Investigating grid-to-rod fretting wear of nuclear fuel claddings using a unique autoclave fretting rig

Journal Article · Thu Jun 28 00:00:00 EDT 2018 · Wear

DOI:https://doi.org/10.1016/j.wear.2018.06.011· OSTI ID:1474636

Lazarevic, Sladjan  ^[1]; Lu, Roger Y. ^[2]; Favede, Cyrille ^[3]; Plint, George ^[3]; Blau, Peter J. ^[4]; Qu, Jun  ^[1]

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1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division
2. Westinghouse Electric Company, Hopkins, SC (United States)
3. Phoenix Tribology Ltd, Hampshire (United Kingdom)
4. Blau Tribology Consulting, Enka, NC (United States)

Fuel rods in a pressurized water nuclear reactor (PWR) experience coolant flow-induced relative motion against the grid support features, a phenomenon called grid-to-rod fretting (GTRF), which may cause progressive fretting wear damage. Wear-through of the cladding on fuel rods will leak the radioactive fuel into the coolant loop to significantly increase the radiation level in PWR primary loop and often trigger expensive Post Irradiation Examination. In this work, a unique autoclave fretting test rig was designed and fabricated to allow studying GTRF using actual cladding and grid materials. Water temperature was up to 220 °C, which was designed to mimic the environment in an industry full-assembly reactor core simulator. The contact load (0.1–1 N), oscillation frequency (20–30 Hz) and stroke (50–150 µm), and work rate (1–3 mW) were defined based on simulations of GTRF in an actual reactor. Using this autoclave GTRF rig, tests were conducted to learn the effect of water temperature and to investigate the wear behavior of different cladding-grid material combinations currently used in actual PWRs, including Zr alloy claddings without and with pre-oxidation against Zr alloy and Inconel grids. Higher water temperature evidently increased the cladding wear, and pre-oxidation of the Zr alloy and/or using the Inconel grid effectively reduced the wear rate. In conclusion, results were correlated with the data of a full-size industry reactor core simulator in terms of both wear rate and wear scar morphology.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE
• Grant/Contract Number: AC05-00OR22725
• OSTI ID: 1474636
• Alternate ID(s): OSTI ID: 1702117
• Journal Information: Wear, Vol. 412-413, Issue C; ISSN 0043-1648
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
Grid-to-rod fretting (GTRF)
Autoclave
Zirconium
Inconel
Wear coefficient