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Title: High-temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High_temperature Gas-Cooled Reactor Applications

Abstract

Incoloy 800HT and Inconel 617 have been selected as candidate structural alloys for the hightemperature gas-cooled reactor (HTGR) concept. Helium, the primary coolant, contains impurities (e.g., H2O and CH4) that can induce corrosion reactions at high temperatures, which in turn can affect the tribological behavior of components in sliding contact such as valves and control-rod drive systems. This paper presents results from the study of the tribological behavior of both alloys before and after conditioning them in either an impure-helium oxidizing environment or in an air environment. Both alloys were conditioned for 22 days at elevated temperatures in a once-through helium loop with 4 ppmv H2O and tested subsequently at elevated temperatures with a pin-on-disk tribometer in an air environment 650 degrees C and 750 degrees C for 800HT; 850 degrees C and 900 degrees C for 617 - with various applied loads - 1 N, 2 N and 5 N. SEM-EDS analysis revealed that conditioning the samples in an oxidizing environment leads to the formation of a mixed Fe/Cr-oxide on alloy 800HT and a Cr-oxide on alloy 617, both increasing the wear resistance compared to that of as-received samples. Alloy 617 exhibited lower steady-state friction coefficients compared to thosemore » of alloy 800HT. There was a significant decrease in the scatter of the steadystate friction coefficient of the conditioned samples compared to that of the unconditioned samples. The steady-state friction coefficient for alloy 800HT and 617 were found to be 0.53 +/- 0.07 and 0.25 +/- 0.04, respectively. The wear resistance of alloy 800HT is approximately one order of magnitude lower than that of alloy 617 in all cases that exhibited measurable wear. After sample conditioning, the wear volumes measured at low loads were undistinguishable from the unworn background, a result attributed to the formation of a compacted glaze layer under high temperatures and high contact stresses.« less

Authors:
; ; ; ; ; ; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy - Nuclear Energy University Programs (NEUP)
OSTI Identifier:
1561517
DOE Contract Number:  
AC02-06CH11357
Resource Type:
Journal Article
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 522
Country of Publication:
United States
Language:
English
Subject:
Alloys 800 and 617; Corrosion; Friction; HTGR; Helium; Tribology; Wear

Citation Formats

Pauly, Valentin, Tesch, Carter, Kern, Joseph, Clark, Malcolm, Grierson, David, Singh, Dileep, Ajayi, Oyelayo, and Sridharan, Kumar. High-temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High_temperature Gas-Cooled Reactor Applications. United States: N. p., 2019. Web. doi:10.1016/j.jnucmat.2019.05.025.
Pauly, Valentin, Tesch, Carter, Kern, Joseph, Clark, Malcolm, Grierson, David, Singh, Dileep, Ajayi, Oyelayo, & Sridharan, Kumar. High-temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High_temperature Gas-Cooled Reactor Applications. United States. doi:10.1016/j.jnucmat.2019.05.025.
Pauly, Valentin, Tesch, Carter, Kern, Joseph, Clark, Malcolm, Grierson, David, Singh, Dileep, Ajayi, Oyelayo, and Sridharan, Kumar. Thu . "High-temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High_temperature Gas-Cooled Reactor Applications". United States. doi:10.1016/j.jnucmat.2019.05.025.
@article{osti_1561517,
title = {High-temperature Tribological Behavior of Structural Materials after Conditioning in Impure-Helium Environments for High_temperature Gas-Cooled Reactor Applications},
author = {Pauly, Valentin and Tesch, Carter and Kern, Joseph and Clark, Malcolm and Grierson, David and Singh, Dileep and Ajayi, Oyelayo and Sridharan, Kumar},
abstractNote = {Incoloy 800HT and Inconel 617 have been selected as candidate structural alloys for the hightemperature gas-cooled reactor (HTGR) concept. Helium, the primary coolant, contains impurities (e.g., H2O and CH4) that can induce corrosion reactions at high temperatures, which in turn can affect the tribological behavior of components in sliding contact such as valves and control-rod drive systems. This paper presents results from the study of the tribological behavior of both alloys before and after conditioning them in either an impure-helium oxidizing environment or in an air environment. Both alloys were conditioned for 22 days at elevated temperatures in a once-through helium loop with 4 ppmv H2O and tested subsequently at elevated temperatures with a pin-on-disk tribometer in an air environment 650 degrees C and 750 degrees C for 800HT; 850 degrees C and 900 degrees C for 617 - with various applied loads - 1 N, 2 N and 5 N. SEM-EDS analysis revealed that conditioning the samples in an oxidizing environment leads to the formation of a mixed Fe/Cr-oxide on alloy 800HT and a Cr-oxide on alloy 617, both increasing the wear resistance compared to that of as-received samples. Alloy 617 exhibited lower steady-state friction coefficients compared to those of alloy 800HT. There was a significant decrease in the scatter of the steadystate friction coefficient of the conditioned samples compared to that of the unconditioned samples. The steady-state friction coefficient for alloy 800HT and 617 were found to be 0.53 +/- 0.07 and 0.25 +/- 0.04, respectively. The wear resistance of alloy 800HT is approximately one order of magnitude lower than that of alloy 617 in all cases that exhibited measurable wear. After sample conditioning, the wear volumes measured at low loads were undistinguishable from the unworn background, a result attributed to the formation of a compacted glaze layer under high temperatures and high contact stresses.},
doi = {10.1016/j.jnucmat.2019.05.025},
journal = {Journal of Nuclear Materials},
number = ,
volume = 522,
place = {United States},
year = {2019},
month = {8}
}