Search NASA⌕ Search

Engineering topics

Natesan, K.

Publications and source records attributed to Natesan, K..

33 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Fatigue Behavior of Pressure Tube Material Zr-2.5Nb in Air and in Simulated CANDU-Reactor Water Environments

An experimental program was initiated to develop fatigue data in strain-controlled conditions for an assessment of the fatigue behavior of Zr-2.5Nb pressure tube material in support of CANDU reactors for Atomic Energy of Canada Limited (AECL)/Chalk River Laboratories (CRL). Fatigue tests were conducted on specimens fabricated from a Zr2.5Nb pressure tube to determine the fatigue performance in air and in simulated CANDUreactor water environments. This report summarizes the fatigue data of Zr-2.5Nb material obtained in both environments.

Li, M.↗
DOE OSTI2013

June 2013 Update: Status Report on Assessment of Environmentally Assisted Fatigue for LWR Extended Service Conditions

This report provides an update on an earlier assessment of environmentally assisted fatigue for light water reactor (LWR) materials under extended service conditions. This quarterly report is a deliverable in April-June 2013 quarter under the work package for environmentally assisted fatigue in the Light Water Reactor Sustainability (LWRS) program. The overall objective of this LWRS project is to assess the degradation by environmentally assisted cracking/fatigue of LWR materials such as various alloy base metals and their welds used in reactor coolant system piping. This effort is to support the Department of Energy LWRS program for developing tools to predict the aging/failure mechanism and to correspondingly predict the remaining life of LWR components for anticipated 60-80 year operation.

Mohanty, S.↗
DOE OSTI2013

Crack Growth Rate and Fracture Toughness Tests on Irradiated Cast Stainless Steels

Cast austenitic stainless steels are used in the cooling system of light water reactors for components with complex shapes, such as pump casings, valve bodies, and coolant piping. In the present study, crack growth rate and fracture toughness JR curve tests were performed on irradiated cast stainless steels in low-corrosion-potential environments (low-dissolved-oxygen high-purity water or simulated pressurized water reactor environment) at 320°C. Both as received and thermally aged materials were included to investigate the combined effect of thermal aging and irradiation embrittlement on the fracture behavior of cast stainless steels. The samples were irradiated to approximately 0.08 dpa at the Halden reactor. Good resistance to corrosion fatigue and stress corrosion cracking was observed for all samples. Thermal aging had little effect on the crack growth behavior at this dose level. Cleavage-like fracture was the dominant cracking morphology during the crack growth rate tests, and the ferrite phase was deformed to a lesser extent compared with the surrounding austenite phase. The fracture toughness results showed a dominant effect of neutron irradiation, and the fracture resistances were decreased considerably for all cast specimens regardless of their thermal aging conditions. The reduction in fracture toughness was more significant in the unaged than thermally aged materials. Nonetheless, the fracture toughness values of thermally aged specimens were about 30% lower than their unaged counterparts, suggesting a combined effect of thermal aging and neutron irradiation in cast stainless steel.

Chen, Y.↗
DOE OSTI2013

Technical Letter Report of Slow-Strain-Rate Tensile Tests on Irradiated Stainless Steels in a PWR Environment

Slow-strain-rate tensile (SSRT) tests have been performed on irradiated specimens in a simulated pressurized water reactor (PWR) environment. The samples are miniature tensile specimens of various austenitic stainless steels (SSs) with different thermal-mechanical treatments commonly used for reactor core internal components. Neutron irradiation was carried out in the BOR-60 reactor, a sodium-cooled, fast breeder reactor in Russia, at ~320°C. The damage doses of the specimens are 5, 10, and 48 dpa (displacements per atom). All irradiated materials show significant irradiation hardening and loss of ductility in the SSRT tests. The yield strengths of cold-worked are higher than that of solution-annealed samples at all doses up to 48 dpa. While the irradiation hardening seems to saturate between ~5 and ~10 dpa, the loss of ductility continues to increase above 10 dpa. Strain softening is also observed for all irradiated materials above 5 dpa. Fractographic examinations show that ductile dimple fracture is the dominant morphology for all SSRT tests in the PWR environment. Small areas of transgranular, mixed mode and cleavage fractures are seen on most fracture surfaces in PWR water tests. Intergranular cracking is also observed in 48-dpa Types 316 and 347 SSs. Cracking susceptibility of the tested materials was evaluated with fracture morphology and time to failure. In general, high-doses cold-worked SSs are more susceptible to transgranular cleavage cracking in the PWR environment. Solution-annealed Type 347 SS is susceptible to intergranular cracking at 45 dpa in the PWR environment.

Chen, Y.↗
DOE OSTI2013

Technical Letter Report on the Cracking of Irradiated Stainless Steels in Low-Corrosion-Potential Environments

Irradiation-assisted stress corrosion cracking (IASCC) is a major degradation mechanism for reactor internal components of austenitic stainless steels (SSs) in light water reactors. Crack growth rate (CGR) and fracture toughness J-R curve tests were performed on four irradiated SS specimens in high-purity water with low dissolved oxygen and in a simulated pressurized water reactor (PWR) environment. The samples had been previously irradiated from ~5 to 8 dpa in the BOR-60 reactor. The materials were cold-worked (CW) 316 and solution-annealed (SA) 304L SSs, which are commonly used in PWR core internals. Cyclic and constant-load CGRs were obtained from these samples to assess their cracking behaviors in low-corrosion-potential environments. A Ti-stabilized specimen, CW 316-Ti SS, was found to be highly vulnerable to IASCC, and its cracking susceptibility increased with neutron dose between ~5 and 8 dpa. The CW 316 and SA 304L SSs showed moderate cracking susceptibilities in the low-corrosion potential environments up to ~7 dpa. Periodic partial unloading (PPU) had a significant impact on the crack growth behavior of all irradiated specimens, and the measured CGRs were more than one order of magnitude higher in the test periods with than without PPU. A “stair-step” crack growth behavior was also observed for all irradiated specimens during the test periods with PPU, suggesting a strong effect of dynamic loading on IASCC. The SA 304L SS showed the best fracture resistance, and the CW 316 SS was the most brittle among the tested specimens. The fracture toughness J values of the two CW 316-Ti SS specimens were nearly identical at ~5 and ~8 dpa.

Chen, Y.↗
DOE OSTI2012

Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR - 60 Reactor

As nuclear power plants age and neutron fluence increases, detrimental effects resulting from radiation damage have become an increasingly important issue for the operational safety and structural integrity of core internal components. In this study, irradiated specimens of reactor core internal components were characterized by transmission electron microscopy. The specimens had been irradiated to 5.5-45 dpa in the BOR-60 reactor at a dose rate close to 10 -6 dpa/s and temperature of about 320°C. No voids were observed in the austenitic stainless steels and nickel alloys at all doses. Despite the possibility that fine voids below the TEM resolution limit may be present, it was clear that void swelling was insignificant in all examined alloys up to 45 dpa. Irradiated microstructures of the studied alloys were dominated by a high density of Frank loops. The mean size and density of the Frank loops varied from one material to another, but saturated with increasing dose above ~10 dpa. While no irradiation-induced precipitations were present below 24.5 dpa, fine precipitates were evident in several alloys at 45 dpa.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Resolution of Qualification Issues for Existing Structural Materials

This report gives a detailed assessment of several key technical issues that needs resolution for the existing structural materials with emphasis on application in liquid metal reactors (LMRs), in particular, sodium cooled fast reactors. The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for the Advanced Fuel Cycle Initiative (AFCI) Reactor Campaign. The report is the second deliverable in FY09 (M2505050201) under the work package “Advanced Materials Code Qualification”. The overall objective of the Advanced Materials Code Qualification project is to evaluate the key technical requirements for the qualification of currently available and future advanced materials for application in sodium reactor systems and the resolution of issues that the U.S. Nuclear Regulatory Commission (NRC) has raised in the past on structural materials in support of the design and licensing of the LMR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. However, the development of sufficient database and qualification of these materials for application in LMRs require considerable amount of time and resources. In the meantime, the currently available materials will be used in the early development of fast reactors. Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants licensing. As the LMR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Section III Subsection NH (Class 1 Components in Elevated Temperature Service). Assessment of materials performance issues and high temperature design methodology issues pertinent to the LMR were presented in an earlier report (Natesan et al. 2008). In a subsequent report (Majumdar et al. 2009), we addressed the needs in high temperature methodologies for design of various high temperature components in sodium cooled fast reactor. The present report addresses several key technical issues for the currently available structural materials such as Type 304 and 316 austenitic stainless steels and ferritic steels such as 2.25Cr-1Mo and modified 9Cr-1Mo. The 60-year design life for the LMR presents a significant challenge to the development of database, extrapolation/prediction of long-term performance, and high temperature structural design methodology. The current Subsection NH is applicable to the design life only up to 34 years. No experimental data contain test durations of 525,000 hours, and it is impractical to conduct such long-term tests in any types of testing. So far the longest creep tests for Grade 91 and Grade 92 steels have run up to 100,000 hours. It has been noted that there is a large drop in creep rupture strength in long-term tests for these high-Cr creep-resistant steels, which may result in overestimation of the long-term creep strength and allowable stress. The report addresses in detail the need for a mechanistic understanding of the structural materials, from the standpoint of the effects of thermal aging, creep deformation, creep fracture, fatigue and creep-fatigue, creep-fatigue predictive models, fatigue and creep crack growth, and fracture toughness. Based on an in-depth assessment of the available data and mechanistic understanding, key technical issues are identified and discussed for each of the property areas. Furthermore, we have proposed viable approaches to resolve the issues and prioritized our recommendations.

Natesan, K.↗
DOE OSTI2012

Code Qualification of Structural Materials for AFCI Advanced Recycling Reactors

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Corrosion performance of advanced structural materials in sodium.

This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.'

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Creep Rupture Testing of Alloy 617 and A508/533 Base Metals and Weldments

The NGNP, which is an advanced HTGR concept with emphasis on both electricity and hydrogen production, involves helium as the coolant and a closed-cycle gas turbine for power generation with a core outlet/gas turbine inlet temperature of 750-1000°C. Alloy 617 is a prime candidate for VHTR structural components such as reactor internals, piping, and heat exchangers in view of its resistance to oxidation and elevated temperature strength. However, lack of adequate data on the performance of the alloy in welded condition prompted to initiate a creep test program at Argonne National Laboratory. In addition, Testing has been initiated to evaluate the creep rupture properties of the pressure vessel steel A508/533 in air and in helium environments. The program, which began in December 2009, was certified for quality assurance NQA-1 requirements during January and February 2010. Specimens were designed and fabricated during March and the tests were initiated in April 2010. During the past year, several creep tests were conducted in air on Alloy 617 base metal and weldment specimens at temperatures of 750, 850, and 950°C. Idaho National Laboratory, using gas tungsten arc welding method with Alloy 617 weld wire, fabricated the weldment specimens. Eight tests were conducted on Alloy 617 base metal specimens and nine were on Alloy 617 weldments. The creep rupture times for the base alloy and weldment tests were up to ~3900 and ~4500 h, respectively. The results showed that the creep rupture lives of weld specimens are much longer than those for the base alloy, when tested under identical test conditions. The test results also showed that the creep strain at fracture is in the range of 7-18% for weldment samples and were much lower than those for the base alloy, under similar test conditions. In general, the weldment specimens showed more of a flat or constant creep rate region than the base metal specimens. The base alloy and the weldment exhibited tertiary creep after 50-60% of the rupture life, irrespective of test temperature in the range of 750-950°C. The results showed that the stress dependence of the creep rate followed a power law for both base alloy and weldments. The data also showed that the stress exponent for creep is the same and one can infer that the same mechanism is operative in both base metal and weldments in the temperature range of the current study. SEM fractography analysis indicated that both base metal and weldment showed combined fracture modes consisting of dimple rupture and intergranular cracking. Intergranular cracking was more evident in the weldment specimens, which is consistent with the observation of lower creep ductility in the weldment than in the base metal.

08 HYDROGEN↗
DOE OSTI2011

Report on Sodium Compatibility of Advanced Structural Materials

This report provides an update on the evaluation of sodium structural materials. The report is a deliverable (level 3) in FY11 (M3A11AN04030403), under the Work Package A-11AN040304, “Sodium Compatibility of Advanced Structural Materials” performed by Argonne National Laboratory (ANL), as part of Advanced Structural Materials Program for the Advanced Reactor Concepts. This work package supports the advanced structural materials development by providing corrosion and tensile data from the standpoint of sodium compatibility of advanced structural alloys. The scope of work involves exposure of advanced structural alloys such as G92, mod.9Cr-1Mo (G91) ferritic-martensitic steels and HT-UPS austenitic stainless steels to a flowing sodium environment with controlled impurity concentrations. The exposed specimens are analyzed for their corrosion performance, microstructural changes, and tensile behavior. Previous reports examined the thermodynamic and kinetic factors involved in the design, fabrication, and construction of a forced convection sodium loop for sodium compatibility studies of advanced materials. This report presents the results on corrosion performance, microstructure, and tensile properties of advanced ferritic-martensitic and austenitic alloys exposed to liquid sodium at 550°C for up to 2700 h and at 650°C for up to 5064 h in the forced convection sodium loop. The oxygen content of sodium was controlled by the cold-trapping method to achieve ~1 wppm oxygen level. Four alloys were examined, G92 in the normalized and tempered condition (H1 G92), G92 in the cold-rolled condition (H2 G92), G91 in the normalized and tempered condition, and hot-rolled HT-UPS. G91 was included as a reference to compare with advanced alloy, G92. It was found that all four alloys showed weight loss after sodium exposures at 550 and 650°C. The weight loss of the four alloys was comparable after sodium exposures at 550°C; the weight loss of ferritic-martensitic steels, G92 and G91 is more significant than that of austenitic stainless steel, HT-UPS after sodium exposures at 650°C. Sodium exposures up to 2700 h at 550°C had no significant influence on tensile properties, while sodium exposures up to 5064 h at 650°C dramatically lowered the tensile strengths of the four alloys. The ultimate tensile strength of H1 G92, H2 G92, and G91 ferritic-martensitic steels was reduced to as much as nearly half of its initial value after sodium exposures at 650°C. Though the uniform elongation was recovered to some extent, these three ferritic-martensitic steels showed considerable strain softening after sodium exposures. The yield stress of HT-UPS austenitic stainless steel increased, the ultimate tensile strength decreased, and the total elongation was reduced after sodium exposures at 650°C. The dynamic strain aging effect observed in the as-received HT-UPS specimens became less pronounced after sodium exposures at 650°C. Microstructural characterization of sodium-exposed specimens showed no appreciable surface deterioration or grain structure changes under an optical microscope, except for the H2 G92 steel, in which the martensite structure transformed to large grain ferrite after sodium exposures at 650°C. TEM observations of the sodium-exposed H2 G92 steel showed significant recrystallization after sodium exposure for 2700 h at 550°C, and transformation of martensite to ferrite and high density of precipitates in nearly dislocation-free matrix after sodium exposures at 650°C. Further microstructural analysis and evaluation of decarburization/carburization behavior is needed to understand the dramatic changes in the tensile strengths of advanced ferritic-martensitic and austenitic steels after sodium exposures at 650°C.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Report on Thermal Aging Effects on Tensile Properties of Advanced Austenitic Steels

This report provides an update on the evaluation of thermal-aging induced degradation of tensile properties of advanced HT-UPS austenitic stainless steels. The report is the second deliverable (level 3) in FY11 (M3A11AN04030103), under the Work Package A-11AN040301, “Advanced Alloy Testing” performed by Argonne National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Concepts. This work package supports the advanced structural materials development by providing tensile data on aged alloys and a mechanistic model, validated by experiments, with a predictive capability on long-term performance. The scope of work is to evaluate the effect of thermal aging on the tensile properties of advanced alloys such as ferritic-martensitic steels, mod.9Cr-1Mo (G91), G92, and advanced austenitic stainless steel, HT-UPS. The aging experiments have been conducted over a temperature of 550-750°C for various time periods to simulate the microstructural changes in the alloys. In addition, a mechanistic model based on thermodynamics and kinetics has been used to address the changes in microstructure of the alloys as a function of time and temperature, which is developed in the companion work package at ANL. The focus of this project is advanced alloy testing and understanding the effects of long-term thermal aging on the tensile properties. The first deliverable of FY11 reported the tensile testing results of thermally-aged G92 in the normalized and tempered condition (H1 G92), G92 in the cold-rolled condition (H2 G92), and reference alloy, G91 ferritic-martensitic steels. This report summarizes the tensile data of thermally aged HT-UPS austenitic stainless steel. Two heats of HT-UPS have been investigated, i.e., solution-annealed HT-UPS (SA HT-UPS) and hot-rolled HT-UPS (HR HT-UPS). Tensile data have been obtained for SA HT-UPS thermally-aged for up to 18031 h at 550°C and for up to 17595 h at 650°C, as well as for HR HT-UPS thermally aged for up to 6909 h at 550°C, 4293 h at 600°C, and 5407 h at 650°C. Major finding are: (1) the solution-annealed HT-UPS showed continued, moderate increase in strength and decrease in ductility with increasing aging time up to 18034 h at 550°C. Thermal aging at 650°C increased the yield stress and decreased the ultimate tensile strength and drastically reduced the ductility in the SA HT-UPS. For the HR HT-UPS, thermal aging at 550°C for up to 6909 h showed little influence on its tensile behavior. Thermal aging at 600°C for 4293 h increased the strength and decreased the ductility. The 650°C-thermally-aged HR HT-UPS specimens showed very different strengthening, strain hardening, and flow localization behavior. Thermal aging at 650°C increased the yield stress but decreased the ultimate tensile stress for the HR HT-UPS. (2) Dynamic strain aging (DSA) was observed in both the solution-annealed HT-UPS and the hot-rolled HT-UPS at temperatures between 550-650°C at a strain rate of 0.001 s -1 . Extensive precipitation in the matrix after thermal aging at 600 and 650°C significantly reduced serrated plastic flow observed in the stress-strain curves, i.e. less effect of DSA. Because the DSA phenomenon can have significant influence on tensile, creep strength, and fatigue resistance, and the DSA effects in HT-UPS steels were observed in the temperature range of interest to the SFR applications, the DSA phenomenon in the HT.

Li, Meimei↗
DOE OSTI2011

Report on Thermal Aging Effects on Tensile Properties of Ferritic-Martensitic Steels.

This report provides an update on the evaluation of thermal-aging induced degradation of tensile properties of advanced ferritic-martensitic steels. The report is the first deliverable (level 3) in FY11 (M3A11AN04030103), under the Work Package A-11AN040301, 'Advanced Alloy Testing' performed by Argonne National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Concepts. This work package supports the advanced structural materials development by providing tensile data on aged alloys and a mechanistic model, validated by experiments, with a predictive capability on long-term performance. The scope of work is to evaluate the effect of thermal aging on the tensile properties of advanced alloys such as ferritic-martensitic steels, mod.9Cr-1Mo, NF616, and advanced austenitic stainless steel, HT-UPS. The aging experiments have been conducted over a temperature of 550-750°C for various time periods to simulate the microstructural changes in the alloys as a function of time at temperature. In addition, a mechanistic model based on thermodynamics and kinetics has been used to address the changes in microstructure of the alloys as a function of time and temperature, which is developed in the companion work package at ANL. The focus of this project is advanced alloy testing and understanding the effects of long-term thermal aging on the tensile properties. Advanced materials examined in this project include ferritic-martensitic steels mod.9Cr-1Mo and NF616, and austenitic steel, HT-UPS. The report summarizes the tensile testing results of thermally-aged mod.9Cr-1Mo, NF616 H1 and NF616 H2 ferritic-martensitic steels. NF616 H1 and NF616 H2 experienced different thermal-mechanical treatments before thermal aging experiments. NF616 H1 was normalized and tempered, and NF616 H2 was normalized and tempered and cold-rolled. By examining these two heats, we evaluated the effects of thermal-mechanical treatments on material microstructures and associated mechanical properties during long-term aging at elevated temperatures. Thermal aging experiments at different temperatures and periods of time have been completed: 550°C for up to 5000 h, 600°C for up to 7500 h, and 650°C for more than 10,000 h. Tensile properties were measured on thermally aged specimens and aging effect on tensile behavior was assessed. Effects of thermal aging on deformation and failure mechanisms were investigated by using in-situ straining technique with simultaneous synchrotron XRD measurements.

36 MATERIALS SCIENCE↗