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DOE OSTI2025

Synthesis of DOTA-Based 43 Sc Radiopharmaceuticals Using Cyclotron-Produced 43 Sc as Exemplified by [ 43 Sc]Sc-PSMA-617 for PSMA PET Imaging

The implementation of theranostics in oncologic nuclear medicine has exhibited immense potential in improving patient outcomes in prostate cancer with the implementation of [ 68 Ga]Ga-PSMA-11 PET and [ 177 Lu]Lu-PSMA-617 into clinical practice. However, the correlation between radiopharmaceutical biodistributions seen with [ 68 Ga]Ga-PSMA-11 PET imaging and downstream [ 177 Lu]Lu-PSMA-617 therapy remains imperfect. This suggests that prostate cancer theranostics could potentially be further refined through the implementation of true theranostics, tandem pairs of diagnostic and therapeutic radiopharmaceuticals that utilize the same ligand and element, thus yielding identical pharmacokinetics. The radioscandiums are one such group of true theranostic radiopharmaceuticals. The radioscandiums consist of two β+ emitting scandium isotopes ( 43 Sc/ 44 Sc), as well as a β − emitting therapeutic isotope ( 47 Sc), which can all conjugate with PSMA-targeting PSMA-617. This potential has led to extensive investigations into the production of the radioscandiums as well as pre-clinical assessments with several ligands; however, there is a lack of literature extensively describing the complete synthesis of scandium radiopharmaceuticals. which therefore limits the accessibility of radioscandium research in theranostics. As such, this work aims to present an easily translatable protocol for the synthesis of [ 43 Sc]Sc-PSMA-617 from a [ 42 Ca]CaCO 3 starting material, including target formation, nuclear production via 42 Ca(d,n) 43 Sc reaction, chemical separation, radiolabeling, solvent reformulation, and target recycling.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

NASA NTRS1984

Response of Inconel 617 superalloy to combined ground-based and STS reentry exposure

Inconel 617 is a nickel-based superalloy which is being considered for heat-shield applications because of its high-temperature strength, good oxidation resistance and high emittance of oxidized surfaces. While the effects of simulated reentry conditions on emittance and oxidation of Inconel 617 have been studied, the combined effects of the ground-based environment with sea salt exposure and the reentry environment have not been evaluated. Experimental results are presented to show the effects of environmental simulation including ground-based and reentry exposure on the emittance and oxidation of Inconel 617. Specimens were exposed to simulated reentry at a surface temperature of 2000 F in the Langley Research Center Hypersonic Materials Environmental Test System (HYMETS) Facility with and without alternate exposures to an atmospheric seashore environment or a laboratory sea salt environment. This paper presents emittance, mass loss, oxide chemistry, and alloy composition data for the specimens.

Clark, R. K.↗

NASA NTRS1984

Response of Inconel 617 superalloy to combined ground-based and STS reentry exposure

Clark, R. K.↗

DOE OSTI2018

A Study of Tensile Flow and Work-Hardening Behavior of Alloy 617

The simple power relationship σ = Κε{sub p}{sup n} satisfactorily expresses the tensile flow behavior of many metals and alloys in their uniform plastic strain regime. However, many FCC materials with low stacking fault energy have opposed such power law relationship. Alloy 617, an age-hardenable Ni-based superalloy is also observed not to obey the simple power law relationship neither in its solution-treated nor in its aged conditions. Various flow relationships were used to obtain the best fit for the tensile data, and different relationships were identified for the different aged conditions. The work-hardening rate (θ) demonstrates three distinct regions for all aged conditions, and there is an obvious change in the trend of θ versus σ. In the initial portion, θ decreases rapidly followed by a gradual increase in the second stage and again a decrease in its third stage is perceived in the Alloy 617. These three-stage characteristics are attributed to a commonly known precipitate, γ′: Ni{sub 3}(Ti, Al) which evolves during aging treatment and well recognized under transmission electron microscopy (TEM) observation. TEM results also reveal a slight degree of coarsening in γ′ over aging. The tensile flow and the work-hardening behavior are well correlated with other microstructural evolution during the aging treatments.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Microstructural investigation of oxide dispersion strengthened alloy 617 after creep rupture at high temperature

Highlights: • The creep related microstructural evolution of ODS-617 at 950 °C under 25 MPa was investigated. • Dispersion of nanoscale Y-Al-O particles contributed to the superior creep properties of the ODS-617. • The creep mechanism of the ODS-617 at 950 °C under 25 MPa was Coble creep. • Cracks preferentially developed around the large rod-like AlN particles, resulting in the brittle creep fracture mode. - Abstract: Oxide dispersion strengthened (ODS) alloys have been widely considered as candidate structural materials for advanced nuclear reactors because of their good high temperature properties and irradiation resistance. In this study, the creep properties of oxide dispersion strengthened Alloy 617 (ODS-617) was examined at 950 °C under 25 MPa in the air. Conventional Alloy 617 tested under the same conditions was included for comparison. The ODS-617 exhibited superior creep resistance, including a much longer rupture life of 8533 h and a steady-state creep rate of 8.9 × 10{sup −8} (1/s), which was two orders of magnitude lower than that of the conventional alloy. The lower steady-state creep rate was related to the dispersion of nano-sized Y-Al-O particles, which may impede grain boundary diffusion. It seems that cracks preferentially developed near some large rod-like AlN particles, resulting in the brittle creep fracture mode.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Elevated temperature mechanical properties of Inconel 617 surface oxide using nanoindentation

Inconel 617 is a principal candidate material for helium gas cooled very-high-temperature reactors with outlet temperatures of 700–950 °C. Recent findings showed that this alloy develops unique surface oxide layers especially at high temperature (HT) helium environment with distinctive wear, friction and contact properties. This study investigates the elevated temperature mechanical properties of Inconel 617 top surface layers aged in HT helium environment. Nanoindentation technique is used to obtain load-displacement graphs of the alloy top surface oxide in temperatures ranging from 25 °C up to 600 °C. In addition, using finite element analysis along with an iterative regression technique, a semi-numerical method is developed to further measure and quantify the material parameters and, in particular, time-independent and creep characteristics of the oxide. Although Young's modulus of the oxide is found to be relatively close to the bulk for the tested temperatures, the yield strength and hardness, in comparison to the bulk material, increase significantly as the material is oxidized after aging. The oxide exhibits significant softening as the temperature increases to 600 °C. Unlike the bulk material, diffusion through the grains is found to be the dominant creep mechanism for the oxide. Considerable difference between the mechanical properties of the oxide and the bulk material shows the need for accurate measurements of near surface mechanical properties, if reliable predictive contact and tribological models are sought at elevated temperatures.

36 MATERIALS SCIENCE↗

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 OSTI2020

Self-welding of Inconel 617 under high-pressure-high-temperature conditions for nuclear reactors

In response to continuously declining availability of fossil fuels, nuclear power has become an important alternative source of power generation. Growing application of nuclear power has relieved the shortage of electricity. However, it is also accompanied with new drawbacks, such as radioactive waste, and potential catastrophic consequences in case of major accidents. For nuclear power plants, severe accidents mostly originate from the reactor, making the reliability of the reactor of paramount importance to the overall safety of a nuclear power plant. Massive heat generation through nuclear fission/fusion and heat exchange render the whole reactor operating at extremely high temperatures (i.e., 900–1000 °C), which, in conjunction with large load from subsystems/ components made of special heavy alloys, renders the reactor under the risk of self-welding, a phenomenon that could lead to strong bonding between components and paralyze the operability of the nuclear reactor. This paper investigates the self-welding behavior of Inconel 617, a primary candidate nickel alloy for key components of the next generation gas cooled nuclear plant, inside a controlled atmospheric furnace with preset static load, simulating the high-pressure high-temperature conditions inside nuclear reactors. Therefore, test results show that Inconel 617 experiences self-welding through common oxide zone at the interface, and it has remarkable bonding strength, requiring 4114 N (57.15 MPa) and 4338 N (60.25 MPa) to break apart a mated sample pair, under an apparent contact pressure of 0.341 MPa after 50-h aging, in helium and air atmospheres, respectively. The higher oxygen presence in air results in slightly stronger oxide bond at the interface, compared to helium, and the bond strength increases with dwell time.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2014

Nonlinear acoustics experimental characterization of microstructure evolution in Inconel 617

Inconel 617 is a candidate material for the intermediate heat exchanger in a very high temperature reactor for the next generation nuclear power plant. This application will require the material to withstand fatigue-ratcheting interaction at temperatures up to 950°C. Therefore nondestructive evaluation and structural health monitoring are important capabilities. Acoustic nonlinearity (which is quantified in terms of a material parameter, the acoustic nonlinearity parameter, β) has been proven to be sensitive to microstructural changes in material. This research develops a robust experimental procedure to track the evolution of damage precursors in laboratory tested Inconel 617 specimens using ultrasonic bulk waves. The results from the acoustic non-linear tests are compared with stereoscope surface damage results. Therefore, the relationship between acoustic nonlinearity and microstructural evaluation can be clearly demonstrated for the specimens tested.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2016

Biaxial Thermal Creep of Alloy 617 and Alloy 230 for VHTR Applications

In this study, we employed pressurized creep tubes to investigate the biaxial thermal creep behavior of Inconel 617 (alloy 617) and Haynes 230 (alloy 230). Both alloys are considered to he the primary candidate structural materials for very high-temperature reactors (VITITRs) due to their exceptional high-temperature mechanical properties. The current creep experiments were conducted at 900 degrees C for the effective stress range of 15-35 MPa. For both alloys, complete creep strain development with primary, secondary, and tertiary regimes was observed in all the studied conditions. Tertiary creep was found to he dominant over the entire creep lives of both alloys. With increasing applied creep stress, the fraction of the secondary creep regime decreases. The nucleation, diffusion, and coarsening of creep voids and carbides on grain boundaries were found to be the main reasons for the limited secondary regime and were also found to be the major causes of creep fracture. The creep curves computed using the adjusted creep equation of the form epsilon= cosh 1(1 rt) + P-sigma ntm agree well with the experimental results for both alloys at die temperatures of 850-950 degrees C.

36 MATERIALS SCIENCE↗

DOE OSTI2016

DETERMINATION OF THE CREEP–FATIGUE INTERACTION DIAGRAM FOR ALLOY 617

Alloy 617 is the leading candidate material for an intermediate heat exchanger for the very high temperature reactor. To evaluate the behavior of this material in the expected service conditions, creep-fatigue testing was performed. Testing has been performed primarily on a single heat of material at 850 and 950°C for total strain ranges of 0.3 to 1% and tensile hold times as long as 240 minutes. At 850°C, increases in the tensile hold duration degraded the creep fatigue resistance, at least to the investigated strain-controlled hold time of up to 60 minutes at the 0.3% strain range and 240 minutes at the 1.0% strain range. At 950°C, the creep-fatigue cycles to failure becomes constant with increasing hold times, indicating saturation occurs at relatively short hold times. The creep and fatigue damage fractions have been calculated and plotted on a creep-fatigue interaction D-diagram. Results from earlier creep-fatigue tests at 800 and 1000°C on an additional heat of Alloy 617 are also plotted on the D-diagram. The methodology for calculating the damage fractions will be presented, and the effects of strain rate, strain range, temperature, hold time, and strain profile (i.e. holds in tension, compression or both) on the creep-fatigue damage will be explored.

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 OSTI2013

Tensile Properties of Alloy 617 Bar Stock

Alloy 617 is approved for non-nuclear construction in the ASME Boiler and Pressure Vessel Code Section I and Section VIII, Division 1 but is not currently qualified for nuclear use in ASME Code Section III. A draft Code Case was developed in the 1980’s to qualify the alloy for nuclear service but efforts were stopped before the approval process was completed. Renewed interest in high temperature nuclear reactors has resulted in a new effort to qualify Alloy 617 for use in nuclear pressure vessels. Tensile properties of the alloy are needed for several aspects of Code qualification. Minimum yield strength and average tensile strength at temperature are used to set time independent allowable stress, Sm, for structural materials in B&PV Code, Section III, Subsection NH. It must also be demonstrated that there is no minimum ductility between room temperatures and 427°C (the upper temperature limit for Subsection NB) to minimize the number of fatigue tests required for developing the fatigue design curve. The bar properties reported here represent data for an additional heat and product form. A forged and rolled plate, designated as a reference material for the program, has been studied extensively and tensile properties are presented for comparison. Both of the product forms that have been characterized in the current program, plate and bar, were produced using current mill practice that includes additional refining using electro-slag re-melting. This additional processing step was not used for some of the materials in the historical database and it is of interest to demonstrate that the properties of contemporary material are comparable to the historical values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Progress Report on Alloy 617 Isochronous Stress-Strain Curves

Isochronous stress-strain curves for Alloy 617 up to a temperature of 1000°C will be required to qualify the material for elevated temperature design in Section III, Division 1, Subsection NH of the ASME Boiler and Pressure Vessel Code. Several potential methods for developing these curves are reviewed in this report. It is shown that in general power-law creep is the rate controlling deformation mechanism for a wide range of alloy heats, test temperatures and stresses. Measurement of the strain rate sensitivity of Alloy 617 indicates that the material is highly strain rate sensitive in the tensile deformation range above about 750°C. This suggests that the concept of a hot tensile curve as a bounding case on the isochronous stress-strain diagrams is problematic. The impact of strain rate on the hot tensile curves is examined and it is concluded that incorporating such a curve is only meaningful if a single tensile strain rate (typically the ASTM standard rate of 0.5%/min) is arbitrarily defined. Current experimentally determined creep data are compared to isochronous stress-strain curves proposed previously by the German programs in the 1980s and by the 1990 draft ASME Code Case. Variability in how well the experimental data are represented by the proposed design curves that suggests further analysis is necessary prior to completing a new draft Code Case.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Monitoring microstructural evolution of alloy 617 with non-linear acoustics for remaining useful life prediction; multiaxial creep-fatigue and creep-ratcheting

The research built upon a prior investigation to develop a unified constitutive model for design-by-analysis of the intermediate heat exchanger (IHX) for a very high temperature reactor (VHTR) design of next generation nuclear plants (NGNPs). Model development requires a set of failure data from complex mechanical experiments to characterize the material behavior. Therefore uniaxial and multiaxial creep-fatigue and creep-ratcheting tests were conducted on the nickel-base Alloy 617 at 850 and 950°C. The time dependence of material behavior, and the interaction of time dependent behavior (e.g., creep) with ratcheting, which is an increase in the cyclic mean strain under load-controlled cycling, are major concerns for NGNP design. This research project aimed at characterizing the microstructure evolution mechanisms activated in Alloy 617 by mechanical loading and dwell times at elevated temperature. The acoustic harmonic generation method was researched for microstructural characterization. It is a nonlinear acoustics method with excellent potential for nondestructive evaluation, and even online continuous monitoring once high temperature sensors become available. It is unique because it has the ability to quantitatively characterize microstructural features well before macroscale defects (e.g., cracks) form. The nonlinear acoustics beta parameter was shown to correlate with microstructural evolution using a systematic approach to handle the complexity of multiaxial creep-fatigue and creep-ratcheting deformation. Mechanical testing was conducted to provide a full spectrum of data for: thermal aging, tensile creep, uniaxial fatigue, uniaxial creep-fatigue, uniaxial creep-ratcheting, multiaxial creep-fatigue, and multiaxial creep-ratcheting. Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), and Optical Microscopy were conducted to correlate the beta parameter with individual microstructure mechanisms. We researched application of the harmonic generation method to tubular mechanical test specimens and pipes for nondestructive evaluation. Tubular specimens and pipes act as waveguides, thus we applied the acoustic harmonic generation method to guided waves in both plates and shells. Magnetostrictive transducers were used to generate and receive guided wave modes in the shell sample and the received signals were processed to show the sensitivity of higher harmonic generation to microstructure evolution. Modeling was initiated to correlate higher harmonic generation with the microstructure that will lead to development of a life prediction model that is informed by the nonlinear acoustics measurements.

36 MATERIALS SCIENCE↗

DOE OSTI2015

Alloy 617 Isochronous Stress-Strain Curves

The development of isochronous stress-strain curves is an important piece of the code case for Alloy 617. A 2-parameter creep equation was used to model creep based on test results at INL. The Ramberg-Osgood and Voce equations were used to generate hot tensile curves based on INL test results at different temperatures. Both the isochronous and hot tensile curves were then shifted to represent the average heat of Alloy 617.

36 MATERIALS SCIENCE↗

DOE OSTI2014

Prediction of the Creep-Fatigue Lifetime of Alloy 617: An Application of Non-destructive Evaluation and Information Integration

A relatively simple method using the nominal constant average stress information and the creep rupture model is developed to predict the creep-fatigue lifetime of Alloy 617, in terms of time to rupture. The nominal constant average stress is computed using the stress relaxation curve. The predicted time to rupture can be converted to number of cycles to failure using the strain range, the strain rate during each cycle, and the hold time information. The predicted creep-fatigue lifetime is validated against the experimental measurements of the creep-fatigue lifetime collected using conventional laboratory creep-fatigue tests. High temperature creep-fatigue tests of Alloy 617 were conducted in air at 950°C with a tensile hold period of up to 1800s in a cycle at total strain ranges of 0.3% and 0.6%. It was observed that the proposed method is conservative in that the predicted lifetime is less than the experimentally determined values. The approach would be relevant to calculate the remaining useful life to a component like a steam generator that might fail by the creep-fatigue mechanism.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Creep-Fatigue Life Prediction of Alloy 617 at 950 deg. C using an Effective Hold Time Approach - 18511

Time Fraction Rule for calculating creep damage fraction is suitable for predicting creep-fatigue life of Alloy 617 at 950 deg. C. However, purely strain-controlled loading waveforms cannot produce creep-dominant interaction due to rapid stress-relaxation of Alloy 617. This study employs a novel loading profile with a combination of strain and force-control to produce varying proportions of creep and fatigue damage. The Time Fraction Rule is found to be inaccurate for life prediction for this novel loading waveform. A new method for calculating the creep damage fraction is proposed with the objective of predicting creep-fatigue life for traditional purely strain-controlled and novel hybrid control loading waveforms. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗