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81 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Compatibility of Type 304H Stainless Steel in Static Molten FLiBe for Inertial Confinement Fusion Reactors: Role of Impurities and Redox Control

Molten fluoride salts, such as FLiBe (LiF-BeF2), are promising candidates for tritium breeding and heat transfer in fusion reactors, but corrosion of structural materials remains a major challenge. This study investigates the corrosion behavior of austenitic stainless steel 304H in purified and NiF2-containing FLiBe at 500°C and 600°C, focusing on the effects of impurities and redox control. Exposure to purified FLiBe resulted in the concurrent depletion of Cr, Mn, and Fe, with corrosion at 500 °C dominated by the combined oxide formation and elemental dissolution, while at 600°C elemental depletion was predominant. The addition of a controlled NiF2 impurity significantly accelerated corrosion at both temperatures, demonstrating the sensitivity of 304H to the salt redox state. Beryllium additions were effective in mitigating corrosion for both baseline and NiF2-containing FLiBe; minimal depletion of Cr, Mn, and Fe occurred with Be additions as low as 2.5 mg (147 wppm), and no NiBe intermetallics formed at 5 mg (294 wppm), indicating that small Be inventories can provide substantial protection without deleterious phase formation. Thermodynamic equilibrium and coupled thermodynamic-kinetic analyses at the salt-alloy interface suggested low corrosion rates in systems with limited hydrogen fluoride (HF) generation, highlighting the importance of salt redox control. Estimates for a Be addition rate were calculated for the HYLIFE-II fusion reactor that can mitigate corrosion-induced degradation, assuming complete conversion of tritium to tritium fluoride (TF). Overall, 304H shows reasonable compatibility with FLiBe under optimized redox conditions. These results provide quantitative guidance for material selection and salt management in fusion blanket and heat exchanger systems and motivate validation under flowing, nonisothermal, and irradiated conditions.

Pillai, Rishi [ORNL] (ORCID:0000000243688197)↗
NASA NTRS1976

High modulus rare earth and beryllium containing silicate glass compositions

Glass compositions having a Young's modulus of at least 16 million psi and a specific modulus of at least 110 million inches consisting essentially of approximately, by weight, 20 to 43% SiO2, 8 to 21% Al2O3, 4 to 10% BeO, 27 to 58% of at least one oxide selected from a first group consisting of Y2O3, La2O3, Nd2O3, Ce2O3, Ce2O3, and the mixed rare earth oxides, and 3 to 12% of at least one oxide selected from a second group consisting of MgO, ZrO2, ZnO and CaO are described. The molar ratio of BeO to the total content of the first group oxides is from 1.0 to 3.0.

Bacon, J. F.↗
NASA NTRS1996

Simultaneous Use Of Zr And Mg Anodes In XPS

Improved x-ray source for x-ray photoelectron spectroscopy (XPS) contains both zirconium anode with beryllium window and magnesium anode with aluminum window. Previously unresolvable peaks of electron-energy spectrum become resolvable. Developed specifically for use in analyzing distributions of chemical constituents in surface layers of specimens of 2219 aluminum alloy and in determining the depths of surface oxide layers and relative proportions of aluminum and oxide in layers. Also used to study chemical constituents of surface layers in other material systems - for example, thin oxide films on silicon-based semiconductor devices, oxide films on alloys, and surface layers affecting adhesion of paints or bonding materials.

Allgeyer, D. F.↗
DOE OSTI2026

Investigation of Ion-Irradiated Beryllium Carbide (Be 2 C) for High-Temperature Nuclear Reactor Moderators

Beryllium carbide (Be 2 C) is an attractive alternative to graphite as a moderator material because of its high melting point, moderating efficiency, and theoretical environmental compatibility in MSRs, FHRs and HTGRs. However, its behavior under neutron irradiation is not yet known. Research on otherwise promising beryllium compounds are restricted because of the toxic nature of the material. For this work, a novel experiment was designed to safely irradiate beryllium-containing samples at The Michigan Ion Beam Laboratory (MIBL). Using this new capability, Be 2 C samples were irradiated with 9 MeV C 3+ ions from 2 dpa to 30 dpa, at temperatures up to 500 °C. Samples were characterized to investigate radiation tolerance. No evidence of phase precipitation, dislocation loops, or amorphization was observed up to 30 dpa, suggesting good radiation tolerance. One sample left in air for ~3 months oxidized slowly and lost structural integrity, while a sample kept in argon did not, suggesting that Be 2 C is not suitable for service in moist environments, but maintains its integrity in dry environments.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1972

Cleaning and activation of beryllium-copper electron multiplier dynodes.

Description of a cleaning and activation procedure followed in preparing beryllium-copper dynodes for electron multipliers used in sounding-rocket experiments to detect auroral electrons. The initial degreasing step involved a 5-min bath in trichloroethylene in an ultrasonic cleaner. This was followed by an ultrasonic rinse in methanol and by a two-step acid pickling treatment to remove the oxides. Additional rinsing in water and methanol was followed by activation in a stainless-steel RF induction oven.

Pongratz, M. B.↗
DOE OSTI2025

Experimental and computational studies on high-entropy carbide MoNbTaVWC 5 under high pressures

High-entropy carbide, MoNbTaVWC 5 , was synthesized from oxide precursors of the constituent metals, mixed with graphite powder in a microwave-generated hydrogen plasma at 26.66 kPa and 2100 °C. Ambient x-ray diffraction analysis confirms the full conversion of oxide precursors into a single-phase, face-centered cubic structure with a lattice parameter a = 4.3309 Å. Nanoindentation measured a hardness of 24.5 ± 1.3 GPa and an elastic modulus of 386 ± 22 GPa. The synthesized sample, mixed with a copper pressure marker, was studied by the radial x-ray diffraction technique with beryllium gasketing in a diamond anvil cell up to 70 GPa. The experimentally measured pressure–volume curve and shear strength were compared with theoretical predictions using the special quasi-random structure technique and density functional theory. MoNbTaVWC 5 achieved a 12% volume compression at 70 GPa and exhibited a high shear strength of 6.6 GPa. The present study demonstrates that the high-entropy carbide MoNbTaVWC 5 exhibits exceptional incompressibility and high strength under extreme conditions.

36 MATERIALS SCIENCE↗
NASA NTRS1972

Notched strength of beryllium powder and ingot sheets.

The effects of notches in thin beryllium sheets were studied as functions of material variables and notch severity. Double edge notched samples having stress concentration factors of 1.0 to 15.4 were prepared by milling to size, etching, and electrical discharge machining the notches. Strength was not reduced greatly by sharp notches, and duller notches were more deleterious than sharp notches. The trend was for reduced strength for dull notches, increased strength for sharper notches, and reduced strength for very sharp notches. Differences in material purity or source of the sheet had little affect on notch sensitivity. The most important factors appear to be oxide content and directionality of the sheet microstructure; high oxide content and highly directional microstructure tend to give more notch sensitivity than low oxide content, and more bidirectional microstructure. Postulated causes of the change in notched/unnotched strength are given.

Moss, R. G.↗
DOE OSTI2020

An electrochemical study of hydrogen in molten 2LiF-BeF 2 (FLiBe) with addition of LiH

The chemical and transport behavior of hydrogen isotopes in molten 2LiF-BeF 2 (FLiBe) is of interest for the design of tritium management systems in nuclear fission and fusion reactors that use FLiBe. The chemical reaction of LiH with FLiBe is used to introduce hydrogen in the molten salt and electrochemical methods are used for in situ studies of hydrogen in FLiBe. LiH reacts with molten FLiBe to generate an electroactive species whose voltammetry peak is proportional to the added quantity of LiH. The cyclic voltammetry reaction potential of 2.009 ± 0.050 V vs Be/Be 2+ and the electron exchange of n = 0.8 ± 0.5 are consistent with the one-electron oxidation of dissolved H in the zero valence state, H o to H + . The concentration of H o is estimated by linear sweep voltammetry at 60–80% of the hydrogen introduced by the reaction of 0.42 mol% LiH with FLiBe and after eleven hours it remains above 40%. It is postulated that covalent BeH 2 is formed in FLiBe upon LiH addition, as a FLiBe-soluble quasi-stable intermediate product. Furthermore, the results provide an evaluation of LiH as a means of introducing dissolved hydrogen in FLiBe, enabling electrochemical methods as tools to advance the understanding of the chemistry of hydrogen isotopes in FLiBe.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Tritium Transport Phenomena in Molten-Salt Reactors: Molten Salt Tritium Transport Experiment Design

Tritium is produced from neutron interactions with both lithium and beryllium. Large quantities of tritium are generated in Molten Salt Reactors (MSRs) which use LiF/BeF2 (FLiBe) as the fuel salt. Tritium is unique among the radionuclide hazards as it readily permeates through metal structural materials at high temperatures. All metal surfaces are potential release paths for tritium. For adequate safety analysis and eventual licensing of new reactors, predictive models for tritium transport and release from MSRs must be developed. These models must account for the multiple transport phenomena involved with tritium: fuel salt phase mass transport, dissociation/recombination reactions on metal surfaces, interstitial diffusion through the metal structure, and salt or gas phase mass transport in the downstream fluid. These models also must also be validated with representative experiments. Our previous report outlined tritium transport phenomena involved in MSRs, made suggestions on gaps in the transport dataset, and proposed an experimental test stand to test combined transport effects – tritium transport through pipe walls in a convective salt flow. In this report, we summarize an updated analysis framework for tritium transport in MSRs, report our results on hydrogen and deuterium permeation through Hastelloy N, and describe the final design of the Molten Salt Tritium Transport Experiment (MSTTE, pronounced “misty”). The MSRE provides the only wholistic experimental data set for tritium transport in MSRs and understanding the transport phenomena involved in the MSRE is crucial for future model development. One set of parameters in our analysis framework was unknown for the MSRE—surface reaction rates for tritium on Hastelloy N. This warranted our hydrogen and deuterium permeation campaign to assess the permeability, diffusivity, and solubility of hydrogen isotopes in clean Hastelloy N. Surface reaction rate constants were probed by low pressure measurements, however, no surface effects were observed in the limits of our permeation apparatus. Permeation experiments on oxidized Hastelloy N were not performed for this report but are planned in future work. The experimental test stand, MSTTE, measures combined transport properties of the salt-metal system. MSTTE is a forced convection FLiBe loop with custom designed test section to measure tritium transport through candidate structural materials. We use MSRE relevant dimensionless numbers to design and scale the test section. Hastelloy N is a candidate loop and test section material due to the relevance for the MSRE and related designs, however, other metals are being considered (e.g. 316H SS) which may better align with current vendor concepts.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗