scholarly journals Simulation of Mechanical Behavior of Fuel Pellets With Different Geometries

2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Amanda Abati Aguiar ◽  
Danilo Faria ◽  
José Berretta ◽  
Paulo Afonso Rodi ◽  
Marcelo Santos ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Exothermic Reaction ◽  
Mechanical Analysis ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Pellets ◽  
Iron Based ◽  
Water Reactors ◽  
Accident Conditions

Typical Pressurized Water Reactors (PWR) fuel rods are manufactured using zirconium-based alloys as cladding and slightly enriched UO2 sintered pellets as fuel. However, in the last years efforts have been made to develop Accident Tolerant Fuels (ATF) focusing mainly in new materials to replace the cladding in order to avoid the exothermic reaction with steam experienced by zirconium-based alloys under accident conditions as observed during the Fukushima Daiichi accident. In this sense, iron-based alloys appear as a possibility to replace conventional zirconium-based alloys, and the effect of the pellet geometry in the performance of iron-based alloys fuel rods shall be investigated. The fuel pellet geometry experiences changes due to irradiation can promote early gap closure, mechanical loadings to the cladding and/or bamboo effects due to the combination of loads and irradiation creep, and all these effects depend also on the cladding properties. The objective of this paper was to address the influence of geometric parameters in the fuel pellet behavior of a stainless steel fuel rod by means of structural mechanical analysis using the well-known ANSYS software. The parameters evaluated in this paper considered fuel pellet with and without chamfer and dish. The data related to the fuel pellet performance under irradiation were obtained using a modified version of the FRAPCON code considering stainless steel as cladding. Results obtained from mechanical evaluation considering the effects through the responses of the axial, radial, plastic deformations, and resulting tensions were evaluated.

scholarly journals Modification of fuel performance code to evaluate iron-based alloy behavior under LOCA scenario

2021 ◽  
Vol 9 (2A) ◽  
Author(s):  
Claudia Giovedi ◽  
Alfredo Abe ◽  
Rafael O. R. Muniz ◽  
Daniel S. Gomes ◽  
Antonio T. Silva ◽  
...  
Keyword(s):  
New Materials ◽  
Fuel Performance ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Fukushima Daiichi ◽  
Iron Based ◽  
Water Reactors ◽  
Accident Scenarios

Accident tolerant fuels (ATF) has been studied since the Fukushima Daiichi accident in the research efforts to develop new materials which under accident scenarios could maintain the fuel rod integrity for a longer period compared to the cladding and fuel system usually utilized in Pressurized Water Reactors (PWR). The efforts have been focused on new materials applied as cladding, then iron-base alloys appear as a possible candidate. The aim of this paper is to implement modifications in a fuel performance code to evaluate the behavior of iron-based alloys under Loss-of-Coolant Accident (LOCA) scenario. For this, initially the properties related to the thermal and mechanical behavior of iron-based alloys were obtained from the literature, appropriately adapted and introduced in the fuel performance code subroutines. The adopted approach was step by step modifications, where different versions of the code were created. The assessment of the implemented modification was carried out simulating an experiment available in the open literature (IFA-650.5) related to zirconium-based alloy fuel rods submitted to LOCA conditions. The obtained results for the iron-based alloy were compared to those obtained using the regular version of the fuel performance code for zircaloy-4. The obtained results have shown that the most important properties to be changed are those from the subroutines related to the mechanical properties of the cladding. The results obtained have shown that the burst is observed at a longer time for fuel rods with iron-based alloy, indicating the potentiality of this material to be used as cladding with ATF purposes.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

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.

scholarly journals An Experimental Programme optimized with Uncertainty Propagation: PETALE in the CROCUS Reactor

2019 ◽  
Vol 211 ◽  
pp. 03003 ◽  
Author(s):  
Vincent Lamirand ◽  
Axel Laureau ◽  
Dimitri Rochman ◽  
Gregory Perret ◽  
Adrien Gruel ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Energy Range ◽  
Present Article ◽  
Uncertainty Propagation ◽  
Nuclear Data ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Experimental Programme ◽  
Iron Nickel ◽  
Water Reactors

The PETALE experimental programme in the CROCUS reactor at EPFL intends to contribute to the validation and improvement of neutron nuclear data in the MeV energy range for stainless steel, particularly in the prospect of heavy reflector elements of pressurized water reactors. It mainly consists of several transmission experiments: first, through metallic sheets of nuclear-grade stainless steel interleaved with dosimeter foils, and, successively, through its elemental components of interest – iron, nickel, and chromium. The present article describes the study for optimizing the response of the dosimetry experiments to the nuclear data of interest.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

Spherical steel containments of pressurized water reactors under accident conditions

1984 ◽  
Vol 82 (1) ◽  
pp. 77-87 ◽  
Author(s):  
R. Krieg ◽  
F. Eberle ◽  
B. Göller ◽  
W. Gulden ◽  
J. Kadlec ◽  
...  
Keyword(s):  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Water Reactors ◽  
Accident Conditions

scholarly journals CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION

2015 ◽  
Vol 55 (6) ◽  
pp. 384 ◽  
Author(s):  
Dávid Halabuk ◽  
Jiří Martinec
Keyword(s):  
Reactor Core ◽  
Fuel Pellet ◽  
Deformation Analysis ◽  
Main Function ◽  
Fuel Cladding ◽  
Fuel Rods ◽  
Radial Temperature ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Stress And Deformation

The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.

Cost Saving When Using Enhanced Conductivity Nuclear Fuel Containing BeO in WWER-1000 Reactors

2014 ◽  
Author(s):  
Tomáš Zahrádka ◽  
Radek Škoda
Keyword(s):  
Thermal Conductivity ◽  
Nuclear Fuel ◽  
Fuel Pellet ◽  
Internal Stresses ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Adverse Conditions ◽  
Advantages And Disadvantages ◽  
Fuel Pin ◽  
Water Reactors

Current pressurized water reactors utilize sintered UO2 that has a number of advantages and disadvantages. Uranium Dioxide’s low thermal conductivity results in a large thermal gradient within the fuel pellet corresponding to higher centerline temperatures compared to other potential fuel forms. These gradients result in non-uniform thermal expansion leading to large internal stresses resulting in cracking of the pellet and fuel-clad interaction, which can lead to loss of the integrity of the fuel pin. Higher fuel temperatures also increase the release of fission gases. Fuels with higher thermal conductivity may alleviate or reduce the severity of these adverse conditions. It is shown that higher thermal conductivity can be obtained by adding BeO to the basic UO2 matrix. This paper focuses on WWER1000 hexagonal fuel geometry. Improvements when using 10% of BeO, as proposed in this paper, reduce the centerline nuclear fuel temperature by 234°C and improve the fuel economy while reducing its cost by 7%. The study was done for NPP Temelín which has two units WWER1000/320.

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