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.

scholarly journals Prediction of thermophysical characteristics of fuel rods based on calculations

2021 ◽  
Vol 7 (2) ◽  
pp. 79-86
Author(s):  
Stepan Lys ◽  
◽  
Igor Galyanchuk ◽  
Tetiana Kovalenko
Keyword(s):  
Reactor Core ◽  
Operating Conditions ◽  
Design Criteria ◽  
Gas Release ◽  
Fuel Cladding ◽  
Thermophysical Characteristics ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Main Factors

The paper analyzes operating conditions, thermophysical characteristics were calculated as applied to WWER-1000 fuel rods in a four-year cycle for unified core. The paper gives a summary of models for calculating gas release, pressure of gases within fuel rod cladding, fuel swelling and thermal conductivity, fuel-cladding gap conductance. The thermophysical condition of fuels in a reactor core is one of the main factors that determine their serviceability. The stress-strained condition of fuel claddings under design operating conditions is closely related to fuel rod temperature, swelling, gas release from fuel pellets and the mode in which they change during the cycle and transients. Aside from this, those parameters are an independent goal of studies since their ultimate values are governed by the system of design criteria.

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.

The Research on Core Melting Process: Oxidation

2013 ◽  
Author(s):  
Jun Wang ◽  
Wenxi Tian ◽  
Jianan Lu ◽  
Yingying Ma ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Reactor Core ◽  
Melting Process ◽  
Coolant Pump ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Design Basis ◽  
Loss Of Coolant ◽  
Design Basis Accident ◽  
Station Blackout

Beyond-design basis accidents in the AP1000 may result in reactor core melting and are therefore termed core melt accidents. The aim of this work is to develop a code to calculate and analyze the oxidation of a single fuel rod with total failures of engineered safeguard systems under a certain beyond-design basis accident such as a gigantic earthquake which can result in station blackout and then total loss of coolant flow. Using the code, the responses of the most dangerous fuel rod in the AP1000 were calculated under the accident. A discussion involving fuel pellets melting, cladding rupture and oxidation, and hydrogen production then was carried out, focused on DNBR during coolant pump coastdown, the cladding intactness under different flow rates in natural circulation, and the delay effect on cladding rupture due to cladding oxidation. By the analysis of calculated results, several suggestions on guaranteeing the security of fuel rods were provided.

Preliminary Investigation of Physical Characteristics in a New Power Ramp Test Irradiation Rig

2016 ◽  
Author(s):  
Liang Zhang ◽  
Liqing Qiu ◽  
Mingyan Tong
Keyword(s):  
Preliminary Investigation ◽  
Reactor Core ◽  
Neutron Field ◽  
Gas Pressure ◽  
Hydraulic Calculation ◽  
Fuel Pellet ◽  
Fuel Rod ◽  
Ramp Test ◽  
Natural Logarithm ◽  
Test Fuel

Power Ramp test (PRT) of a fuel element is generally conducted with a PRT irradiation rig within a research reactor, in order to study the fuel’s behavior and validate its safety under power transient. Neutronics characteristics of a new PRT irradiation rig within a typical HFETR (High Flux Engineering Test Reactor) core and its components’ heat generation rates are calculated with MCNP code in this paper. The range of the test fuel rod power is obtained with a coupled Neutronic-Thermal-Hydraulic calculation method which combines MCNP and CFX code. The results show that changing the density of 3He gas can vary the test fuel rod power effectively, and the 3He gas layer influences the neutron field intensely by reducing the thermal neutron current into the layer and decreasing the neutron flux in and near the irradiation rig. The test fuel rod power varies from 5.80kW to 15.3kW while decreasing the 3He gas pressure from 4.5MPa to 0.13MPa, along with 0.231$ reactivity addition. Power of the fuel pellet in the test rod increases monotonically along with the 3He gas pressure reducing, and its calculation results have good agreement with the curve fitting by a natural logarithm function.

Axial Fission Gas Transport in Nuclear Fuel Rods

2009 ◽  
Vol 283-286 ◽  
pp. 262-267
Author(s):  
M.T. del Barrio ◽  
Luisen E. Herranz
Keyword(s):  
Nuclear Fuel ◽  
Gas Transport ◽  
Fuel Pellet ◽  
Gas Release ◽  
Gas Mixing ◽  
Fuel Performance ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fission Gas ◽  
Fission Gas Release

Fission of fissile uranium or plutonium nucleus in nuclear fuel results in fission products. A small fraction of them are volatile and can migrate under the effect of concentration gradients to the grain boundaries of the fuel pellet. Eventually, some fission gases are released to the rod void volumes by a thermally activated process. Local transients of power generation could distort even further the already non-uniform axial power and fission gas concentration profiles in fuel rods. Most of the current fuel rod performance codes neglects these gradients and the resulting axial fission gas transport (i.e., gas mixing is considered instantaneous). Experimental evidences, however, highlight axial gas mixing as a real time-dependent process. The thermal feedback between fission gas release, gap composition and fuel temperature, make the “prompt mixing assumption” in fuel performance codes a key point to investigate due to its potential safety implications. This paper discusses the possible scenarios where axial transport can become significant. Once the scenarios are well characterized, the available database is explored and the reported models are reviewed to highlight their major advantages and shortcomings. The convection-diffusion approach is adopted to simulate the axial transport by decoupling both motion mechanisms (i.e., convection transport assumed to be instantaneous) and a stand-alone code has been developed. By using this code together with FRAPCON-3, a prospective calculation of the potential impact of axial mixing is conducted. The results show that under specific but feasible conditions, the assumption of “prompt axial mixing” could result in temperature underestimates for long periods of time. Given the coupling between fuel rod thermal state and fission gas release to the gap, fuel performance codes predictions could deviate non-conservatively. This work is framed within the CSN-CIEMAT agreement on “Thermo-Mechanical Behaviour of the Nuclear Fuel at High Burnup”.

RBMK Fuel Rod Modelling and Investigation of Process During Operation Cycle

2010 ◽  
Author(s):  
Tadas Kaliatka ◽  
Ausˇra Marao ◽  
Renatas Karalevicˇius ◽  
Eugenijus Usˇpuras
Keyword(s):  
Normal Operation ◽  
Spent Fuel ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
The Status ◽  
Fuel Storage ◽  
Spent Fuel Pool ◽  
Operation Cycle ◽  
Rbmk Fuel

This paper presents the analysis determining the status of fuel rods after whole normal operation. The FEMAXI–6 code was selected for such analysis. Evaluating the specifics of RBMK fuel rods, the adaptation of code was provided. After the adaptation of FEMAXI-6 code, the single fuel rod model of RBMK-1500 was developed and the processes, which occur during whole life of fuel rods, were analyzed. For this analysis the fuel rod from fuel channel with average initial power (2.5 MW) was selected. After (normal) operation the fuel rods from the reactor are transferred to the spent fuel pool and the state of the fuel rods (intactness of cladding, residual stresses in the cladding and fuel pellets, gap between cladding and pellets and etc.) is very important, because fuel rod cladding is one of the safety barriers. In this paper the stresses in cladding, plastic deformation of cladding and other parameters were calculated using FEMAXI-6 and method of final elements. The performed analysis demonstrates possibility to identify state of fuel rods after normal operation that is necessary for long-term fuel storage in spent fuel pools.

Analysis of the Control Rod Drop Accident (CRDA) for Lungmen ABWR

2014 ◽  
Author(s):  
Ai-Ling Ho ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin ◽  
Chunkuan Shih
Keyword(s):  
High Temperature ◽  
Fuel Pellet ◽  
Output Data ◽  
Two Phase ◽  
Temperature Oxidation ◽  
Control Rod ◽  
Melting Temperatures ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Full Power

The object of this paper is to understand the realistic behavior in Lungmen ABWR (Advanced Boiling Water Reactor) during a control rod drop accident (CRDA) transient. The CRDA transient would lead the reactor through an extremely fast and localized power excursion, requiring an accurate core modeling. The CRDA analysis for Lungmen ABWR was performed by coupling the 3D neutron kinetic code, PARCS, and two-phase thermal-hydraulic (T-H) code, TRACE. After TRACE/PARCS coupling calculation, the output data from TRACE/PARCS would be inputted into FRAPTRAN code as a function of time-dependent fuel rod power and coolant boundary conditions to calculate the fuel damage. The CRDA analysis for Lungmen ABWR was performed for two conditions: a) case1: hot-full-power (HFP) at beginning of cycle (BOC); b) case2: hot-zero-power (HZP) at BOC. Under these conditions, the damage mechanisms of fuel rod are: 1) cladding ballooning and burst; 2) embrittlement and failure by high-temperature oxidation; 3) melting of cladding and/or fuel pellets. And the relevant quantities for fuel performance are the maximum fuel enthalpy and the melting temperatures of cladding and fuel pellet. The results of CRDA analysis show that a) case1: no fuel failure occurs under HFP condition at BOC; b) case2: the fuel rod nearby the dropped control rod failed under HZP condition at BOC, and the FRAPPTRAN data exposes that the main reason of rod failure is the cladding high temperature.

scholarly journals Predicting behavior of AP-1000 nuclear reactor fuel rod under steady state operating condition by using FRAPCON-4.0 software

2021 ◽  
Vol 8 (2) ◽  
pp. 43-50
Author(s):  
Van Tung Nguyen ◽  
Trong Hung Nguyen ◽  
Thanh Thuy Nguyen ◽  
Duy Minh Cao
Keyword(s):  
Steady State ◽  
Nuclear Reactor ◽  
Fuel Pellet ◽  
Coolant Temperature ◽  
Oxide Surface ◽  
Gas Temperature ◽  
Operating Condition ◽  
Fuel Rod ◽  
Fuel Pellets ◽  
Nuclear Reactor Fuel

This paper reports the results on the predictions of behavior of AP-1000 nuclear reactorfuel rod under steady state operating condition by using FRAPCON-4.0 software. The predictive items were the temperature distribution in the fuel rod, including fuel centerline temperature, fuel pellet surface temperature, gas temperature, cladding inside and outside temperature, oxide surface and bulk coolant temperature; and gap conductance and thickness.The predictive items also include deformation of fuel pellets, fission gas release and rod internal pressure, cladding oxidation and hydration. The predictive data were suggested the fuel rod behavior image in nuclear reactor.

Improvement of SAMPSON Code for BWR Severe Accident Analysis

2013 ◽  
Author(s):  
Maolong Liu ◽  
Yuki Ishiwatari ◽  
Koji Okamoto
Keyword(s):  
Stainless Steel ◽  
Oxidation Reaction ◽  
Reactor Core ◽  
Severe Accident ◽  
Accident Analysis ◽  
Light Water Reactors ◽  
Fuel Cladding ◽  
Fuel Rod ◽  
Water Reactors ◽  
The Impact

The SAMPSON code has been developed in the IMPACT project in Japan to investigate severe accident phenomena for light water reactors. It integrates various analysis modules into a single code. The authors improved the fuel rod heat-up module of SAMPSON code by modeling the oxidation reaction of various core structures, including Zircaloy, stainless steel and B4C. And the creep failures of the Zircaloy fuel cladding and stainless steel monitoring guide tubes of the source range monitor (SRM) in the reactor core was also modeled for severe accident analysis.

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.

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