Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors

Giulia B. Mazzei, Rosa Lo Frano, James F. Stubbins

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Abstract

Many R&D programs are being conducted worldwide to study and, in some cases, develop materials with the properties required by Gen IV reactor systems. These advanced reactor systems require materials having sufficient dimensional stability and mechanical properties to meet the anticipated near- and long-term service demands under extreme conditions. These conditions are most challenging in environments with high levels of irradiation damage coupled with requirements for corrosion resistance and severe mechanical loading conditions. Based on past experience, austenitic alloys based around 316SS compositions, 15% Cr-15% Ni-Ti, have not been able to meet these demanding conditions at elevated temperatures. Ferritic-martensitic steels and ferritic-martensitic ODS steels have been proposed for these conditions. However, long-term fast neutron irradiation results show that austenitic alloys of the 15%Cr-25%Ni-Ti-Nb class have excellent swelling resistance at temperatures up to 600°C and at doses larger than 150 dpa. A major reason for this excellent behaviour is the thermal-mechanical treatment of the alloy and the presence of both Ti and Nb to doubly stabilize the precipitate structure. This material could represent an effective option for claddings in SFR and other advanced reactor applications. For this reason, in consideration of 60-y-of-service requirement, predictions of long-term materials behaviour will be carried out taking into account ageing effects (from 500° to 700°C). A complete analysis aiming to characterize the microstructure of the as-received material, cold-worked at 20%, and also after thermal treatment is under way. Ageing tests as well as TEM and SEM analysis of the microstructural evolution as a function of aging condition will be performed. The evolution of some phases in the common nucleation and growth sites will be presented.

Original languageEnglish (US)
Title of host publicationStudent Paper Competition
PublisherAmerican Society of Mechanical Engineers (ASME)
ISBN (Electronic)9780791850053
DOIs
StatePublished - Jan 1 2016
Event2016 24th International Conference on Nuclear Engineering, ICONE 2016 - Charlotte, United States
Duration: Jun 26 2016Jun 30 2016

Publication series

NameInternational Conference on Nuclear Engineering, Proceedings, ICONE
Volume5

Other

Other2016 24th International Conference on Nuclear Engineering, ICONE 2016
CountryUnited States
CityCharlotte
Period6/26/166/30/16

Fingerprint

Austenitic stainless steel
Heat treatment
Martensitic steel
Aging of materials
Ferritic steel
Neutron irradiation
Dimensional stability
Microstructural evolution
Swelling
Corrosion resistance
Precipitates
Nucleation
Irradiation
Transmission electron microscopy
Mechanical properties
Temperature
Microstructure
Scanning electron microscopy
Chemical analysis

ASJC Scopus subject areas

  • Nuclear Energy and Engineering

Cite this

Mazzei, G. B., Lo Frano, R., & Stubbins, J. F. (2016). Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors. In Student Paper Competition [V005T15A052] (International Conference on Nuclear Engineering, Proceedings, ICONE; Vol. 5). American Society of Mechanical Engineers (ASME). https://doi.org/10.1115/ICONE24-60710

Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors. / Mazzei, Giulia B.; Lo Frano, Rosa; Stubbins, James F.

Student Paper Competition. American Society of Mechanical Engineers (ASME), 2016. V005T15A052 (International Conference on Nuclear Engineering, Proceedings, ICONE; Vol. 5).

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Mazzei, GB, Lo Frano, R & Stubbins, JF 2016, Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors. in Student Paper Competition., V005T15A052, International Conference on Nuclear Engineering, Proceedings, ICONE, vol. 5, American Society of Mechanical Engineers (ASME), 2016 24th International Conference on Nuclear Engineering, ICONE 2016, Charlotte, United States, 6/26/16. https://doi.org/10.1115/ICONE24-60710
Mazzei GB, Lo Frano R, Stubbins JF. Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors. In Student Paper Competition. American Society of Mechanical Engineers (ASME). 2016. V005T15A052. (International Conference on Nuclear Engineering, Proceedings, ICONE). https://doi.org/10.1115/ICONE24-60710
Mazzei, Giulia B. ; Lo Frano, Rosa ; Stubbins, James F. / Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors. Student Paper Competition. American Society of Mechanical Engineers (ASME), 2016. (International Conference on Nuclear Engineering, Proceedings, ICONE).
@inproceedings{86b7867075b44350ac07eb2d5435f138,
title = "Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors",
abstract = "Many R&D programs are being conducted worldwide to study and, in some cases, develop materials with the properties required by Gen IV reactor systems. These advanced reactor systems require materials having sufficient dimensional stability and mechanical properties to meet the anticipated near- and long-term service demands under extreme conditions. These conditions are most challenging in environments with high levels of irradiation damage coupled with requirements for corrosion resistance and severe mechanical loading conditions. Based on past experience, austenitic alloys based around 316SS compositions, 15{\%} Cr-15{\%} Ni-Ti, have not been able to meet these demanding conditions at elevated temperatures. Ferritic-martensitic steels and ferritic-martensitic ODS steels have been proposed for these conditions. However, long-term fast neutron irradiation results show that austenitic alloys of the 15{\%}Cr-25{\%}Ni-Ti-Nb class have excellent swelling resistance at temperatures up to 600°C and at doses larger than 150 dpa. A major reason for this excellent behaviour is the thermal-mechanical treatment of the alloy and the presence of both Ti and Nb to doubly stabilize the precipitate structure. This material could represent an effective option for claddings in SFR and other advanced reactor applications. For this reason, in consideration of 60-y-of-service requirement, predictions of long-term materials behaviour will be carried out taking into account ageing effects (from 500° to 700°C). A complete analysis aiming to characterize the microstructure of the as-received material, cold-worked at 20{\%}, and also after thermal treatment is under way. Ageing tests as well as TEM and SEM analysis of the microstructural evolution as a function of aging condition will be performed. The evolution of some phases in the common nucleation and growth sites will be presented.",
author = "Mazzei, {Giulia B.} and {Lo Frano}, Rosa and Stubbins, {James F.}",
year = "2016",
month = "1",
day = "1",
doi = "10.1115/ICONE24-60710",
language = "English (US)",
series = "International Conference on Nuclear Engineering, Proceedings, ICONE",
publisher = "American Society of Mechanical Engineers (ASME)",
booktitle = "Student Paper Competition",

}

TY - GEN

T1 - Investigation on thermal treatments of high-resistance austenitic stainless steels employed as structural materials in generation IV reactors

AU - Mazzei, Giulia B.

AU - Lo Frano, Rosa

AU - Stubbins, James F.

PY - 2016/1/1

Y1 - 2016/1/1

N2 - Many R&D programs are being conducted worldwide to study and, in some cases, develop materials with the properties required by Gen IV reactor systems. These advanced reactor systems require materials having sufficient dimensional stability and mechanical properties to meet the anticipated near- and long-term service demands under extreme conditions. These conditions are most challenging in environments with high levels of irradiation damage coupled with requirements for corrosion resistance and severe mechanical loading conditions. Based on past experience, austenitic alloys based around 316SS compositions, 15% Cr-15% Ni-Ti, have not been able to meet these demanding conditions at elevated temperatures. Ferritic-martensitic steels and ferritic-martensitic ODS steels have been proposed for these conditions. However, long-term fast neutron irradiation results show that austenitic alloys of the 15%Cr-25%Ni-Ti-Nb class have excellent swelling resistance at temperatures up to 600°C and at doses larger than 150 dpa. A major reason for this excellent behaviour is the thermal-mechanical treatment of the alloy and the presence of both Ti and Nb to doubly stabilize the precipitate structure. This material could represent an effective option for claddings in SFR and other advanced reactor applications. For this reason, in consideration of 60-y-of-service requirement, predictions of long-term materials behaviour will be carried out taking into account ageing effects (from 500° to 700°C). A complete analysis aiming to characterize the microstructure of the as-received material, cold-worked at 20%, and also after thermal treatment is under way. Ageing tests as well as TEM and SEM analysis of the microstructural evolution as a function of aging condition will be performed. The evolution of some phases in the common nucleation and growth sites will be presented.

AB - Many R&D programs are being conducted worldwide to study and, in some cases, develop materials with the properties required by Gen IV reactor systems. These advanced reactor systems require materials having sufficient dimensional stability and mechanical properties to meet the anticipated near- and long-term service demands under extreme conditions. These conditions are most challenging in environments with high levels of irradiation damage coupled with requirements for corrosion resistance and severe mechanical loading conditions. Based on past experience, austenitic alloys based around 316SS compositions, 15% Cr-15% Ni-Ti, have not been able to meet these demanding conditions at elevated temperatures. Ferritic-martensitic steels and ferritic-martensitic ODS steels have been proposed for these conditions. However, long-term fast neutron irradiation results show that austenitic alloys of the 15%Cr-25%Ni-Ti-Nb class have excellent swelling resistance at temperatures up to 600°C and at doses larger than 150 dpa. A major reason for this excellent behaviour is the thermal-mechanical treatment of the alloy and the presence of both Ti and Nb to doubly stabilize the precipitate structure. This material could represent an effective option for claddings in SFR and other advanced reactor applications. For this reason, in consideration of 60-y-of-service requirement, predictions of long-term materials behaviour will be carried out taking into account ageing effects (from 500° to 700°C). A complete analysis aiming to characterize the microstructure of the as-received material, cold-worked at 20%, and also after thermal treatment is under way. Ageing tests as well as TEM and SEM analysis of the microstructural evolution as a function of aging condition will be performed. The evolution of some phases in the common nucleation and growth sites will be presented.

UR - http://www.scopus.com/inward/record.url?scp=84995793711&partnerID=8YFLogxK

UR - http://www.scopus.com/inward/citedby.url?scp=84995793711&partnerID=8YFLogxK

U2 - 10.1115/ICONE24-60710

DO - 10.1115/ICONE24-60710

M3 - Conference contribution

AN - SCOPUS:84995793711

T3 - International Conference on Nuclear Engineering, Proceedings, ICONE

BT - Student Paper Competition

PB - American Society of Mechanical Engineers (ASME)

ER -