scholarly journals Completion of a Chlorination Test Using 250 grams of High-Burnup Used Fuel Cladding from a North Anna Pressurized Water Reactor

2018 ◽  
Author(s):  
Emory D. Collins ◽  
Tom D. Hylton ◽  
Guillermo Daniel DelCul ◽  
Barry B. Spencer ◽  
Rodney Dale Hunt ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Burnup

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

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.

CFD Simulation of Fission Gas Release Under Fuel Defects Condition in Pressurized Water Reactor

2018 ◽  
Author(s):  
Bing Dong ◽  
Leihao Li ◽  
Chenyue Li ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Cfd Simulation ◽  
Three Dimensional ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Two Phase ◽  
Release Process ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fission Gas ◽  
Failure Monitoring

During normal pressurized water reactor (PWR) operation, the fuel cladding is inevitably defective. It is important to develop a model to evaluate the fission gas (FG) release from the fuel-cladding gap into the coolant, which benefits the fuel failure monitoring and improves the reactor safety. The objective of this paper is to give an overview of three-dimensional two-phase transient CFD simulation based on the Volume of Fluid (VOF) method to evaluate the fission gas migrating inside the gap and escaping through the defects. The topics of this paper include the behavior and release rate of FG in the fuel-cladding gap when the fuel cladding is defective. An analysis is presented on the suitability of the CFD simulation and it is shown that three-dimensional two-phase transient CFD simulation can be utilized in evaluating the FG release process. The results show that water enters the gap immediately after an abrupt cladding defect. The entering water flashes and causes pressure pulsation, which induces the transient FG release.

High Temperature Oxidation of the Hot Rolled ZrNbMoGe Alloy for Fuel Cladding Material of Pressurized Water Reactor

2015 ◽  
Vol 1123 ◽  
pp. 356-359 ◽  
Author(s):  
Bandriyana ◽  
A.K. Rivai ◽  
J.H. Prajitno ◽  
A. Dimyati
Keyword(s):  
High Temperature ◽  
Oxide Layer ◽  
High Temperature Oxidation ◽  
Pressurized Water Reactor ◽  
Fuel Cladding ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Water Reactor ◽  
Hot Rolled ◽  
Cladding Material

The Zr-2.5Nb-0.5Mo-0.1Ge alloy were developed for application as fuel cladding material in an advanced Pressurized Water Reactor (PWR) with higher burn-up at higher service temperature. Oxidation behaviour of the alloy at high temperature similar to condition related to the operating with Lost of Cooling Accident (LOCA) conditions was investigated. The sample ingot was synthesized in an arc melting furnace followed by hot rolling at 850 °C down to 3 mm of thickness. Each one alloy ingot and one alloy sheet were subjected to the high temperature oxidation test in the Magnetic Suspension Balance (MSB) at 500 °C. Light optical microscope (LIOM) and Scanning Electron Microscope (SEM) equipped with X-ray Diffraction Spectroscopy (EDS) were used to characterize the microstructure of the oxide layer. The Vickers hardness tester was used to evaluate the hardness of the alloy matrix before and after oxidation processes. The results concluded that both samples showed oxidation rate characteristic which follows the parabolic phenomenon. However the hot rolled sample had lower rate. The oxide layer was indicated as ZrO2 .

Sign in / Sign up

Export Citation Format

Share Document