant of pressurized water reactors (PWRs) with both solid and dual-cooled annular fuel. In this regard, first, theoretical and experimental models for calculating thermo-physical properties of nanofluids including density, specific heat, thermal conductivity, viscosity, and heat transfer coefficient were reviewed and the effects of different variables such as particle volume concentration, particle size, shape, and temperature on these parameters were determined. In the following, research related to neutronic, thermal-hydraulic, and safety investigations of nanofluids as the main coolant of PWRs with solid and dual-cooled annular fuel were reviewed. Corresponding methods, codes, and results were discussed and compared. Based on studies, it was concluded that nanofluids' effects on key neutronic parameters (keff, neutron flux and radial and axial power peaking factor) depend on the used nanoparticle, volume concentration, and nanoparticle deposition on the cladding, while nanofluids can increase heat transfer, critical heat flux, and minimum departure from nuclear boiling ratio (MDNBR). Also, it was found that nanofluids with low volume concentration in reactor core with dual-cooled annular fuel can replicate the same advantages. Furthermore, using dual-cooled annular fuel instead of solid fuel increases MDNBR margin and reduces coolant flow which leads to a more compact design of the reactor core and a reduction of capital cost. Finally, incoming challenges and future works were presented. © 2022 John Wiley & Sons Ltd.
Progress in Nuclear Energy (01491970)143
A nuclear reactor is an attractive choice in space applications such as planetary and deep space exploration. This paper deals with an optimal design for a 100 kWe integrated space nuclear reactor core based on the dynamic analysis. This type of reactor has advantages such as compact structure, high power density, lower cost, higher safety and reliability, and proven technology. In this study, neutronic analysis is performed considering different fuel compositions to obtain excess reactivity, radial power peaking factor, fuel reactivity temperature coefficient, and coolant reactivity temperature coefficient. An artificial neural network is trained considering the extracted data of neutronic codes. Genetic Algorithm (GA) and Particle Swarm Optimization (PSO) are implemented to find an optimal design of the fuel composition based on an objective function. The trained neural network is considered in the optimization process. The simulation result shows that the optimized fuel composition has better performance in comparison with the existing design. Also, the optimal design is validated using neutronic calculations. © 2021 Elsevier Ltd
Nuclear Engineering and Design (00295493)385
In this paper, a new fresh Small Modular Nuclear Reactor core with changes on the Fuel Assembly-type, fuel enrichment, and gadolinium (Gd) concentration of the Fuel Assemblies (FAs) is designed and proposed to improve and enhance the NuScale nuclear reactor as a typical PWR-Small Modular Reactor. An algorithm is additionally presented to redesign the core consisting of square lattice FAs into a core including hexagonal FAs. This algorithm is implemented based on the utilization of MCNP code which classifies the design criteria considering simplified modeling and complicated modeling with high computational cost. The algorithm is helpful for a designer to manage the computational cost of modeling during the design process. Core design Basis Limits including Neutronic and Thermal-hydraulics (TH) analysis, are performed and verified for the first cycle of the presented new Small Modular Nuclear Reactor as a Hexagonal NuScale (HNu). Via ANSYS-FLUENT& MCNP codes, TH and Neutronic analysis have been coupled. Also, the burnup calculations for the presented core have been done to analyze the core first cycle length. The excess reactivity and power peaking factor have been assessed during the first-cycle length. Average core fuel depletion has been taken into account, and masses of produced fission fragments and build-up of plutonium isotopes and the remaining mass of uranium isotopes (and Gd) have been computed at the end of the cycle (EOC). The presented new core (HNu) has the benefits of both the TVS-2M FA and NuScale. Compared to the NuScale, HNu has several advantages and superiority, such as increasing MDNBR, decreasing the maximum temperature of fuel and clad, more fuel burn-up, longer working cycle length, and decreasing the maximum power peaking factor. © 2021 Elsevier B.V.
Progress in Nuclear Energy (01491970)141
Load-following control of a nuclear reactor core is very crucial. The main challenge in examining power distribution control of the reactor core is to perform axial power distribution control rather than handling radial power distribution. In this work, a new model consisting of fractional order derivatives and two nodes of the reactor core with reactivity feedbacks is presented named as the two-point fractional neutron point kinetics (TPFNPK) model. According to neutron diffusion phenomena, this modeling method has a better physical interpretation for a quick change of neutron flux in load-following operation mode. For load-following control, a fractional order PID controller is implemented. Optimization of the parameters of this controller (KP, KI, λ, KD, and μ) is carried out utilizing a genetic algorithm (GA) which is an evolutionary algorithm. The objective is the minimization of the weighted sum of the integral of time-weighted absolute error (ITAE), and the overshoot or the undershoot (MP). The new reactor core model is simulated in a MATLAB/SIMULINK environment coupled with a genetic algorithm. Different load patterns (the assumed tracking signal) considering different values of the model parameter (α) are introduced to address the effectiveness of the implemented control method. The simulation result demonstrates that the optimized control methods (GA-PID and GA-FOPID) have satisfactory and suitable functionality during load pattern tracking for all selected values of the model parameter. However, the GA-FOPID controller has a relatively better performance compared to the GA-PID controller. In all cases, it has lower costs and faster convergence. In addition, core normalized axial offset (AO) is always maintained within a certain limit during load-following operation for all the cases. Furthermore, in all the cases, the normalized axial xenon oscillation index (AXOI) remains bounded during the power transient. © 2021 Elsevier Ltd
Progress in Nuclear Energy (01491970)138
In recent years, “Nanofluid” and “Energy-Efficient Cooling” have been achieved growing attention for use in new nuclear reactor technology. Also, the assurance of reactor safety has become of top priority in nuclear energy development because of the significant role of nuclear power in providing worldwide electricity. The use of nanofluid coolant is an effective way to enhance safety margins and heat transfer performance. Studies also prove that nanofluids are promising cooling fluids for applying as a future coolant in nuclear reactors. The current paper investigates the nanofluid effects as a coolant on the velocity of bulk, pressure drop, heat transfer coefficient, power peaking factors, and minimum departure from nucleate boiling ratio in a NuScale reactor, a natural circulation pressurized water reactor (PWR). It also analyzes the thermal characteristics of Al2O3 nanofluid with specific attention to their boiling heat transfer behavior. The study first presents the NuScale designed core with the use of water and different concentrations of Al2O3 nanofluid (0.001–10% volume fractions and 10–90 nm particle sizes). Then it shows the simulated equivalent cell with surrounding coolant utilizing computational fluid dynamic code (CFD). The study also describes the outcomes of replacing water with nanofluid coolant on natural circulation parameters, heat transfer coefficient, and minimum departure from nucleate boiling ratio with changes in nanoparticle concentrations and sizes. The results demonstrate the potential of this innovative coolant to enhance the minimum departure from nucleate boiling ratio and heat transfer coefficient. Overall, the presence of Alumina nanoparticles in water improves thermal performance and safety. The investigation also shows that the nanoparticles do not change the power peaking factor and neutronic performance significantly. © 2021 Elsevier Ltd
Kerntechnik (21958580)86(6)pp. 419-436
The purpose of this study is to display the neutronic simulation of nanofluid application to reactor core. The variations of VVER-1000 nuclear reactor primary neutronic parameters are investigated by using different volume fraction of nanofluid as coolant. The effect of using nanofluid as coolant on reactor dynamical parameters which play an important role in the dynamical analysis of the reactor and safety core is calculated. In this paper coolant and fuel temperature reactivity coefficients in a VVER-1000 nuclear reactor with nanofluid as a coolant are calculated by using various volume fractions and different sizes of TiO2 (Titania) nanoparticle. For do this, firstly the equivalent cell of the hexagonal fuel rod and the surrounding coolant nanofluid is simulated. Then the thermal hydraulic calculations are performed at different volume fractions and sizes of the nanoparticle. Then, using WIMS and CITATION codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated. For doing optimization, an artificial neural network is trained in MATLAB using the observed data. The different sizes and various volume fractions are inputs, fuel and coolant temperature reactivity coefficients are outputs. The optimal size and volume fraction is determined using the neural network by implementing the genetic algorithms. In the optimization, volume fraction of 7% and size 77 nm are optimal values. © 2021 Walter de Gruyter GmbH, Berlin/Boston, Germany.
Annals of Nuclear Energy (03064549)161
This investigation studies the optimization of water-based alumina nanofluid to advance heat transfer and safety performance of the NuScale natural circulation reactor. First, comprehensive CFD and neutronic simulation is employed to design a reactor core using nanofluid coolant (0.001–10% volume fractions and 10–90 nm particle sizes). Consequently, the outlined results prove a sufficient enhancement of safety and heat transfer parameters by applying nanofluid coolant. Next, a developed Artificial Neural Network (ANN), utilizing the obtained data, predicts the thermal–hydraulic and neutronic parameters of the NuScale reactor core with Al2O3/Water nanofluid. Achieving the optimal vol% and size of nanoparticles by implementing Genetic Algorithms (GA), Particle Swarm Optimization (PSO), and Hybridized PSO-GA, based on the developed ANN results, are the main goals of this work. These optimization algorithms, which have a significant ability to attain the best solutions, also determine the optimal values of natural circulation parameters (Vmax/Vavg, Vout-Vin, and pressure drop), heat transfer coefficient, MDNBR, RPPF, and excess reactivity, for obtained vol% and size. The validation results demonstrate the efficiency of the developed ANN and these three evolutionary computation algorithms for optimization. The differences between the outcomes of implemented algorithms, focusing on how each works and affects optimal solutions in problem space, are also described. Finally, this paper compares the results of optimal design with conventional NuScale, which uses water coolant. This comparison indicates the potential of the proposed nanofluid coolant to increase thermal and safety performance. © 2021 Elsevier Ltd
Annals of Nuclear Energy (03064549)161
Pressurizer is one of the most important components in a pressurized water nuclear power plant. Since it is responsible for coolant mass balance and non-boiling heat transfer in the primary circuit. The pressurizer water level control system in a nuclear power plant with the Pressurized Water Reactor (PWR) is responsible for the coolant mass balance. The main control goal is to stabilize the water level at a reference value and to suppress the effect of time-varying disturbances such as the coolant leakage in the primary circuit pipeline system. In the process of PWR power plant operation, the incorrect pressurizer water level may disturb pressure control or may cause damage to electric heaters which could threaten plant security and stability. In modern reactors, standard Proportional–Integral–Derivative (PID) controllers are used to control the water level in a pressurizer. In this paper, for the first time, in order to improve the performance and transient response of the pressurizer control system, an observer-based dynamic sliding mode control using Lyapunov-approach is designed and applied for the level control problem of the pressurizer in a pressurized water nuclear power plant based on the validated nonlinear model. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. The comparison between the Dynamic Sliding Mode Control (DSMC) and the PID controller shows a significant improvement in water set-point tracking and increased ability in disturbance rejection for the proposed observer-based DSMC. Simulation results confirm the improvement in the transient response obtained by using the proposed controller. © 2021 Elsevier Ltd
Annals of Nuclear Energy (03064549)138
Recently, a kind of fuel called “dual-cooled annular fuel” is used in the nuclear reactors, which allows the coolant to flow from the inside and outside of the fuel rod and finally causes a substantial increase in power density. The dual-cooled annular fuel rod has both inner and outer coolant channels. Also, Small Modular Reactors (SMRs) are the innovative design of nuclear reactors making remarkable interest during recent years. Since there is not enough available operating experience on SMRs, it might be possible to initiate extensive investigations on these types of reactors for the purpose of improving the current performance level of these systems, significantly. In this paper, for the first time, the core of a Small Modular Reactor (SMR) is designed based on the use of internally and externally cooled annular fuels. And then, neutronics and natural circulation parameters of this type of reactor are analyzed. For this purpose, at the first, the dual-cooled annular fuel under clean and cold conditions is modeled for the different inner clad diameter with same distance gap to gap of fuel rod and effective multiplication factor has been calculated. Then, these annular fuels under full power conditions are modeled for each inner clad diameter and power peaking factor has been calculated. Finally, natural circulation parameters calculations are performed for a simulated fuel rod in the hot channel using computational fluid dynamics simulation codes. These calculations are compared with a conventional NuScale reactor that does not use this kind of fuel. One of the most prominent of its advantages is the ability to moderate more of the neutrons in the reactor core with this type of fuel, which can increase the core effective multiplication, reduce total power peaking factor and improve natural circulation parameters. Increasing of the effective multiplication coefficient at cold and clean condition leads to increase the excess reactivity of the core which increases the operation cycle without refueling and it is useful from fuel management view-point. Also, it was found that by increasing the inner radius by more than 0.5 (cm), the power peaking factor increases, which is due to the more increasing of the pitch lattice relative to the mean free-path and excessive reduction of the mean power-density in the core and, also, increasing the flux gradient. © 2019 Elsevier Ltd
Annals of Nuclear Energy (03064549)148
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. © 2020 Elsevier Ltd
Annals of Nuclear Energy (03064549)142
In nuclear reactor, spatial oscillations in neutron flux distribution resulting from xenon reactivity feedback are a matter of concern. If the oscillations in power distribution are not controlled, power density and rate of change of power at some locations in the reactor core may exceed their respective limits causing the nuclear power plant instability. Therefore, during the design stages of any Pressurized Water Nuclear Reactor (P.W.R), it is necessary to identify the existence of xenon oscillation and to design suitable control strategy for regulating the spatial power distribution. In each nuclear power plant, Load-following control is one of the most important techniques for nuclear reactor regulation. A novel nonlinear controller called observer-based adaptive robust feedback-linearization controller for VVER-1000 nuclear reactors is presented in this paper. This novel control strategy is then applied to the Axial power distribution control during load following operation in the VVER-1000 nuclear reactors. The reactor core is simulated based on the validated four nodes kinetics reactor model and three groups of delayed neutron precursor's concentration based on the Skinner-Cohen model. Considering the limitations of the xenon concentrations and delayed neutrons precursor's densities physical measurement, an adaptive sliding mode observer is designed to estimate their values and finally adaptive robust feedback-linearization based on the adaptive sliding mode observer and nodal kinetics reactor model is presented to AO control during load following operation for nuclear reactors. The stability analysis is given by means Lyapunov approach, thus the designed control system is guaranteed to be stable within a large range. Simulation results show that robust control and state estimation with adaptive robust feedback-linearization and adaptive Sliding mode methodologies can be achieved in nuclear plant systems with diverse applications including control and estimation in the presence of model uncertainties and external disturbances. One significant finding to emerge from this paper is that the Observer based Adaptive robust feedback-linearization control method provides a robust, high-performance and automatic control mechanism at all the power ranges of operation for P.W.R control system. © 2020 Elsevier Ltd
Nuclear Engineering and Design (00295493)360
Small Modular Reactors (SMRs) are the creative design of nuclear reactors that have received considerable attention in recent years. Since there is insufficient operational experience in SMRs, extensive research on these types of reactors may be needed to improve the current level of these systems performance. Also, one of the approaches which can help the enhancement of a reactor power is changing its fuel geometry. For this purpose, as well as decreasing the maximum fuel temperature in the reactors, the technology of annular fuels with the ability of internal and external cooling shows its importance. The nuclear SMRs with annular fuel which is cooled internally and externally has the potential to increase high power density while maintaining a sufficient safety margin. The benefit of the dual cooled annular fuel design is that heat can be transferred to the coolant at both the outer and inner channels. Hence, in this study, for the first time, dual cooled annular fuel in the Small Modular Reactors is considered and studied thoroughly. In this paper, neutronics and natural circulation parameters are calculated for an SMR with Annular Fuel which is cooled internally and externally, with changes of the internal radius. By analyzing the fuel internal radius changes in a specific range via the neutronics and computational fluid dynamics (CFD) simulation codes, the effects of power peaking factor, effective multiplication coefficient and natural circulation parameters are investigated. For the purpose of data fitting, an artificial neural network is trained using the observed data. The input consists of different internal radiuses which yields to output consisting effective multiplication factor, power peaking factor and natural circulation parameters. The Optimal geometry of annular fuel is determined using the neural network by implementing the genetic algorithms based on these neutronics and natural circulation parameters. Finally, using a simulated optimal geometry in the hot channel by neutronics and CFD simulations, thermal–hydraulic and neutronics calculations are accomplished, and these parameters are compared with the conventional SMR which uses solid (conventional) fuel. One of the main results of the evaluation is that optimal internally and externally cooled fuel in the presented SMR has a satisfactory margin for the departure from nucleate boiling (DNB) and maximum fuel rod temperature relative to the solid fuel which is used in the conventional SMRs. Indeed, fuel center temperature of conventional SMR is much more than that of presented optimal SMR. Also, presented optimal geometry increases the core effective multiplication coefficient which leads to increase in the excess reactivity. Also, reduces total power peaking factor and improves natural circulation parameters compared to the conventional SMR. © 2020 Elsevier B.V.
Annals of Nuclear Energy (03064549)148
One of the approaches which can help the enhancement of a reactor power is changing its fuel geometry. For this purpose, as well as decreasing the maximum fuel temperature in PWR reactors, the technology of annular fuels with ability of internal and external cooling shows its importance and has been considered widely. Such fuels are investigated in western PWR and VVER-1000 reactors. Hence, in this study, this fuel in VVER-1000 reactors are considered and studied thoroughly. In this paper, fuel and coolant reactivity temperature coefficients are calculated for a VVER-1000 Nuclear Reactor with Dual Cooled Annular Fuel with changes of internal radius and different power levels and effects of the fuel internal radius on the reactivity temperature coefficients are investigated. By analyzing the fuel internal radius changes in a specific range in the neutronic code, the effects of effective multiplication factor are investigated. For purpose of data fitting, an artificial neural network is trained using the observed data. The input consists of different internal and external radiuses, outputs consist of pitch, fuel and coolant reactivity temperature coefficients. Finally, the optimal geometry of fuel is determined using the neural network by implementing the genetic algorithms based on these dynamic-coefficients. In the optimization process, it has been shown that having an internal radius 2.67 mm and external radius 6.95 mm provides an optimal geometry. Also, validation of the designed artificial neural network and genetic algorithm has been done using neutronic and Thermal hydraulic calculations. © 2020 Elsevier Ltd
Annals of Nuclear Energy (03064549)139
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. © 2019 Elsevier Ltd
Annals of Nuclear Energy (03064549)133pp. 623-636
Increasing efficiency and improving energy consumption in the nuclear power plants have always been of interest for researchers. So, they try to improve the heat transfer in the applied systems. Hence, extensive research has been performed to apply an alternative fluid which has more suitable thermal properties instead of the conventional fluids such as water. One of these efforts is application of nanoparticles in a coolant fluid. The most important advantage of the nanoparticles is increase of the thermal conductivity and heat transfer coefficient. Considering the importance of the nanofluids effect as a coolant in the nuclear reactors on the reactor dynamic parameters, in this paper, for the first time, fuel and coolant temperature reactivity coefficients which have important contribution in the dynamic analysis and safety requirements of the nuclear reactors, are calculated in a VVER-1000 Nuclear Reactor with nanofluid as a coolant. In this study, using different volumetric percentages and sizes of Al2O3 (Alumina) nanoparticle, the important and fundamental parameters of the VVER-1000 reactor, including dynamic reactor parameters such as temperature reactivity coefficients are calculated. For this purpose, at the first, the equivalent cell of the fuel rod and the surrounding coolant nanofluid are simulated in the hexagonal fuel cell of the VVER-1000 reactor. Then, the thermal hydraulic calculations are carried out at different concentrations and sizes of the nanoparticle and their effects on the heat transfer parameters such as the heat transfer coefficient, temperature of coolant and fuel are assessed. Also, using the neutron calculating codes, the reactor core is simulated and the effect of coolant and fuel temperature changes on the effective multiplication factor is calculated and analyzed. Through the thermal hydraulic and neutronics calculations, the fuel and coolant temperature reactivity coefficients are calculated and analyzes versus variation of the concentration and size. © 2019 Elsevier Ltd
IEEE Transactions on Nuclear Science (15581578)66(7)pp. 1804-1812
Core power control of a nuclear reactor during power maneuvering transients is an important issue in a nuclear power plant. This paper, for the first time, presents an optimized PID controller for a nuclear research reactor with a new model of the reactor core, which is based on the fractional neutron point kinetics (FNPK) equations. This model preserves the main dynamic properties of the neutron movement in which the relaxation time related to a quick change in the neutron flux includes a fractional order, acting as an exponent of the relaxation time, to show the dynamic characteristics of the nuclear reactor in the best way. The physical justification of the fractional exponent is associated with nonFickian impressions from the neutron diffusion equation point of view. In addition, in this paper, for the first time, feedback reactivity effects are considered for FNPK model. Genetic algorithm (GA) is used to optimize gains of the PID controller. The objective function is according to the minimization of the integral of time-weighted absolute error (ITAE). The simulation result demonstrates that this optimized control method has satisfactory performance and stability during output-tracking process for higher values of fractional order. © 1963-2012 IEEE.
Progress in Nuclear Energy (01491970)117
After the Fukushima Daiichi nuclear accident, all EU nuclear power plants became subject to safety assessments as stress test programs. The objective was to check and improve the capability of nuclear safety systems in withstanding damage from hazards caused by flooding, earthquakes, terrorist attack or aircraft collision. The two most important nuclear accidents which are considered after the Fukushima accident are loss of ultimate heat sink (LUHS) and station blackout. The LUHS is one of the most important accidents, which can be recognized as a severe accident and is caused by severe natural phenomena. In this paper, for the first time, the LUHS accident and its management as a stress test program is assessed in a VVER-1000/V446 reference nuclear reactor. By applying some portable diesel pumps which can provide long term cooling of the reactor core, steam generators and spent fuel pool with the available NPP site demineralized water for long duration; the stress test strategy is assessed. To estimate the effect of portable equipment on the NPP safety, the probabilistic safety assessment (PSA) is adopted. In this method, fault trees of engineering safety features including the proposed portable equipment and event trees of different initiating events are applied to estimate the frequency of each accident sequence. The safety of the nuclear power plant is assessed by the risk measure of core damage frequency (CDF). The calculated CDF values of LUHS accident with and without deploying the portable diesel pumps are 1.01E-11(yr−1) and 1.419E-8 (yr−1), respectively. The results indicate a significant decrease in CDF of the VVER-1000/V446 nuclear reactor for the LUHS accident by applying portable equipment, introduced as a stress test program. © 2019 Elsevier Ltd
International Journal of Nuclear Energy Science and Technology (17416361)13(3)pp. 274-294
Global core calculations use the diffusion equation to predict theoretically the nuclear reactor behaviour. However, this equation is not valid in strong absorbing media where the neutron spectrum is a rapidly varying function of the position, such as control rods or burnable poisons. In this paper, to overcome this misleading, the Monte Carlo simulation has been performed and the VVER-1000 reactor core in the MCL (Reactor Minimally Controlled Power Level) condition is modelled using the MCNPX code to calculate the reactivity worth of the control rod groups. The calculations in this model are divided into four steps. At first, the integral and differential worth are calculated for control groups 8, 9 and 10 with 50% overlapping and shadowing effect is considered. And in three other steps the integral and differential reactivity worths of control groups 8, 9 and 10 are calculated separately (without overlapping). In each step, the core is maintained critical by variation of the boron concentration. In these processes, the boric acid coefficient is achieved while the core is critical. The results are compared with experimental values and are in good agreement with them. © 2019 Inderscience Enterprises Ltd.
Progress in Nuclear Energy (01491970)117
Fractional neutron point kinetics (FNPK) model has been known for some researchers in the field of nuclear science and engineering for less than a decade. This modeling has a better approximation and physical interpretation for some phenomena, specially, when large variations of neutron cross sections occur and the dynamic behavior of neutron is different. This paper, for the first time, presents a fractional neutron point kinetics model which consists of three delayed neutron groups and the reactivity feedback effects (temperature feedbacks and poison feedbacks). Also, Stability Analysis of linear FNPK model considering reactivity feedback effects for a Research Nuclear Reactor is done. The model has been linearized around the equilibrium operating point. Then, the closed-loop linear fractional neutron point kinetics model considering reactivity feedback effects is transformed from S-domain to W-domain. For stability evaluation, two strategies have been implemented. The first strategy is the assessment of the region of closed-loop poles in the principal Riemann sheet (PRS) for the transformed system. The second one is the step response evaluation of the linear fractional-order closed-loop model for different values of the model parameters (including relaxation times and anomalous diffusion coefficients). Simulation results demonstrate that the stability of the system depends on the values of these model parameters and for one scenario, the closed-loop poles are located in the unstable region of PRS and the step response is unstable. However, for the others, it is stable and settles at an equilibrium value. In addition, a smaller anomalous diffusion coefficient (α), leads to a more unstable closed-loop model (have poles closer or inside the unstable region). Comparative analysis is shown that reactivity feedback effects are considerable on the stability of the research nuclear reactor. Specially, feedback by xenon has a negative effect on the stability and reduces the stability of system. Also, the stability highly depends on the values of α and τ. © 2019 Elsevier Ltd
Annals of Nuclear Energy (03064549)127pp. 53-67
Thermal-hydraulic model is an important tool in deterministic safety analysis of the Nuclear Power Plant (NPP) which would confirm the adequacy and efficiency of provisions in the defense in depth concept to cope with challenges of the plant safety. In this paper, for the first time, a validated thermal-hydraulic model is presented for Deterministic Safety Analysis of the portable equipment to be applied in VVER-1000 nuclear reactor during severe accident. Therefore, the VVER-1000 as a case study reactor model is assessed and evaluated through RELAP5/MOD3.2 thermal hydraulic code during the loss of ultimate heat sink accident. Since LUHS accident is absent in the Final Safety Analysis Report of the nuclear power plant, this accident scenario is analyzed for the first time in this study. The simulation is done for two situations of with and without safety modification. This modification is related to portable equipment application which would maintain the primary or secondary circuit cooling in long term. The results of these simulations indicate that LUHS would lead to fuel failure and finally core meltdown. But, the modified LUHS simulation shows that this designed portable equipment can remove the decay heat of the core or transfer the heat into the ultimate heat sink in a continuous manner until the safety systems recovery become possible. Indeed, the most important achievements of this paper consist of: preparation of emergency instructions, accident management at the time of loss of ultimate heat sink and the training of NPP's personnel. © 2018 Elsevier Ltd
Annals of Nuclear Energy (03064549)121pp. 382-405
Load-following is the most important operation in the nuclear reactors. In the nuclear reactors, imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained bounded within acceptable limits. Otherwise, the nuclear reactor could become unstable. Therefore, bounded imbalance of axial power distribution is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, xenon concentrations and delayed neutrons precursors densities must be available. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the xenon concentrations and delayed neutrons precursors densities and ensure imbalance of axial power distribution are kept bounded within acceptable limits during Load-following operation, an Adaptive Observer based Adaptive Sliding Mode Control based on the two-point kinetics reactor model is presented. The sliding mode method exhibits acceptable tracking performance in the presence of parametric uncertainty only at the expense of high gains and control chattering. Therefore, parameter uncertainties match envisaged by designing adaptive Algorithm for both the control and observer and high chattering authority is inhibited. The adaptation laws to online parameter adaptation and observer gains adaptation are generated using the Lyapunov approach. Also, the stability analysis is given by means of Lyapunov approach, thus the system is guaranteed to be stable within a large range. Simulation results are presented to demonstrate the effectiveness of the proposed controller for the load-following operation, in terms of the robustness and stability. Also, results show that the xenon oscillations kept bounded in the given region and the adaptive parameters are bounded during load following operation and observer follows the actual system variables accurately in the presence of the parameters uncertainties and external disturbances. © 2018 Elsevier Ltd
Nuclear Engineering and Technology (2234358X)50(6)pp. 877-885
Various controllers such as proportional–integral–derivative (PID) controllers have been designed and optimized for load-following issues in nuclear reactors. To achieve high performance, gain tuning is of great importance in PID controllers. In this work, gains of a PID controller are optimized for power-level control of a typical pressurized water reactor using particle swarm optimization (PSO) algorithm. The point kinetic is used as a reactor power model. In PSO, the objective (cost) function defined by decision variables including overshoot, settling time, and stabilization time (stability condition) must be minimized (optimized). Stability condition is guaranteed by Lyapunov synthesis. The simulation results demonstrated good stability and high performance of the closed-loop PSO–PID controller to response power demand. © 2018
Annals of Nuclear Energy (03064549)118pp. 107-121
Load following is an importance topic in the Nuclear Power Plants (NPPs). One of the conventional and simplest ways is the use of Proportional-Integral-Derivative (PID) controller. The reactor power is simulated based on the point kinetic model. PID gains of a nonlinear time-varying system (a PWR NPP) are optimized and scheduled using real-coded genetic algorithm (GA). To this end, the objective function of the decision variables, include the overshoot, settling time and stabilization time (based on the Lyapunov approach) of the system is minimized. The presented control system track demand power level change within a wide range of time. The simulation results demonstrate good stability of this method and show high performance of the optimized PID gains to adapt any changes in the output power. © 2018
Annals of Nuclear Energy (03064549)112pp. 158-169
Reactor power control is one of the most important problems in a nuclear power plant. Considering the importance of the negative feedback reactivity from neutron absorber poisons such as xenon and samarium in the design of the nuclear reactor control system and regarding the limitations of the xenon and samarium reactivity measurement, in this paper, for the first time, a modified higher-order sliding mode observer is presented to estimate the xenon and samarium reactivity in the P.W.R Nuclear Reactors. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and Thermal-hydraulic) and three delayed neutrons groups. Traditional sliding mode technique has intrinsic problem of chattering. To cope with this problem, Higher Order Sliding Mode (HOSM) is used. The employed method is easy to implement in practical applications and moreover, the higher order sliding mode control exhibits the desired dynamic behavior during the entire output-tracking process. Simulation results are presented to demonstrate the effectiveness of the proposed observer in terms of performance, robustness and stability and show that the HOSM observer follows the actual system variables accurately and is satisfactory in the presence of the parameters uncertainties and disturbances. © 2017 Elsevier Ltd
Annals of Nuclear Energy (03064549)112pp. 808-828
In this paper, for the first time, numerical technique using Method of Lines (MOL) based on the Generalized Differential Quadrature method (GDQ) is developed and applied for two energy groups of reactor kinetics and one group of precursor delayed neutrons. Also, the presented method (MOL-GDQ) has been developed for 3D transient benchmark nuclear reactor with two neutron energy groups and six delayed precursor groups. The basic idea of the MOL is to replace the spatial (boundary value) derivatives in the Partial Differential Equations (PDE) with algebraic approximations. In other words, with only one remaining independent variable, we have a system of Ordinary Differential Equations (ODEs) that approximates the original PDE. The advantages of the GDQ method lie in its easy use and flexibility with regard to arbitrary grid spacing. Compared to the conventional low-order numerical techniques such as the finite element and finite difference methods, the GDQ method can yield accurate solutions with relatively much fewer grid points. The numerical technique is applied to three-dimensional space-time neutron diffusion equations with average one group of delayed neutrons in the different nuclear reactors. Also, the presented method (MOL-GDQ) is developed for 3D transient benchmark nuclear reactor with two neutron energy groups and six delayed precursor groups. Simulation results are presented to demonstrate the effectiveness of the proposed method in terms of performance and stability. The results of numerical technique are discussed and compared with the results of traditional methods and strong citations to demonstrate the accuracy of it. © 2017 Elsevier Ltd
Nuclear Engineering and Technology (2234358X)50(1)pp. 97-106
Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping. © 2017
Nuclear Engineering and Technology (2234358X)50(5)pp. 654-664
Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established. © 2018
International Journal of Nuclear Energy Science and Technology (17416361)12(3)pp. 283-293
In this paper, modelling of the Tehran Research Reactor is done using Recurrent Neural Network (RNN) in Loss of Flow Accident (LOFA). TRANS code is calculated as training data mode for each of the scenarios. Supervised recurrent neural network is chosen for modelling and identification system, classified system data and appropriate parameters for modelling function of system have been chosen, then data is classified. In the next step, we choose variant networks to train and compare with each other. Next, an optimised network is chosen according to mean square error parameter and correlation among educational data from TRANS code and network output data. Finally, entrance data related to the unforeseen accident was entered to the system and the predicted results by model and output data of TRANS code were compared. Results demonstrate the appropriate conformity between extraction data of TRANS code and extraction data of the model, which shows appropriate function of the model. Copyright © 2018 Inderscience Enterprises Ltd.
Annals of Nuclear Energy (03064549)103pp. 251-264
Control of the nuclear reactors during load-following operation is the most important problem in nuclear power plants due to safety reasons. In the nuclear reactor, imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. In this paper, for the first time, in order to ensure these oscillations are kept bounded within allowable limits during load-following operation, an adaptive robust control based on the multipoint kinetics reactor model is presented for P.W.R nuclear reactors. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermal-hydraulic) and three delayed neutrons groups. The adaptation laws for updating the reactor parameters are generated using the Lyapunov approach. The stability analysis is given by means Lyapunov approach, then the control system is guaranteed to be stable within a large range. Simulation results are presented to demonstrate the effectiveness of the proposed adaptive control in terms of performance, robustness and stability and show that the adaptive parameters are bounded and stable in the presence of the parameters uncertainties and disturbances. © 2017 Elsevier Ltd
Annals of Nuclear Energy (03064549)108pp. 277-300
In nuclear power plants, imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits; otherwise, the nuclear power plant could become unstable. Therefore, axial xenon oscillations must be considered during the load-following operation. Besides, xenon concentration and axial xenon oscillations are not measurable in practice; therefore, they should be estimated using suitable observer. In this paper, in order to estimate the axial xenon oscillations for the load-following operation, an adaptive estimator is presented based on the multipoint kinetics reactor model. Besides, regarding the limitations of the xenon concentration and delayed neutron precursors densities measurement to design the nuclear reactor control systems, the presented observer can estimate their values, carefully. The adaptation laws for updating the reactor parameters are generated using the Lyapunov approach. In the proposed adaptive estimator, both parameters and state variables of the nuclear reactor are estimated simultaneously. Also, the stability analysis is given by means of Lyapunov approach, thus the system is guaranteed to be stable within a large range. Simulation results are presented to demonstrate the effectiveness of the proposed observer in terms of performance, robustness and stability and show that the adaptive estimator follows the actual system variables and parameters accurately and are satisfactory in the presence of the parameters uncertainties and disturbances. © 2017 Elsevier Ltd
Annals of Nuclear Energy (03064549)101pp. 576-585
An important problem in nuclear power plants is reactor power control. Considering the importance of the neutron absorber poisons such as xenon and samarium in design of the nuclear reactor control system and regarding the limitations of the xenon and samarium concentrations measurement, in this paper, an Extended Kalman Filter (E.K.F) is presented to estimate the xenon & samarium concentrations. Besides, Precursors produce delayed neutrons which are most important in specification of reactor period and control of nuclear reactor, but precursors densities cannot be measured directly and the designed Extended Kalman Filter can estimate delayed neutrons precursors densities. The reactor core is simulated based on the point kinetics equations and three delayed neutron groups. Simulation results are presented to demonstrate the effectiveness of the proposed observer in terms of performance, robustness and stability and show that the Extended Kalman Filter (E.K.F) follows the actual system variables accurately and is satisfactory in the presence of the parameters uncertainties and disturbances. © 2016 Elsevier Ltd
Kerntechnik (21958580)82(5)pp. 586-597
In nuclear reactor imbalance of axial power distribution induces xenon oscillations. These fluctuations must be maintained bounded within allowable limits. Otherwise, the nuclear power plant could become unstable. Therefore, bounded these oscillations is considered to be a restriction for the load following operation. Also, in order to design the nuclear reactor control systems, poisons concentrations, especially xenon must be accessible. But, physical measurement of these parameters is impossible. In this paper, for the first time, in order to estimate the axial xenon oscillations and ensures these oscillations are kept bounded within allowable limits during load-following operation, a robust observer based nonlinear control based on multipoint kinetics reactor model for pressurized-water nuclear reactors is presented. The reactor core is simulated based on the multi-point nuclear reactor model (neutronic and thermalhydraulic). Simulation results are presented to demonstrate the effectiveness of the proposed observer based controller for the load-following operation. © 2017 Carl Hanser Verlag, München.
International Journal of Nuclear Energy Science and Technology (17416361)11(1)pp. 22-55
In this paper, Sliding Mode Control (SMC) which is a robust nonlinear controller is designed to control the pressurised-water nuclear reactor power for the load-following operation problem that ensures xenon oscillations are kept bounded within acceptable limits. Considering neutron absorber poisons and regarding the limitations of the xenon concentration and delayed neutrons precursors densities measurement, a sliding mode observer is designed to estimate their values and finally an SMC based on the sliding mode observer is presented to control the reactor core power. The reactor core is simulated based on the two-point nuclear reactor model and three delayed neutrons groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Copyright © 2017 Inderscience Enterprises Ltd.
Annals of Nuclear Energy (03064549)101pp. 118-127
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. © 2016 Elsevier Ltd
Annals of Nuclear Energy (03064549)88pp. 280-300
The water level control problem of steam generators has been a main cause of unexpected shutdowns of nuclear power plants which must be considered for plant safety and availability. The control problem is challenging, especially at low power levels due to shrink and swell phenomena and flow measurement errors. Moreover, the dynamics of steam generator vary as the power level changes. Therefore, it is necessary to improve the water level control system of SG. In this paper, at the first a nonlinear model based on the fundamental conservation equations for mass, energy and momentum is presented for the nuclear steam generator which is validated with other computer programs and experimental results and then, an adaptive dynamic sliding mode control method is applied for the level control problem of U-tube steam generators based on the presented nonlinear model. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Simulation results confirm the improvement in transient response obtained by using the proposed controller. © 2015 Elsevier Ltd. All rights reserved.
Nuclear Engineering and Design (00295493)296pp. 1-8
In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a fast nuclear reactor. The reactor core is simulated based on the point kinetics equations and one delayed neutron group. Considering the limitations of the delayed neutron precursor density measurement, a sliding mode observer is designed to estimate it and finally a sliding mode control based on the sliding mode observer is presented. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. © 2015 Elsevier B.V. All rights reserved.
Annals of Nuclear Energy (03064549)87(P2)pp. 39-47
Nowadays, many efforts have been made to improve the efficiency of nuclear power plants. One of which is use of the dual cooled annular fuel which is an internally and externally cooled annular fuel with many advantages in heat transfer characteristics. Another is the use of nanoparticle/water (nanofluid) as coolant. In this paper, by combining these two methods, the change in neutronic parameters of the VVER-1000 nuclear reactor core with dual cooled annular fuel attributable to the use of nanoparticle/water (nanofluid) as coolant is presented. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local power peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. As a result of changing the effective multiplication factor and PPF calculations for six types of nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper, Titania, and Zirconia with different volume fractions, it can be concluded that at low concentration (0.03 volume fraction), Zirconia and Alumina are the optimum nanoparticles for normal operation. The maximum radial and axial PPF are found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on the outer and inner clad, a flux and Keff depression occurred and ZrO2 and Al2O3 have the lowest rate of drop off. © 2015 Elsevier Ltd. All rights reserved.
Nuclear Technology (00295450)195(1)pp. 105-109
Dual-cooled annular nuclear fuel, which is an internally and externally cooled annular fuel, has many advantages for heat transfer. One of the most prominent of these advantages is the ability to harvest more of this type of fuel, which can increase the thermal power of nuclear plants. In this technical note, the core of a VVER-1000 reactor is designed based on the use of internally and externally cooled annular fuels. The thermal-hydraulic parameters of the fuel rods in this type of reactor are analyzed. In addition, the uprate of the thermal power in a VVER-1000 reactor using annular fuels is investigated. For this purpose, first, the proper pitch length of fuel rods in the core is designed under clean and cold conditions using cell and core neutronics calculation codes. Then, thermal-hydraulic calculations are performed for a simulated fuel rod in a hot channel using computational fluid dynamics simulation codes. These calculations are compared with a conventional VVER-1000 reactor that does not use this kind of fuel. One of the most important results of the analysis is that annular fuel shows a sufficient margin for the departure from nucleate boiling and fuel pellet temperature relative to cylindrical fuel. The margin seems viable in accommodating a 129% power uprate. © 2016, American Nuclear Society. All rights reserved.
Energy (18736785)98pp. 1-14
In this paper, thermal-hydraulic effects of nanofluid as coolant in VVER-1000 nuclear reactor with annular fuel are investigated. At the first, 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 are analyzed. From the neutronic viewpoint, Alumina is the best nanoparticle for normal operation at low concentration. In this paper, for this nanoparticle, fuel assembly is simulated in the hot channel using CFD (Computational Fluid Dynamics) simulation codes and thermal-hydraulic calculations (maximum fuel temperature, fluid outlet, MDNBR (Minimum Departure from Nucleate Boiling Ratio), etc.) are done. As one of the most important results of the analysis, using the nanoparticles, the heat transfer coefficient in outer coolant, which was already decreased using annular fuel, is increased. Also, by applying the nanoparticle with smaller size and major concentration, MDNBR is increased. © 2016 Elsevier Ltd.
International Journal of Nuclear Energy Science and Technology (17416361)10(1)pp. 28-58
In this paper, an adaptive sliding mode control system is presented to control the Pressurised-Water Nuclear Reactor (PWR) core power. Sliding mode controller is a robust non-linear control but chattering in this method is undesirable; large parameter uncertainties accommodation envisaged by designing adaptive mechanisms for the controller and high chattering authority are inhibited. The reactor core is simulated based on the point kinetics equations and one delayed neutrons group. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The adaptive sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of the perturbations. Copyright opy; 2016 Inderscience Enterprises Ltd.
Journal of Process Control (09591524)46pp. 84-91
Power control of the nuclear reactor is one of the most important subjects in each nuclear power plant. In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a Traveling Wave Nuclear Reactor (TWR) power. The reactor core is simulated based on the point kinetics equations and six delayed neutron groups. Considering the limitations of the delayed neutron precursors densities measurement, a sliding mode observer is designed to estimate their values and finally a sliding mode control based on the sliding mode observer is presented to control the reactor core power. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. Since it has systematic design procedure, it is one of the most powerful solutions to design many practical control systems. The designed control system is evaluated in the presence of disturbances and uncertainties. The results show the robustness and performance of the used control system. © 2016 Elsevier Ltd
International Journal of Nuclear Energy Science and Technology (17416361)10(2)pp. 164-182
Nanoparticles have been investigated during the recent decade. The most important advantage of nanoparticles is increased thermal conductivity coefficient and convection heat transfer coefficient. In this paper, the effects of the combined nanofluid (Al2O3-TiO2/water) on the heat transfer characteristics such as thermal conductivity coefficient, heat transfer coefficient, fuel clad and fuel centre temperatures in VVER-1000 reactor are investigated numerically, based on the K-ωSST Turbulence model. Therefore, at first, the cell equivalents for a fuel rod and its surrounding coolant fluid are obtained in the hexagonal fuel assembly of VVER-1000 reactor. Then, a fuel rod is simulated in the hot channel using CFD simulation codes, and thermo-hydraulic calculations (maximum fuel temperature, fluid outlet, MDNBR, etc.) are done and compared with a VVER-1000 reactor without nanoparticles. As one of the most important results of the conducted analysis, it could be observed that heat transfer and thermal conductivity coefficient increased and DNBR of the nanoparticle combination state was better than that of nanofluid and nanoparticle state. Copyright © 2016 Inderscience Enterprises Ltd.
International Journal of Nuclear Energy Science and Technology (17416361)10(4)pp. 302-312
The most widely used mathematical description of the neutron distribution in nuclear reactors is provided by neutron diffusion theory. If fission source does not balance the leakage and absorption terms, to use steady state diffusion equation we multiply the source term by a constant 1/k, where k is multiplication factor. The equation may be rewritten as an eigenvalue equation. In this paper, Generalised Differential Quadrature (GDQ) method is presented to solve the multi-groups neutron diffusion equations for nuclear reactors as an eigenvalue problem. The main idea of the GDQ is that the derivative of a function at a sample point can be approximated as a weighted linear summation of the value of the function at all of the points in the domain. The comparison between GDQ and Finite Difference (FD) methods shows a significant improvement in the convergence rate for GDQ method with a smaller number of grid points. © Copyright 2016 Inderscience Enterprises Ltd.
Nuclear Science and Techniques (22103147)27(2)
This paper presents findings on the sliding mode controller for a nuclear reactor. One of the important operations in nuclear power plants is load following. In this paper, a sliding mode control system, which is a robust nonlinear controller, is designed to control the pressurizedwater reactor power. The reactor core is simulated based on the point kinetics equations and six delayed neutron groups. Considering neutron absorber poisons and regarding the limitations of the xenon concentration measurement, a sliding mode observer is designed to estimate its value, and finally, a sliding mode control based on the sliding mode observer is presented to control the core power of reactor. The stability analysis is given by means Lyapunov approach; thus, the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications, and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed observerbased controller in terms of performance, robustness and stability. © Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Chinese Nuclear Society, Science Press China and Springer Science+Business Media Singapore 2016.
Nuclear Engineering and Technology (2234358X)47(7)pp. 814-826
The most important advantage of nanoparticles is the increased thermal conductivity coefficient and convection heat transfer coefficient so that, as a result of using a 1.5% volume concentration of nanoparticles, the thermal conductivity coefficient would increase by about twice. In this paper, the effects of a nanofluid (TiO2/water) on heat transfer characteristics such as the thermal conductivity coefficient, heat transfer coefficient, fuel clad, and fuel center temperatures in a VVER-1000 nuclear reactor are investigated. To this end, the cell equivalent of a fuel rod and its surrounding coolant fluid were obtained in the hexagonal fuel assembly of a VVER-1000 reactor. Then, a fuel rod was simulated in the hot channel using Computational Fluid Dynamics (CFD) simulation codes and thermohydraulic calculations (maximum fuel temperature, fluid outlet, Minimum Departure from Nucleate Boiling Ratio (MDNBR), etc.) were performed and compared with a VVER-1000 reactor without nanoparticles. One of the most important results of the analysis was that heat transfer and the thermal conductivity coefficient increased, and usage of the nanofluid reduced MDNBR. © 2015.
Annals of Nuclear Energy (03064549)77pp. 1-22
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. © 2014 Elsevier Ltd. All rights reserved.
Annals of Nuclear Energy (03064549)75pp. 728-735
Reactor power control is one of the most important problems in a nuclear power plant. This paper presents the higher order sliding mode controller (H.O.S.M.C.) which is a robust nonlinear controller for a nuclear research reactor considering the effect of xenon concentration during load following operation. Sliding mode controllers for nuclear reactors were developed before. Traditional sliding mode technique has intrinsic problem of chattering. To cope with this problem higher order sliding mode (HOSM) is used. The nonlinear model of a research reactor (Pakistan Research Reactor-1) has been used for higher order sliding mode controller design and performance evaluation. The reactor core is simulated based on the point kinetics equations and three delayed neutron groups. The model assumes feedback from lumped fuel and coolant temperatures. The effect of xenon concentration is also included. The employed method is easy to implement in practical applications and moreover, the higher order sliding mode control exhibits the desired dynamic behavior during the entire output-tracking process. Simulation results show the effectiveness of the proposed controller in terms of performance, stability and robustness against disturbances. © 2014 Elsevier Ltd. All rights reserved.
Nuclear Engineering and Technology (2234358X)47(7)pp. 838-848
One of the most important operations in nuclear power plants is load following, in which an imbalance of 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. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region. © 2015 Published by Elsevier Korea LLC on behalf of Korean Nuclear Society.
Jabbari, M.,
Hadad, K.,
Ansarifar, G.R.,
Tabadar z., Z.,
Hashemi-tilehnoee m., Annals of Nuclear Energy (03064549)77pp. 129-132
In this study, an in-operation VVER-1000 reactor thermal power is analyzed by a thermal method for the primary circuit and a heat balance procedure for the secondary circuit. The results are compared with the measurements by in-core and out-core neutron flux (power) monitoring instruments. The calculated values of reactor thermal power by the thermal method are comparable with the reactor power measured by the instruments. The thermal method, which is used in the average power calculation, leads to a smaller error in comparison with the other applied methods. © 2014 Elsevier Ltd. All rights reserved.
Nuclear Engineering and Technology (2234358X)47(1)pp. 94-101
This paper presents findings on the second-order sliding-mode controller for a nuclear research reactor. Sliding-mode controllers for nuclear reactors have been used for some time, but higher-order sliding-mode controllers have the added advantage of reduced chattering. The nonlinear model of Pakistan Research Reactor-1 has been used for higherorder sliding-mode controller design and performance evaluation. The reactor core is simulated based on point kinetics equations and one delayed neutron groups. The model assumes feedback from lumped fuel and coolant temperatures. The effect of xenon concentration is also considered. The employed method is easy to implement in practical applications, and the second-order sliding-mode control exhibits the desired dynamic properties during the entire output-tracking process. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. © 2015, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society.
Annals of Nuclear Energy (03064549)75pp. 611-619
Reactor power control is one of the most important problems in a nuclear power plant. In this paper, a sliding mode control system which is a robust nonlinear controller is designed to control the Pressurized-Water Nuclear Reactor (PWR) Power. The reactor core is simulated based on the point kinetics equations and three delayed neutron groups. Considering neutron absorber poisons and regarding the limitations of the xenon concentration and delayed neutron precursors densities measurement, a sliding mode observer is designed to estimate their values and finally a sliding mode control based on the sliding mode observer is presented to control the reactor core power. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. © 2014 Elsevier Ltd. All rights reserved.
Annals of Nuclear Energy (03064549)76pp. 209-217
One of the important operations in nuclear power plants is load-following in which imbalance of 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 considered to be a constraint for the load-following operation. In this paper, sliding mode control (SMC) which is a robust nonlinear controller is designed to control the Pressurized-Water Nuclear Reactor (PWR) power for the load-following operation problem that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to maintain xenon oscillations to be bounded. The constant AO is a robust state constraint for load-following problem. The reactor core is simulated based on the two-point nuclear reactor model and one delayed neutron group. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Results show that the proposed controller for the load-following operation is sufficiently effective so that the xenon oscillations are kept bounded in the considered region. © 2014 Elsevier Ltd. All rights reserved.
Progress in Nuclear Energy (01491970)79pp. 104-114
One of the important operations in nuclear power plants is load-following in which imbalance of 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 considered to be a constraint for the load-following operation. In other hands, precursors produce delayed neutrons which are most important in control of nuclear reactor, but xenon concentration & precursor density cannot be measured directly. In this paper, 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 to estimate the xenon concentration & delayed neutron precursor density of the Pressurized-Water Nuclear Reactor (PWR) using reactor power measurement. The stability analysis is given by means Lyapunov approach, thus the system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications. 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 parameters uncertainties & disturbances. © 2014 Elsevier Ltd.
Progress in Nuclear Energy (01491970)76pp. 36-43
Nuclear steam generator is a critical component of the pressurized-water nuclear power plant that plays an important role in security and efficiency of the nuclear power plant. Therefore, using assured thermo-hydraulic model to simulate the nuclear steam generator has particular importance. In this paper, the numerical solution of void fraction in horizontal nuclear steam generator in the steady state analysis is presented using Drift-Flux Model. For a two phase mixture of a gas or vapor and liquid flowing together in a channel, different internal flow geometries or structure can occur depending on the size or orientation of the flow channel, the magnitudes of the gas and liquid flow parameters, the relative magnitudes of this flow parameter, and on the fluid properties of the two phases. The Drift-Flux Model (DFM) is able to predict the void fraction in different geometries. The drift velocity in various two phase flow regimes for small and big diameter pipes is explained. VVER-1000 nuclear steam generator is simulated by DFM using the FLUENT 6.3.26 code. It is explained that void fraction in horizontal steam generator is strongly affected by using a perforated sheet in the top of the horizontal hot tubes. Validity and superiority of the DFM compared to the other two-phase models is proved. Simulation results are compared with RELAP5 and BAGIRA codes results. The calculated void fraction is in good agreement with measured data. The accuracy of the prediction shows that it is possible to use the DFM for thermal-hydraulic analysis in advanced models in nuclear power plant and other industries. This model can be used for assessment of experimental data and licensing processes. © 2014 Published by Elsevier Inc.
Progress in Nuclear Energy (01491970)56pp. 61-70
Steam Generator (SG) is a crucial component of nuclear power plant. The proper water level control of a nuclear steam generator is of great importance in order to secure the sufficient cooling source of the nuclear reactor and to prevent damage of turbine blades. The water level control problem of steam generators has been a main cause of unexpected shutdowns of nuclear power plants which must be considered for plant safety and availability. The control problem is challenging, especially at low power levels due to shrink and swell phenomena and flow measurement errors. Moreover, the dynamics of steam generator vary as the power level changes. Therefore, it is necessary to improve the water level control system of SG. In this paper, an adaptive estimator-based dynamic sliding mode control method is developed for the level control problem. The proposed method exhibits the desired dynamic properties during the entire output tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability. Simulation results confirm the improvement in transient response obtained by using the proposed controller. © 2011 Elsevier Ltd. All rights reserved.
Communications in Nonlinear Science and Numerical Simulation (10075704)17(1)pp. 414-425
This paper presents a new algorithm for designing dynamic sliding-mode controllers. The proposed controller is based on dynamic sliding manifolds to circumvent the difficulties associated with the conventional sliding mode controllers in the face of non-minimum phase systems. Unlike previous works, a proper and easy to implement algorithm is presented for designing the dynamic sliding manifold which facilitates the design of the controller. The output tracking problem in nonlinear non-minimum phase systems with matched and unmatched disturbances and matched nonlinearities is addressed. Then, the performance of the dynamic sliding mode controller is significantly improved by combining the given dynamic sliding manifold with online parameter adaptation. Simulations results are presented to demonstrate the effectiveness of the proposed sliding mode controller in terms of performance, robustness and stability. © 2011 Elsevier B.V.
Progress in Nuclear Energy (01491970)53(6)pp. 651-663
U-Tube Steam Generator (UTSG) is one of the most important facilities in a pressurized-water nuclear reactor. Poor control of the Steam Generator (SG) water level in the secondary circuit of a nuclear power plant can lead to frequent reactor shutdowns or damage of turbine blades. The control problem is challenging, especially at low power levels due to shrink and swell phenomena and flow measurement errors. In addition, the dynamics of steam generator vary as the power level changes. Therefore, designing a suitable controller for all power levels is a necessary step to enhance the plant availability factor. The purpose of this paper is to design, analyze and evaluate a water level controller for U-tube steam generators using dynamic sliding mode control. The employed method is easy to implement in practical applications and moreover, the dynamic sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Gain scheduling is used to obtain a global water level controller. Simulation results are presented to demonstrate the performance, robustness, and stability of the proposed controller. Computer simulations show that the proposed controller improves the transient response of steam generator water level and demonstrates its superiority to existing controllers. © 2011 Elsevier Ltd. All rights reserved.
A novel method is presented to extract the important parameters of semiconductor detectors using artificial neural networks (ANN). We have designed a feed-forward (FF) Multi-layer Perceptron (MLP) ANN, with supervised training based on Levenberg-Marquardt (LM) back propagation algorithm (BP). In this method the sampling points of the transient current are the inputs of the network and the outputs are some important parameters of the detector, which are selected by sensitivity analysis. The training set is obtained by uniform sampling of the parameter space within the range of typical experimental data. These data are collected from drift-diffusion model for the transient current of laser illuminated hydrogenated amorphous silicon p+-i-n+ diode in reversed bias. At final step the trained ANN is used to extract the parameters of a real detector by using the transient current technique (TCT) measurements. It is observed that the simulated transient current with the extracted parameters by ANN exhibit an excellent agreement with experimental results. The new developments in semiconductor technology and the necessity of having a more accurate estimation of the parameters of advanced semiconductor devices to predict their behavior and also the capabilities of ANN in this field, is explained in section 1. In section 2 a brief review of ANN and its historical applications in designing and extracting the fabrication parameters of semiconductors devices is presented. The new advances in fabrication of amorphous devices, the theoretical bases and advantages of TCT and the signal induction are discussed in section 3. The complexity of hydrogenated amorphous silicon detectors, the trapping model, the simulation steps and also the required equations is the subject of section 4. In section 5 by using the Finite Element Method (FEM), the two dimensional simulation of the transient current of simple p+-i-n+ diode is performed and a method for reduction the execution time is presented. The strategy for extracting the transient current of the detector from the preamplifier output by using the Laplace transform is introduced in section 6. In section 7 the transient current of the detector is used to extract the electric field profile, the electron mobility, and the ionized dangling bonds and electrons collection time. In this section, the effect of lifetime of electrons and holes on the behavior of such detectors is also estimated. In section 8 the ANN structure and modeling scheme and some practical notes for optimizing the performance of ANN structure is presented. The final results and the performance of ANN in validation and testing steps and its excellent ability in extracting the parameters of a real amorphous silicon detector is also discussed in this section. © 2011 by Nova Science Publishers, Inc. All rights reserved.
Design and synthesis of a nonlinear supersonic missile longitudinal dynamics control for angle-of-attack output tracking are presented. In addition a sliding mode observ er is designed to estimate the angle-of-attack which is difficult to measure in practice. The employed method is simple to implement in practical applications and enables the sliding mode control design to exhibit the desired dynamic properties during the entire output-tracking process independent of perturbations. Results of simulations are presented to demonstrate the performance, robustness, and stability of the considered autopilot.
