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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Comparing the deformation mechanism in 316 L stainless steel fabricated by hybrid and additive manufacturing

This study rationalizes the relatively higher elongation with minimal strength drop of 316L stainless steel (316LSS) manufactured using the hybrid approach in comparison to a fully additive approach followed by machining. It was found that twinning was the dominant deformation mechanism in 316LSS produced by both additive and hybrid approaches. Additionally, the relatively lower fraction of twins in the as-fabricated state of hybrid approach compared to a fully additive approach resulted in delaying the onset of steady-state strain hardening rate to higher strain levels, thereby resulting in an overall increase in elongation to failure.

36 MATERIALS SCIENCE↗
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 OSTI2023

An Initial Evaluation of the Elevated-Temperature Cyclic Properties of Optimized 316H Stainless Steel Fabricated by Powder Metallurgy Hot Isostatic Pressing

Powder metallurgy hot isostatic pressing (PM-HIP) is a mature advanced manufacturing technology, which can consolidate metallic powder under high temperature and pressure to generate near-net shape components. This process can lower production costs and reduce component lead times for nuclear reactor construction and can be readily deployed in the near term for microreactor components because of their smaller size. The purpose of this work is to continue the evaluation of elevated-temperature cyclic material properties for PM-HIP 316H stainless steel. Prior scoping results have shown that the commercially procured PM-HIP stainless-steel materials have reduced creep-fatigue performance compared to wrought 316 stainless steels. To better understand how PM-HIP processed material properties compare to traditional manufacturing methods, additional billets of PM-HIP 316H stainless steel were procured and tested. This testing is necessary to understand how elevated-temperature properties are influenced by the PM-HIP process and what data are needed for incorporating PM HIP 316H into the American Society of Mechanical Engineers Boiler Pressure Vessel Code, Section III, Division 5 for high temperature reactor construction. Specifically, an additional PM-HIP 316H stainless steel billet with lower oxygen and nitrogen concentrations was fabricated by the Nuclear Advanced Manufacturing Research Centre of the United Kingdom and tested to understand if elevated-temperature cyclic properties can be improved. The creep-fatigue resistance of the material from this new billet did not show appreciable improvement compared to prior testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Mechanical and Corrosion Response of 316SS in Supercritical CO 2

The supercritical carbon dioxide (s-CO 2 ) Brayton cycle is currently being explored as a replacement for the steam Rankine cycle due to its potential for higher efficiency and lower cycle cost. 316 stainless steel is a candidate alloy for use in s-CO 2 up to roughly 600 °C, but the mechanical effects of prolonged exposure of base and welded material in s-CO 2 have not been analyzed. The potential for carburization makes this an important concern for the implementation of 316 and similar austenitic stainless steels in the s-CO 2 environment. In this study, welded and base material of two types of 316–316L and 316H–were exposed in either s-CO 2 or argon at 550 °C or 750 °C for 1000 h. 550 °C s-CO 2 exposure yielded a thin (< 1 µm) Cr oxide with occasional nodules of duplex Fe oxide and Fe–Cr spinel that were up to 5 microns thick. However, tensile results from s-CO –2 exposure matched those of 550 °C thermal aging in Ar, indicating that no mechanically detrimental carburization occurred in either 316 variant after 1000 h exposure. Conversely, 750 °C s-CO 2 exposure produced roughly 10 × the oxide thickness, with a more substantial Fe oxide (3–5 µm) on the majority of the surface and nodules of up to 40 µm thick. In comparison to aged samples, tensile testing of 750 °C CO 2 -exposed samples revealed ductility loss attributed to carburization. As a result, projections of 316L performance in s-CO 2 indicate that mechanically detrimental carburization—equal to that shown here for 750 °C, 1000 h—will likely be present after 7–14 years of service at 550 °C.

316 stainless steel↗
DOE OSTI2020

Microstructural banding of directed energy deposition-additively manufactured 316L stainless steel

The microstructures of 316 L stainless steel created by rapid solidification are investigated here by comparing the similar microstructures of individual hatches of directed energy deposition additive manufacturing (DED-AM) and those of single, laser surface-melted tracks formed on a solid plate. High recoil pressure, which is exponentially dependent on the laser beam power density, induces convection of the melt pool, which causes formation of microstructural bands in the as-solidified microstructure. The microstructural bands are associated with changes in the chromium concentration and are a significant component of the inhomogeneous microstructure of DED-AM.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Neutron Irradiation Experiments

This presentation discusses the ongoing and upcoming neutron irradiation experiments within the Advanced Materials and Manufacturing Technologies (AMMT) program for the 2024 AMMT program review meeting. Ongoing and upcoming neutron irradiation of additively manufactured 316 stainless steel and A709 in the High Flux Isotope Reactor and Advanced Test Reactor are described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1971

Creep-fatigue analysis by strain-range partitioning.

The framework of a new method is outlined for treating creep-fatigue behavior of metals. Inelastic strain-ranges are partitioned into the components of (1) completely reversed plasticity, (2) tensile plasticity reversed by compressive creep, or tensile creep reversed by compressive plasticity, and (3) completely reversed creep. Each of these components is shown to be related to cyclic life by a Manson-Coffin type power-law equation. A linear life fraction rule is used to combine the damaging effects of the individual components enabling the prediction of life. Test results are presented for a 2.25 Cr-1 Mo steel as well as limited information for a Type 316 stainless steel.

Manson, S. S.↗
NASA NTRS1997

Sample Diluter for Detecting Hypergolic Propellants and Other Toxic or Hazardous Gases

Hardware was developed to dilute vapor samples of purged hypergolic propellants (with air) into the range of existing instruments for detection of such toxic vapors. Since these detectors are normally used to monitor at the threshold limit value (TLV), most do not have quantitative capability at percent levels which relate to lower explosion limit (LEL) and fire hazards. For example, the upper limits of Energetic Sciences (ESI) 6000 series detectors used at KSC are 200 parts per million (ppm) for monomethyl hydrazine (MMH) and 500 ppm for nitrogen dioxide (NO2) arising from decomposition of nitrogen tetroxide (N2O4). Orbiter Processing Facility (OPF) personnel servicing Shuttle thrusters need to measure up to 250 ppm MMH and 7,500 ppm NO2 with portable, intrinsically safe instruments. Our objective was to quickly fabricate a sample diluter out of existing materials as a temporary measure while other parallel efforts were conducted to provide a commercial or in-house-developed instrument to detect high propellant levels. A 3 to 1 diluter would bring 500 ppm MMH into the range of the existing fuel ESI, and a 30 to 1 diluter would do the same for NO2. In this way, familiar equipment already available would be used, resulting in minimal paperwork, safety, and training impacts and low cost. An MMH vapor sample-diluter was constructed from a 1/4-inch Kynar tee, along with specially designed lengths of sample and dilution tubing. The sample line was 3 feet of Bev-A-Line 4, 1/4-inch tube leading to the straight run of the tee. The side run of the tee had a 17-inch length of Bev-A-Line 4, 1/4-inch tube, for nominal 3 to 1 dilution. A gas sample bag was prepared and assayed at 113 ppm ppm MMH, and diluted vapor samples were assayed at 39.5 ppm, or a measured dilution of 2.9 to 1. For NO2, a 316 stainless steel (SS) 1/8-inch tee with 49.5 inches of coiled, 1/8-inch outside diameter (OD) 316 SS tubing was used as the sampling end of the dilution system. The side run of the tee was open. The measured dilution ratio, based on the input value of 6,480 ppm NO2, and the average output value of 233 ppm, was 28 to 1. Thus, sample-diluters were successful in diluting concentrated hypergolic propellant vapors, both MMH and N2O4 into the ranges of existing TLV detectors.

Barile, R. G.↗
NASA NTRS1997

Sample Diluter for Detecting Hypergolic Propellants and Other Toxic or Hazardous Gases

Hardware was developed to dilute vapor samples of purged hypergolic propellants (with air) into the range of existing instruments for detection of such toxic vapors. Since these detectors are normally used to monitor at the threshold limit value (TLV), most do not have quantitative capability at percent levels which relate to lower explosion limit (LEL) and fire hazards. For example, the upper limits of Energetic Sciences (ESI) 6000 series detectors used at KSC are 200 parts per million (ppm) for monomethyl hydrazine (MMH) and 500 ppm for nitrogen dioxide (NO2) arising from decomposition of nitrogen tetroxide (N2O4). Orbiter Processing Facility (OPF) personnel servicing Shuttle thrusters need to measure up to 250 ppm MMH and 7500 ppm NO2 with portable, intrinsically safe instruments. Our objective was to quickly fabricate a sample diluter out of existing materials as a temporary measure while other parallel efforts were conducted to provide a commercial or in-house-developed instrument to detect high propellant levels. A 3 to 1 diluter would bring 500 ppm MMH into the range of the existing fuel ESI, and a 30 to 1 diluter would do the same for NO2. In this way, familiar equipment already available would be used, resulting in minimal paperwork, safety, and training impacts and low cost. An MMH vapor sample-diluter was constructed from a 1/4-inch Kynar tee, along with specially designed lengths of sample and dilution tubing. The sample line was 3 feet of Bev-A-Line 4, 1/4 inch tube leading to the straight run of the tee. The side run of the tee had a 17-inch length of Bev-A-Line 4, 1/4-inch tube, for nominal 3 to 1 dilution. A gas sample bag was prepared and assayed at 113 ppm MMH, and diluted vapor sarnples were assayed at 39.5 ppm, or a measured dilution of 2.9 to 1. For NO2, a 316 stainless steel (SS) 1/8-inch tee with 49.5 inches of coiled, 1/8-inch outside diameter (OD) 316 SS tubing was used as the sarnpling end of the dilution system. The side run of the tee was open. The measured dilution ratio, based on the input value of 6,480 ppm NO2 and the average output value of 233 ppm, was 28 to 1. Thus, sample-diluters were successful in diluting concentrated hypergolic propellant vapors, both MMH and N2O4, into the ranges of existing TLV detectors.

Barile, R. G.↗
DOE OSTI2020

Voltammetry measurements in lithium chloride-lithium oxide (LiCl–Li 2 O) salt: An evaluation of working electrode materials

Instrumentation to provide process monitoring (PM) and safeguards for the oxide reduction (OR) step in the electrochemical processing of used oxide nuclear fuel is necessary to ensure equipment is operated as declared. Cyclic voltammetry (CV) has been proposed for real-time monitoring of the operation of an OR process for safeguards purposes. In this study, different materials including 316 stainless steel (SS), tantalum (Ta), molybdenum (Mo), tungsten (W), platinum (Pt), and iridium (Ir) were evaluated as potential working electrode (WE) materials based on their chemical inertness, corrosion resistance, and sensitivity in detecting Li 2 O and other chloride salts in the OR electrolyte. Of the electrodes evaluated, 316 SS, Ir, and Pt all performed reasonably well in the LiCl-Li 2 O electrolyte. Here, stainless steel was operated in the cathodic potential and had reasonable corrosion resistance and is relatively inexpensive. Iridium could be operated in both the cathodic and anodic potentials and was the most corrosion resistant of those evaluated. Platinum is limited to the anodic potential range, had reasonable corrosion resistance, and was the most sensitive to Li 2 O concentrations in the salt. In the development of a stand-alone safeguards instrumentation, our recommendation is a CV sensor fitted with both Ir and Pt WEs. The results in this study may also be helpful for anode material selection in other electrometallurgy industries, such as molten salt electrolysis.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
NASA NTRS2003

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Martin, James↗
DOE OSTI2022

Corrosion Compatibility of Stainless Steels and Nickel in Pyrolysis Biomass-Derived Oil at Elevated Storage Temperatures

Corrosion compatibility of stainless steels and nickel (Ni200) was assessed in fast pyrolysis bio-oil produced from pyrolysis of high ash and high moisture forest residue biomass. Sample mass change, ICP-MS and post-exposure electron microscopy characterization was used to investigate the extent of corrosion. Among the tested samples, type 430F and type 316 stainless steels (SS430F and SS316) and Ni200 (~98.5% Ni) showed minimal mass changes (less than 2 mg∙cm−2) after the bio-oil exposures at 50 and 80 °C for up to 168 h. SS304 was also considered to be compatible in the bio-oil due to its relatively low mass change (1.6 mg∙cm−2 or lower). SS410 samples showed greater mass loss values even after exposures at a relatively low temperature of 35 °C. Fe/Cr values from ICP-MS data implied that Cr enrichment in stainless steels would result in a protective oxide layer associated with corrosion resistance against the bio-oil. Post exposure characterization showed continuous and uniform Cr distribution in the surface oxide layer of SS430F, which showed a minimal mass change, but no oxide layer on a SS430 sample, which exhibited a significant mass loss.

09 BIOMASS FUELS↗
DOE OSTI2025

Direct Imaging of Hydrogen‐Driven Dislocation and Strain Field Evolution in a Stainless Steel Grain

Hydrogen embrittlement (HE) poses a significant challenge to the durability of materials used in hydrogen production and utilization. Disentangling the competing nanoscale mechanisms driving HE often relies on simulations and electron-transparent sample techniques, limiting experimental insights into hydrogen-induced dislocation behavior in bulk materials. This study employs in situ Bragg coherent X-ray diffraction imaging to track three-dimensional (3D) dislocation and strain field evolution during hydrogen charging in a bulk grain of austenitic 316 stainless steel. Tracking a single dislocation reveals hydrogen-enhanced mobility and relaxation, consistent with dislocation dynamics simulations. Subsequent observations reveal dislocation unpinning and climb processes, likely driven by osmotic forces. Additionally, nanoscale strain analysis around the dislocation core directly measures hydrogen-induced elastic shielding. These findings experimentally validate theoretical predictions and offer mechanistic insights into hydrogen-driven dislocation behavior. The quantified nanoscale phenomena serve as critical inputs for multiscale modeling frameworks to predict bulk material responses and accelerate the development of HE-resistant alloys.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2023

Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Lach, Tim↗
NASA NTRS1992

Cumulative creep-fatigue damage evolution in an austenitic stainless steel

A model of cumulative creep-fatigue damage has been developed which is based on the use of damage curve equations to describe the evolution of creep-fatigue damage for four basic creep-fatigue cycle types. These cycle types correspond to the four fundamental cycles of the Strain Range Partitioning Life Prediction approach of Manson, Halford, and Hirschberg. A concept referred to as Damage Coupling is introduced to analytically account for the differences in the nature of the damage introduced by each cycle type. For application of this model, the cumulative creep-fatigue damage behavior of type 316 stainless steel at 816 C has been experimentally established for the two-level loading cases involving fatigue and creep-fatigue, in various permutations. The tests were conducted such that the lower life (high strain) cycling was applied first, for a controlled number of cycles, and the higher life (lower strain) cycling was conducted at the second level, to failure. The proposed model correlated the majority of the observed cumulative creep-fatigue data.

Mcgaw, Michael A.↗
DOE OSTI2022

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↗
DOE OSTI2023

High-Strength, High-Ductility, High Entropy Alloys with High-Efficiency Native Oxide Solar Absorbers for Concentrating Solar Power Systems

This EPSCoR Project has been investigating the synergy between the excellent high-temperature mechanical behavior of FeMnNiAlCr high entropy alloys (HEA) and the high solar absorptance of their native oxides for high efficiency concentrated solar thermal power (CSP) systems working at >700°C. While HEAs have attracted substantial interest in recent years, most investigations have focused on their applications as structural materials rather than functional materials. This EPSCoR project discovered that FeMnNiAlCr HEAs can potentially be applied synergistically as both a structural and functional material for high-efficiency concentrating solar thermal power (CSP) systems working at >700°C. The HEA itself would be used in high-temperature tubing to carry molten salts or supercritical CO 2 , while its surface oxide would act as a high-efficiency solar thermal absorber. With Fe and Mn being the major components in these HEAs (adding up to ~70 at.% of the alloy), these materials are much more cost-effective than the Ni-based superalloys currently being investigated for high-temperature CSP systems. Through this research, these Fe-Mn based HEAs have demonstrated yield strengths 2-3x greater than that of stainless steel at 700°C and a creep lifetime >800 h at 700ºC under a typical CSP tubing mechanical load of 35 MPa. Their Mn-rich surface oxides maintain a high optical-to-thermal conversion efficiency of ~87% under 1000x solar concentration ratio for 20 simulated day-night thermal cycles between 750ºC and room temperature. In preliminary corrosion studies, these HEAs have sustained immersion in unpurified bromide molten salts for 14 days at 750°C with <2% weight loss, in contrast to 70% weight loss from a 316 stainless steel reference. The simultaneous achievement of promising mechanical, optical, and thermochemical properties in this FeMnNiAlCr system opens the door to new applications of HEAs in solar energy harvesting. Partnerships with Ames Laboratory and Oak Ridge National Laboratory (ORNL) also advanced our understanding of the fundamental structure-property relationships through atomic scale material characterization and first-principles computational modeling. The key research results in this project can potentially be extended to other HEAs and their native oxides. In terms of applications, the proposed FeMnNiAlCr HEA/native oxide system could potentially exceed the mechanical and the optical performance of existing tubing and solar coating materials under EERE’s CSP program at lower cost, which also aligns well with the EPSCoR Science and Technology strategies of New Hampshire in boosting the deployment of renewable energy.

14 SOLAR ENERGY↗
NASA NTRS2022

Sample Materials Considerations for Curating and Processing Pristine MSR Samples

The perseverance rover is collecting and caching samples of Mars as part of the Mars 2020 mission, which represents the first leg of a multi-mission Mars Sample Return Campaign. The MSR Campaign is an international partnership that will result in delivery of the first martian samples to Earth that were not delivered through meteoritic infall. All meteorites, regardless of how they were handled from recovery to curation, have experienced uncontrolled entry and exposure to the terrestrial environment. Whilst meteorite deliveries are serendipitous, they are also unplanned events that require reactionary responses for recovery and curation. However, with the direct return of pristine astromaterials from another body, we are afforded the ability to design a facility in advance of sample delivery to keep those samples in a pristine (i.e., as returned) state for an indefinite period of time. Given that the curation and processing infrastructure needs to be made out of something, it is important to choose materials for the pristine curation environment that will optimize between the need to effectively process samples and the need to minimize contamination of the samples. The Johnson Space Center (JSC) has an optimized list of materials that have been used in previous sample return missions that includes 304/316 Stainless Steel, Teflon, and T6061 Aluminum (1). This set of materials are compatible with inorganic, organic, and biological cleanliness requirements and protocols. Furthermore, only these materials are permitted to come in contact with pristine samples. We note that JSC uses Neoprene and Hypalon for the gloves on their gloveboxes, but the glove material never comes in direct contact with the samples, only the approved materials. The MSR sample tubes will be made of Ti, so Ti may be an acceptable material for making tools, but the minor and trace element abundances of 304 and 316 stainless steel are well known and do not inhibit scientific investigations of metals, including HSE (2). More work is needed to determine whether the same is true for Ti alloys. In addition to defining the materials in the pristine environment, one must also choose whether the pristine environment will be under vacuum or under a specific atmospheric composition and pressure. Although JAXA has successfully implemented pristine curation vacuum chambers for their Hayabusa and Hayabusa2 samples (3), a vacuum environment is not appropriate for martian samples because it may drive deliquescence of mineral phases in the samples that are sensitive to pressure and relative humidity (4). Consequently, the pristine environment for the martian samples should be under an inert gas. It will be crucial to minimize the number of gases that come into direct contact with samples and these gases will need to be high purity and consistent throughout the pristine isolators. Samples at JSC are stored under high purity gaseous nitrogen (1). Dry N2 gas has not been a problem for N isotope studies for high-T release phases, but an additional inert atmosphere like Ar may be needed for samples where there is a particular concern about low-T release of N from bulk sample analysis. References: (1) McCubbin FM, et al. (2019) Space Science Reviews, 215, 1-81. (2) Day JMD, et al. (2018) Meteorit. Planet. Sci. 53:1283-1291. (3) Yada, T., et al., (2014). Meteorit. Planet. Sci. 49, 135-153. (4) Tosca NJ, et al. (2021). Astrobiology, in press, doi:10.1089/ast.2021.0115.

F M McCubbin↗