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A forced convection PbLi loop has been designed and constructed to test tritium extraction from molten PbLi. This Tritium Extraction eXperiment (TEX) is designed to operate at temperatures and tritium concentrations relevant to the helium cooled lead lithium, water cooled lead lithium, and dual coolant lead lithium breeding blankets. The major systems consist of a moving magnet pump, source permeator, supply tank, test section, and plenum. Additional subsystems, including flowmeters, pressure gauges, valves, and heaters are described. The versatile design of TEX will enable code validation using simple tubular membranes, as well as determining tritium extraction efficiencies for alternative test sections. [ABSTRACT FROM AUTHOR]

Fusion and fission reactions are profoundly dependent on hydrogen for sustained reactions. Fusion is fueled by the isotopes of hydrogen, and predominantly hydrogen-based moderators slow fission neutrons to propagate chain reactions. Intentional tritium production in fusion and concomitant tritium production in fission reactors introduce challenges. The technology to efficiently extract, harvest, and quantify tritium in and from advanced fission molten salt coolants and fusion molten tritium breeder materials will require more pronounced research and development. Measuring and quantifying hydrogen is necessary in all areas of nuclear materials. Although many characterization techniques cannot directly detect hydrogen, numerous techniques provide the necessary information to understand the behavior of hydrogen in nuclear materials. [ABSTRACT FROM AUTHOR]

Dense vanadium-based membranes offer high permeability and perfect selectivity to hydrogen isotopes, maintain favorable neutronic properties, and are compatible with liquid metals such as PbLi. These properties make vanadium membranes a promising fusion fuel cycle technology for processes such as tritium extraction from PbLi and exhaust processing. Surface contamination has a deleterious effect on the gas-phase hydrogen permeation through vanadium, and the reported permeabilities range from 10-14 to 10-7 mol m-1 s-1 Pa-0.5. Thin dense films of palladium applied to clean vanadium surfaces enable a consistently high hydrogen permeability. In this study, uncoated vanadium resulted in deuterium permeabilities ranging from 2.8 × 10-11 to 6.4 × 10-9 mol m-1 s-1 Pa-0.5 at 300 °C–700 °C, respectively. Post-test analysis revealed a VO x surface layer and VC x subsurface layer formed on the feed side, while the as-received surface oxide dissolved leaving a submonolayer oxide on the permeate surface. The Pd-coated V resulted in a maximum deuterium permeability of 2.1 × 10-7 mol m-1 s-1 Pa-0.5 at 375 °C upon activation of the Pd surface by oxidation and reduction. The deuterium permeation declined upon heating to 500 °C due to intermetallic diffusion between the Pd and V. The Mo 2 C-coated V resulted in deuterium permeabilities ranging from 2.7 × 10-10 to 1.8 × 10-9 at 500 °C–700 °C, respectively, and a post-test analysis found the carbon in the Mo 2 C layer had dissolved into the V near the interface. Image 1 • D 2 permeation through uncoated V is controlled by surface oxide and carbide layers. • D 2 permeation through Pd coated V has high permeability in activated surface state. • Mo 2 C coated V resulted in carbon transfer from Mo to V forming Mo/VC x. • Hydrogen isotope permeation through V membranes is reviewed. [ABSTRACT FROM AUTHOR]

Abstract The Idaho National Laboratory (INL) Fusion Safety Program (FSP) is the Office of Fusion Energy Sciences' (FES) lead laboratory for Magnetic Fusion Energy (MFE) Safety. Our mission is to assist the US and international fusion communities in developing the inherent safety and environmental potential of fusion power by: 1) Developing fusion licensing data and analysis tools, 2) Participating in national and international collaborations and design studies, and 3) Assisting the US and international fusion community in licensing activities and guidance in operational safety. To achieve our mission, the FSP maintains core competencies in the several areas, two of which are: fusion safety code development and tritium retention and permeation in fusion plasma facing component (PFC) and blanket materials. This article details recent accomplishments of our program in these areas and future directions for fusion safety research and development at INL. [ABSTRACT FROM AUTHOR]