scholarly journals Modal Analysis and Testing for a Middle Spacer Grid of a Nuclear Fuel Rod

2012 ◽  
Vol 61 (12) ◽  
pp. 1948-1952 ◽  
Author(s):  
Bong-Jo Ryu ◽  
Kyung-Wan Koo
Keyword(s):  
Modal Analysis ◽  
Nuclear Fuel ◽  
Spacer Grid ◽  
Fuel Rod

Prediction of the Nuclear Fuel Rod Abrasion in the Probability Sense

2016 ◽  
Vol 821 ◽  
pp. 317-324
Author(s):  
Vladimír Zeman ◽  
Zdeněk Hlaváč
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Fretting Wear ◽  
Primary Circuit ◽  
Grid Cells ◽  
Operational Parameters ◽  
Spacer Grid ◽  
Mean Values ◽  
Fuel Rod ◽  
Lower Limits

The paper deals with the upper and lower limits estimation of the friction work and fretting wear in the contact of nuclear fuel rods with fuel assembly (FA) spacer grid cells. The friction work is deciding factor for the prediction of the fuel rod cladding abrasion caused by FA vibration. Design and operational parameters of the FA components are understood as random variables defined by mean values and standard deviations. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of the friction work and fretting wear in particular contact surfaces between the fuel rod cladding and some of spacer grid cells. The fuel assembly vibration is excited by pressure pulsations of the cooling liquid generated by main circulation pumps in the coolant loops of the NPP primary circuit. The method is applied for hexagonal type nuclear fuel assembly in the VVER type reactors.

Finite Element Model to Simulate the Spacer Grid Buckling Characteristics

2020 ◽  
Author(s):  
Youngik Yoo ◽  
Joongjin Kim ◽  
Kyongbo Eom ◽  
Hyeongkoo Kim
Keyword(s):  
Finite Element ◽  
Finite Element Model ◽  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Grid Cell ◽  
Element Model ◽  
Spacer Grid ◽  
Static Compression ◽  
Fuel Rod ◽  
Nuclear Fuel Assembly

Abstract The purpose of this study is to develop a finite element model that accurately describes the buckling behavior of a spacer grid. The spacer grid is the most important component of a nuclear fuel assembly and supports the fuel rod with a structurally sufficient buckling strength. Therefore, the development of a reliable spacer grid model is essential to evaluate the mechanical integrity of a nuclear fuel assembly. To achieve this objective, a three-dimensional finite element model was proposed to simulate the buckling characteristics and mechanical behavior of a PWR spacer grid. To simulate the exact mechanical properties of the spacer grid cell, the parameter values required for the model were determined by conducting a fuel rod drag test and spacer grid spring/dimple stiffness test. Finally, a spacer grid static compression test and dynamic impact test were performed according to the gap size of the spacer grid cell, and the model was verified by comparing the test and analysis results. The results obtained using the developed spacer grid finite element model agreed well with the mechanical test results, and it was confirmed that both the buckling characteristics and mechanical behaviors of the model were accurately simulated by the proposed model.

Nonlinear Response Structural Optimization of a Spacer Grid Spring for a Nuclear Fuel Rod Using the Equivalent Loads

2007 ◽  
Vol 31 (12) ◽  
pp. 1165-1172 ◽  
Author(s):  
Do-Won Kim ◽  
Hyun-Ah Lee ◽  
Ki-Nam Song ◽  
Yong-ll Kim ◽  
Gyung-Jin Park
Keyword(s):  
Structural Optimization ◽  
Nuclear Fuel ◽  
Nonlinear Response ◽  
Spacer Grid ◽  
Fuel Rod

3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

2015 ◽  
Vol 53 (2) ◽  
pp. 232-239 ◽  
Author(s):  
Chang-Hak Kang ◽  
Sung-Uk Lee ◽  
Dong-Yol Yang ◽  
Hyo-Chan Kim ◽  
Yong-Sik Yang
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Nuclear Fuel ◽  
Element Analysis ◽  
3D Finite Element ◽  
3D Finite Element Analysis ◽  
Fuel Rod

Incorporation of Taguchi approach with CFD simulations on laser welding of spacer grid fuel rod assembly

2021 ◽  
Vol 269 ◽  
pp. 115182
Author(s):  
G. Satyanarayana ◽  
K.L. Narayana ◽  
B. Nageswara Rao
Keyword(s):  
Laser Welding ◽  
Taguchi Approach ◽  
Cfd Simulations ◽  
Spacer Grid ◽  
Fuel Rod

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.

Probabilistic finite element analysis in heat transfer to a nuclear fuel rod bumper support

2004 ◽  
Vol 218 (12) ◽  
pp. 1499-1505
Author(s):  
Rama Subba Reddy Gorla
Keyword(s):  
Heat Transfer ◽  
Finite Element ◽  
Nuclear Fuel ◽  
Cost Effective ◽  
Distribution Functions ◽  
Cumulative Distribution ◽  
Element Analysis ◽  
Transfer Rates ◽  
Fuel Rod ◽  
Design Variables

Heat transfer from a nuclear fuel rod bumper support was computationally simulated by a finite element method and probabilistically evaluated in view of the several uncertainties in the performance parameters. Cumulative distribution functions and sensitivity factors were computed for overall heat transfer rates due to the thermodynamic random variables. These results can be used to identify quickly the most critical design variables in order to optimize the design and to make it cost effective. The analysis leads to the selection of the appropriate measurements to be used in heat transfer and to the identification of both the most critical measurements and the parameters.

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.

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