Stephen N. Paglieri

Preparation and characterization of Pd-Cu composite membranes for hydrogen separation

Pd–Cu composite membranes were made by successive electroless deposition of Pd and then Cu onto various tubular porous ceramic supports. Ceramic filters used as supports included symmetric -alumina (nominal 200 nm in pore size), asymmetric zirconia on-alumina (nominal 50 nm pore size), and asymmetric -alumina on -alumina (nominal 5 nm pore size). The resulting metal/ceramic composite membranes were heat-treated between 350 and 700 ◦ C for times ranging from 6 to 25 days to induce intermetallic diffusion and obtain homogeneous metal films. Pure gas permeability tests were conducted using hydrogen and nitrogen. For an 11 m thick, 10 wt.% Cu film on a nominal 50 nm pore size asymmetric ultrafilter with zirconia top layer, the flux at 450 ◦ C and 345 kPa H2 feed pressure was 0.8 mol/m 2 s. The ideal hydrogen/nitrogen separation factor was 1150 at the same conditions. The thickness of the metallic film was progressively decreased from 28 md own to 1–2m and the alloy concentration was increased to 30 wt.% Cu. Structural factors related to the ceramic support and the metallic film chemical composition are shown to be responsible for the differences in membrane performance. Among the former are the support pore size, which controls the required metal film thickness to insure a leak-free membrane and the internal structure of the support (symmetric or asymmetric) which changes the mass transfer resistance. The support with the 200 nm pores required more Pd to plug the pores than the asymmetric membranes with smaller pore sizes, as was expected. However, leak-free films could not be deposited on the support with the smallest pore size (5 nm -alumina), presumably due to surface defects and/or a lack of adhesion between the metal film and the membrane surface.

Development of membranes for hydrogen separation: Pd coated V–10Pd

Abstract Numerous group IVB and VB alloys were prepared and tested as potential membrane materials, but most of these materials were brittle or exhibited cracking during hydrogen exposure. One of the more ductile alloys, V–10Pd (at.-%), was fabricated into a thin foil (107 μm thick) composite membrane coated with 100 nm of Pd on each side. The material was tested for hydrogen permeability, resistance to hydrogen embrittlement, and long term hydrogen flux stability. The hydrogen permeability ϕ of the V–10Pd membrane was 3·86×10–8 mol m–1 s–1 Pa–0·5 (average of three different samples) at 400°C, which is slightly higher than the permeability of Pd–23Ag at that temperature. A 1400 h hydrogen flux test at 400°C demonstrated that the rate of metallic interdiffusion was slow between the V–10Pd foil and the 100 nm thick Pd coating on the surface. However, at the end of testing, the membrane cracked at 118°C because of hydrogen embrittlement.

Anomalous yield reduction in direct-drive deuterium/tritium implosions due to H3e additiona)

Glass capsules were imploded in direct drive on the OMEGA laser [Boehly et al., Opt. Commun. 133, 495 (1997)] to look for anomalous degradation in deuterium/tritium (DT) yield and changes in reaction history with H3e addition. Such anomalies have previously been reported for D/H3e plasmas but had not yet been investigated for DT/H3e. Anomalies such as these provide fertile ground for furthering our physics understanding of inertial confinement fusion implosions and capsule performance. Anomalous degradation in the compression component of yield was observed, consistent with the “factor of 2” degradation previously reported by Massachusetts Institute of Technology (MIT) at a 50% H3e atom fraction in D2 using plastic capsules [Rygg, Phys. Plasmas 13, 052702 (2006)]. However, clean calculations (i.e., no fuel-shell mixing) predict the shock component of yield quite well, contrary to the result reported by MIT but consistent with Los Alamos National Laboratory results in D2/H3e [Wilson et al., J. Phys.: Conf....

MEASUREMENT OF THE 3HE PERMEABILITY OF DT-FILLED FUSED SILICA INERTIAL CONFINEMENT FUSION (ICF) TARGETS TO STUDY THE EFFECTS OF 3HE ON NEUTRON EMISSION DURING IMPLOSION

Abstract A set of laser implosion experiments were conducted at the OMEGA laser at the University of Rochester, Laboratory for Laser Energetics (LLE) to study the effect of 3He concentration in DT-filled target shells on fusion yield in ICF implosions.. Eleven laser fusion shells consisting of 1100-μm diameter, hollow, fused silica spheres with 4.6 to 4.7-μm-thick walls were loaded with 520 kPa of deuterium-tritium (DT) and then with 3He (101.3 or 520 kPa). The 3He permeabilities of the shells were determined by measuring the pressure rate of rise into a system with known volume. A mathematical method was developed that relied on the experimental fill pressure and time, and the rate of rise data to solve differential equations using MathCAD to simultaneously calculate 3He permeability and initial 3He partial pressure inside the shell. Because of the high permeation rate for 3He out of the shells compared to that for DT gas, shells had to be recharged with 3He immediately before being laser imploded or “shot” at LLE. The 3He partial pressure in each individual shell at shot time was calculated from the measured 3He permeability. Two different partial pressures of 3He inside the shell were shown to reduce neutron and gamma yields during implosion.

High-Concentration Tritium Sensor

A bi-layer device was fabricated and tested for the direct collection of electrons emitted by tritium beta decay. The sensor functions at high pressures and concentrations where previously no simple and cost effective direct measurement technique existed for tritium. A polished KOVARTM (Fe-Ni-Co alloy) rod was coated with a 1-μm thick insulating layer of alumina using electron-beam evaporation, physical vapor deposition (PVD) of alumina with oxygen dosing. The alumina deposition process was optimized to minimize pinholes and obtain a stable coating with high resistivity. The detector exhibited a nanoampere electrical response over a few decades of tritium concentration, up to pure tritium at 200 kPa. The sensor has been in service for several months now without showing signs of degradation and no discernible physical damage or change in efficiency has been observed.

A TECHNIQUE FOR SYNTHESIZING METAL TRITIDE STANDARDS

Abstract Before surplus plutonium pits can be decommissioned and converted into metal oxides to be used as reactor fuels, residual tritium must be reduced to an acceptable level. We have developed two analytical methods involving melting and acid dissolution, combined with liquid scintillation counting as a quantitative and sensitive technique for measuring residual tritium in Pu metal. The detection limit, linearity, and reproducibility of these analytical methods must be validated with a series of metal tritide standards. Since there are no commercially available metal tritide standards, we have developed a technique for their synthesis. The synthesis of these low-level metal tritide standards is accomplished by charging cerium powder with a known amount of tritium to form a master cerium tritide alloy and then by aliquoting from this master alloy and diluting with pure cerium powder to form a series of standards with different tritium concentrations. The major difficulty in synthesizing these standards is that the samples contain extremely low levels of tritium, which span over three decades of concentrations. The synthesis technique and initial data obtained for cerium hydride samples will be presented.

Design, fabrication, and testing of a getter-based atmosphere purification and waste treatment system for a nitrogen-hydrogen-helium glovebox

Abstract A system containing a combination of getters (Zr-Mn-Fe, SAES St909; and Zr2Fe, SAES St198) was used to process the nitrogen-hydrogen-helium atmosphere in a glovebox used for handling metal tritide samples. During routine operations, the glovebox atmosphere is recirculated and hydrogenous impurities (i.e. CQ4, Q2O, and NQ3, where Q [is equivalent to] H, D, T) are decomposed (cracked) and removed by Zr-Mn-Fe without absorbing elemental hydrogen isotopes. If the tritium content of the glovebox atmosphere becomes unacceptably high, the getter system can rapidly strip the glovebox atmosphere of all hydrogen isotopes by absorption on the Zr2Fe, thus lessening the burden on the facility waste gas treatment system. The getter system was designed for high flowrate (> 100 l/min), which is achieved by using a honeycomb support for the getter pellets and 1.27-cm diameter tubing throughout the system for reduced pressure drop. The novel getter bed design also includes an integral preheater and copper liner to accommodate swelling of the getter pellets, which occurs during loading with oxygen and carbon impurities. Non-tritium functional tests were conducted to determine the gettering efficiencies at different getter bed temperatures and flowrates by recirculating gas through the system from a 6-m3 glovebox containing known concentrations of impurities.