scholarly journals Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

2011 ◽  
Vol 2011 ◽  
pp. 1-11 ◽  
Author(s):  
Armando C. Marino
Keyword(s):  
Nuclear Fuel ◽  
Extreme Case ◽  
Computer Code ◽  
National Commission ◽  
Fuel Rod ◽  
Safety Margins ◽  
Starting Point ◽  
Fuel Design ◽  
Advanced Fuel ◽  
Fuel Elements

The BaCo code (“Barra Combustible”) was developed at the Atomic Energy National Commission of Argentina (CNEA) for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity), probabilistic (or statistic) analysis plus the analysis of the fuel performance (full core analysis) are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

Identification of Non-Linear Vibration Parameters of a Nuclear Fuel Rod

2018 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Stanislas Le Guisquet ◽  
Kostas Karazis ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Harmonic Balance Method ◽  
Resonance Phenomenon ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Non Linear

In Pressurized Water Reactors (PWR), fuel assemblies are composed of fuel rods, long slender tubes filled with uranium pellets, bundled together using spacer grids. These structures are subjected to fluid-structure interactions, due to the flowing coolant surrounding the fuel assemblies inside the core, coupled with large-amplitude vibrations in case of external seismic excitation. Therefore, understanding the non-linear response of the structure and, particularly, its dissipation, is of paramount importance for the choice of safety margins. To model the nonlinear dynamic response of fuel rods, the identification of nonlinear stiffness and damping parameters is required. The case of a single fuel rod with clamped-clamped boundary conditions was investigated by applying harmonic excitation at various force levels. Different configurations were implemented testing the fuel rod in air and in still water; the effect of metal pellets simulating nuclear fuel pellets inside the rods was also recorded. Non-linear parameters were extracted from some of the experimental response curves by means of a numerical tool based on the harmonic balance method. The axisymmetric geometry of fuel rods resulted in the presence of a one-to-one internal resonance phenomenon, which has to be taken into account modifying accordingly the numerical identification tool. The internal motion of fuel pellets is a cause of friction and impacts, complicating further the linear and non-linear dynamic behavior of the system. An increase of the equivalent viscous-based modal damping with excitation amplitude is often shown during geometrically non-linear vibrations, thus confirming previous experimental findings in the literature.

Flow Induced Vibration and Fretting Wear: An Integrated Approach

2002 ◽  
Author(s):  
D. V. Paramonov ◽  
S. J. King ◽  
M. Y. Young ◽  
R. Y. Lu
Keyword(s):  
Flow Field ◽  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Wear Behavior ◽  
Vibration Response ◽  
Fretting Wear ◽  
Analysis Model ◽  
Fuel Rod ◽  
Safety Margins ◽  
Test Loop

Fuel assemblies are exposed to severe thermal, mechanical and radiation loads during operation. Global core and local fuel assembly flow fields typically result in fuel rod vibration. Under certain conditions, this vibration, when coupled with other factors, might result in excessive cladding fretting wear. This phenomenon is of the concern for nuclear fuel designers, especially in light of the need for higher burnup, longer cycle lengths, and operational safety margins in fuel designs. Understanding of (1) the fretting wear margins for a particular nuclear fuel design, (2) the probability of a fuel assembly exposed to a particular set of thermal, mechanical, flow and radiation conditions being at risk of excessive wear, and (3) the factors affecting fretting wear resistance, are important in order to better guide design, testing, and operational flexibility. In this paper, an integrated method to estimate fretting margin of nuclear fuel is presented, including its formulation, benchmark against experimental data and example application to in-core conditions. The major features of the method are as follows: • flow and rod vibration response are coupled through a linear structural analysis model, • flow field is determined using a sub-channel thermal-hydraulic code, • wear progression is treated as a time-dependent process, through taking into account impact of resulting rod-to-support clearance, • a possibility of a fluid-elastic instability is accounted for. Supporting data on basic wear mechanisms, flow field and fuel assembly fretting wear behavior obtained at a number of experimental facilities at Westinghouse Electric Company and Atomic Energy of Canada Limited are also presented. These facility include: • VIPER hydraulic test loop data where vibration response and wear are measured under prototypical flow conditions, and • autoclave fretting-wear machine steam employed to determine fretting-wear coefficients of fuel rod and grid-support designs.

Experiments on Nonlinear Vibrations of a Nuclear Fuel Rod Supported by Spacer Grids in Air and Submerged in Water

2019 ◽  
Author(s):  
Marco Amabili ◽  
Prabakaran Balasubramanian ◽  
Giovanni Ferrari ◽  
Giulio M. Franchini ◽  
Francesco Giovanniello ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Vibration Response ◽  
Nonlinear Phenomena ◽  
Internal Resonances ◽  
Pressurized Water ◽  
Spacer Grids ◽  
Fuel Rod ◽  
Safety Margins ◽  
Fuel Assemblies ◽  
Stiffness And Damping

Abstract For safety reasons, the nuclear fuel assemblies of Pressurized Water Reactors (PWR) must be able to withstand external excitations ranging from large amplitude seismic motions of the reactor to flow-induced vibrations from the surrounding coolant water. A nuclear fuel assembly is composed of long slender tubes, most of them filled with uranium pellets, maintained in a square array by spacer grids. The spacer grids provide a nonlinear flexible boundary condition with friction and micro-impacts that complicates the nonlinear dynamics. In order to improve safety margins in the design of nuclear fuel assemblies, it is of great interest to understand the influence of the spacer grids, as it relates to the overall structural stiffness and damping properties. In particular, the evolution of the vibration amplitude with increasing excitation forces is still undetermined. In order to understand the nonlinear vibration response of a zirconium fuel rod filled with nuclear fuel pellets and supported by spacer grids, experiments were carried out in water and in air. They consisted of measuring the vibration response of the rod under a step-sine harmonic excitation at different force amplitude levels in the frequency neighborhood of the fundamental mode. If the excitation is large enough, the response of the rod displays nonlinear phenomena such as the shift of the resonant frequencies, multiple solutions with instabilities (jumps) and hysteresis, and one-to-one internal resonances. These experiments were carried out on zirconium tubes filled with axially unconstrained as well as axially blocked metallic pellets, which simulate the nuclear fuel. The zirconium tubes were tested both in air and immersed in water. The experimental data will be processed in the future by means of an identification procedure to extract the nonlinear stiffness and damping parameters of the system. An increase of the equivalent viscous damping with the excitation amplitude level is expected.

scholarly journals IMPROVEMENT OF FUEL ELEMENT SHELL CONTROL METHODS TO INCREASE NUCLEAR REACTOR SAFETY

2020 ◽  
pp. 26-31
Author(s):  
P. F. Budanov ◽  
K. Yu. Brovko ◽  
Е. А. Khomiak ◽  
О. А. Tymoshenko
Keyword(s):  
Fuel Element ◽  
Nuclear Fuel ◽  
Geometric Parameters ◽  
Material Structure ◽  
Shell Material ◽  
Fuel Element Cladding ◽  
Fuel Rod ◽  
Inner Surface ◽  
Fuel Elements ◽  
Cladding Material

The analysis of the existing methods of control of the surface of the fuel element cladding material was carried out, which showed that their use for detecting surface and internal defects, such as local inhomogeneities, micro- and macropores, various cracks, axial looseness, etc. is characterized by low efficiency, is a laborious process that requires additional surface treatment, material of the fuel elements cladding. In addition, the investigated methods of controlling the surface of the fuel element cladding material make it possible to visually identify only rough external cracks and large slag inclusions, small cracks and non-metallic inclusions invisible under the slag layer. It is proposed to assess the quality of the surface of the shell material in case of its damage and destruction, the use of a computational apparatus based on the method of the theory of fractals. It is proposed to use the fractal properties of the shell material structure and a quantitative fractal value – the fractal dimension, which makes it possible to determine the degree of filling of the volume of the shell material structure during fuel element depressurization. A mathematical model of damage to the structure of the fuel element cladding material is developed depending on the simultaneous effect of high temperature and internal pressure caused by the accumulation of nuclear fuel fission products between the nuclear fuel pellet and the inner surface of the fuel element cladding, taking into account the fractal increases in the geometric parameters of the fuel element cladding. It is shown that damaged structures of the fuel rod cladding material depend on the pressure and temperature inside the fuel rod cladding, as well as the fractal increase in geometric parameters, such as: volume and surface area, outer and inner diameters, height and cross-sectional area, cladding length and height of nuclear pellets, gap between the inner surface of the cladding and nuclear fuel. A criterion for assessing the integrity of the fuel rod cladding is determined, which depends on the change in geometric values in the event of damage and destruction of the structure of the fuel rod cladding material. Practical recommendations are given on the use of the proposed method for monitoring the tightness of the fuel element cladding for processing information obtained from the computational module of the system for monitoring the tightness of the cladding for the automated process control system of the nuclear power plant power unit, which makes it possible to detect the depressurization of fuel elements at an earlier stage in comparison with the standard procedure.

Experimental Research and Development of Safety Analysis Systems for Advanced Types of Fuel Rods

2018 ◽  
Author(s):  
Alexander I. Maximkin ◽  
Ivan S. Kryukov ◽  
Alexander N. Ableev ◽  
Alexander V. Berestov ◽  
Ilya I. Rodko
Keyword(s):  
Experimental Data ◽  
Nuclear Energy ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Gas Leak ◽  
Fuel Assemblies ◽  
Advanced Fuel ◽  
Monitoring And Diagnostics ◽  
Fuel Elements ◽  
New System

According to the development of the concept of “zero failure” or “zero fuel element defect”, accepted in 2011, which consists in reducing the number of fuel elements that are depressurized in the process of operation to the reached level in the leading countries in nuclear energy (10−6–10−5 defective fuel rods) and avoidance of fuel assemblies with non-hermetic cladding of fuel rods for further operation, including defects with a “gas leak” type, new promising fuels are being developed and introduced, including methods for justifying their safety. Thus, to ensure reliability and safety of new fuel types, it is necessary to provide procedures for monitoring current performance characteristics at all stages of the life cycle of fuel rods. In this paper, experience is given on the development and implementation of instrumentation and methods for monitoring of fuel rods with advanced types of nuclear fuel for VVER reactors that ensure the reliability, safety and competitiveness of technologies associated with the use of advanced fuel rod types, and the implementation of associated components, systems and equipment for monitoring and diagnostics. The features of the applied techniques are presented, and the new system of requirements for the implemented equipment created on their basis. This research continues, and the analysis of intermediate experimental data is carried out in this article.

Enhancement of Heat Transfer Performance in Nuclear Fuel Rod Using Nanofluids and Surface Roughness Technique

2015 ◽  
Author(s):  
Kang Liu ◽  
Titan C. Paul ◽  
Leo A. Carrilho ◽  
Jamil A. Khan
Keyword(s):  
Heat Transfer ◽  
Surface Roughness ◽  
Heat Transfer Coefficient ◽  
Transfer Coefficient ◽  
Nuclear Fuel ◽  
Three Dimensional ◽  
Pressurized Water ◽  
Bulk Fluid ◽  
Fuel Rod ◽  
Dimensional Surface

The experimental investigations were carried out of a pressurized water nuclear reactor (PWR) with enhanced surface using different concentration (0.5 and 2.0 vol%) of ZnO/DI-water based nanofluids as a coolant. The experimental setup consisted of a flow loop with a nuclear fuel rod section that was heated by electrical current. The fuel rod surfaces were termed as two-dimensional surface roughness (square transverse ribbed surface) and three-dimensional surface roughness (diamond shaped blocks). The variation in temperature of nuclear fuel rod was measured along the length of a specified section. Heat transfer coefficient was calculated by measuring heat flux and temperature differences between surface and bulk fluid. The experimental results of nanofluids were compared with the coolant as a DI-water data. The maximum heat transfer coefficient enhancement was achieved 33% at Re = 1.15 × 105 for fuel rod with three-dimensional surface roughness using 2.0 vol% nanofluids compared to DI-water.

Rod Bow Ultrasonic Measurements Possibilities

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marcin Kopeć ◽  
Martina Malá
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactors ◽  
Research Centre ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly ◽  
Fuel Assemblies ◽  
Ultrasonic Measurements ◽  
History Of ◽  
Fuel Particles

The ultrasonic (UT) measurements have a long history of utilization in the industry, also in the nuclear field. As the UT transducers are developing with the technology in their accuracy and radiation resistance, they could serve as a reliable tool for measurements of small but sensitive changes for the nuclear fuel assembly (FA) internals as the fuel rods are. The fuel rod bow is a phenomenon that may bring advanced problems as neglected or overseen. The quantification of this issue state and its probable progress may help to prevent the safety-related problems of nuclear reactors to occur—the excessive rod bow could, in the worst scenario, result in cladding disruption and then the release of actinides or even fuel particles to the coolant medium. Research Centre Rez has developed a tool, which could serve as a complementary system for standard postirradiation inspection programs for nuclear fuel assemblies. The system works in a contactless mode and reveals a 0.1 mm precision of measurements in both parallel (toward the probe) and perpendicular (sideways against the probe) directions.

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