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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Building blocks for a digital twin of additive manufacturing

Properties and serviceability of additively manufactured components are affected by their geometry, microstructure and defects. These important attributes are now optimized by trial and error because the essential process variables cannot currently be selected from scientific principles. A recourse is to build and rigorously validate a digital twin of the additive manufacturing process that can provide accurate predictions of the spatial and temporal variations of metallurgical parameters that affect the structure and properties of components. Key building blocks of a computationally efficient first-generation digital twin of laser-based directed energy deposition additive manufacturing utilize a transient, three-dimensional model that calculates temperature and velocity fields, cooling rates, solidification parameters and deposit geometry. The measured profiles of stainless steel 316L and Alloy 800H deposits as well as the secondary dendrite arm spacing (SDAS) and Vickers hardness measurements are used to validate the proposed digital twin. The predicted cooling rates, temperature gradients, solidification rates, SDAS and micro-hardness values are shown to be more accurate than those obtained from a commonly used heat conduction calculation. Finally, these metallurgical building blocks serve as a phenomenological framework for the development of a digital twin that will make the expanding knowledge base of additive manufacturing usable in a practical way for all scientists and engineers.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Thermomechanical treatment for improved neutron irradiation resistance of austenitic alloy (Fe-21Cr-32Ni)

An optimized thermomechanical treatment (TMT) applied to austenitic alloy 800H (Fe-21Cr-32Ni) had shown significant improvements in corrosion resistance and basic mechanical properties. This study examined its effect on radiation resistance by irradiating both the solution-annealed (SA) and TMT samples at 500 degrees C for 3 dpa. Microstructural characterization using transmission electron microscopy revealed that the radiation-induced Frank loops, voids, and y'-Ni3(Ti,Al) precipitates had similar sizes between the SA and TMT samples. The amounts of radiation-induced defects and more significantly y' precipitates, however, were reduced in the TMT samples. These reductions would approximately reduce by 40.9% the radiation hardening compared to the SA samples. This study indicates that optimized-TMT is an economical approach for effective overall property improvements.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Emulation of neutron-irradiated microstructure of austenitic 21Cr32Ni model alloy using dual-ion irradiation

Here, in this study, the capability of heavy-ion irradiation to emulate neutron irradiation was demonstrated on an austenitic 21Cr32Ni type ternary model alloy. The model alloy used in this study is chemically analogous but compositionally simpler than of alloy 800H, which is a candidate austenitic Ni alloy which has been proposed for use in Generation IV reactors. The microstructure of the 21Cr32Ni model alloy irradiated in the BOR-60 fast reactor to 17.1 dpa and 35 dpa at ~380°C was characterized using transmission electron microscopy (TEM). The 17.1 dpa BOR-60 irradiated microstructure was then compared with the microstructure of the same material developed under dual-ion (DI) irradiation using various He/dpa ratios between 0.1 and 16.6 appm He/dpa in the temperature range of 430°C-500°C. The results showed that both neutron and DI irradiation of 21Cr32Ni model alloy produced dislocations in the form of a dislocation network as well as {111}-type faulted dislocation loops, cavities, and radiation-induced Ni enrichment at radiation-induced sinks. When the dose and the He/dpa ratio were kept similar to those in neutron irradiation, DI irradiation of the 21Cr32Ni model alloy at 460°C resulted in over-nucleation of small cavities and in a high density of faulted dislocation loops compared to those observed in the fast-neutron irradiated alloy of the same heat irradiated at ~380°C. The optimal condition for reproducing the neutron-irradiated microstructure was DI irradiation at 460°C and 0.1 appm He/dpa. In that case, the faulted loop and cavity size distributions in the BOR-60 irradiated 21Cr32Ni model alloy samples closely matched with those measured in the DI irradiated 21Cr32Ni model alloy sample. The fact that the He/dpa is an order of magnitude smaller than the helium generation rate for fast neutron irradiation, stops over nucleation and allows for the development of a similar microstructure as for neutron irradiation.

21Cr32Ni model alloy↗
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 OSTI2025

Damage modeling of power tower receiver tubes using the SRLIFE tool

Concentrating Solar Power (CSP) molten-salt central receivers are subject to high, transient incident flux during daily operation. The resulting creep-fatigue damage impacts the receiver’s reliability and restricts the permissible incident flux distribution for a given receiver. This paper aims to reduce CSP plants’ levelized cost of electricity by developing a methodology to predict lifetime and identifies the primary damage mechanism (creep vs fatigue) for any given fluid temperature and temperature gradient. Results are presented in the form of a damage map that serves as a valuable operation guide and design tool. Damage maps can be used to reduce maintenance costs by improving reliability and reduce receiver capital costs by better utilizing the receiver area. FEA simulation and damage modeling of tubes subject to asymmetrical flux conditions is performed in the open-source receiver design tool srlife. Parametric studies are performed over a range of inner tube temperatures and thermal gradients for A230, 316H, 740H, A282, A617, and 800H high temperature alloys. Damage maps are presented for each alloy. A parametric, FEA-based methodology is presented for comparison of fatigue-creep ratios and prediction of tube lifetime based on the critical thermal operating conditions. Fatigue is found to be negligible compared to creep for almost every case. Here, this finding suggests that fatigue effects associated with cloud events are insignificant compared to creep at these high temperature operating conditions. Additionally, lifetime predictions identify thermal conditions where small changes in operating conditions can result in large changes in predicted lifetime.

14 SOLAR ENERGY↗
DOE OSTI2026

Thermomechanical analysis and modeling of a high-temperature light trapping planar cavity solar receiver

The development of durable particle-based high-temperature solar receivers is critical for advancing concentrating solar-thermal (CST) technologies to enable high-efficiency power generation and industrial process heat. Here, this study presents a computational framework to evaluate the thermomechanical performance of a proposed enclosed light-trapping planar cavity receiver designed for particle-based thermal energy systems. The receiver incorporates absorptive cavities and fluidized particle-bed channels to enhance heat capture and reduce thermal losses. Finite element analysis (FEA) is employed to assess stress, strain, and creep-fatigue behavior under concentrated solar flux using realistic thermal boundary conditions derived from coupled system models and experimental assembly parameters. The analysis investigates the influence of particle-to-wall heat transfer coefficients (HTC) ranging from 800 to 1800 W/m 2 .K on the thermomechanical response of six candidate high-temperature alloys: Alloy 740H, Alloy 282, Alloy 617, 316H, Alloy 230, and 800H. Results show that increasing HTC reduces thermal gradients, leading to lower stresses and strains and extended minimum predicted creep life. While all materials satisfy fatigue life requirements under the investigated conditions, significant differences in creep resistance are observed. Alloy 740H consistently exhibits the longest minimum predicted creep life and the most favorable durability margins, followed by Alloy 282, with the remaining materials showing reduced creep resistance under identical loading. The reported creep lives are conservative lower-bound estimates intended for comparative material evaluation. This framework highlights the critical roles of material selection and geometry optimization in improving mechanical durability and reliability of solar-thermal receivers, forming a foundation for future experimental validation and design optimization.

14 SOLAR ENERGY↗
DOE OSTI2016

Corrosion of alloys in a chloride molten salt (NaCl-LiCl) for solar thermal technologies

Next-generation solar power conversion systems in concentrating solar power (CSP) applications require high-temperature advanced fluids in the range of 600–800 °C. Current commercial CSP plants use molten nitrate salt mixtures as the heat transfer fluid and the thermal energy storage (TES) media while operating with multiple hours of energy capacity and at temperatures lower than 565 °C. At higher temperatures, the nitrates cannot be used because they decompose. Molten chloride salts are candidates for CSP applications because of their high decomposition temperatures and good thermal properties; but they can be corrosive to common alloys used in vessels, heat exchangers, and piping at these elevated temperatures. In this article, we present the results of the corrosion evaluations of several alloys in eutectic 34.42 wt% NaCl – 65.58 wt% LiCl at 650–700 °C in nitrogen atmosphere. Electrochemical evaluations were performed using open-circuit potential followed by a potentiodynamic polarization sweep. Corrosion rates were determined using Tafel slopes and Faraday's law. A temperature increase of as little as 50 °C more than doubled the corrosion rate of AISI stainless steel 310 and Incoloy 800H compared to the initial 650 °C test. These alloys exhibited localized corrosion. Inconel 625 was the most corrosion-resistant alloy with a corrosion rate of 2.80±0.38 mm/year. For TES applications, corrosion rates with magnitudes of a few millimeters per year are not acceptable because of economic considerations. Additionally, localized corrosion (intergranular or pitting) can be catastrophic. Furthermore, corrosion-mitigation approaches are required for advanced CSP plants to be commercially viable.

47 OTHER INSTRUMENTATION↗
DOE OSTI2016

Corrosion evaluation of alloys and MCrAlX coatings in molten carbonates for thermal solar applications

Here, stainless steels (SS) 310, 321, 347, Incoloy 800H (In800H), alumina-forming austenitic (AFA-OC6), Ni superalloy Inconel 625 (IN625), and MCrAlX (M: Ni, and/or Co; X: Y, Hf, Si, and/or Ta) coatings were corroded in molten carbonates in N 2 and bone-dry CO 2 atmospheres. Electrochemical tests in molten eutectics K 2 CO 3 -Na 2 CO 3 and Na 2 CO 3 -K 2 CO 3 -Li 2 CO 3 at temperatures higher than 600 °C were evaluated using an open-circuit potential followed by a potentiodynamic polarization sweep to determine the corrosion rates. Because the best-performing alloys at 750 °C were In800H followed by SS310, these two alloys were selected as the substrate material for the MCrAlX coatings. The coatings were able to mitigate corrosion in molten carbonates environments. The corrosion of substrates SS310 and In800H was reduced from ~2500 um/year to 34 um/year when coated with high-velocity oxyfuel (HVOF) NiCoCrAlHfSiY and pre-oxidized (air, 900 °C, 24 h, 0.5 °C/min) before molten carbonate exposure at 700 °C in bone-dry CO 2 atmosphere. Metallographic characterization of the corroded surfaces showed that the formation of a uniform alumina scale during the pre-oxidation seems to protect the alloy from the molten carbonate attack.

14 SOLAR ENERGY↗
DOE OSTI2021

An Analytical Solution for the Initiation and Early Progression of Fretting Wear in Spherical Contacts

Abstract This article derives analytical solutions to calculate the wear volume at the initiation of fretting motion and its early progression over the first few oscillation cycles. The Archard-based model considers a deformable hemisphere that is contact with a deformable flat block. The material pairs investigated are special alloys, the Inconel 617/Incoloy 800H, and Inconel 617/Inconel 617. The analytical study begins with a unidirectional frictional sliding contact, where the local interfacial sliding distance and the nominal sliding distance at the initiation of gross slip are derived. The obtained analytical expressions for unidirectional sliding are then used to derive the corresponding wear volume for the initiation and early progression of gross slip and the wear volume for a general fretting cycle under elastic conditions. These analytical derivations are all verified by the finite element analysis (FEA). The FEA method and the analytical solutions render virtually identical results for both similar and dissimilar material pairs. The effects of plasticity on the wear volume under elastic–plastic conditions are also investigated. It is found that the fretting wear volumes obtained from the FEA simulations, which include plasticity, are close to those obtained from the analytical expressions for purely elastic regimes. All the results are presented in normalized forms, which can be easily generalized and applied to three-dimensional fretting wear of other material pairs.

Engineering↗
DOE OSTI2012

Diffusion Welding of Alloys for Molten Salt Service – Status Report

The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900°C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700°C. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.

02 PETROLEUM↗
DOE OSTI2012

Diffusion Welding of Alloys for Molten Salt Service - Status Report

The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 and 800H. The MSR output temperature is also materials limited, and is projected at about 700 °C. (RR E) A different set of alloys, such as Alloy N and 242, are needed to handle molten salts at this temperature. The diffusion welding development work described here builds on techniques developed during the NGNP work, as applied to these alloys. There is also the matter of dissimilar metal welding, since alloys suitable for salt service are generally not suited for service in gaseous oxidizing environments, and vice versa, and welding is required for the Class I boundaries in these systems, as identified in the relevant ASME codes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Corrosion and Creep of Candidate Alloys in High Temperature Helium and Steam Environments for the NGNP

This project aims to understand the processes by which candidate materials degrade in He and supercritical water/steam environments characteristic of the current NGNP design. We will focus on understanding the roles of temperature, and carbon and oxygen potential in the 750-850 degree C range on both uniform oxidation and selective internal oxidation along grain boundaries in alloys 617 and 800H in supercritical water in the temperature range 500-600 degree C; and examining the application of static and cyclic stresses in combination with impure He environments in the temperature rang 750-850 degree C; and examining the application of static and cyclic stresses in combination with impure He environments in the temperature range 750-850 degree Cover a range of oxygen and carbon potentials in helium. Combined, these studies will elucidate the potential high damage rate processes in environments and alloys relevant to the NGNP.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments

The helium coolant in high-temperature reactors inevitably contains low levels of impurities during steady-state operation, primarily consisting of small amounts of H 2 , H 2 O, CH 4 , CO, CO 2 , and N 2 from a variety of sources in the reactor circuit. These impurities are problematic because they can cause significant long-term corrosion in the structural alloys used in the heat exchangers at elevated temperatures. Currently, the primary candidate materials for intermediate heat exchangers are Alloy 617, Haynes 230, Alloy 800H, and Hastelloy X. This project will evaluate the role of impurities in helium coolant on the stress-assisted grain boundary oxidation and creep crack growth in candidate alloys at elevated temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Corrosion in Supercritical Carbon Dioxide: Materials, Environmental Purity, Surface Treatments, and Flow Issues (Project No. 10-872)

The supercritical CO 2 Brayton cycle is gaining importance for power conversion in the Generation IV fast reactor system because of its high conversion efficiencies. When used in conjunction with a sodium fast reactor, the supercritical CO 2 cycle offers additional safety advantages by eliminating potential sodium-water interactions that may occur in a steam cycle. In power conversion systems for Generation IV fast reactors, supercritical CO 2 temperatures could be in the range of 30°C to 650°C, depending on the specific component in the system. Materials corrosion primarily at high temperatures will be an important issue. Therefore, the corrosion performance limits for materials at various temperatures must be established. The proposed research will have four objectives centered on addressing corrosion issues in a high-temperature supercritical CO 2 environment: Task 1: Evaluation of corrosion performance of candidate alloys in high-purity supercritical CO 2 : The following alloys will be tested: Ferritic-martensitic Steels NF616 and HCM12A, austenitic alloys Incoloy 800H and 347 stainless steel, and two advanced concept alloys, AFA (alumina forming austenitic) steel and MA754. Supercritical CO 2 testing will be performed at 450°C, 550°C, and 650°C at a pressure of 20 MPa, in a test facility that is already in place at the proposing university. High purity CO 2 (99.9998%) will be used for these tests. Task 2: Investigation of the effects of CO, H 2 O, and O 2 impurities in supercritical CO 2 on corrosion: Impurities that will inevitably present in the CO 2 will play a critical role in dictating the extent of corrosion and corrosion mechanisms. These effects must be understood to identify the level of CO 2 chemistry control needed to maintain sufficient levels of purity to manage corrosion. The individual effects of important impurities CO, H 2 O, and O 2 will be investigated by adding them separately to high purity CO 2 . Task 3: Evaluation of surface treatments on the corrosion performance of alloys in supercritical CO 2 : Surface treatments can be very beneficial in improving corrosion resistance. Shot peening and yttrium and aluminum surface treatments will be investigated. Shot peening refines the surface grain sizes and promotes protective Cr-oxide layer formation. Both yttrium and aluminum form highly stable oxide layers (Y 2 O 3 and Al 2 O 3 ), which can get incorporated in the growing Fe-oxide layer to form an impervious complex oxide to enhance corrosion resistance. Task 4: Study of flow-assisted corrosion of select alloys in supercritical CO 2 under a selected set of test conditions: To study the effects of flow-assisted corrosion, tests will be conducted in a supercritical CO 2 flow loop. An existing facility used for supercritical water flow studies at the proposing university will be modified for use in this task. The system is capable of flow velocities up to 10 m/s and can operate at temperatures and pressures of up to 650°C and 20 MPa, respectively. All above tasks will be performed in conjunction with detailed materials characterization and analysis using scanning electron microscopy/energy dispersive spectroscopy (SEM-EDS), x-ray diffraction (XRD), Auger electron spectroscopy (AES) techniques, and weight change measurements. Inlet and outlet gas compositions will be monitored using gas chromatography-mass spectrometry (GCMS).

36 MATERIALS SCIENCE↗
DOE OSTI2014

Compliant Glass Seals for SOFC Stacks

This report summarizes results from experimental and modeling studies performed by participants in the Solid-State Energy Conversion Alliance (SECA) Core Technology Program, which indicate that compliant glass-based seals offer a number of potential advantages over conventional seals based on de-vitrifying glasses, including reduced stresses during stack operation and thermal cycling, and the ability to heal micro-damage induced during thermal cycling. The properties and composition of glasses developed and/or investigated in these studies are reported, along with results from long-term (up to 5,800h) evaluations of seals based on a compliant glass containing ceramic particles or ceramic fibers.

36 MATERIALS SCIENCE↗
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 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 OSTI2019

Compatibility studies of cladding candidates and advanced low-Cr superalloys in molten NaCl-MgCl 2

To mitigate corrosion of salt-facing components in molten salts, the use of cladding on structural materials is under consideration. A second approach is the use of low-Cr alloys similar to Alloy-N, but with improved properties. To examine the compatibility of candidate materials for these purposes, potential cladding materials, ORNL-developed low-Cr nickel alloys, and commercial alloys were exposed to static NaCl-MgCl 2 (63-37 mol%) at 700° for 500h. Alloys 617 and 800H had low corrosion rates compared to alloy-N and 316L stainless steel. The ORNL developed alloys had varying corrosion performance, all losing less than 10 mg/cm 2 , but did not significantly outperform commercial alloys with higher Cr contents. Cladding candidates tungsten, TZM, and tantalum had very low mass change, while alloy-200 samples (99% Ni) had by far the highest mass change, 20-25 mg/cm 2 .

36 MATERIALS SCIENCE↗