Israth Jahan

ANALYSIS OF STRUCTURAL SAFETY PARAMETERS AND THERMAL BEHAVIOR OF THE FUEL ASSEMBLY OF A GENERIC VVER-1200 NUCLEAR REACTOR USING ANSYS SHAH NUSRAT JAHAN SHANTA TASNIA UMMA TAHIA ISRATH JAHAN ASHA B.Sc. ENGINEERING THESIS DEPARTMENT OF NUCLEAR SCIENCE AND ENGINEERING MILITARY INSTITUTE OF SCIENCE AND TECHNOLOGY DHAKA, BANGLADESH MAY 2025 ACKNOWLEDGEMENTS The authors would like to express their deep gratitude to Professor Dr. Abdus Sattar Mollah, their research supervisor, for his patient guidance, enthusiastic encouragement and useful critiques of this research work. The authors would also like to thank their respected teachers, for their advice and assistance in keeping the authors updated regarding the recent knowledge needed for the research. The authors acknowledge the generosity of the Head of the Department Colonel Khalid Mahmud, for financial allocation that made this study possible. The authors would also like to thank all the laboratory assistants for their support in using the software. Finally, the authors wish to thank their encouragement throughout the sturdy. II friends for their support and ABSTRACT Analysis of Structural Safety Parameters and Thermal Behavior of The Fuel Assembly of a Generic VVER-1200 Nuclear Reactor Using ANSYS This study focuses on analyzing the thermo-mechanical behavior of fuel assembly in VVER-1200 nuclear reactors to ensure their safe and efficient operation. Nuclear energy is considered a clean source because it emits minimal carbon dioxide and other greenhouse gases, helping to reduce air pollution from fossil fuels. To address global warming, the energy sector must shift to a low-carbon system within the next 20 to 30 years. Bangladesh aims to support this by operating two VVER-1200 nuclear power plants at Rooppur by 2025/2026, aligning with the UN's goal for affordable and clean energy (SDG 7). The efficient and safe operation of nuclear reactors significantly depends on the performance and integrity of their fuel rods and core. Anticipating the thermo-mechanical behavior of fuel rod and fuel assembly in a nuclear power plant is essential to designing them effectively and averting failures during operation. The core of a nuclear reactor is usually constructed from cylindrical fuel elements that enclose fuel rods, a helium gas gap, and cladding material. And a fuel assembly is the combination of fuel rods. The objective of this study is to analyze the structural behavior and calculate safety factors. Also by using thermal analysis, temperature distribution along the radial axis of the fuel assembly is analyzed. The mechanical parameters such as deformation, stress distribution of fuel assembly have been evaluated with ANSYS software. The study develops a detailed geometric model of the UO₂ fuel rod with a zirconium alloy cladding and helium gas gap, reflecting accurate physical and material properties. Using ANSYS, distribution simulations under steady-state conditions reveal critical temperature gradients and peak temperatures. Structural analysis assesses stress and strain from thermal expansion and operational loads, identifying high-stress regions and potential deformation. The combined and structural analysis offers a comprehensive understanding of the fuel rod's behavior in the VVER reactor by safety factor and temperature distribution profile. III TABLE OF CONTENTS DECLARATION I ACKNOWLEDGEMENTS II ABSTRACT III CHAPTER 1: INTRODUCTION 1 1.1 Background 1 1.2 Objectives 8 1.3 Organization 8 CHAPTER 2: LITERATURE REVIEW 9 2.1 Introduction 9 2.2 Overview of Existing Research 9 2.2.1 Thermal Analysis Studies 9 2.2.2 Structural Analysis Studies 11 2.2.3 Coupled Thermo-Mechanical Studies 12 2.2.4 Advanced Numerical Modeling 13 2.3 Research Gaps 15 2.4 Summary 15 2.5 Research Questions 16 2.6 Research Hypothesis 16 CHAPTER 3: METHODOLOGY 3.1 17 Methodology for Assessing Structural Integrity and Thermal Behavior Operation Load 17 3.2 Computational Model 19 3.3 Material Properties 21 3.4 Theory and Mathematics for Analysis 22 3.4.1 Structural Analysis 22 3.4.2 Thermal Analysis 3.5 22 Mesh Configuration 23 V 3.5.1 Mesh Configuration of Structural Analysis 23 3.5.2 Mesh Configuration of Thermal Analysis 25 3.6 Initial Boundary Conditions 25 CHAPTER 4: RESULTS AND DISCUSSION 29 4.1 3D Structural Analysis Results and Analysis 29 4.2 Temperature Distribution 35 CHAPTER 5: CONCLUSION 5.1 41 Limitations 42 APENDIX 43 Journal Paper 43 Conference Papers 43 REFERENCES 44 VI LIST OF FIGURES Fig. 1.1: Major components of Pressurized Water Reactor Fuel Assembly. Fig. 3.1: Working flow chart. 2 18 Fig. 3.2: Model of a single fuel rod with spacer grid created in ANSYS Design Modular. 20 Fig. 3.3: Model of a partial fuel assembly. 21 Fig. 3.4: Finite element modeling (a) Single fuel rod and (b) Assembly. 24 Fig. 3.5: Structural boundary conditions (a) fixed support (b,c) internal pressure and (d) outer pressure and for assembly all the conditions and identical. 26 Fig. 3.6: Thermal boundary conditions and for assembly all the conditions and identical (a, b, c, d). 28 Fig. 4.1: Von-Mises stress. 30 Fig. 4.2: Representation of stress for a single fuel rod in previous work. 30 Fig. 4.3: Experimental value of structural analysis for a single fuel rod with considering spacer grid. 32 Fig. 4.4: Experimental value of structural analysis for partial fuel assembly with 19 fuel rod and considering spacer grid (a, b, c and d). 33 Fig. 4.5: Radial temperature distribution in the fuel rod. 36 Fig. 4.6: Radial temperature distribution graph of a fuel rod using ANSYS. 37 Fig. 4.7: Radial temperature distribution graph using COMSOL. 38 Fig. 4.8: Radial temperature distribution in the assembly. 39 Fig. 4.9: Radial temperature distribution of the assembly along the x-axis. 39 VII LIST OF TABLES Table 3.1: Different parameter of the fuel element of VVER-1200 19 Table 3.2: Technical Data Considered for Analysis 21 Table 3.3: Skewness and mesh quality list 23 Table 3.4: Mesh statistics 24 Table 3.5: Mesh statistics 25 Table 3.6: Parameters of Initial Boundary Conditions 26 Table 4.1: Validation of Structural analysis results of Von-Mises Stress simulation with previous research work. 31 Table 4.2: Estimated results of structural analysis 34 Table 4.3: Safety factor 34 Table 4.4: Validation of Temperature distribution along the fuel rod with previous research work 38 VIII CHAPTER 1 INTRODUCTION 1.1 Background A nuclear power plant is a form of thermal power plant that derives its heat from nuclear fission within a reactor core. A steam turbine produces electricity using the steam created from this heat, which is connected to a generator. Nuclear power plants (NPPs) are the most reliable and cost-effective energy generation options in the globe. Nuclear power generation is low-cost, low-carbon, and allows contemporary civilization to satisfy its growing electricity demand. The world needs far more energy in the future, especially renewable energy. Electricity consumption is increasing at double the pace at which global energy consumption is increasing, and is anticipated to more than quadruple by 2040. Nuclear energy accounts for around 10% of worldwide energy production and 18% of electricity generation in OECD countries. Almost all prominent estimates for the future of energy supply indicate that nuclear power will play a greater role as an environmentally beneficial method of supplying reliable electricity on a large scale (Kang et al., 2016). Especially for nuclear power facilities, the VVER-1200 fuel assembly is an essential feature of the design of a Pressurized Water Reactor (PWR). It gives the reactor's core a dependable and durable fuel source. The spacer grids, upper and lower end fittings, and fuel rods are the main parts of fuel assemblies for Pressurized Water Reactors (PWRs). The fuel rods are around 12 feet long and contain ceramic fuel pellets. At the top of the fuel rod, there is a hole to collect any gases that are produced by the fission process. For pressurized water reactor, a square matrix consisting of 17 x 17 rods is used. The spacer grids use springy metal pieces to divide the various rods. This gives the assemblies their stiffness and permits coolant to freely circulate around the fuel rods and up through the assemblies. A typical Pressurized Water Reactor (PWR) assembly with grid is shown in Fig. 1.1 (M. S. and Xueyang, 2003). The Nuclear power plants (NPPs) may be twice as expensive to run than coal-fired power stations, and by the time they are shut down, they have degraded into a kind of radioactive waste. Nuclear fuel rods are a component of nuclear power reactors that are made up of nuclear fuel and cladding (Eskandari et al., 2012). The coolant flows through the cladding 1 and the gap, which is the space between the pellet and the cladding, transferring the heat generated inside the pellet to it. The turbine is powered by the steam generated by the nuclear fuel. Nuclear power plant catastrophes, such as the 1986 Chernobyl nuclear power plant accident in Ukraine and the Fukushima nuclear power plant explosions in Japan, which burst in 2011, caused widespread devastation. Each of these incidents was caused by overheated nuclear fuel rods. An important factor to address is the safety of nuclear fuel rods in a transient state. Nuclear power plant regulation has been strengthened in response to the disastrous accidents. Thermal conductance between the pellet and the cladding is a well-known and distinct characteristic of nuclear fuel rods. The heat generated by the pellets is transported to the cladding through the space between them. Fig. 1.1: Major components of Pressurized Water Reactor Fuel Assembly. 2 Nuclear rods release energy for power generation through heat exchangers and turbines and have a tendency to fail for a variety of reasons. The behavior of the nuclear fuel rod (thermal characteristics) is one of the most significant reasons. As time goes on, the pellet swells and occasionally comes into contact with cladding. Therefore, it becomes necessary to forecast the temperature changes in the nuclear fuel rod so that the predicted result can be studied and analyzed. The design of the rod must be in a way so that heat transfer is enhanced to minimize rod failure. For the purpose of analysis and temperature distribution prediction, fuel rods are taken into consideration (K. M. Pandey and Amrit Sarkar, 2011). The study of the temperature behavior of nuclear fuel rods and components is a critical part of nuclear safety. The reality is that when the temperature of the fuel element exceeds its critical value (extreme value), it melts, causing widespread destruction and loss (Eskandari et al., 2012). Nuclear fuels must withstand extreme conditions during operation, such as corrosive environments, mechanical stress, high temperatures, and intense radiation in the reactor core (M. Yassin et al., 2023). Research on the behavior of the fuel rod exposed to radiation shows significant changes in their geometry, dimensions, chemical composition, and microstructure both during and after exposure to radiation (P.R. Roy and D.N. Sah, 1985). The VVER 1200 is the most sophisticated version of the VVER reactor series that has progressed from its initial VVER 400 versions that includes passive safety features and a higher output (1200 MWe). Recent research has used thermal hydraulic simulations such as CFD analysis, annular fuels and nanofluids to enhance performance and safety (ElMorshedy, 2023). Spacer Grid of Fuel Assembly is a crucial structural component of a nuclear fuel assembly, particularly in Pressurized Water Reactors (PWRs) like the VVER-1200. Its primary function is to maintain precise spacing and alignment of the individual fuel rods within the fuel assembly. These grids are typically made of high-strength, corrosion-resistant alloys such as stainless steel, ensuring durability and resistance to the extreme operating conditions inside the reactor core. Each fuel assembly may contain several spacer grids arranged at intervals along the length of the fuel rods. These grids serve multiple purposes: • Mechanical Support: They prevent fuel rods from vibrating or buckling under the influence of coolant flow and neutron flux, ensuring the rods remain aligned. 3 CHAPTER 2 LITERATURE REVIEW 2.1 Introduction Fuel assemblies play a crucial role in the performance and safety of nuclear power plants, especially in water-cooled reactors like the VVER-1200. Operating under high thermal loads and mechanical stresses, the assemblies need to behave correctly and understanding their mechanical response under such conditions is critical to keep the structural integrity and operational efficiency of the reactor. Over the years many researchers have studied the working of fuel rod, assembly behavior from temperature distribution and material deformation up to vibration analysis during reactor operation and accident response. Due to the development of ANSYS simulation tools, it becomes possible to perform more detailed and precise analyses. These tools let engineers assess and predict how the fuel assemblies will behave under the prescribed operating conditions as well steam supply accidents. 2.2 Overview of Existing Research A wide range of studies has been conducted to analyze the thermal and mechanical behavior of nuclear fuel elements under operating and accident conditions. These studies span analytical models, finite element methods (FEM), and advanced multi-physics simulations using tools like ANSYS, ABAQUS, RELAP5, and BISON. The reviewed literature reveals a growing emphasis on the importance of accurately predicting fuel performance metrics such as temperature distribution, structural deformation, and safety factors—parameters that directly influence the operational reliability and lifespan of fuel assemblies in power reactors. 2.2.1 Thermal Analysis Studies Dall’Osso and Braun (2018) incorporated the Nodal Expansion Method (NEM) into their work to assess radial temperature distribution in nuclear fuel rods with the goal of increasing the speed of calculations relative to the traditional methods that use finite 9 difference or volume techniques. They created a second-order NEM solver in Mathematica for modeling heat conduction in the fuel, gap, and cladding interfaces, which was later transferred to FORTRAN. Their work proved that NEM could achieve analytical precision with greater efficiency in computation and coarser meshes even in the presence of temperature-dependent conductivity. In relation to finite volume techniques, NEM was more precise and less computationally expensive NEM’s superiority over finite volumetric methods provided enhanced accuracy with reduced computation needs. Even if applicable only for steady-state situations, the research underlines NEM’s capabilities regarding thermal modeling for nuclear systems, encouraging further work on dynamic modeling for transient analysis and advanced techniques for lower-order time responses. Dinh Van Thin, Bui Van Loat, and Bui Thi Hong (2017) investigated the thermal behavior of the VVER-1200 reactor's heat channel using computational fluid dynamics (CFD). Determining the effects of varying heat flux levels on coolant channel temperature and velocity was the aim of the investigation. The authors used computational fluid dynamics (CFD) to simulate a system with four fuel rods and twelve spacer grids under typical operating parameters of 16.2 MPa pressure and 571.16 K inlet temperature. Their findings showed that all values were within safety limits and that the outlet temperature and cladding surface temperature increased as the heat flow increased from 0.3 × 10⁶ to 1.486 × 10⁶ W/m². Polynomial formulae that were curve fitted to the temperature data were reported in the study. The study, however, assumed steady-state condition without considering transient behaviors or accident scenarios which are crucial for comprehensive safety assessments. Odii et al. (2018) focused on modeling the temperature profile in the radial direction of a nuclear fuel element with special consideration to the heat transfer from the fuel center line to its cladding and the helium gas gap in between. The objective was to evaluate how the thickness of the gas gap would alter the temperature distribution along the fuel rod. The authors performed the simulation in ANSYS APDL and implemented an analytical model of classical heat conduction for the gas gap. They noted an increase in the temperature at the centerline due to the volume of gas gap present because helium’s low thermal conductivity limits its ability to transfer heat. This shows why in reactor design a very small gas gap is required. The validation of the model was successfully achieved due to the agreement of results from both the analytical side and the numerical side through the 10 CHAPTER 3 METHODOLOGY 3.1 Methodology for Assessing Structural Integrity and Thermal Behavior Operation Load The methodology followed in this research involves a simulation-based analytical approach using ANSYS Workbench to evaluate both the structural safety and radial temperature distribution within the fuel assembly of a VVER-1200 nuclear reactor. The process commenced with selecting ANSYS Workbench as the simulation platform due to its robust capabilities in coupled thermal-structural analysis. Material selection was carried out for each component of the fuel assembly, including uranium dioxide (UO₂) for the fuel pellet, zirconium alloy for the cladding, and water as the coolant. Accurate thermo-mechanical properties were defined for these materials, sourced from validated nuclear engineering databases and reactor-specific documentation. The fuel assembly geometry was modeled in accordance with VVER-1200 design specifications, with particular focus on the axial and radial structure of the fuel rods. Meshing was performed to discretize the geometry into finite elements suitable for thermal and stress analysis. Boundary conditions were applied, including convection boundaries at the outer cladding surface and symmetry conditions where applicable. Thermal loads, such as heat generation due to fission, and structural loads, including internal gas pressure and external coolant pressure, were imposed to mimic reactor operating conditions. Upon completion of the simulation setup, the model was solved to compute the temperature distribution across the fuel pellet radius, extending to the cladding. Structural analysis was also performed to determine the equivalent stress and safety factor under operational loading conditions. Post-processing tools in ANSYS facilitated the extraction and visualization of results, which were critically analyzed to assess the thermal behavior and mechanical stability of the fuel rod. This methodology enabled a comprehensive evaluation of fuel assembly performance under typical VVER-1200 reactor conditions. 17 Start Select Workbench Specify Material Define Material Properties Boundary Condition Application Mesh Conversion Geometry Creation Load Application Result Extraction End Fig. 3.1: Working flow chart. Fig. 3.1 illustrates the step-by-step workflow adopted in ANSYS for the safety factor calculation and radial temperature distribution analysis of the fuel assembly of the VVER1200 nuclear reactor. The process begins with the selection of the ANSYS Workbench environment, which serves as the integrated simulation platform for structural and thermal analysis. Following this, the material selection step ensures that the relevant fuel, cladding, and coolant materials are specified accurately. Subsequently, the material properties such as thermal conductivity, specific heat, Young’s modulus, and Poisson’s ratio are defined according to the operating conditions of the VVER-1200 reactor. With material data established, the geometry of the fuel rod or fuel assembly is modeled to match the physical dimensions and configurations of the reactor components. Once the geometry is finalized, it is converted into a finite element mesh, allowing discretization of the model for numerical simulation. The boundary conditions are then applied, which include thermal boundaries such as convective heat transfer and fixed temperature points, as well as structural constraints. Load application follows, involving the imposition of operating thermal and mechanical loads such as coolant temperature and internal fuel pressure. After all constraints and loads are defined, the simulation is run, and results such as radial 18 barrier to retain fission products and provide mechanical support under high-pressure reactor conditions. The model also features a spacer grid, which is structurally significant in maintaining the proper alignment and spacing of fuel rods within the fuel assembly. This component contributes to coolant flow optimization and prevents fuel rod vibration or displacement under reactor operation. The fuel rod, as a whole, is the composite cylindrical structure formed by these concentric layers and external components, accurately modeled to match the dimensional and material specifications of the VVER-1200 design. The orientation axes (X, Y, Z) shown in the cross-sectional view aid in defining boundary conditions and applying loads in subsequent simulation steps. This model forms the basis for meshing, boundary condition application, and analysis of radial temperature distribution and mechanical stresses in the thermal-structural simulation process using ANSYS. Spacer Grid Cladding Fuel Fuel Rod Fig. 3.2: Model of a single fuel rod with spacer grid created in ANSYS Design Modular. 20 Fig. 3.3: Model of a partial fuel assembly. 3.3 Material Properties To specify the materials used, the structural and thermal properties of Zr-4 and UO₂ must be entered into the engineering data library as user-defined materials. However, they are already defined in the ANSYS data library. Table 3.2 (Mihaela et al., 2013; Pandey, K.M. and Sarkar, 2011) lists the material properties of Zr-4 and UO₂ used in this work. Table 3.2: Technical Data Considered for Analysis Material Property E110 Uranium Dioxide (UO₂) Density (kg/m³) 6.505 10.97 Young's Modulus (GPa) 88 182 Thermal Conductivity (W/m·K) 22.6 29 Poisson's Ratio 0.34 0.295 Bulk Modulus (GPa) 91.1 265 Shear Modulus (GPa) 33 0.95 Heat Generation (w/m3) Not applicable 8x108 Thermal co-efficient of expansion 20x10-6 10.8x10-6 21 CHAPTER 4 RESULTS AND DISCUSSION 4.1 3D Structural Analysis Results and Analysis The objective of this work is to perform structural and thermal analysis of a fuel rod from the VVER-1200 reactor under normal operating conditions, evaluate its safety and integrity, and present the temperature distribution across the fuel, gap, and cladding. First, to validate the results, the analysis of a single fuel rod without a spacer grid is compared with results from a previous study, as shown in Fig. 4.2 (Kwon et al., 2016). Fig. 4.1 Von-Mises Stress presents the distribution of equivalent (Von-Mises) stress across a segment of the fuel rod in the VVER-1200 reactor assembly, as simulated using ANSYS under static structural analysis conditions. The stress values, given in Mega-pascals (MPa), range from a minimum of approximately 2.32 MPa to a maximum of 34.23 MPa, with the highest stress concentrations occurring near the upper edge of the rod. The color gradient indicates that the red zones represent the regions of maximum stress, likely corresponding to areas of geometric discontinuity or mechanical constraint, while the blue and green regions signify lower stress magnitudes distributed across the cylindrical body. The presence of stress concentration bands suggests the influence of structural features or boundary interactions that affect load distribution during reactor operation. The results of the averaged and non averaged values represent the accuracy and reliability of the analysis. Deviation of the two values = (34.233 − 34.211) × 100% 34.233 = 0.067% The deviation between these two values is 0.067%, indicating that the results are reliable with respect to the mesh selection and meshing method. 29 (a) Average (b) Unaverage Fig. 4.1: Von-Mises stress. Fig. 4.2: Representation of stress for a single fuel rod in previous work. 30 CHAPTER 5 CONCLUSION This study successfully conducted a comprehensive structural and thermal analysis of the VVER-1200 nuclear reactor fuel rod and partial fuel assembly using ANSYS 19.2, a finite element method (FEM)-based simulation platform. The primary objectives were to evaluate the radial temperature distribution and assess the mechanical integrity of the reactor’s fuel components under operational loading conditions. The methodology included accurate 3D modeling of the fuel rod and partial assembly, application of realistic boundary conditions, and validation through comparative analysis with existing literature and experimental data. The thermal analysis revealed that the maximum temperature within the fuel rod reaches 2093.1 K at the centerline and decreases radially to 612.46 K at the cladding outer surface. These results align closely with previously published data, with a minimal deviation of only 0.27%, confirming the thermal accuracy of the simulation model. Similarly, the axial temperature profile along the fuel rod's length displayed periodic behavior consistent with the arrangement of fuel pellets and structural features, providing valuable insight into heat distribution in the reactor core. On the structural side, the simulation results demonstrated that both the single fuel rod and the partial fuel assembly remain well within allowable deformation and stress limits. The calculated safety factors are 8.81 for the single fuel rod and 5.8 for the partial assembly, both significantly exceeding the industry-accepted minimum of 2 for nuclear components. These high safety margins confirm that the fuel assembly is structurally robust and capable of withstanding the mechanical loads encountered during normal reactor operation. Furthermore, all observed deformations 1.67 µm for the single rod and 1.40 µm for the assembly, were well below the 4-5 µm threshold reported in literature, reinforcing the mechanical stability of the components. The comparative analysis with COMSOL-based studies and earlier experimental work further validates the reliability of the ANSYS simulations, indicating that the current approach provides accurate predictions of reactor behavior. Notably, this research not only confirms the integrity of the VVER-1200 fuel assembly under steady-state conditions but also contributes to the body of knowledge regarding thermal-mechanical performance of modern nuclear fuel designs. 41 Overall, this thesis provides a robust framework for the thermal and structural assessment of nuclear reactor components. The findings are especially valuable for countries adopting VVER-1200 technology, offering practical insights into reactor core behavior and contributing to safer and more efficient nuclear power generation. The analysis also serves as a foundational reference for future research focused on optimizing fuel designs and enhancing reactor safety. 5.1 Limitations Few constraints were faced throughout this study, even though structural and thermal simulations of a fuel rod and a section of fuel assembly (having 19 fuel rods) of the VVER1200 reactor were accomplished successfully with the assistance of ANSYS Workbench. The following discussion of these constraints is presented: In order to decrease the simulation time and computational intensity of the partial assembly and fuel rods in this investigation, these were simulated with simplified geometries. The real fuel rod geometries have such minute details as spacer grids, end fittings, and rough surfaces that were not modeled. Such variations could result in simulated and real reactor behavior discrepancies. Also, for simplification, material parameters of the simulation (thermal conductivity, Young's modulus, Poisson's ratio, specific heat) were considered constant or linear temperature-dependent. These material properties were not fully modeled herein as they are extremely temperaturedependent in actual cases and may fluctuate because of burn-up effects, irradiation, and fuel-cladding interaction. 42 REFERENCES Bang, I. C., Lee, S., and Kim, H. (2011). Performance of UO2-graphene composite fuel and SIC cladding during LOCA. 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