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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

FY24 progress report on A709 mechanical properties data development and A709 thermal aging status

The report provides the status of the creep, fatigue, and creep-fatigue testing to date conducted at Idaho National Laboratory to generate the data package. This data package evaluates the material performance from three commercial heats to support the Alloy 709 qualification in American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 5. First procured heat was manufactured by G.O. Carlson with heat number 58776. Second and third heats were fabricated by ATI Specialty Rolled Products with heat numbers 529900 and 530843, respectively. A series of creep and cyclic specimens were fabricated from these three heats and tests were performed. A list of finished, ongoing, and planned test are presented for creep, fatigue, and creep-fatigue tests. Cyclic properties of three commercial heats are compared.

36 MATERIALS SCIENCE↗
DOE OSTI2025

FY24 progress report on A709 mechanical properties data development and A709 thermal aging status

The report provides the status of the creep, fatigue, and creep-fatigue testing to date conducted at Idaho National Laboratory to generate the data package. This data package evaluates the material performance from three commercial heats to support the Alloy 709 qualification in American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 5. First procured heat was manufactured by G.O. Carlson with heat number 58776. Second and third heats were fabricated by ATI Specialty Rolled Products with heat numbers 529900 and 530843, respectively. A series of creep and cyclic specimens were fabricated from these three heats and tests were performed. A list of finished, ongoing, and planned test are presented for creep, fatigue, and creep-fatigue tests. Cyclic properties of three commercial heats are compared.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

Interim Report on FY22 ORNL A709 Welding Research and Testing of Production Welds in Support of Developing ASME A709 Code Case Data Package

As part of the Alloy 709 ASME Code Case development effort under the Advanced Reactor Technologies (ART) Program, this work covers the development of the technical basis for weld fabrication and weld qualification of Alloy 709. This report summarizes the Alloy 709 welding research conducted at Oak Ridge National Laboratory (ORNL) in FY 2022. Two new production welds were fabricated on two commercial heats of Alloy 709 of different phosphorus (P) levels using Alloy 709 filler metal with P content less than 20 wppm with gas tungsten arc welding (GTAW). Both production welds successfully passed ASME Section IX weld qualification tests, and this concludes the Alloy 709 welding procedure development to scale up to 2-in thick plates. In FY 2022, we also demonstrated the success in welding of high P commercial Alloy 709 plates with weld wires having higher P content at 30 wppm. A test weld fabricated with the 30 wppm weld wire on the first commercial heat (140 wppm P) passed all weld qualification tests without issues. Additionally, experiment setup and testing procedure of the circular patch weldability test has been developed, for evaluating the P effect in weld wire on solidification cracking susceptibility of Alloy 709 weld, with the preliminary results summarized in this report. Research on further relaxing the P level restriction beyond 30 wppm are planned in FY 2023. The preliminary cross-weld creep tests results continue to show little or no creep strength reduction relative to the base metal.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Progress Report on Performance of A709 and G91 Steels in Sodium

Specimens in six different processing and heat treatment conditions of A709 H58776 were exposed to sodium at 550, 600, and 650°C, respectively for various exposure times. G91 base metal was tested in sodium at 550, 600, and 650°C, and G91 weldment at 550 and 600°C, respectively. Parallel thermal aging experiments were conducted on G91 and A709 to obtain thermal aging data for comparison with sodium exposure results to separate the thermal and sodium effects. The corrosion data obtained on A709 steel continue to show low corrosion rates, which indicate good compatibility of A709 with sodium when oxygen content is controlled. G91 has also shown acceptable corrosion rates over the temperature range of 550-650°C in sodium. Thermal aging or sodium exposures reduced the tensile strength, uniform elongation, and total elongation for A709 ESR, AOD specimens. These specimens consistently showed higher yield stress and tensile strength and lower uniform and total elongations after sodium exposure than after thermal aging under comparable testing conditions. It implies that there was an additional effect of sodium exposure that gave rise to an increase in tensile strength of A709. Dynamic strain aginginduced flow serrations were completely removed by either thermal aging or sodium exposure at 650°C, which imply that carbon or nitrogen in solution was removed from the solution and formed precipitates during aging or sodium exposures. The tensile data ruled out the possibility of decarburization under the sodium exposure conditions. The HOMO specimens behaved somewhat differently from the ESR and AOD specimens in sodium. While thermal aging and sodium exposures at 550 and 600°C decreased the yield stress and the ultimate tensile strength of G91, there was virtually no additional effect resulting from sodium exposures at these two temperatures. In contrast, sodium exposures at 650°C had a drastic effect on the yield stress and the ultimate tensile strength of G91. It is suggested that G91 experienced decarburization in the 650°C sodium environments. Carbon concentrations in sodium in the SMT-1 and SMT-2 loops were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8-1.2 ppm in the SMT-1 loop and 0.3-0.7 ppm in the SMT-2 loop. The carbon activity in sodium in the SMT-2 loop was estimated to be 0.03-0.08 at 600°C, and 0.08-0.2 at 550°C. The carbon activity in sodium at 650°C in the SMT-1 loop was 0.04-0.07. Equilibrium simulation of the carburization – decarburization processes was conducted for A709 and G91 steels exposed in sodium environments at temperatures of 550-700°C. The carbon activity-concentration relationship for G91 was re-evaluated by considering four phases in G91, i.e. bcc ferrite, M 23 C 6 , NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M 23 C 6 carbides at high carbon activities, while NbC carbides dominated the process at low carbon activities. Formation of NbC increases the decarburization resistance of G91 steel. It remains to be understood whether the decarburization resistance provided by MC carbides in G91 can be maintained during long-term operations of SFRs. The carbon concentration-activity relationship for A709 was calculated based on the equilibrium of fcc-austenite and M 23 C 6 carbide phase. The calculations showed that A709 decarburizes at 650°C in the SMT-1 loop environments, which is different from the experimental findings. The carbon concentration-activity relationship in A709 was further evaluated by considering four phases, including fcc-austenite, M 23 C 6 , NbC, and TiC carbide phases. TiC is the most stable carbides among the three carbide phases. The carburization-decarburization process in A709 is dictated by M 23 C 6 carbides at high carbon activities and by TiC at low carbon activities. Formation of TiC increases the decarburization resistance of A709 steel. Because of the complexity of the precipitation process in A709 during sodium exposures, detailed characterization of precipitates in sodium-exposed specimens and improved thermodynamic models are needed to understand the carburization-decarburization behavior of A709. Kinetic analysis of carbon transfer will be investigated for G91 and A709 in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Swelling resistance of advanced austenitic alloy A709 and its comparison with 316 stainless steel at high damage levels

Alloy A709 is an austenitic alloy developed for power boiler applications in thermal power plants and is being considered as a candidate structural material for Generation IV reactors and fusion reactors, based especially on its increased creep strength over Type 316 stainless steel. However, improved thermal creep properties would not necessarily imply improved radiation resistance, especially with respect to void swelling. Since there are currently no neutron irradiation data on A709 to high fluences, A709 and cold-worked 316 were irradiated in the present study under identical conditions to doses between 100 and 400 peak dpa using irradiation by 3.5 MeV Fe +2 ions. The swelling behavior of 316 is well-known for both neutron and ion irradiation, thus the relative swelling behavior of A709 and 316 under ion irradiation might provide an indication of the swelling behavior of A709 in neutron irradiation environment. Swelling of A709 under ion irradiation was observed over the range of 500-600 ⁰C, peaking at 575 ⁰C. Both A709 and 316 eventually swelled at a rate of ~1 %/dpa under ion irradiation at 575 ⁰C, consistent with the swelling of 316 observed during neutron irradiation. But A709 had a significantly longer transient regime than 316 at any given dpa rate, demonstrating enhanced swelling resistance of A709 over 316 under self-ion irradiation. The swelling levels reached at 400 peak dpa were 55% and 90% for A709 and 316, respectively. The assigned local dpa levels were adjusted both temporally and spatially at each dose for swelling-induced increases in the ion range and concomitant decreases in density. The duration of the transient regimes in each alloy was also observed to increase as the local dpa rate increased, an observation also consistent with neutron irradiation behavior of 316 stainless steel.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Microstructural Characterization of A709 Commercial Heats with Precipitation Treatment

This report discusses work conducted at Idaho National Laboratory in fiscal year 2022 associated with the precipitates in Alloy 709 (A709). The purpose of this work is to support the development of the A709 Code Cases to qualify A709 in Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. This would permit the use of A709 for elevated-temperature nuclear construction. This work encompassed two broad goals. The first goal was to understand the microstructure of the second commercial heat of plate A709 purchased through the Advanced Reactors Technologies Program in both the solution-annealed and precipitation-treated conditions. Transmission electron microscopy was utilized to characterize these microstructures. Direct comparisons were then made between the solution-annealed and precipitation-treated conditions as well as between the first and second commercial heats of plate A709. The second goal was to understand the impact of variations in the time and temperature of the precipitation treatment on the properties of A709. The precipitation treatment temperature was varied from 750°C to 800°C for times ranging from 3 hours to 30 hours. The hardness of each precipitation treatment variation was measured. All of the precipitation treatment variations investigated besides a specimen from the second commercial heat of plate A709 precipitation treated at 800? for 9 hours met the room-temperature hardness requirement specified in ASTM A213 and SA-213 for UNS S31025. The hardness measured for the specimen from the second commercial heat of plate A709 that was precipitation treated at 800? for 9 hours is considered an outlier. The next step is to look at the extremes of the precipitation treatment variations investigated and conduct elevated-temperature mechanical testing. The purpose of this testing would be to assess if these time and temperature variations have any impact on the mechanical performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mechanical Properties of Aged A709

A709 plate is in the process of being developed and qualified through a collaboration by Argonne National Laboratory (ANL), Idaho National Laboratory (INL), and Oak Ridge National Laboratory (ORNL). The goal is to qualify A709 plate in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). This would permit the construction of A709 plate for fast-reactor structural applications. A tradeoff between creep and creep-fatigue properties was established in prior work. A processing route was identified as providing the optimal balance between creep-rupture and elevated-temperature cyclic properties for the solution annealed condition. This condition, however, resulted in a shorter than desired creep-rupture life. Aging is probed to evaluate its potential at further improving mechanical properties. Aging statically precipitates the solutes in solution prior to service. The processing routes identified as providing the best creep-rupture properties and best balance in mechanical properties are investigated. This report discloses the results from tensile, creep, and elevated-temperature cyclic testing performed at INL on the aged material. The aged properties are compared to solution annealed properties. A709 offers improvement in performance compared to 316H particularly at high temperatures; 316H is a material qualified in Section III, Division 5 of the ASME BPVC. For certain high-temperature fast-reactor design conditions, 316H cannot be used while A709 can.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Irradiation And Post-Irradiation Examination Plan For A709

The Advanced Materials and Manufacturing Technologies program is proposing a neutron irradiation campaign for alloy 709 (A709) plate product developed by the Advanced Reactor Technologies program. This irradiation campaign will support the deployment of A709 (included welded components) by providing data of engineering importance, such as data that can be used for establishing end-of-life design parameters and knockdown factors for irradiated materials properties versus properties for non-irradiated material. It will also advance the scientific understanding of A709 irradiation behaviors and material property evolution, which is vital for deploying A709 components into reactors with different neutron spectra, neutron fluxes, and irradiation temperatures as well as accelerating the deployment of new product forms and weld procedures. This document presents the plan for the neutron irradiation and post-irradiation examination for A709 and the scientific and technical basis for the plan.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Status of INL Aged A709 Mechanical Testing

A709 was selected to be qualified in Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Idaho National Laboratory (INL), Argonne National Laboratory and Oak Ridge National Laboratory are collaborating to develop and qualify A709 plate. Aging A709 to form beneficial precipitates prior to service is investigated. One tensile test, five creep-rupture tests, and eight cyclic tests have been completed on aged A709. Two creep-rupture tests are in progress.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Sodium Exposure Tests of A709 Steel (FY2020 Progress Report)

Sodium exposure tests were conducted on the first commercial heat, H58776 of A709 steel. Six different processing and heat treatment conditions were investigated, including three processing conditions, namely, Argon-Oxygen-Decarburization (AOD), Electro-slag Remelting (ESR), and Electro-slag Remelting with subsequent homogenization (HOMO), and two heat treatment conditions, i.e. solution annealing at 1100°C (SA1100°C) or 1150°C (SA1150°C). Specimens of AOD-SA1100°C, AOD-SA1150°C, ESR-SA1100°C, ESR-SA1150°C, HOMO-SA1100°C, and HOMO-SA1150°C were tested in sodium at 650°C and for times up to 10,152 h. Specimens in the AOD-SA1100°C and AOD-SA1150°C conditions were also tested in sodium at 550 and 600°C. Parallel thermal aging experiments were carried out for comparison with sodium exposure tests to separate the thermal and sodium effects. It was found that A709 H58776 generally showed a weight loss after sodium exposures at 650°C except a few data points of the ESR-SA1100°C specimens. The corrosion rate showed large variations in the initial stage and decreased with increasing exposure time. Long-term exposure tests are needed to determine its steady-state corrosion rate. The AOD-SA1100°C and HOMO-SA1100°C specimens showed similar tensile responses after thermal or sodium exposures at 650°C. The yield stress increased, the ultimate tensile strength decreased, and the uniform and total elongations were reduced after thermal or sodium exposures. A longer exposure time had a minimal effect on the tensile strength but further reduced the tensile ductility. The sodium-exposed specimens had higher tensile strength than the thermally-aged specimens. In contrast, the ESR-SA1100°C specimens showed an increase in the yield stress and the ultimate tensile strength after thermal or sodium exposures at 650°C and ~10,000 h. The AOD-SA1150°C and ESR-SA1150°C specimens showed similar tensile behavior after thermal or sodium exposures at 650°C.Thermal or sodium exposures at 650°C increased the yield stress, decreased the ultimate tensile strength, and reduced uniform and total elongations. There were minimal differences in tensile properties between the sodium-exposed and thermally-aged specimens after an ~3000 h exposure. After an exposure for ~10,000 h the sodium-exposed specimens showed higher yield stress and ultimate tensile strength and lower uniform and total elongations than the thermally-aged specimens. For the HOMO-SA1150°C specimens, thermal or sodium exposures at 650°C significantly increased the yield stress but had a minimal effect on the ultimate tensile strength. Both uniform and total elongations were significantly reduced at an exposure time of ~3,000 h, and they were further reduced after an exposure of ~10,000 h. The sodium-exposed specimens had lower tensile ductility than the thermally-aged specimens. Before exposure, the stress-strain curves of the specimens in all six processing-heat treatment conditions showed serrations indicating the dynamic aging effect at 650°C. Thermal aging or sodium exposures at 650°C after ~10,000 h completely removed the serrations of the flow curves in A709 H58776, indicating strong precipitation during thermal or sodium exposures. Longer-term sodium exposure tests are needed to determine the steady-state corrosion rates in sodium for A709 H58776 and fully understand the influence of sodium exposures on its mechanical performance.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Preliminary Experimental Results in Support of the Development of EPP and SMT Design Methods and Viscoplastic Model for A709

The ASME code qualification effort for Alloy 709 (A709) is currently underway to qualify it for Class A construction in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5. The United States (U.S.) Department of Energy (DOE) national laboratories are collaborating in the advanced materials development initiative to investigate the mechanical performance of A709 in support of its code qualification. As part of the A709 code qualification effort, this report summarizes ORNL’s initial experimental findings that support the integration of A709 into the elastic-perfectly plastic (EPP) strain limits code case. It also covers thermomechanical fatigue testing conducted to develop viscoelastic material models, along with the preliminary results of creep-fatigue experiments at 816°C using the Simplified Model Test (SMT) method.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Interim Creep, Fatigue and Creep-Fatigue Data from FY 2022 INL Testing of A709 with Precipitation Treatment for ASME Code Case Data Package

This report provides the status of creep, fatigue, and creep-fatigue testing that transpired in fiscal year 2022 at Idaho National Laboratory (INL). This testing is being conducted to develop the data package to qualify A709 in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). This would permit the use of A709 for elevated temperature nuclear construction. Preliminary results continue to demonstrate the improved creep and fatigue resistance of A709 compared to 316H stainless steel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A709 Code Case Design Parameters

Joint ART Materials/AMMT Program Review to be given at DOE Headquarters on June 5-8, 2023. A709 Code Case Design Parameters regarding preliminary efforts have been started to develop design parameters from test data. Thus far, most of the data evaluated were from the first two commercial heats. Data from these commercial heats continue to support the creep strength advantage of A709 over 316H. Design parameters will be refined as more data are available. Work on the development of other design parameters will begin as data become available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A709 Procurement and ASTM Specification Status

This is a presentation regarding A709 Procurement and ASTM Specification Status for the Joint ART Materials/AMMT Program Review meeting. The A709 team consists of Argonne National Laboratory, Idaho National Laboratory, Oak Ridge National Laboratory, and Subject matter expert John Grubb and Richard Wright. Discussion topics will discuss incorporation of 709 into ASTM Standard A 240 for plate material in the solution annealed condition with a precipitation treatment as a supplemental requirement is underway, Incorporation of Alloy 709 into ASME, Section II, Specification SA 240 will be made upon the completion of the ASTM task, The required commercial scale heats of Alloy 709 for property characterization to provide the technical basis for Code qualification have been procured, and Laboratory testing of properties from these three heats is underway as will be discussed separately at this review meeting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Complete the Planned FY24 Design Parameters Development at INL to Support the Drafting of the A709 Near-Term Code Case

This report presents the thermophysical properties measurements conducted in the Fiscal Year 2024 to extend the database up to 1000?. The new values are compared against properties of high Molybdenum steels in the ASME Section II Part D tables. This report also shows the comparison between new measurements against the previously measured values in literature and provides recommended thermophysical properties for A709 code case. These values will support the thermophysical property tables in near term code case of A709.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Initial Characterization of the Second A709 Commercial Heat

In fiscal year 2021, Idaho National Laboratory (INL) received the second commercial heat of A709 in the plate-product form, heat 529900. The chemistry, microstructure, and mechanical properties of this heat were characterized. Heat 529900 met the chemistry and mechanical property requirements specified for UNS S31025 in American Society for Testing and Materials (ASTM) A213 and American Society of Mechanical Engineers SA 213 for seamless tube which A709 was derived from. The microstructure of heat 529900 was comprised of equiaxed, recrystallized grains and elongated, unrecrystallized grains. The frequency of the unrecrystallized grains varied across the thickness of the plate with the highest frequency observed at the surface and none observed at the center. The equiaxed grain size of heat 529900 ranged from an ASTM grain size number 4 to an ASTM grain size number 6. The size of the equiaxed grains varied across the thickness of the plate with the coarsest grains located at the center of the plate. Additionally, the creep-fatigue properties of the second commercial heat were better than those of the first commercial heat.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

FY19 Report on the Status of Sodium Exposure Tests of the A709 First Commercial Heat

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy A709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The focus of the FY19 program was (a) to complete the sodium exposure of sub-sized specimens from A 709 lab heats and 316H (a reference material) and (b) to continue sodium exposure of sub-sized specimens from A 709 commercial heat 58776.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Microstructural Investigation of Precipitation Treated A709 Test Samples

This report presents the microstructural characterization using scanning transmission electron microscopy (STEM) of two creep-tested ESR1150-PT specimens, one short-term and one long-term. The results show that during creep testing, the microstructure of the A709 ESR1150-PT material continues to evolve as expected due to the synergistic effect of strain, temperature and time. Key features include dislocation forest and cell structures, precipitates on grain boundaries, precipitates randomly within grains, and precipitates associated with dislocations. In addition, the STEM study of one ESR1100-SA specimen that has been tested to rupture at a relatively low temperature (550°C) for a relatively long term (16226 h) is also presented. The results show that there is a high density of very fine nano precipitates form on dislocations.

36 MATERIALS SCIENCE↗