Effects of high-pressure hydrogen on storage vessel materials.
Tensile tests in high pressure hydrogen on various steels, discussing strength and ductility of notched and unnotched specimens
Chandler, W. T.↗
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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
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Liquid-Level Monitor for Pressurized Vessels
Technique for monitoring water levels in pressurized stainless-steel cylinders, based on differences in gamma-ray attenuation coefficients in water and air, developed. Full-scale laboratory prototype system constructed to test technique. Technique usable with liquids other than water, since linear attenuation coefficients for intermediate-energy gamma rays in air considerably lower than in liquids. Also adaptable for continuous monitoring of liquid levels in resevoir systems and in underground storage tanks.
Singh, J. J.↗
TECHNICAL BACKGROUND FOR THE UPDATE AND EXTENSION OF THE TIME-DEPENDENT ALLOWABLE STRESSES OF TYPE 304H AND 316H STAINLESS STEELS
Type 304H and 316H stainless steels are codified in the ASME Boiler and Pressure Vessel Code for Section III, Division 5, Class A construction for up to 1500F (816C) and 300,000 hours. Extension of the design lifetimes to 500,000 hours has been undertaken by various research projects and by the ASME Section III, Division 5 Code Committees. Due to a long-standing issue related to non-classical creep behavior on the use of the time to the onset of tertiary creep as one of the time-dependent allowable stress criteria, little progress was made in the extension of the time-dependent allowable stresses for Type 304H and 316H stainless steels. A recent effort by Dabrow and Nestell (Impact of Tertiary Creep on Time Dependent Allowable Stresses for Type 304H and 316H Stainless Steels, Report 0300-0003-RPT-001, MPR Associates, Alexandria, VA, 2020) has provided a rational method for the treatment of the highly variable data for the time to onset of tertiary creep. In this paper, the background formulas for determining the time-dependent allowable stress values of Type 304H and 316H stainless steels, as assembled from relevant project work, are presented. These updates and extensions are being recommended to the ASME Code committees for approval, and are subject to change upon feedback from ASME Section III Code committees.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
TECHNICAL BACKGROUND FOR THE UPDATE AND EXTENSION OF THE TIME-DEPENDENT ALLOWABLE STRESSES OF TYPE 304H AND 316H STAINLESS STEELS
Type 304H and 316H stainless steels are codified in the ASME Boiler and Pressure Vessel Code for Section III, Division 5, Class A construction for up to 1500F (816C) and 300,000 hours. Extension of the design lifetimes to 500,000 hours has been undertaken by various research projects and by the ASME Section III, Division 5 Code Committees. Due to a long-standing issue related to non-classical creep behavior on the use of the time to the onset of tertiary creep as one of the time-dependent allowable stress criteria, little progress was made in the extension of the time-dependent allowable stresses for Type 304H and 316H stainless steels. A recent effort by Dabrow and Nestell (Impact of Tertiary Creep on Time Dependent Allowable Stresses for Type 304H and 316H Stainless Steels, Report 0300-0003-RPT-001, MPR Associates, Alexandria, VA, 2020) has provided a rational method for the treatment of the highly variable data for the time to onset of tertiary creep. In this paper, the background formulas for determining the time-dependent allowable stress values of Type 304H and 316H stainless steels, as assembled from relevant project work, are presented. These updates and extensions are being recommended to the ASME Code committees for approval, and are subject to change upon feedback from ASME Section III Code committees.
304/316 life extension↗
Effect of Heat Treatment on Microstructure and Mechanical Property of 316L Stainless Steel Produced by Laser Powder Bed Fusion
The advanced non-light water reactor designs (Gen IV reactors), including molten salt/ very high temperature/ sodium-cooled and lead-cooled fast reactors, typically operate at higher temperatures and more extreme radiation conditions than light water reactors. An intrinsic part of the deployment and progress of Gen IV reactor designs is selecting the most suitable structural material for a specific application. Additive manufacturing (AM), a fairly new process of making physical, three-dimensional objects from a computer design file, is going to completely change the way of design, build and certify nuclear systems. It offers a range of opportunities to produce complex geometries from existing materials, offers new routes for processing of previously difficult to process materials, allows for design of new high-performance materials, and finally facilitates hybridization of dissimilar materials. This emerging technology has successfully produced cars, wind turbine blade molds and even live cells. It could also open up big opportunities for the nuclear industry to quickly deploy technologies at a fraction of the cost. So far, AM techniques have been preliminarily applied in the field of nuclear reactors, including the classical parts such as the pressure vessel of a small reactor with 508-III steel, the bottom nozzle of a fuel assembly with 304L steel, the fuel cladding with zirconium alloy and the integrated impeller of a pump and the multi-channel valve body with 316L steel [6,7]. The AM applications for operating nuclear reactors started in auxiliary plant components and have slowly migrated to metallic reactors and core components, but many of these are not safety critical components. Although many parts used for nuclear reactors have been fabricated by AM techniques, practical applications in engineering are still a long way off due to the uncertainty factors focused on the processing, material properties, analysis methods and application standards, which feeds the safety and life-cycle of the nuclear reactor. Due to rapid, repeated heating and cooling during production, a high dislocation density was present in the AM material. This microstructure feature is unstable at elevated temperature while high temperature is one of the typical operation environments for nuclear reactors. Thus, it is important to understand the thermal effect on the microstructure of AM material. The objectives of this study are to investigate the effect of heat treatment on the microstructure and mechanical properties of 316L stainless steel produced by laser powder bed fusion additive manufacturing, and to determine an appropriate heat treatment practice that will be applied to the lightweight AM lattice-structured material with the same chemistry. The heat treatment study consisted of annealing the samples at a temperature range of 800 to 1200 oC with a 50 oC increment for different times (1-24 hours), followed by vacuum or air cooling. Microstructural characterization was carried out by Scanning Electron Microscope (SEM). Grain size and crystallographic orientation were investigated by Electron Backscatter Diffraction (EBSD). Vickers hardness tests with a 0.5 kg load were employed to determine the hardness of samples after different heat treatments. After heat treatment, the random crystallographic orientation was preserved, and the volume fraction of high-angle grain boundaries (grain boundary misorientation =15 oC) remained the same. The dislocation density decreased with annealing temperature due to recovery. The fine subgrain structures in the as-printed specimen were quite stable up to 1200 oC. Minimal recrystallization was observed up to 1200 oC. Recrystallization initiated only after 8.5 hours at 1200 oC. The SEM images did not show obvious dependence of microstructure on cooling rate. The hardness of the specimens decreased with increasing annealing temperature as a result of the decrease in dislocation density. It is interesting to note that the AM material showed very similar hardness to the wrought material when annealing at similar temperature, although the microstructures are very different. Annealing at 1050 oC for 1 hour followed by air cooling was selected as the heat treatment procedure for the lattice designed lightweight AM 316L material.
36 MATERIALS SCIENCE↗
Modified 316H constitutive model updated with high temperature stress relaxation test data
This report describes the calibration of a new high temperature constitutive model for 316H stainless steel, suitable for use with the ASME Boiler & Pressure Vessel Section III, Division 5, Class A rules for design by inelastic analysis. The model retains the same mathematical form used by the reference model included in Nonmandatory Appendix Z of the Code, but refits the model to an expanded dataset including all the data used to fit the original model plus seven new stress relaxation tests. The addition of these high temperature stress relaxation tests improves the model's accuracy in predicting relaxation at temperatures greater than 700 ⁰C, without compromising the accuracy of the model versus the original calibration data.
36 MATERIALS SCIENCE↗
Strain Analysis of Graphite/Epoxy Vessels
NASA Technical Paper discusses stiffness parameters of filament-wound graphite/epoxy pressure vessels. In some situations, lightweight composite substituted for steel.
Verderaime, V.↗
HYDRODYNAMIC TESTING OF ZIRCALOY TO STEEL HUB COUPLINGS FOR USE IN ADVANCED TEST REACTOR I-LOOP TUBE
The ATR I-positions' isolation from the flux trap region provides an opportunity for a test facility that could create conditions representative of commercial light water reactors. Such a facility requires piping couplings and joints that withstand extreme thermal and pressure cycling. Also, transitioning from steel to zircaloy piping inside the reactor vessel allows the neutronic properties of zircaloy to be leveraged while using cost-effective stainless steel outside of the core. Grayloc-style hub couplings will act as transition points between stainless steel piping and zircaloy piping inside the reactor vessel. These couplings have not been sufficiently tested for leakage where there is a transition between materials. This transition is a concern because the thermal expansion coefficients of the materials differ. Seal integrity between the two components may be greatly affected by thermal expansion at the connection. Thus, the effect of thermal cycling must be assessed. Thermal and pressure cycling and hydrodynamic testing will be conducted to verify seal integrity of the steel to zircaloy hub coupling when subjected to ATR conditions.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Interaction of beryllium with 316H stainless steel in molten Li 2 BeF 4 (FLiBe)
The increased demand in renewable energy resources has led to a renewed interest in nuclear reactors including a new generation of molten salt reactors. One of the issues encountered in handling of molten chloride or fluoride salts is the inability of metals to form and retain a protective oxide surface layer, generally chromium oxide. The low chromium content, nickel-based alloy INOR-8, later marketed as Hastelloy N, was specifically developed to contain molten fluorides. This alloy had some deficiencies, but its composition resulted in it incurring far less chromium depletion by the molten fluoride salts than alloys like 316 stainless steel. For the next generation of molten salt reactors, an alloy with pressure vessel code approval and higher temperature capability than Hastelloy N is required. Consequently, several reactor designers have reportedly chosen to build the reactor containment vessel from 316H stainless steel. In order to minimize corrosion of the stainless steel by the molten fluoride salt (2LiF-BeF 2 ), it is proposed to add beryllium metal to the salt to react with impurities and provide a means to lower the oxidation potential thus making the salt less corrosive. Even so, there is a concern whether the beryllium would react with components of the stainless steel. To address this concern, static capsule tests were conducted in which selected amounts of beryllium were added to capsules containing 2LiF-BeF 2 (FLiBe) salt and 316H stainless steel and Hastelloy N samples. Following exposure, the samples were cleaned and examined using a variety of analysis techniques including, optical microscopy, SEM-EDS, XPS, LIBS, XRD and EPMA. Tensile samples were also exposed in capsules, and those were subjected to tensile testing. Notably, It was found that for the level of beryllium additions used in this study, intermetallic compounds were formed which could be detrimental to the long-term performance of the 316H stainless steel.
36 MATERIALS SCIENCE↗
Effects of high-pressure hydrogen on storage vessel materials
Tensile tests on welded plates of three grades of steel indicate an appreciable decrease of tensile strength and ductility of notched specimens and a decrease in ductility of unnotched samples. Surface cracking, evident in unnotched specimens, is conducive to breakdown in tensile strength and ductility.
Chandler, W. T.↗
Experimental Validation of Exact Burst Pressure Solutions for Thick-Walled Cylindrical Pressure Vessels
Burst pressure is one of the critical strength parameters used in the design and operation of pressure vessels because it represents the maximum pressure that a vessel can withstand before failing. Historically, the Barlow formula was used as a design base for estimating burst pressure. However, it does not consider the plastic flow response for ductile steels and is applicable only to thin-walled cylinders (i.e., the diameter to thickness ratio D/t ≥ 20). A new multiaxial plastic yield theory was developed to consider the plastic flow response, and the associated theoretical (i.e., Zhu–Leis) solution of burst pressure was obtained and has gained extensive applications in the pipeline industry because it was validated by different full-scale burst test datasets for large-diameter, thin-walled pipelines in a variety of steel grades from Grade B to X120. The Zhu–Leis flow theory of plasticity was recently extended to thick-walled pressure vessels, and the associated exact flow solution of burst pressure was obtained and is applicable to both thin and thick-walled cylindrical shells. Many full-scale burst tests are available for thin-walled line pipes in the pipeline industry, but limited pressure burst tests exist for thick-walled vessels. To validate the newly developed exact solutions of burst pressure for thick-walled cylinders, this paper conducts a series of burst pressure tests on small-diameter, thick-walled pipes. In particular, six burst tests are carried out for three thick-walled pipes in Grade B carbon steel. These pipes have a nominal diameter of 2.375 inches (60.33 mm) and three nominal wall thicknesses of 0.154, 0.218, and 0.344 inches (3.91, 5.54, and 8.74 mm), leading to D/t = 15.4, 10.9, and 6.9, respectively. With the burst test data, comparisons show that the Zhu–Leis flow solution of burst pressure matches well the burst test data for thick-walled pipes. Thus, these burst tests validate the accuracy of the Zhu–Leis flow solution of burst pressure for thick-walled cylindrical vessels.
42 ENGINEERING↗
Cyclic-load crack growth in ASME SA-105 grade II steel in high-pressure hydrogen at ambient temperature
ASME SA-105 Grade II steel, which is used in high-pressure hydrogen compressor systems, is similar to steels used or considered for use in high-pressure hydrogen storage vessels and pipelines. This paper summarizes the results of a program conducted to provide cyclic-load crack growth rate (da/dN) data for a fracture mechanics analysis of a 15,000 psi hydrogen compressor facility which contains pulse quieter and after-cooler separator vessels constructed of the ASME SA-105 Grade II steel. Included in the program were tests performed to assist in establishing operating procedures that could minimize the effect of hydrogen on crack growth rates during operation.
Walter, R. J.↗
Time-Dependent 600 °C Post-Weld Heat Treatment Response of Laser-Wire-Directed Energy Deposited Austenitic Stainless Steel Claddings on Carbon Steel
Austenitic stainless steel overlay claddings are often used to protect nuclear reactor pressure vessels from corrosion. Arc welding processes are traditionally employed, which require a high heat input to bond the dissimilar metals, creating a large heat-affected zone and high residual stress that demands a lengthy stress-relieving post-weld heat treatment, up to 48 hours. High energy beam (e.g., laser, electron beam) directed energy deposition-based overlay cladding processes show promise for fabricating claddings with superior performance compared to arc welded ones because of their localized heat input; however, their necessity and time of post-weld heat treatment are currently unclear. In this research, we investigate the time dependence of a 600 C stress-relieving post-weld heat treatment of austenitic stainless steel clad on low carbon steel by laser-wire-directed energy deposition, comparing the microstructure and mechanical properties of the as-fabricated condition to 2-, 10-, and 48-hour heat treatment times. Diffusion simulations and microscopy show a gradual carburization of the stainless steel cladding and decarburization of the heat-affected zone/base metal. Nanoindentation testing shows up to a 1.8–2.69 embrittlement in the cladding and a 0.33 to 0.69 softening in the heat-affected zone after the 48-hour post-weld heat treatment. Tensile testing of the cladding composites reveals that increasing post-weld heat treatment time raises clad cracking susceptibility due to intergranular carbide formation and shows that laser-wire-directed energy deposition claddings may not require post-weld heat treatment.
Coating↗
A709 Qualification Plan Update and Mechanical Properties Data Assessment
This report provides a summary of the development effort for the qualification of Alloy 709, an advanced austenitic stainless steel, in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors. It provides an assessment of the mechanical properties data generated to date from the tensile, creep, fatigue and creep-fatigue tests of the precipitation-treated Alloy 709 from two commercial heats in plate product form. It was concluded that the mechanical properties of Alloy 709 with the precipitation treatment continued to outperform those of Type 316 stainless steel. This affirms the recommendation to Code qualification this alloy as a replacement for Type 316 stainless steel to support the objective of reducing the construction and operating costs to incentivize advanced reactor deployment. The data also demonstrated that the precipitation treatment is effective in enhancing the creep-fatigue resistance of Alloy 709 while maintaining a significant creep strength advantage over Type 316 stainless steel. This report also provides an update to the test conditions for the creep, fatigue and creep-fatigue test matrices in order to cover the Code Case data package more effectively. Finally, it is recommended to continue the Alloy 709 Code Case Testing Program to develop the data package needed for the determination of the material-specific design parameters for inclusion in the Alloy 709 Code Case.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Detection of Cracks at Welds in Steel Tubing Using Flux Focusing Electromagnetic Probe
The inspection of weldments in critical pressure vessel joints is a major concern in the nuclear power industry. Corrosive environments can speed the fatigue process and access to the critical area is often limited. Eddy current techniques have begun to be used to help overcome these obstacles [1]. As direct contact and couplants are not required, remote areas can be inspected by simply snaking an eddy current coil into the intake tube of the vessel. The drawback of the eddy current method has been the high sensitivity to small changes in the conductivity and permeability of the test piece which are known to vary at weldments [1]. The flaw detection mechanism of the flux focusing electromagnetic probe can help alleviate these difficulties and provide a unique capability for detecting longitudinal fatigue cracks in critical tube structures. The Flux Focusing Electromagnetic Flaw Detector, originally invented for the detection of fatigue and corrosion damage in aluminum plates [2-3], has been adapted for use in testing steel tubing for longitudinal fatigue cracks. The modified design allows for the probe to be placed axisymmetrically into the tubing, inducing eddy currents in the tube wall. The pickup coil of the probe is fixed slightly below the primary windings and is rotated 90 so that its axis is normal to the tube wall. The magnetic flux of the primary coil is focused through the use of ferromagnetic material so that in the absence of fatigue damage there will be no flux linkage with the pickup coil. The presence of a longitudinal fatigue crack will cause the eddy currents induced in the tube wall to flow around the flaw and directly under the pickup coil. The magnetic field associated with these currents will then link the pickup coil and an unambiguous increase in the output voltage of the probe will be measured. The use of the flux focusing electromagnetic probe is especially suited for the detection of flaws originating at or near tube welds. The probe is shown to discriminate against signals due solely to the weld joint so that flaw signals are not hidden in the background in these locations. Experimental and finite element modeling results are presented for the flaw detection capabilities of the probe in stainless steel tubes.
Wincheski, Buzz↗
The Elevated-Temperature Cyclic Properties of Powder Metallurgy-Hot Isostatic Pressed 316H and 316L Stainless Steel
Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code covers construction rules for elevated-temperature nuclear components. Microreactor developers have expressed a need for advanced manufacturing processes to fabricate microreactor components to reduce manufacturing costs. Components fabricated using manufacturing processes other than conventional techniques are not currently qualified in Section III, Division 5. An expeditious approach to qualifying an advanced manufacturing process for alloys whose wrought-product form is already qualified in Division 5 is to demonstrate the resultant properties from the advanced manufacturing process are equivalent or superior to the wrought-product form. Powder metallurgy-hot isostatic pressing (PM-HIP) is a mature technology that offers many advantages that are attractive to the microreactor industry. Preliminary data show that the elevated-temperature creep-fatigue properties of PM-HIP 316H stainless steel (SS) are reduced compared to the Wrought 316H SS, which is qualified in Section III, Division 5. Work is ongoing to identify the mechanisms responsible for the reduced creep-fatigue properties and to establish acceptance criteria to confirm the adequacy of the component for service.
316H stainless steel↗
NASA Ground-based Layered Pressure Vessels Materials Report
A significant portion of Layered Pressure Vessels (LPVs) manufactured for NASA in the 1950s and ’60s are of non-code construction. These vessels were constructed to the standards of individual manufacturers rather than an industry or consensus standard and often contain legacy or proprietary materials, which have poor or no characterization. Obtaining correct material properties is the first step in assuring vessel safety and predicting life and critical failure modes before they happen. This report details the efforts of the LPV Project related to material properties and characterization. The goal in testing materials from deconstructed vessels was to provide material characterization utilizing modern test methods that would give a strong probabilistic base as an option for future analysis. ASTM E1820, Standard Test Method for Measurement of Fracture Toughness was used for fracture toughness; ASTM E1921, Standard Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range for transition temperature; and ASTM E647, Standard Test Method for Measurement of Fatigue Crack Growth Rates for fatigue crack growth. Particular focus was placed on determination of transition temperature through evaluation of the T0 reference temperature in accordance with ASTM E1921. Transition temperature testing allows for a probabilistic analysis of vessel materials that may be operating at temperatures below the transition from ductile tearing to brittle cleavage failure. This test method was not available at the time of construction and has proved invaluable to the characterization of older ferritic steels. The values reported are specific to the largest body of materials found in the current NASA fleet, but do not claim to be all inclusive. As demonstrated in this report, some materials contain extreme lot-to-lot variability. In some cases, this can be comfortably bounded by worst case probabilistic methods, however it is always recommended that vessel specific data be collected in as much depth as possible to ensure correct assumptions and avoid unduly restricting vessel usage.
Layered Pressure Vessel↗
Environmental crack-growth behavior of high strength pressure vessel alloys
Results of sustained-load environmental crack growth threshold tests performed on six spacecraft pressure vessel alloys are presented. The alloys were Inconel 718, 6Al-4V titanium, A-286 steel, AM-350 stainless steel, cryoformed AISI 301 stainless steel; and cryoformed AISI 304L steel. The test environments for the program were air, pressurized gases of hydrogen, oxygen, nitrogen, and carbon dioxide, and liquid environments of distilled water, sea water, nitrogen tetroxide, hydrazine, aerozine 50, monomethyl hydrazine, and hydrogen peroxide. Surface flaw type specimens were used with flaws located in both base metal and weld metal.
Forman, R. G.↗