Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

Jun Qu; Kevin M. Cooley; Austin H. Shaw; Roger Y. Lu; Peter J. Blau

doi:10.1016/j.wear.2016.02.020

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

Jun Qu, Kevin M. Cooley, Austin H. Shaw, Roger Y. Lu, Peter J. Blau

Surface Engineering and Tribology

Research output: Contribution to journal › Article › peer-review

41 Scopus citations

Abstract

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. This study attempts to measure the wear coefficients of zirconium claddings without and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.

Original language	English
Pages (from-to)	17-22
Number of pages	6
Journal	Wear
Volume	356-357
DOIs	https://doi.org/10.1016/j.wear.2016.02.020
State	Published - Jun 15 2016

Funding

This research was supported by the Consortium for Advanced Simulation of Light Water Reactors ( http://www.casl.gov ), an Energy Innovation Hub ( http://www.energy.gov/hubs ) for Modeling and Simulation of Nuclear Reactors, U.S. Department of Energy . The authors thank Randy J. Parten and William C. Barnhill from ORNL for fretting machine modification and assistance on wear volume quantification, respectively. Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract no. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

Keywords

Grid-to-rod-fretting (GTRF)
Nuclear zirconium claddings
Pre-oxidation
Stage-wise wear model
Wear coefficient

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10.1016/j.wear.2016.02.020

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title = "Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation",

abstract = "In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. This study attempts to measure the wear coefficients of zirconium claddings without and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.",

keywords = "Grid-to-rod-fretting (GTRF), Nuclear zirconium claddings, Pre-oxidation, Stage-wise wear model, Wear coefficient",

author = "Jun Qu and Cooley, {Kevin M.} and Shaw, {Austin H.} and Lu, {Roger Y.} and Blau, {Peter J.}",

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year = "2016",

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N2 - In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. This study attempts to measure the wear coefficients of zirconium claddings without and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.

AB - In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. This study attempts to measure the wear coefficients of zirconium claddings without and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.

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Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

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Keywords

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