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 Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

2015 ◽  
Author(s):  
Sharon M. Robinson ◽  
Marc Rhea Chattin ◽  
Joseph Giaquinto ◽  
Robert Thomas Jubin
Keyword(s):  
Reactor Fuel ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Tritium Content ◽  
Pressurized Water ◽  
Water Reactor

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.

Estimation of Pressure Loads During Ex-Vessel Steam Explosion

2008 ◽  
Author(s):  
Matjazˇ Leskovar
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Phase 1 ◽  
Interaction Process ◽  
Radioactive Material ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central and side melt pours), the cavity water sub-cooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplets freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.

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