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Title: Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel

Journal Article · · Wear

In pressurized water reactors (PWRs), water flow induced vibrations cause contact and rubbing between the fuel rods and the supporting grid, a phenomenon known as Grid-to-Rod-Fretting (GTRF). GTRF may produce progressive wear damage on the fuel claddings leading to subsequent leakage of radioactive fission products. Various accident-tolerant fuel (ATF) concepts are being developed for higher resistance to the high temperature steam and one approach is to apply a cladding coating. Here, fretting wear behavior of a candidate Cr-coating was investigated using a unique bench-scale autoclave testing rig mimicking the environment in an industrial full-assembly PWR simulator. The contact was under a realistically low load (~0.5 N) lubricated by deionized water at a temperature of 204 °C under a pressure of 20-23 bars. Results demonstrated that the Cr-coating significantly improved the cladding's wear resistance when tested against a commercial ZIRLO grid with or without pre-oxidization. In addition, the Cr-coating also reduced wear on the non-oxidized ZIRLO grid but slightly increased the wear on the pre-oxidized grid.

Research Organization:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Organization:
USDOE Office of Nuclear Energy (NE), Nuclear Fuel Cycle and Supply Chain. Advanced Fuel Campaign; USDOE
Grant/Contract Number:
AC05-00OR22725
OSTI ID:
1756262
Alternate ID(s):
OSTI ID: 1776259
Journal Information:
Wear, Vol. 466-467; ISSN 0043-1648
Publisher:
ElsevierCopyright Statement
Country of Publication:
United States
Language:
English

References (18)

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation journal June 2016
Grid-to-rod flow-induced impact study for PWR fuel in reactor journal August 2016
Fretting-wear of zirconium alloys journal April 2002
Research on performance enhancement of nuclear fuel with SiC cladding by using high thermal conductivity fuels journal June 2020
A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding journal January 2020
A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel journal October 2015
Accident tolerant fuel cladding development: Promise, status, and challenges journal April 2018
Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding journal December 2019
Effects of amplitude and frequency on the wear mode change of Inconel 690 SG tube mated with SUS 409 journal May 2014
A multi-stage wear model for grid-to-rod fretting of nuclear fuel rods journal May 2014
Cracking and spalling of the oxide layer developed in high-burnup Zircaloy-4 cladding under normal operating conditions in a PWR journal December 2018
Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes journal March 2018
Enhanced wear resistance of CrAl-coated cladding for accident-tolerant fuel journal September 2019
AREVA NP's enhanced accident-tolerant fuel developments: Focus on Cr-coated M5 cladding journal March 2018
Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors journal September 2016
A study on fretting fatigue characteristics of Inconel 690 at high temperature journal October 2011
Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors journal April 2019
Investigating grid-to-rod fretting wear of nuclear fuel claddings using a unique autoclave fretting rig journal October 2018