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Title: In-Situ Proton Irradiation-Corrosion Study of ATF Candidate Alloys in Simulated PWR Primary Water

Abstract

Irradiation enhanced corrosion behavior of Accident Tolerant Fuel candidate alloys T91 and Fe15Cr4Al were evaluated using in-situ proton irradiation-corrosion experiments in hydrogenated pure water (at 320 °C, 3 wppm H2) with a 5.4 MeV proton beam. The thin sample acted as a “window” to allow protons to fully penetrate the sample while maintaining system pressure. The area of the samples exposed to the proton beam experienced effects from displacement damage and radiolysis products. The aim of the study was to characterize the effect of radiation on the kinetics and character of oxidation caused by accelerated waterside corrosion under irradiation. Samples irradiated with protons for total displacement damage of *0.1 dpa (dose rate in water, 400 kGy/s) at an exposure time of 24 h were compared. Oxide morphology, phase structure, and composition of the oxide, metal and the metal/oxide interface were investigated using TEM and EDS and are related to the test conditions. The oxidation rate or resulting oxide thickness is dependent on the alloy Cr content; the oxidation rate increased as the Cr content decreased. The resulting oxide consists of an inner layer of Cr-rich spinel oxide and outer magnetite crystals in the unirradiated region; while the irradiated region consistsmore » of Cr-rich inner oxide spinel that was partially dissolved and coverage of outer non-faceted hematite precipitates.« less

Authors:
;
Publication Date:
Research Org.:
Notre Dame University
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
OSTI Identifier:
1474322
Report Number(s):
2
DOE Contract Number:  
NE0008272
Resource Type:
Conference
Journal Name:
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, The Minerals, Metals & Materials Series
Additional Journal Information:
Conference: 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, Portland, Oregon, August 13-17, 2017
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; Accident tolerant fuel, Irradiation, Proton, Corrosion

Citation Formats

Was, Gary, and Wang, Peng. In-Situ Proton Irradiation-Corrosion Study of ATF Candidate Alloys in Simulated PWR Primary Water. United States: N. p., 2017. Web.
Was, Gary, & Wang, Peng. In-Situ Proton Irradiation-Corrosion Study of ATF Candidate Alloys in Simulated PWR Primary Water. United States.
Was, Gary, and Wang, Peng. Sun . "In-Situ Proton Irradiation-Corrosion Study of ATF Candidate Alloys in Simulated PWR Primary Water". United States. https://www.osti.gov/servlets/purl/1474322.
@article{osti_1474322,
title = {In-Situ Proton Irradiation-Corrosion Study of ATF Candidate Alloys in Simulated PWR Primary Water},
author = {Was, Gary and Wang, Peng},
abstractNote = {Irradiation enhanced corrosion behavior of Accident Tolerant Fuel candidate alloys T91 and Fe15Cr4Al were evaluated using in-situ proton irradiation-corrosion experiments in hydrogenated pure water (at 320 °C, 3 wppm H2) with a 5.4 MeV proton beam. The thin sample acted as a “window” to allow protons to fully penetrate the sample while maintaining system pressure. The area of the samples exposed to the proton beam experienced effects from displacement damage and radiolysis products. The aim of the study was to characterize the effect of radiation on the kinetics and character of oxidation caused by accelerated waterside corrosion under irradiation. Samples irradiated with protons for total displacement damage of *0.1 dpa (dose rate in water, 400 kGy/s) at an exposure time of 24 h were compared. Oxide morphology, phase structure, and composition of the oxide, metal and the metal/oxide interface were investigated using TEM and EDS and are related to the test conditions. The oxidation rate or resulting oxide thickness is dependent on the alloy Cr content; the oxidation rate increased as the Cr content decreased. The resulting oxide consists of an inner layer of Cr-rich spinel oxide and outer magnetite crystals in the unirradiated region; while the irradiated region consists of Cr-rich inner oxide spinel that was partially dissolved and coverage of outer non-faceted hematite precipitates.},
doi = {},
journal = {Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, The Minerals, Metals & Materials Series},
number = ,
volume = ,
place = {United States},
year = {2017},
month = {8}
}

Conference:
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