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Title: Optimization and testing results of Zr-bearing ferritic steels

Abstract

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, asmore » well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.« less

Authors:
 [1];  [1];  [2];  [2]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Univ. of Wisconsin, Madison, WI (United States)
Publication Date:
Research Org.:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Center for Nanophase Materials Sciences (CNMS)
Sponsoring Org.:
USDOE
OSTI Identifier:
1159486
Report Number(s):
ORNL/TM-2014/440
NT0104000; NENT016
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE

Citation Formats

Tan, Lizhen, Yang, Ying, Tyburska-Puschel, Beata, and Sridharan, K. Optimization and testing results of Zr-bearing ferritic steels. United States: N. p., 2014. Web. doi:10.2172/1159486.
Tan, Lizhen, Yang, Ying, Tyburska-Puschel, Beata, & Sridharan, K. Optimization and testing results of Zr-bearing ferritic steels. United States. doi:10.2172/1159486.
Tan, Lizhen, Yang, Ying, Tyburska-Puschel, Beata, and Sridharan, K. Mon . "Optimization and testing results of Zr-bearing ferritic steels". United States. doi:10.2172/1159486. https://www.osti.gov/servlets/purl/1159486.
@article{osti_1159486,
title = {Optimization and testing results of Zr-bearing ferritic steels},
author = {Tan, Lizhen and Yang, Ying and Tyburska-Puschel, Beata and Sridharan, K.},
abstractNote = {The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. Ferritic-martensitic (FM) steels are important structural materials for nuclear reactors due to their advantages over other applicable materials like austenitic stainless steels, notably their resistance to void swelling, low thermal expansion coefficients, and higher thermal conductivity. However, traditional FM steels exhibit a noticeable yield strength reduction at elevated temperatures above ~500°C, which limits their applications in advanced nuclear reactors which target operating temperatures at 650°C or higher. Although oxide-dispersion-strengthened (ODS) ferritic steels have shown excellent high-temperature performance, their extremely high cost, limited size and fabricability of products, as well as the great difficulty with welding and joining, have limited or precluded their commercial applications. Zirconium has shown many benefits to Fe-base alloys such as grain refinement, improved phase stability, and reduced radiation-induced segregation. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of a new generation of Zr-bearing ferritic alloys to be fabricated using conventional steelmaking practices, which have excellent radiation resistance and enhanced high-temperature creep performance greater than Grade 91.},
doi = {10.2172/1159486},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Sep 01 00:00:00 EDT 2014},
month = {Mon Sep 01 00:00:00 EDT 2014}
}

Technical Report:

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  • The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization.« less
  • The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less
  • A program is in progress in which the physical properties of 316 stainless steel and 21/4 Cr-- 1 Mo steel are compared in dynamic sodium systems at 1200 deg F and 1100 deg F respectively. This report presents the results of the first test series in which the fatigue, tensile, creep and creep-torupture properties of Type 316 ss were compared in 1200 deg F sodium, air, and helium environments. The results of these tests revealed: (1) The stainless steel specimens to be used throughout the program were representative of an average Type 316 ss heat. (2) The high temperature (1200more » deg F) fatigue life of 316 ss in helium is longer than in air. The fatigue life of 316 ss in 1200 deg F sodium is the same as in air at high cyclic strains but is the same as in 1200 deg F helium at the low cyclic strains. Short time exposure to 1200 deg F sodium prior to high cyclic strain tests did not appreciably effect the life in sodium and helium but shortened the life in air. (3) No significant differences in the creep-rupture properties of 316 ss were discernible in air, sodium, or helium at 1200 deg F. Specimens exposed for 4000 hours (unstressed) in 1200 deg F sodium also showed no significant change in creep-rupture properties. (4) The creep rates of 316 ss appear to be consistentiy higher in 1200 deg F helium and sodium compared to 1200 deg F air; however, the results show the stresses to produce a minimum creep rate of 1% in 10,000 hours in all three environments to be within a 15% spread. (5) Tensile properties showed no significant differences when tested in air or helium regardless of whether the specimens were exposed for 4000 hours to sodium or whether the exposed specimens were washed or unwashed prior to testing. (auth)« less