Robust design of fuel rods against potential structural damage by flow-induced vibration

2019 ◽  
Vol 33 (01n03) ◽  
pp. 1940008
Author(s):  
Huan-Huan Qi ◽  
Nai-Bin Jiang ◽  
Yi-Xiong Zhang ◽  
Zhi-Peng Feng ◽  
Xuan Huang
Keyword(s):  
Structural Damage ◽  
Flow Distribution ◽  
Fretting Wear ◽  
Leaf Spring ◽  
Wear Depth ◽  
Pressurized Water Reactor ◽  
Flow Induced Vibration ◽  
Vibration Displacement ◽  
Pressurized Water ◽  
Fuel Rod

We studied the flow-induced vibration (FIV) and fretting wear of fuel rod with grid relaxation. According to the flow distribution around a type of pressurized water reactor (PWR) fuel rod, the power spectral density (PSD) is obtained to characterize the turbulence excitation. By combining the correlation of PSD test parameters, the mean square value of the vibration displacement of each rod mode is found, and then the wear depth of dimple position is calculated based on the ARCHARD wear formula. The grids may relax due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of grid clamping force would significantly influence the rod mode and thereby changes its FIV responses. Simulation results show that the failure of the leaf spring has negligible effect on the rod natural frequency whereas the dimple failure near the location with larger FIV amplitude has a much significant effect. The lateral flow velocities at the inlet and outlet of the core are larger. For the fully clamped fuel rod, the responses amplitude of turbulent excitation at the bottom and top of the fuel rod are larger. This is even more obvious with a failed dimple at these locations. Comparatively, the effect of dimple support failure in the middle is less influential. The influence of dimple support failure on the rod wear depth depicts basically the same trend as on the maximum FIV amplitude.

Effect of Clamping Failure on Flow Induced Vibration and Fretting Wear of Fuel Rods

2018 ◽  
Author(s):  
Qi Huan-huan ◽  
Feng Zhi-peng ◽  
Xiong Fu-rui ◽  
Jiang Nai-bin ◽  
Huang Qian ◽  
...  
Keyword(s):  
Natural Frequency ◽  
Lateral Flow ◽  
Fretting Wear ◽  
Leaf Spring ◽  
Wear Depth ◽  
Flow Induced Vibration ◽  
Vibration Displacement ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Fuel Rod

Fuel rods are subjected to both axial and lateral flow in the reactor core. In this study, we present a study on the flow induced vibration (FIV) and fretting wear of fuel rod with failed clamping at grids. First, according to the flow distribution around a type of pressurized water reactor (PWR) fuel rod, the power spectral density (PSD) is obtained to characterize the turbulence excitation. Next, by combining the correlation PSD test parameters, the mean square value of the vibration displacement of each rod mode is found, and then the wear depth of dimple position is calculated based on the ARCHARD wear formula. The clamping of fuel rod at various grids may fail due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of clamping force would significantly influence the rod mode and thereby changes its FIV responses. Simulation results show that the failure of the leaf spring has negligible effect on the rod natural frequency whereas the dimple failure near the location with larger FIV amplitude has a much significant effect. The lateral flow velocities at the inlet and outlet of the core are larger. For the fully clamped fuel rod, the response amplitude of turbulent excitation at the bottom and top of the fuel rod is larger. This is even more obvious with a failed dimple at these locations. Comparatively, the effect of dimple support failure in the middle is less influential. The influence of dimple support failure over the rod wear depth depicts basically the same trend as on the maximum FIV amplitude. In addition to the FIV amplitude, wear depth is also related to rod natural frequency. By examining the multiplication of amplitude and frequency at the top and bottom grid, we found that the dimple failure has greater impact at the top grid.

Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

MRS Advances ◽  
2016 ◽  
Vol 1 (35) ◽  
pp. 2495-2500
Author(s):  
Thomas Winter ◽  
James Huggins ◽  
Richard Neu ◽  
Preet Singh ◽  
Chaitanya S. Deo
Keyword(s):  
Nuclear Reactor ◽  
Wear Behavior ◽  
Sem Analysis ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Rod

ABSTRACTIn support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.

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.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

An Evaluation of Fuel Rod Fretting Wear in Spacer Grid With Conformal Shape

2006 ◽  
Author(s):  
Young Ki Jang ◽  
Nam Kyu Park ◽  
Jae Ik Kim ◽  
Kyu Tae Kim ◽  
Chong Chul Lee ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Calculation Method ◽  
Power Plants ◽  
Failure Time ◽  
Fretting Wear ◽  
Measuring System ◽  
Wear Volume ◽  
Wear Depth ◽  
Volume Calculation ◽  
Fuel Rod

Turbulent flow-induced vibration in nuclear fuel may cause fretting wear of fuel rod at grid support locations. An advanced nuclear fuel for Korean PWR standard nuclear power plants (KSNPs), has been developed to get higher performance comparing to the current fuel considering the safety and economy. One of the significant features of the advanced fuel is the conformal shape in mid grid springs and dimples, which are developed to diminish the fretting wear failure. Long-term hydraulic tests have been performed to evaluate the fretting wear of the fuel rod with the conformal springs and dimples. Wear volume is a measure to predict the fretting wear performance. The shapes of a lot of scars are non-uniform such as wedge shapes, and axially non-symmetric shapes, etc., depending on the contact angle between fuel rod and springs/dimples. In addition, conformal springs and dimples make wear scars wide and thin comparing to conventional ones with convex shape. It is found that wear volumes of these kinds of non-uniform wear scars are over-predicted when the traditionally used wear depth-dependent volume calculation method is employed. In order to predict wear volume more accurately, therefore, the measuring system with high accuracy has been used and verified by the known wear volumes of standard specimens. The wear volumes of the various wear scars have been generated by the measuring system and used for predicting the fretting wear-induced failure time. Based on new evaluation method, it is considered that the fretting wear-induced fuel failure duration with this conformal grid has increased up to 8 times compared to the traditionally used wear depth-dependent volume calculation method.

Experimental Investigation of the Flow-Induced Vibrations of a Rod Cluster Control Assembly Inside Guides With Enlarged Gaps

2019 ◽  
Author(s):  
Pierre Moussou ◽  
Vincent Fichet ◽  
Luc Pastur ◽  
Constance Duhamel ◽  
Yannick Tampango
Keyword(s):  
Experimental Investigation ◽  
Flow Velocity ◽  
Nuclear Power ◽  
Axial Flow ◽  
Fretting Wear ◽  
Pressurized Water Reactor ◽  
Test Rig ◽  
Pressurized Water ◽  
Control Assembly ◽  
Control Rods

Abstract In order to better understand the mechanisms of fretting wear damage of guide cards in some Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP), an experimental investigation is undertaken at the Magaly facility in Le Creusot. The test rig consists of a complete Rod Cluster with eleven Guide Cards, submitted to axial flow inside a water tunnel. In order to mimic the effect of fretting wear, the four lower guide cards have enlarged gaps, so that the Control Rods are free to oscillate. The test rig is operated at ambient temperature and pressure, and Plexiglas walls can be arranged along its upper part, and a series of camera records the vibrations of the control rods above and below the guide cards. The vertical flow velocity is in the range of a few m/s. Beam-like pinned-pinned modes at about 5 Hz are observed, and oscillations of several mm of the central rods are measured, which come along with impacts at the higher flow velocities. A simple non-linear calculation reveals that the main effect of the impacts between Control Rods and Guide Cards is an increase of the natural frequency of the rods by about 10%. Furthermore, as the vibration spectra collapse remarkably well with the flow velocity, the experiments prove that turbulent forcing is responsible for the large oscillations of the control rods, no other mechanism being involved.

Physico-chemical characterization of a spent UO2 fuel with respect to its stability under final disposal conditions

2014 ◽  
Vol 1665 ◽  
pp. 283-289 ◽  
Author(s):  
Ernesto González-Robles ◽  
Detlef H. Wegen ◽  
Elke Bohnert ◽  
Dimitrios Papaioannou ◽  
Nikolaus Müller ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Chemical Characterization ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Final Disposal ◽  
Fuel Rod ◽  
Physico Chemical ◽  
The Stability ◽  
Fission Gas Release

ABSTRACTTwo adjacent fuel rod segments were irradiated in a pressurized water reactor achieving an average burn-up of 50.4 GWd/tHM. A physico-chemical characterisation of the high burn-up fuel rod segments was performed, to determine properties relevant to the stability of the spent nuclear fuel under final disposal conditions. No damage of the cladding was observed by means of visual examination and γ-scanning. The maximal oxide layer thickness was 45 µm. The relative fission gas release was determined to be (8.35 ± 0.66) %. Finally, a rim thickness of 83.7 µm and a rim porosity of about 20% were derived from characterisation of the cladded pellets.

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