According to developers at the GAIN 2019 microreactors workshop, advanced manufacturing (AM) is needed to fabricate microreactor components.A methodology has not been established to qualify AM for elevated-temperature nuclear construction.A potential approach is to use the wrought design methodology by establishing the AM properties as equivalent to or better than wrought.This investigation aims to understand the relationship between powder metallurgy hot isostatic pressing (PM HIP) and material characteristics for Alloy 316L as they relate to elevated-temperature component design and construction.The elevated-temperature cyclic behavior of Alloy 316L manufactured by PM HIP differs from wrought.Of particular significance is the reduced number of cycles to failure for the PM HIP Alloy 316L compared to wrought.Additional work to identify the microstructural feature that control the elevated-temperature cyclic behavior is needed.

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.

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

Creep behavior of Alloy 617 has been extensively characterized to support the development of a draft Code Case to qualify Alloy 617 in Section III division 5 of the ASME Boiler and Pressure Vessel Code. This will allow use of Alloy 617 in construction of nuclear reactor components at elevated temperatures and longer periods of time (up to 950°C and 100,000 hours). Prior to actual use, additional concerns not considered in the ASME code need to be addressed. Code Cases are based largely on uniaxial testing of smooth gage specimens. In service conditions, components will generally be under multi axial loading. There is also the concern of the behavior at discontinuities, such as threaded components. To address the concerns of multi axial creep behavior and at geometric discontinuities, notched specimens have been designed to create conditions representative of the states that service components experience. Two general notch geometries have been used for these series of tests: U notch and V notch specimens. The notches produce a tri axial stress state, though not uniform across the specimen. Characterization of the creep behavior of the U notch specimens and the creep rupture behavior of the V notch specimens provides a good approximation of the behavior expected of actual components. Preliminary testing and analysis have been completed and are reported in this document. This includes results from V notch specimens tested at 900°C and 800°C. Failure occurred in the smooth gage section of the specimen rather than at the root of the notch, though some damage was present at the root of the notch, where initial stress was highest. This indicates notch strengthening behavior in this material at these temperatures.

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.

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 themore » 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.« less

Alloy 617 has been selected as the primary candidate for heat exchanger applications in advanced reactors. For the VHTR this application could require extended service up to a reactor outlet temperature of 950°C. A key hurdle to using this alloy in the VHTR heat exchanger application is qualifying the alloy for Section III of the ASME Boiler and Pressure Vessel Code. In order to Code qualify the material it is necessary to characterize the influence of long term aging on the mechanical behavior. Alloy 617 has been aged at 650 and 750°C for times up to 5300 hours. The microstructure after aging has been characterized using optical and transmission electron microscopies. It has been determined that in addition to carbides, a significant volume fraction of ?’ phase (Ni3Al) is formed at these temperatures. The ?’ does not contribute significantly to changing the tensile or impact properties of the aged material. It does, however, appear to increase creep resistance and impede creep crack growth.

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 °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 mechanisms and failure life. Continuous cycle fatigue and creep-fatigue testing of Alloy 617 was conducted at 950 °C and 0.3% and 0.6% total strain in air to simulate damage modes expected in a VHTR application. Continuous cycle fatigue specimens exhibited transgranular cracking. Intergranular cracking was observed in the creep-fatigue specimens and the addition of a hold time at peak tensile strain degraded the cycle life. This suggests that creep-fatigue interaction occurs and that the environment may be partially responsible for accelerating failure.

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°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. Furthermore, previous work on corrosion of nickel base alloys in impure helium has suggested that this environment is far from inert with respect to Alloy 617. Continuous cycle fatigue and creep-fatigue testing of Alloy 617 was conducted at 950°C and 0.3% and 0.6% total strain in air to simulate damage modes expected in a VHTR application. Continuous cycle and creep-fatigue specimens exhibited intergranular cracking, but did not show evidence of grain boundary cavitation. 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 creep-fatigue interaction may occur by a different mechanism or that the environment may be partially responsible for accelerating failure.

Several nickel based solid solution alloys are under consideration for application in heat exchangers for very high temperature gas cooled reactors. The principal candidates being considered for this application by the Next Generation Nuclear Plant (NGNP) project are Inconel 617 and Haynes 230. While both of these alloys have an attractive combination of creep strength, fabricability, and oxidation resistance a good deal remains to be determined about their environmental resistance in the expected NGNP helium chemistry and their long term response to thermal aging. A series of experiments has been carried out in a He loop with controlled impurity chemistries within the range expected for the NGNP. The influence of oxygen partial pressure and carbon activity on the microstructure and mechanical properties of Alloys 617 and 230 has been characterized. A relatively simple phenomenological model of the environmental interaction for these alloys has been developed.

HVOF coatings have shown high resistance to corrosion in fossil energy applications and it is generally accepted that mechanical failure, e.g. cracking or spalling, ultimately will determine coating lifetime. The high velocity oxygen-fuel method (HVOF) for applying coatings is one of the most commercially viable and allows the control of various parameters including powder particle velocity and temperature which influence coating properties, such as residual stress, bond coat strength and microstructure. The mechanical durability of coatings is being assessed using a dual eddy current coil method to monitor crack formation in real time during thermal cycling. Absolute impedence signals from two coils, which interrogate two different areas on the sample, are collected. Crack detection can be determined from the differential signal generated from these absolute signals. The coils are operated at two different frequencies, resulting in two differential signals used for crack detection. Currently this crack detection method is being used to elucidate the influence of thermal cycling temperature and coating thickness on cracking. Recent results (cycles to failure) will be presented for FeAl coatings thermally sprayed (HVOF) onto carbon steel to two coating thicknesses (160 microns and 250 microns thick) and subsequently cycled at temperatures up to 700oC. Thinnermore » coatings exhibit greater resistance to cracking. Ultimately the resistance to cracking will be used to explore the relationship between HVOF spraying parameters, the mechanical properties of the coating and coating bond strength to develop optimized thermal spray parameters. To this end thermal spray coatings (FeAl and Fe3Al) have been applied to additional alloy substrates (Grade 91 steel, 316 SS, etc.) relevant to the fossil industry. Future plans also include a direct comparison to conventional weld overlay coatings currently used in the industry as well as exploration of new coatings. The room temperature mechanical strength and coating adhesion to the substrate is also of considerable importance. Eddy current methods are being developed to detect coating failure during room temperature tensile tests to optimize surface preparation as well as aid in the optimization of the HVOF thermal spray parameters.« less

Coatings of iron based aluminides have been deposited on steel substrates using the high velocity oxygen- fuel method (HVOF). The coatings are highly dense, contain relatively low fraction of oxides and have compressive residual stress in the as-deposited condition. While the microstructures are stable during corrosion testing in either oxidizing or simulated combustion gas environments the state of stress in the coating and the bonding to the substrate can change significantly as a result of corrosion or thermal cycling and this potentially leads to coating failure. Mechanical properties of free standing HVOF processed specimens have been characterized and compared to material that remained attached to substrates to separately determine the influence of microstructure evolution and constraint due to residual stress arising from the coating-substrate interaction on the cracking of coatings. A series of thermal cycling tests is being developed and the influence on corrosion resistance and mechanical properties will be determined.