Engineering topics

Lu, Roger Y.

Publications and source records attributed to Lu, Roger Y..

7 records

DOE OSTI2021

Grid-to-rod fretting wear study of SiC/SiC composite accident-tolerant fuel claddings using an autoclave fretting bench test

Grid-to-rod-fretting (GTRF) in pressurized water reactors (PWRs) is known to cause wear and surface damage on the fuel claddings, potentially leading to radioactive leakage. One of the accident-tolerant fuel (ATF) concepts is to use advanced cladding materials that could withstand higher temperatures. Here, we investigated the wear behavior of candidate silicon carbide (SiC)-based composite claddings with different levels of surface finish in fretting against a commercial ZIRLO alloy grid using a unique bench-scale autoclave GTRF rig. The experiments mimicked the environment in an industrial full-assembly PWR simulator. Fretting tests were conducted with a realistic load (~0.5 N) in deionized water under a pressure of 20–23 bar at 204 °C for 100 h. While the SiC/SiC composite claddings showed significantly higher wear resistance than the commercial ZIRLO alloy cladding as expected, the smoother versions experienced surprisingly higher wear than the much softer counterface, ZIRLO grid. The wear mechanism of the SiC/SiC cladding was attributed to the SiC wear debris that was trapped at the fretting interface causing both 3-body and 2-body (embedded into the grid surface) abrasion of the cladding. Rougher SiC/SiC claddings had less material loss but caused more wear on the ZIRLO grid. Finally, pre-oxidized ZIRLO grid showed better compatibility with the SiC/SiC cladding to protect both the cladding and grid as a result of reduced wear debris trapping.

36 MATERIALS SCIENCE↗

DOE OSTI2020

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

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

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

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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Summary report describing CASL accomplishments in support of the GTRF Challenge Problem

Coolant flow-induced relative motions between structural components in the cores of lightwater nuclear reactor cores, a phenomenon called Grid-To-Rod Fretting (GTRF), can result in progressive wear damage. Wear-through of the cladding on fuel rods can introduce radioactive species into the coolant loop. Such an event is termed a “leaker” and may cause premature and expensive reactor shut-down. GTRF results in wear scars of various sizes, shapes, and depths on the sides of the fuel rod cladding. Fretting damage may also occur on the springs and dimples that hold the fuel rods in place, resulting in the formation of gaps. Due to variability in the operating conditions of the reactor (burn-up rates and operating campaigns) and the associated variable cladding surface oxide film growth and removal rates, the wear coefficient is unlikely to remain constant over time. Therefore, constant wear rate models do not accurately account for transients in the operating conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

GTRF Cladding Wear Modeling and Experimental Validation

Coolant flow-induced relative motions between certain structural components in the cores of lightwater nuclear reactor cores, a phenomenon called Grid-To-Rod Fretting (GTRF), can result in progressive wear damage. Wear-through of the cladding on fuel rods can introduce radioactive species into the coolant loop. Such an event is termed a “leaker” and may cause premature and expensive reactor shut-down. GTRF results in wear scars of various sizes, shapes, and depths on the sides of the fuel rod cladding. Fretting damage may also occur on the springs and dimples that hold the fuel rods in place, resulting in the formation of gaps. Due to variability in the operating conditions of the reactor (burn-up rates and operating campaigns) and the associated variable cladding surface oxide film growth and removal rates, the wear coefficient is unlikely to remain constant over time. Therefore, constant wear rate models do not accurately account for transients in the operating conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Grid-to-rod flow-induced impact study for PWR fuel in reactor

The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effective way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

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

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. In this paper, we attempt 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. Finally, 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.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗