Multiphysics Modeling of Pressurized Water Reactor Fuel Performance

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chungyu ◽  
Yuan Cenxi
Keyword(s):  
Three Dimensional ◽  
Fission Products ◽  
Axial Direction ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Nuclear Fuel Rods ◽  
Reactivity Insertion

Nuclear fuel rods operate under complex radioactive, thermal, and mechanical conditions. Nowadays, fuel rod codes usually make great simplifications on analyzing the multiphysics behavior of fuel rods. The present study develops a three-dimensional (3D) module within the framework of a general-purpose finite element solver, i.e., abaqus, for modeling the major physics of the fuel rods. A typical fuel rod, subjected to stable operations and transient conditions, is modeled. The results show that the burnup levels have an important influence on the thermomechanical behavior of fuel rods. The swelling of fission products causes a dramatically increasing strain of pellets. The variation of the stress and the radial displacement of the cladding along the axial direction can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress in the outer regime of the pellet and may cause further fragmentation to the pellets. Fission products migration effects and differential thermal expansion become more severe if there are flaws or imperfections on the pellet.

Three Dimensional Modeling of the Thermo-Mechanical Performance of the Fuel Rods of a PWR

2017 ◽  
Author(s):  
Wang Zhu ◽  
Zhang Chunyu ◽  
Li Aolin ◽  
Yuan Cenxi
Keyword(s):  
Mechanical Performance ◽  
Three Dimensional ◽  
Fission Products ◽  
General Purpose ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod ◽  
Three Dimensional Modeling ◽  
Water Reactors

The fuel rods of pressurized water reactors operate under complex radioactive, thermal and mechanical conditions. Multiphysics has to be taken into account in order to evaluate their performance. Many existing fuel rod codes make great simplifications on analyzing the behavior of fuel rods. The present study develops a three dimensional module within the framework of a general-purpose finite element solver, i.e. ABAQUS, for modeling the thermo-mechanical performance of the fuel rods. A typical fuel rod is modeled and the temperature as well as the stress within the pellets are computed. The results show that the burnup levels have an important influence on the fuel temperature. The swelling of fission products cause dramatically increasing of pellet strain. The change of the cladding stress and radial displacement with the axial length can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress to the outer regime of the pellet and may cause further fragmentation to the pellets.

Modeling of Steam Explosion in Partially Flooded PWR Reactor Cavity

2005 ◽  
Author(s):  
Bosˇtjan Koncˇar ◽  
Matjazˇ Leskovar ◽  
Leon Cizelj
Keyword(s):  
Steam Explosion ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Fission Products ◽  
General Purpose ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fit For Purpose ◽  
Computational Fluid Dynamics Cfd ◽  
Reactor Accidents

When the hot molten core comes into contact with the water in the reactor cavity a steam explosion can occur. The steam explosion might be triggered during some scenarios of severe nuclear reactor accidents, when extremely hot molten nuclear fuel interacts with the coolant water. A highly energetic steam explosion in a nuclear power plant could cause the containment failure and the release of radioactive fission products to the environment. The purpose of the performed analysis is to provide a first estimation of the expected pressure loadings on the typical PWR cavity structures during a steam explosion. To achieve this, the fit-for-purpose steam explosion model is proposed, followed by a Computational Fluid Dynamics (CFD) analysis. In the present work two steam explosion scenarios in the partially flooded Pressurized Water Reactor (PWR) cavity were simulated with the general purpose code CFX-5 [1] to estimate pressure loadings on cavity walls.

Study of the release of the fission gases (Xe and Kr) and the fission products (Cs and I) under anoxic conditions in bicarbonate water

2015 ◽  
Vol 1744 ◽  
pp. 35-41 ◽  
Author(s):  
Ernesto González-Robles ◽  
Markus Fuß ◽  
Elke Bohnert ◽  
Nikolaus Müller ◽  
Michel Herm ◽  
...  
Keyword(s):  
Gas Phase ◽  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Collaborative Project ◽  
Fuel Rod ◽  
Release Ratio

ABSTRACTFor safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of fission products to the instant release fraction (IRF). During the last three years the EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF.Within CP FIRST-Nuclides, a leaching experiment with a cladded SNF pellet was performed in bicarbonate water (19 mM NaCl + 1 mM NaHCO3) under Ar /H2 atmosphere over 333 days. The cladded SNF pellet was obtained from a fuel rod segment which was irradiated in the Gösgen pressurized water reactor; the average burn-up of the segment was 50.4 MWd/kgUO2. In the multi-sampling experiment, gaseous and liquid samples were taken periodically. The moles of the fission gases Kr and Xe released in the gas phase and those of 129I and 137Cs released in solution were measured. Cumulative release fractions of (1.6 ± 0.2)·10-1 fission gases, (1.6 ± 0.1)·10-1129I and (3.9 ± 0.2)·10-2 137Cs, respectively, were achieved after 333 days of leaching. Accordingly the release ratio of fission gases to 129I was 1:1 and the release ratio of fission gases to 137Cs was 4:1, respectively.

scholarly journals Effects of Loading Conditions and Types of Motion on PWR Fuel Rod Cladding Wear

2002 ◽  
Author(s):  
T. P. Joulin ◽  
F. M. Gue´rout ◽  
A. Lina ◽  
D. Moinereau
Keyword(s):  
Three Dimensional ◽  
Fretting Wear ◽  
Primary Circuit ◽  
Loading Conditions ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Wear Process ◽  
Fuel Rod ◽  
Support Cell ◽  
Wear Tests

The objective of this study was to investigate the effects of types of motion and loading conditions on the wear of Pressurized Water Reactor (PWR) fuel rod cladding made of Zircaloy-4 in contact with a grid support cell. Fretting-wear tests, for various combinations of motion and preload, were conducted at 310°C and 11.7 MPa using primary circuit water chemistry. Wear coefficients, derived from three-dimensional profilometry, were used to assess the severity of the wear process. The types of motion and the loading conditions were found to have a significant interdependent effect on fuel rod wear coefficients. Scanning Electron Microscope (SEM) examinations were performed on the worn fuel rod cladding specimens to identify wear mechanisms.

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.

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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