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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Investigating the Theranostic Potential of Elementally Matched [ 43 Sc]Sc-PSMA-617 and [ 47 Sc]Sc-PSMA-617

The theranostic approach, which employs diagnostic radiopharmaceuticals to select patients who would benefit from targeted radiotherapy agents, has become an invaluable strategy for effective medical care. Scandium radionuclides offer the advantage of forming elementally matched and chemically identical diagnostic and therapeutic compounds, making them ideal candidates for this strategy. PSMA-617 is an established prostate-specific membrane antigen targeting agent and can be used as a proof of concept to investigate 43 Sc, the diagnostic nuclide, and 47 Sc, the therapeutic nuclide, as a theranostic pair. Methods: Cellular uptake, competitive binding assays, and internalization studies were carried out using LNCaP or PC-3 cell lines. [ 43 Sc]Sc-PSMA-617 was used in PET imaging studies in LNCaP or PC-3 tumor models, with time points ranging from 1–9 h. LNCaP tumor-bearing mice injected with [ 47 Sc]Sc-PSMA-617 were imaged using SPECT up to 48 h. A longitudinal study was carried out using LNCaP tumor-bearing mice imaged with [ 43 Sc]Sc-PSMA-617 prior to receiving a therapeutic dose of [ 47 Sc]Sc-PSMA-617. Results: 43 Sc and 47 Sc were incorporated into PSMA-617 at radiochemical yields of >99%. Cellular uptake studies demonstrated high uptake and specificity to PSMA receptors for [ 47 Sc]Sc-PSMA-617. In vivo PET studies showed specificity of [ 43 Sc]Sc-PSMA-617 while SPECT studies demonstrated tumor retention of [ 47 Sc]Sc-PSMA-617 up to 48 h. [ 47 Sc]Sc-PSMA-617 demonstrated therapeutic efficacy by delaying tumor growth and increasing survival rates from a single administered dose in xenograft models. More importantly, the PET results from [ 43 Sc]Sc-PSMA-617 PET were highly correlated with the therapeutic response from [ 47 Sc]Sc-PSMA-617, showing that 43 Sc PET data can predict therapeutic outcomes in individual animals from 47 Sc agents, even in animals sharing a genetic background and implanted with tumors from the same cell line. Conclusions: Two chemically identical, PSMA-targeting radioscandium pharmaceuticals demonstrated in vivo stability, specificity and retention in PSMA+ tumor models. A theranostic study showed that a higher 43 Sc PET SUVmean was strongly correlated to therapeutic response from the 47 Sc agent, demonstrating that 43 Sc and 47 Sc can be used as an elementally matched theranostic pair.

Biodistribution↗
DOE OSTI2016

Investigation of thermal aging effects on the tensile properties of Alloy 617 by in-situ synchrotron wide-angle X-ray scattering

The nickel-base Alloy 617 has been considered as the lead candidate structural material for the intermediate heat exchanger (IHX) of the Very-High-Temperature Reactor (VHTR). In order to assess the long-term performance of Alloy 617, thermal aging experiments up to 10,000 h in duration were performed at 1000 degrees C. Subsequently, in-situ synchrotron wide-angle X-ray scattering (WAXS) tensile tests were carried out at ambient temperature. M23C6 carbides were identified as the primary precipitates, while a smaller amount of M6C was also observed. The aging effects were quantified in several aspects: (1) macroscopic tensile properties, (2) volume fraction of the M23C6 Phase, (3) the lattice strain evolution of both the matrix and the M23C6 precipitates, and (4) the dislocation density evolution during plastic deformation. The property-microstructure relationship is described with a focus on the evolution of the M23C6 phase. For aging up to 3000 h, the yield strength (YS) and ultimate tensile strength (UTS) showed little variation, with average values being 454 MPa and 787 MPa, respectively. At 10,000 h, the YS and UTS reduced to 380 MPa and 720 MPa, respectively. The reduction in YS and UTS is mainly due to the coarsening of the M23C6 Precipitates. After long term aging, the volume fraction of the M23C6 phase reached a plateau and its maximum internal stress was reduced, implying that under large internal stresses the carbides were more susceptible to fracture or decohesion from the matrix. Finally, the calculated dislocation densities were in good agreement with transmission electron microscopy (TEM) measurements. The square roots of the dislocation densities and the true stresses displayed typical linear behavior and no significant change was observed in the alloys in different aging conditions.

Alloy 617↗
DOE OSTI2018

Formation of Lamellar Carbides in Alloy 617-HAZ and Their Role in the Impact Toughness of Alloy 617/9%Cr Dissimilar Welded Joint

The relationship between the microstructure and impact toughness, especially the formation of the lamellar carbides and its effect on the impact toughness of the heat-affected zone of Alloy 617 (Alloy 617-HAZ), was systematically investigated. Impact test was performed on Alloy 617/9%Cr dissimilar welded joints aged for different time at 620 °C. The results showed that the characteristic zones of the whole welded joint revealed decreased impact toughness with increasing the aging time. The impact energies in weld metal showed high stability despite the aging process, while the Alloy 617-HAZ for all testing specimens presented the lowest impact toughness with various aging time. Meanwhile, the mixed grains were found in the base metal (BM) and heat-affected zone of Alloy 617, together with the twins with different widths located in the interior of the large grains. The columnar and equiaxed grain zones appear alternately due to the multilayer and multi-pass welding technique. In particular, compared with Alloy 617-BM, coarsened and even lamellar carbides were observed in Alloy 617-HAZ. These special carbides can separate the grains and therefore result in the formation of intergranular fracture morphology in Alloy 617-HAZ. Besides, the microhardness distribution was tested with a proper load, and the results exhibited that the weld possessed the highest hardness for all specimens with various aging time. The higher microhardness in Alloy 617-HAZ compared with that in Alloy 617-BM was attributed to the high density of the dislocations, which could be detected by transmission electron microscope. As a result, the coarsened carbides aggregated in the grain boundaries to form the lamellar layer are deemed as the main factor to reduce the impact toughness in Alloy 617-HAZ.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Thermophysical Properties of Alloy 617 from 25°C to 1000°C

Key thermophysical properties needed for the successful design and use of Alloy 617 in steam generator and heat exchanger applications have been measured experimentally, and results are compared with literature values and results obtained from some other commercial Ni–Cr alloys and model materials. Specifically, the thermal diffusivity, thermal expansion coefficient, and specific heat capacity have been measured for Alloy 617 over a range of temperatures, allowing calculation of thermal conductivity up to 1000 degrees C. It has been found that the thermal conductivity of Alloy 617 exhibits significant deviation from monotonic behavior in the temperature range from 600 degrees C to 850 degrees C, the temperatures of interest for most heat transfer applications. The non-linear behavior appears to result primarily from short-range order/disorder phenomena known to occur in the Ni–Cr system. Similar deviation from monotonic behavior was observed in the solid solution Ni–Cr-W Alloy 230, and lesser deviations were observed in iron based Alloy 800H and an austenitic stainless steel. Measured thermophysical property data are provided for four different heats of Alloy 617, and it is shown that property variations between the four different heats are not significant. Measurements were also obtained from Alloy 617 that was aged for up to 2000 h at 750 degrees C, and it was found that this aging treatment does not significantly influence the thermophysical properties.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Creep and Creep-Fatigue of Alloy 617 Weldments

Alloy 617 is the primary candidate material for the heat exchanger of a very high temperature gas cooled reactor intended to operate up to 950°C. While this alloy is currently qualified in the ASME Boiler and Pressure Vessel Code for non-nuclear construction, it is not currently allowed for use in nuclear designs. A draft Code Case to qualify Alloy 617 for nuclear pressure boundary applications was submitted in 1992, but was withdrawn prior to approval. Prior to withdrawal of the draft, comments were received indicating that there was insufficient knowledge of the creep and creep-fatigue behavior of Alloy 617 welds. In this report the results of recent experiments and analysis of the creep-rupture behavior of Alloy 617 welds prepared using the gas tungsten arc process with Alloy 617 filler wire. Low cycle fatigue and creep-fatigue properties of weldments are also discussed. The experiments cover a range of temperatures from 750 to 1000°C to support development of a new Code Case to qualify the material for elevated temperature nuclear design. Properties of the welded material are compared to results of extensive characterization of solution annealed plate base metal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Application of Combined Sustained and Cyclic Loading Test Results to Alloy 617 Elevated Temperature Design Criteria

Alloy 617 is a reference structural material for very high temperature components of advanced-gas cooled reactors with outlet temperatures in the range of 900-950°C . In order for designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. A plan has been developed to submit a draft code for Alloy 617 to ASME Section III by 2015. However, the current rules in Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200°F (650°C). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep deformation, which is the basis for the current simplified rules. This temperature, 1200 °F, is well below the temperature range of interest for this material in High Temperature Gas Cooled Reactor (HTGR) applications. The only current alternative is, thus, a full inelastic analysis which requires sophisticated material models which have been formulated but not yet verified. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods and which are expected to be applicable to very high temperatures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Report on FY15 Alloy 617 SMT Creep-Fatigue Test Results

For the temperature range of 900°C − 950°C, Alloy 617 is a candidate IHX structural material for high temperature gas reactors (HTGRs) because of its high temperature creep properties. Also, its superior strength over a broad temperature range also offers advantages for certain component applications. In order for the designers to be able to use Alloy 617 for these high temperature components, Alloy 617 has to be approved for use in Section III (the nuclear section) of the ASME (American Society of Mechanical Engineers) Boiler and Pressure Vessel Code. A plan has been developed to propose a Code Case for use of Alloy 617 at elevated temperature in Section III of the ASME Code by September 2015. There has not been a new high temperature material approved for use in Section III for almost 20 years. The Alloy 617 Code Case effort would lead the way to establish a path for Code qualification of new high temperature materials of interest to other advanced SMRs. Creep-fatigue at elevated temperatures is the most damaging structural failure mode. In the past 40 years significant efforts have been devoted to the elevated temperature Code rule development in Section III, Subsection NH* of the ASME Boiler and Pressure Vessel Code, to ascertain conservative structural designs to prevent creep-fatigue failure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Updated Draft ASME Boiler and Pressure Vessel Code Case for Use of Alloy 617 for Construction of Nuclear Components for Section III Division 5

Early in the United States’ very high temperature reactor program, several candidate nickel alloys were considered for use in construction of the intermediate heat exchanger. Based primarily on technical maturity, a downselection was made to focus on Alloy 617. After this downselection, the primary goal of the research and development program was to develop sufficient information on the high temperature properties of the material to qualify it for construction of high temperature nuclear components in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code. Alloy 617 is a solid solution strengthened material with nominal Ni Cr Co Mo composition, originally developed for aerospace applications such as burner can liners for turbine engines by Huntington Alloys. (Huntington Alloys is now Special Metals Division of Precision Castparts, Inc.) The ASME BPV Code allows use of Alloy 617 for construction of non nuclear pressure vessels, and Alloy 617 is used in fossil fired power plants. Section III, Division 1, Subsection NB of the ASME BPV Code was developed for construction of nuclear components in light water reactors and allows use of ferritic materials up to 700°F and austenitic alloys up to 800°F. Subsection NH of Section III Division 1 was written to allow higher temperature construction with a primary focus on sodium cooled reactors. Recently, a new Division 5 of Section III was published specifically for high temperature reactors (regardless of the primary working fluid) and incorporates both Subsections NB and NH. There are only five alloys currently allowed for use in high temperature nuclear components: 2.25Cr 1 Mo and V modified 9Cr 1Mo steels, Types 304 and 316 stainless steels, and the high nickel austenitic Alloy 800H. This very sparse set of allowed materials is in contrast with the collection of more than 150 materials that are allowed for use in non nuclear pressure vessel construction. A draft Code Case to add Alloy 617 to the list of qualified alloys for use in high temperature nuclear design was submitted to ASME in the early 1990s, but it was withdrawn prior to formal Code action.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Initial High Temperature Inelastic Constitutive Model for Alloy 617

This report describes progress on a Alloy 617 inelastic constitutive model slated for inclusion in the ASME Boiler & Pressure Vessel Code in Code Case N-898 “Use of Alloy 617 (UNS N06617) for Class A Elevated Temperature Construction, Section III, Division 5” as part of an appendix on inelastic modeling. This appendix will provide a reference material model for A617 for use in conjunction with the ASME design by inelastic analysis criteria. This report focuses on the implementation and evaluation of an existing Alloy 617 model developed to represent the data collected at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) as part of the Alloy 617 Code Case experimental campaign. While the model accurately captures the cyclic stress/strain behavior of Alloy 617, it fails to capture the long-term creep and lower temperature tensile data. In addition, the model form is complicated and has over 50 configurable parameters. As such, we recommend the development of a simplified model, incorporating key features of the ORNL model but reducing the model complexity.

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 OSTI2013

Low Cycle Fatigue and Creep-Fatigue Behavior of Alloy 617 at High Temperature

Alloy 617 is the leading candidate material for an intermediate heat exchanger (IHX) application of the Very High Temperature Nuclear Reactor (VHTR), expected to have an outlet temperature as high as 950 degrees C. Acceptance of Alloy 617 in Section III of the ASME Code for nuclear construction requires a detailed understanding of the creep-fatigue behavior. Initial creep-fatigue work on Alloy 617 suggests a more dominant role of environment with increasing temperature and/or hold times evidenced through changes in creep-fatigue crack growth mechanism/s and failure life. Continuous cycle fatigue and creep-fatigue testing of Alloy 617 was conducted at 950 degrees C and 0.3% and 0.6% total strain in air to simulate damage modes expected in a VHTR application. Continuous cycle specimens exhibited transgranular cracking. Intergranular cracking was observed in the creep-fatigue specimens, although evidence of grain boundary cavitation was not observed. Despite the absence of grain boundary cavitation to accelerate crack propagation, the addition of a hold time at peak tensile strain was detrimental to cycle life. This suggests that creepfatigue interaction may occur by a different mechanism or that the environment may be partially responsible for accelerating failure.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Creep Rupture of Pressurized Alloy 617 Tubes

Creep behavior of Alloy 617 under multi-axial loading has been characterized at 950°C using thin walled tubes with internal gas pressure. Tubes were machined from Alloy 617 plate material to ensure comparability of material properties with extensive testing carried out in the VHTR Research and Development program. Properties of this alloy under multi-axial loading are of interest for elevated temperature design in advanced nuclear heat transfer systems. A constitutive model in the form of a Norton relationship was developed for Alloy 617 plate to allow development of a suitable creep specimen using finite element simulation. It was determined that proper description of the minimum creep rate at 750°C required a threshold stress formalism. The magnitude of the threshold stress was determined to be 75MPa for creep tests using the conditions examined. This threshold stress arises from strengthening by the γ’(Ni 3 Al,Ti) intermetallic phase; above this temperature this phase is unstable and Alloy 617 behaves as a solid solution.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

The Effect of Cold Work on Properties of Alloy 617

Alloy 617 is approved for non-nuclear construction in the ASME Boiler and Pressure Vessel Code Section I and Section VIII, but is not currently qualified for nuclear use in ASME Code Section III. A draft Code Case was submitted in 1992 to qualify the alloy for nuclear service but efforts were stopped before the approval process was completed.1 Renewed interest in high temperature nuclear reactors has resulted in a new effort to qualify Alloy 617 for use in nuclear pressure vessels. The mechanical and physical properties of Alloy 617 were extensively characterized for the VHTR programs in the 1980’s and incorporated into the 1992 draft Code Case. Recently, the properties of modern heats of the alloy that incorporate an additional processing step, electro-slag re-melting, have been characterized both to confirm that the properties of contemporary material are consistent with those in the historical record and to increase the available database. A number of potential issues that were identified as requiring further consideration prior to the withdrawal of the 1992 Code Case are also being re-examined in the current R&D program. Code Cases are again being developed to allow use of Alloy 617 for nuclear design within the rules of the ASME Boiler and Pressure Vessel Code. In general the Code defines two temperature ranges for nuclear design with austenitic and nickel based alloys. Below 427°C (800°F) time dependent behavior is not considered, while above this temperature creep and creep-fatigue are considered to be the dominant life-limiting deformation modes. There is a corresponding differentiation in the treatment of the potential for effects associated with cold work. Below 427°C the principal issue is the relationship between the level of cold work and the propensity for stress corrosion cracking and above that temperature the primary concern is the impact of cold work on creep-rupture behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Creep-Fatigue Behavior of Alloy 617 at 850°C

Creep-fatigue deformation is expected to be a significant contributor to the potential factors that limit the useful life of the Intermediate Heat Exchanger (IHX) in the Very High Temperature Reactor (VHTR) nuclear system. The IHX of a high temperature gas reactor will be subjected to a limited number of transient cycles due to start-up and shut-down operations imparting high local stresses on the component. This cycling introduces a creep-fatigue type of interaction as dwell times occur intermittently. The leading candidate alloy for the IHX is a nickel-base solid solution strengthened alloy, Alloy 617, which must safely operate near the expected reactor outlet temperature of up to 950 °C. This solid solution strengthened nickel-base alloy provides an interesting creep-fatigue deformation case study because it has characteristics of two different alloy systems for which the cyclic behavior has been extensively investigated. Compositionally, it resembles nickel-base superalloys, such as Waspalloy, IN100, and IN718, with the exception of its lower levels of Al. At temperatures above 800 °C, the microstructure of Alloy 617, however, does not contain the ordered γ’ or γ’’ phases. Thus microstructurally, it is more similar to an austenitic stainless steel, such as 316 or 304, or Alloy 800H comprised of a predominantly solid solution strengthened matrix phase with a dispersion of inter- and intragranular carbides. Previous studies of the creep-fatigue behavior of Alloy 617 at 950 °C indicate that the fatigue life is reduced when a constant strain dwell is added at peak tensile strain. This results from the combination of faster crack initiation occurring at surface-connected grain boundaries due to oxidation from the air environment along with faster, and intergranular, crack propagation resulting from the linking of extensive interior grain boundary cracking. Saturation, defined as the point at which further increases in the strain-controlled hold time duration no longer decreases the cycle life, has been observed for Alloy 617 at 950 °C at least to the investigated hold times, as illustrated through a plot of cycles to failure v. hold time in Figure 1. The 950 °C creep-fatigue data set generated by Totemeier and Tian at the 0.3% and 1.0% strain range is consistent in magnitude in terms of the cycles to failure data of that of Carroll et al., however, 0.3% strain range data did not exhibit saturation at hold times of up to 10 min. At 1.0% total strain, saturation in the number of cycles to failure was observed within the investigated peak tensile hold times of up to 10 min. The data of Carroll et al. in Figure 1 and Totemeier and Tian is also consistent in magnitude with the data of Rao and coworkers investigated at the 0.6% strain range. It should be noted that saturation in the number of cycles to failure is not present in the data published by Rao and coworkers for tensile hold times of up to 120 min. The latter testing was in a simulated primary-circuit helium gas as opposed to air and a single data point is reported for the longer hold time conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fatigue and Creep Crack Propagation behaviour of Alloy 617 in the Annealed and Aged Conditions

The crack propagation behaviour of Alloy 617 was studied under various conditions. Elevated temperature fatigue and creep-fatigue crack growth experiments were conducted at 650 and 800 degrees C under constant stress intensity (triangle K) conditions and triangular or trapezoidal waveforms at various frequencies on as-received, aged, and carburized material. Environmental conditions included both laboratory air and characteristic VHTR impure helium. As-received Alloy 617 displayed an increase in the crack growth rate (da/dN) as the frequency was decreased in air which indicated a time-dependent contribution component in fatigue crack propagation. Material aged at 650°C did not display any influence on the fatigue crack growth rates nor the increasing trend of crack growth rate with decreasing frequency even though significant microstructural evolution, including y’ (Ni3Al) after short times, occurred during aging. In contrast, carburized Alloy 617 showed an increase in crack growth rates at all frequencies tested compared to the material in the standard annealed condition. Crack growth studies under quasi-constant K (i.e. creep) conditions were also completed at 650 degrees C and a stress intensity of K = 40 MPa9 (square root)m. The results indicate that crack growth is primarily intergranular and increased creep crack growth rates exist in the impure helium environment when compared to the results in laboratory air. Furthermore, the propagation rates (da/dt) continually increased for the duration of the creep crack growth either due to material aging or evolution of a crack tip creep zone. Finally, fatigue crack propagation tests at 800 degrees C on annealed Alloy 617 indicated that crack propagation rates were higher in air than impure helium at the largest frequencies and lowest stress intensities. The rates in helium, however, eventually surpass the rates in air as the frequency is reduced and the stress intensity is decreased which was not observed at 650 degrees C.

36 MATERIALS SCIENCE↗