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Title: A Novel Approach to Materials Development for Advanced Reactor Systems

Abstract

OAK - B135 The goals of this project are to advance our understanding of the mechanism of irradiation assisted stress corrosion cracking (IASCC) in austenitic stainless steels using proton irradiation, and to apply proton irradiation to Zircaloys and reactor pressure vessel (RPV) steels to determine if neutron irradiation effects in light water reactors can be effectively emulated with proton irradiation. For austenitic stainless steels, the project focused on determining whether the source of hardening is important in the observed IASCC behavior. This was accomplished by producing set samples of 304L SS with various combinations of cold work and irradiation dose such that the overall hardness was fixed but the relative contribution of the two sources of hardening varied. Results showed that indeed, hardening that was predominantly due to irradiation was considerably more effective in causing IASCC in BWR normal water chemistry at 288°C than was hardening that came principally from cold work. Because the total dose was below the level needed to cause grain boundary segregation, chromium depletion was not a factor. A second study was aimed at determining whether chromium depletion was a key factor in IASCC by creating a sample with enriched grain boundary chromium such that themore » boundary level is depleted to the bulk level during irradiation, thus producing a dislocation loop microstructure typical of that in reactor without chromium depletion at the grain boundaries. Results of these experiments showed that indeed, IASCC occurred even though the grain boundary chromium concentration was not below bulk level. This result provides further evidenced that grain boundary chromium concentration is not the sole determining factor in IASCC even in oxidizing environments typical of BWR normal water chemistry. The two experiments, taken together, show that the irradiated microstructure is an important element in the IASCC process. Proton irradiation was conducted on Zircaloy 4 to determine if the microstructure of the alloy resembled that of the alloy irradiated in-core. Irradiations were conducted using 2 MeV protons to doses of up to 7 dpa at temperatures of 310°C and 350°C and with Ne ions at a temperature of 350°C. Measurements of dislocation loop size and density, hardness and precipitate response were made. Results showed that 2 MeV proton irradiation of Zircaloy 4 at 310 °C leads to Fe loss and amorphization in Zr(Fe,Cr)2 precipitates similar to those observed in the neutron-irradiated alloy at 288-300 °C. State-of-the-art EFTEM reveals Fe accumulation outside the precipitate. Further work will be required in order to compare this result with the redistribution in neutron-irradiated alloys.« less

Authors:
; ; ;
Publication Date:
Research Org.:
University of Michigan
Sponsoring Org.:
USDOE Office of Nuclear Energy, Science and Technology (NE)
OSTI Identifier:
812022
DOE Contract Number:  
FG03-99SF21927
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; proton irradiation; stainless steels; Zircaloys; reactor pressure vessel steels; dose; hardening; irradiation assisted stress corrosion cracking; microstructure; microchemistry; radiation induced segregation

Citation Formats

Was, Gary S, Atzmon, Michael, Wang, Lumin, and Hash, Mark. A Novel Approach to Materials Development for Advanced Reactor Systems. United States: N. p., 2003. Web.
Was, Gary S, Atzmon, Michael, Wang, Lumin, & Hash, Mark. A Novel Approach to Materials Development for Advanced Reactor Systems. United States.
Was, Gary S, Atzmon, Michael, Wang, Lumin, and Hash, Mark. Mon . "A Novel Approach to Materials Development for Advanced Reactor Systems". United States.
@article{osti_812022,
title = {A Novel Approach to Materials Development for Advanced Reactor Systems},
author = {Was, Gary S and Atzmon, Michael and Wang, Lumin and Hash, Mark},
abstractNote = {OAK - B135 The goals of this project are to advance our understanding of the mechanism of irradiation assisted stress corrosion cracking (IASCC) in austenitic stainless steels using proton irradiation, and to apply proton irradiation to Zircaloys and reactor pressure vessel (RPV) steels to determine if neutron irradiation effects in light water reactors can be effectively emulated with proton irradiation. For austenitic stainless steels, the project focused on determining whether the source of hardening is important in the observed IASCC behavior. This was accomplished by producing set samples of 304L SS with various combinations of cold work and irradiation dose such that the overall hardness was fixed but the relative contribution of the two sources of hardening varied. Results showed that indeed, hardening that was predominantly due to irradiation was considerably more effective in causing IASCC in BWR normal water chemistry at 288°C than was hardening that came principally from cold work. Because the total dose was below the level needed to cause grain boundary segregation, chromium depletion was not a factor. A second study was aimed at determining whether chromium depletion was a key factor in IASCC by creating a sample with enriched grain boundary chromium such that the boundary level is depleted to the bulk level during irradiation, thus producing a dislocation loop microstructure typical of that in reactor without chromium depletion at the grain boundaries. Results of these experiments showed that indeed, IASCC occurred even though the grain boundary chromium concentration was not below bulk level. This result provides further evidenced that grain boundary chromium concentration is not the sole determining factor in IASCC even in oxidizing environments typical of BWR normal water chemistry. The two experiments, taken together, show that the irradiated microstructure is an important element in the IASCC process. Proton irradiation was conducted on Zircaloy 4 to determine if the microstructure of the alloy resembled that of the alloy irradiated in-core. Irradiations were conducted using 2 MeV protons to doses of up to 7 dpa at temperatures of 310°C and 350°C and with Ne ions at a temperature of 350°C. Measurements of dislocation loop size and density, hardness and precipitate response were made. Results showed that 2 MeV proton irradiation of Zircaloy 4 at 310 °C leads to Fe loss and amorphization in Zr(Fe,Cr)2 precipitates similar to those observed in the neutron-irradiated alloy at 288-300 °C. State-of-the-art EFTEM reveals Fe accumulation outside the precipitate. Further work will be required in order to compare this result with the redistribution in neutron-irradiated alloys.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2003},
month = {6}
}

Technical Report:
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