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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Results of FY 2023 Alloy 617 and Alloy 709 High-Temperature Crack-Growth Testing

This report summarizes the work performed at Idaho National Laboratory under the “Creep, fatigue and creep-fatigue crack growth tests” task of the “Long-Term VHTR Material Qualification – INL” work package. Work was performed this year to validate the crack growth monitoring setup used with the test frames against both continuous crack length monitoring using an optical camera, as well as post-mortem analysis of marker bands on the fracture surface. Delays prevented the use of a gauge to measure load-line displacement for early tests, and so the creep-fatigue crack growth setup was evaluated using the actuator displacement for a load-line displacement setup. While not ideal, this allowed examination of the shortcomings with the current software setup that was originally designed for performing stress corrosion crack growth rate studies. The method for data collection was modified to link the crack growth monitoring software with Instron’s Wave Matrix software. Once the load-line displacement gauge arrived, the crack growth equipment was successfully modified to permit continuous monitoring of both load-line displacement and crack length. This is critical for creep-fatigue and creep crack growth rate studies for ductile material, as it permits the C* and Ct analyses. Test results are shown for Alloy 617 fatigue and creep-fatigue (without the gauge for load-line displacement measurements), as well as creep-fatigue of Alloy 709, which was performed with the completed test setup, allowing for crack length and load-line displacement monitoring.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Summary of Alloy 617 and Alloy 709 Elevated Temperature Crack Growth Test Results from the Planned FY24 Crack Growth Test Activities

This report covers the results from crack growth rate testing performed at Idaho National Laboratory in the fiscal year 2024. It includes two creep crack growth rate tests and two creep-fatigue crack growth rate tests. Results are shown for each of these tests, included crack growth measured during the test using direct current potential drop, and measurements based off of the fracture surface. Load line displacement, as well as load, were also collected during the tests. Detailed analysis of the results will continue into the next fiscal year, however, a number of potential improvements to the equipment and testing procedures were identified that will increase testing reliability and throughput, as well as decrease complications with the analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Assessment of Overmatched Filler (Alloy 617) to Improve Alloy 800H Stress Rupture Factors

Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) specifies rules for elevated temperature nuclear reactors. Currently, only six metals are qualified in Section III, Division for construction of Class A metallic pressure boundary components for elevated-temperature service, one of these being Alloy 800H. There are only two permissible filler metals Alloy 800H is qualified to be welded with: Alloy A and Alloy 82. The stress rupture factors for these two filler metals are low near the maximum temperatures and service lives Alloy 800H is qualified for. This may preclude vendors from being able to use Alloy 800H for elevated-temperature nuclear construction. In this work, an overmatched filler metal, Alloy 617 is assessed to determine its potential to offer improved stress rupture factors. Scoping creep-rupture tests of cross-welds with Alloy 800H base metal and Alloy 617 filler metal were conducted. Preliminary results do not indicate that Alloy 617 will offer significantly improved stress rupture factors. Consequently, a matching filler metal, UTP A 2133, is now being investigated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Effects of Geometric Discontinuities on Creep Behavior of Alloy 617

Due to the excellent long-term high-temperature performance of Alloy 617, the Alloy has been identified as the primary structural material for the intermediate heat exchanger of the Very High Temperature Reactor (VHTR). Both the base and welded alloy has been qualified to be used up to 950 °C up to 100,000 hours by the American Society of Mechanical Engineer (ASME) Boiler and Pressure Vessel Code (BPVC). The qualification properties of the material were exclusively determined by uniaxial creep tests. The components may, however, experience multiaxial stress conditions in service due to the reactor geometry. This study used notched creep tests to investigate the effect of geometric discontinuities on creep behavior of both the base and welded Alloy 617. The V-notch tests reveal that the geometric discontinuities do not reduce the creep rupture live under high and intermediate stress conditions. The U-notch tests show that a larger radius notch behave in a similar way as a straight gauge section of a uniaxial test. The creep rupture live of the welded alloy was slightly longer than the base metal.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

ELECTRIC FIELD-ASSISTED DIFFUSION WELDING TO FABRICATE ALLOY 617 COMPACT HEAT EXCHANGERS

Compact heat exchangers are of interest for a number of applications including advanced reactors. Alloy 617 is one of the top candidate materials for the gas-cooled reactor intermediate heat exchanger. Previous endeavors to diffusion weld Alloy 617 utilized hot pressing (HP). It was reported that grain boundary migration across the interface was hindered by extensive precipitation. Bonds of this nature have been observed to reduce the elevated-temperature mechanical properties compared to the wrought-product form. It was hypothesized that the electric current applied during electric-field-assisted sintering (EFAS) can overcome these challenges, resulting in improved diffusion welding (DW). This study investigated DW of Alloy 617 via EFAS. Stacks composed of three sheets that were 20 mm in diameter were welded using EFAS. Specimens were welded with an applied electric current, a pressure of 30 MPa, hold time of 30 min, and temperatures of 1050°C, 1100°C, and 1150°C. DW using HP as the zero-current analog of EFAS was also performed at the most promising EFAS conditions. Results revealed that both the applied electric current and temperature played a key role in precipitation and grain boundary migration in diffusion-welded Alloy 617. Precipitates were observed at the interface of the hot-pressed samples which limited grain boundary migration. Electric current was found to prevent precipitate formation along the interface at 1150°C. The electric current coupled with a temperature of 1150°C during EFAS resulted in significant grain boundary migration across the interface.

36 - MATERIALS SCIENCE↗
DOE OSTI2022

Results of Initial Alloy 617 High Temperature Crack Growth Testing

Crack propagation data can provide valuable insights when performing a safety evaluation for a component. Alloy 617 is qualified in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear service up to 950°C. Little is known, however, about subcritical crack-growth phenomena in Alloy 617. Previous crack-growth studies of Alloy 617 did not investigate temperatures across the range which is currently qualified in Section III, Division 5. High-temperature crack-growth testing in air and in reactor-grade helium can provide data for establishing the crack-growth correlations in support of an ASME BPVC Section XI high-temperature flaw evaluation Code Case. Idaho National Laboratory (INL) has previously performed crack-growth testing, but the equipment requires revitalization. This report provides the status of crack-growth testing in Alloy 617 at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Draft ASME Boiler and Pressure Vessel Code Cases and Technical Bases for Use of Alloy 617 for Constructions of Nuclear Component Under Section III, Division 5

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code currently only allows five materials for use in construction of nuclear components for high temperature service. These are: 2.25Cr-1Mo and V-modified 9Cr-1Mo steels, Types 304 and 316 stainless steels and the high nickel Alloy 800H. Since 2005, the US high temperature gas-cooled reactor program has been characterizing elevated temperature mechanical properties of Alloy 617 as the leading candidate construction material for the intermediate heat exchanger. After analysis of these experimental results, along with historical data and additional results available through the Generation IV International Forum, Very High Temperature Reactor, Materials Program Management Board Materials Handbook, a draft ASME Code Case to allow nuclear construction with Alloy 617 for temperatures up to 1750°F (954°C) has been developed. This report contains the Code Case for Low Temperature Service Construction of Section III, Division 5, Subsection HB, Subpart A, Class A and Subsection HC, Subpart A, Class B components, which has been approved in Section II, Materials, and Section III, Rules for Construction of Nuclear Facility Components. Supporting technical justification for the low temperature Code Case is also included. This Code Case allows use of Alloy 617 up to 800°F (425°C). This report also contains an updated draft of a Section III, Division 5, Subsection HB, Subpart B, Class A Code Case for Alloy 617 to qualify it for use in construction of nuclear components up to 1750°F (954°C) for service life up to 100,000 hours. The draft contained in Appendix 4, subject to editorial revision and approval by the ASME Special Task Group on Alloy 617 Code Qualification, will be submitted for approval by letter ballot by the appropriate ASME Committees. The technical justification supporting the Code Case is presented in Appendix 5 of this report. This background document is part of the information package that will be submitted with the Code Case for ballot.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Report on FY 2023 Research and Development on Specially Designed Creep-fatigue Experiments on Alloy 617 in Support of Improving Creep-fatigue Evaluation Approaches

Experimental and numerical studies in support of developing the integrated Elastic–Perfectly Plastic (EPP) plus Simplified Model Test (SMT) design methodology, referred to as the EPP+SMT method, continued in FY 2023. This report focuses on the methods for extrapolating the EPP+SMT creep-fatigue life curves at long hold times and low strain ranges at elevated temperatures. In this work, the available uniaxial creep-fatigue failure data on Alloy 617 at temperatures of 950°C and 850°C were analyzed to provide a guidance on the development of the extrapolation method and the creep-fatigue failure criteria. A viscoplastic constitutive model for Alloy 617 was adopted to extrapolate the mechanical responses to low strain ranges and long hold times. A set of design curves of Alloy 617 at temperatures of 950°C, 850°C, and 800°C with tensile hold times of 1 hr, 100 hr, and 1,000 hr are developed. Furthermore, creep-fatigue testing on two notch specimen geometries, shallow-notch and sharp V-notch, on Alloy 617 was performed 950°C to understand the multiaxial stress relaxation behavior. The experimental and numerical results on the notch specimens were compared with those on the standard uniaxial smooth bar specimens. The effect of multi-axial stress state combined with elastic follow-up on the stress relaxation behavior was investigated in this report.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Enhanced diffusion bonding of alloy 617 using electric field-assisted sintering

The development of compact heat exchangers (CHXs) has gained increasing interest in many industries owing to their high thermal efficiency and reduced size. Diffusion bonding (DB) is an advantageous technique for fabricating CHXs. Alloy 617 is a candidate for manufacturing CHXs for high-temperature advanced nuclear reactors due to its elevated-temperature properties. Previous endeavors in DB of Alloy 617 were conducted by hot pressing (HP), which reported precipitates at the diffusion-bond interface, limited grain boundary (GB) migration, and significantly reduced high-temperature mechanical properties. To overcome these challenges, this study investigated DB of Alloy 617 using electric field-assisted sintering (EFAS). Stacks composed of three sheets were bonded with EFAS using different temperatures, pressures, and hold times. DB using HP as the zero-current analog of EFAS was also performed for comparison. The result shows that Cr- and Mo-rich precipitates were formed at the interface of the hot-pressed samples. The electric current and temperature in EFAS play a significant role in precipitation and GB migration. The electric current coupled with correct temperatures can effectively prevent precipitate formation at the interface and achieve excellent GB migration. Nanoscale Al-rich oxide was formed at the interface of the samples made by both HP and EFAS, but grain boundaries can ignore the nanoscale Al-oxide and migrate across the interface. The temperature, pressure, and hold time also affected diffusion. The temperature is a prerequisite for a successful GB migration, and GB migration can be enhanced by increasing pressure and hold time.

36 MATERIALS SCIENCE↗
DOE OSTI2022

AN INITIAL ASSESSMENT OF THE CREEP-RUPTURE STRENGTHS FOR WELDMENTS WITH ALLOY 800H BASE METAL AND ALLOY 617 FILLER METAL

In Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Alloy 800H is qualified for elevated-temperature nuclear construction for temperatures up to 760°C (1,400°F) and a maximum service life of 300,000 hours. There are two permissible filler metals for Alloy 800H weldments specified in Division 5: ENiCrFe-2 (Alloy A) and ERNiCr-3 (Alloy 82). Low creep-rupture strengths of these weldments at the upper limits of the qualified temperatures and service lives may restrict the design envelope for elevated-temperature nuclear construction with Alloy 800H. As a result, an alternative filler metal is desired to improve the creep-rupture strengths of Alloy 800H weldments for the qualified temperatures and service lives. This work investigates an overmatched filler metal. Specifically, a weldment with Alloy 800H base metal and Alloy 617 filler metal fabricated by semiautomated gas tungsten arc welding is investigated. A scoping creep-rupture test program was conducted of cross-weld specimens at temperatures ranging from 750 to 1,000°C (1,292 to 1,832°F). Preliminary results on the creep-rupture strengths of the Alloy 800H weldment with Alloy 617 filler metal do not show significant improvement compared to the filler metals currently qualified in Division 5 for Alloy 800H weldments. Consequently, work is in progress to investigate a matching filler metal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Electric field enhanced diffusion welding of alloy 617: Microstructural characteristics and mechanical properties

This study investigated the microstructural characteristics and mechanical behavior of diffusion welded nickel-based Alloy 617 obtained by electric field-assisted sintering (EFAS) using various parameters. The interfacial microstructure exhibited different characteristics including good grain boundary (GB) migration across the interface in the samples diffusion-welded at 1100 °C and a flat interface in the samples joined at 1000 °C and 1050 °C. The interface consisted of fine Al 2 O 3 oxides, while precipitation of interfacial M 23 C 6 carbides was not observed. Grain boundaries migrated across the Al 2 O 3 oxides, leaving these oxides within the grains. Graded grain size was observed, with grain coarsening being more significant near the sample surface due to the temperature gradient induced by EFAS. Tensile testing revealed that the specimens fractured in the matrix away from the interface, indicting strong diffusion-welded joints. Further, the peak tensile strength of 807 MPa was obtained in the samples welded at 1000 °C due to minimal grain growth. The materials obtained at 1100 °C exhibited reduced tensile strength but improved ductility. Strain maps revealed by digital image correlation showed alternating high and low strain segments in the samples produced at 1000 °C and 1050 °C, indicating that the flat interfaces with no GB migration were less ductile compared to the matrix. A greater strain uniformity was observed along the bond interfaces with improved GB migration. The hardness reduced near the sample surfaces due to enlarged grains induced by temperature gradient. This study demonstrates that GB migration and enhanced mechanical strength can be achieved in diffusion-welded Alloy 617.

36 MATERIALS SCIENCE↗
DOE OSTI2022

An extrapolation method for strain ranges and hold times in developing the EPP+SMT creep-fatigue design curves for Alloy 617

Experimental and numerical studies in developing the integrated Elastic–Perfectly Plastic (EPP) plus Simplified Model Test (SMT) design methodology, referred to as the EPP+SMT method, continued in FY2022. This report focuses on the methods for extrapolating the EPP+SMT creep-fatigue (CF) design curves at long hold times and low strain ranges. In this study, the available CF failure data on Alloy 617 at 950°C were analyzed to determine a set of CF failure criteria. At very low strain ranges and long hold times, CF failure data are not accessible by experiments because of the extraordinarily long test durations and the inability of the test machines to accurately control these small strain ranges. A CF experimental approach with the concept of block-strain range CF testing protocol was developed. Tests using this protocol were conducted to generate the needed information for calibrating material parameters of the numerical material models. The Time Fraction based method and Dissipated Energy method were used to extrapolate the CF life curves to low strain ranges and long hold times. Based on the new experimental approach and CF life prediction methods, the CF life curves with various hold times were developed for Alloy 617 at 950°C. In addition, an experiment was designed and is being performed to verify the predicted CF curves at 950°C. The extrapolation procedure will be applied at lower temperatures to complete the development of the EPP+SMT CF design curves for Alloy 617 in F2023.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Microstructural Characteristics, Mechanical Properties, and Corrosion Behavior of Field-Assisted Diffusion Welded Alloy 617

This study investigated the microstructural characteristics, mechanical response, and corrosion resistance of Alloy 617 diffusion welded by electric field-assisted sintering (EFAS). The microstructure exhibited good grain boundary (GB) migration across the interface and flat interface in the samples bonded at different conditions. Graded grain size was observed, with grain coarsening being more significant near sample surface due to EFAS-induced temperature gradient. Tensile testing revealed that the specimens fractured in the matrix away from the interface. The peak strength of 807 MPa was obtained in the samples welded at 1000°C. The materials obtained at 1100°C exhibited reduced tensile strength but improved ductility. Strain maps revealed by digital image correlation showed the flat interfaces were less ductile compared to the matrix. The hardness reduced near sample surfaces due to enlarged grains induced by temperature gradient. The interfaces with GB migration exhibited improved corrosion resistance in molten salt compared to the flat interfaces.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Sectioning of four Alloy 617 sheets

This is the Statement of Work (SOW) file. The file is used for subcontractor to section four Alloy 617 sheets according to the design specified in this SOW.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Design and Scoping Tests on Alloy 617 Using Notched Specimen Geometry to Validate Methods for Multiaxial Stress Relaxation

In FY 2022, a development effort was initiated at the US Department of Energy’s Oak Ridge National Laboratory (ORNL) to examine the multiaxial stress-relaxation behavior and multiaxial stress-state effect on the creep-fatigue (CF) performance for Alloy 617 at elevated temperatures. This effort supported the development of the design rules in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 5. In this work, two types of the notch specimen geometries were designed. An inelastic constitutive model was used to investigate the stress triaxiality and elastic follow-up effects on the notch specimens under CF deformation in the finite element (FE) simulations. The numerical study demonstrated that the stress triaxiality and elastic follow-up caused by the notches both play significant roles in the stress-relaxation behavior. In addition, Alloy 617 CF experiments were designed, and the testing on the notch specimens is ongoing. CF test failure data will be generated on the specimens with notches under various conditions in FY 2023, and the results will be used to validate methods for accounting for the multiaxial stress relaxation effect in the design code.

36 MATERIALS SCIENCE↗
DOE OSTI2022

GCR: Alloy 617 Crack Growth Testing Status

Environment can have an appreciable effect on crack growth. This presentation provides updates on environmental crack growth testing of Alloy 617 at INL.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Reference constitutive model for Alloy 617 and 316H stainless steel for use with the ASME Division 5 design by inelastic analysis rules

This report describes the development, calibration, and validation of an inelastic constitutive model describing the monotonic, cyclic, and creep behavior of Alloy 617 from room temperature to 983 °C . The model is intended for incorporation into a new Nonmandatory Appendix to Section III, Division 5, Subsection HB, Subpart B of the ASME Boiler & Pressure Vessel Code providing guidance on developing models for use with the ASME design by inelastic analysis provisions as well as a listing of acceptable material models for each Class A material that designers can use without further validation. The report described the development of the A617 model as well as providing an updated draft appendix, ready for ASME ballot, incorporating the A617 and and the previously-developed 316H model. The development of this appendix and the related material models will promote the widespread application of the generally more efficient inelastic analysis rules rules.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗