G.W. Hollenberg

Tritium/hydrogen barrier development

Abstract A review of the hydrogen permeation barriers which can be applied to the structural metals used in fusion power plants is presented. Both implanted and chemically available hydrogen isotopes must be controlled in fusion plants. The need for permeation barriers appears strongest in Pb-17Li blanket designs, although barriers are also necessary for other blanket and coolant systems. Barriers which provide greater than a 1000- fold reduction in the permeation of structural metals are desired. In laboratory experiments, aluminide and titanium ceramic coatings provide permeation reduction factors (PRFs) of 1000 to over 100 000 with a wide range of scatter. The rate-controlling mechanism for hydrogen permeation through these barriers may be related to the number and type of defects in the barriers. Although these barriers appear robust and resistant to liquid metal corrosion, irradiation tests which simulate blanket environments result in very low PRFs in comparison with laboratory experiments, i.e. less than 150. It is anticipated from fundamental research activities that the radiation- and electric-field-induced enhancement of hydrogen diffusion in oxides may contribute to the lower PRFs during in-reactor experiments.

Cation disorder in high dose neutron irradiated spinel

Abstract The crystal structures of MgAl 2 O 4 spinel single crystals irradiated to high neutron fluences (> 5 x 10 26 n/m 2 ( E n > 0.1 MeV)), were examined by neutron diffraction. Crystal structure refinement of the highest dose sample indicated that the average scattering strength of the tetrahedral crystal sites decreased by ∼ 20% while increasing by ∼ 8% on octahedral sites. Since the neutron scattering length for Mg is considerably larger than for Al, this result is consistent with site exchange between Mg 2+ ions on tetrahedral sites and Al 3+ ions on octahedral sites. Least-squares refinements also indicated that, in all irradiated samples, at least 35% of Mg 2+ and Al 3+ ions in the crystal experienced disordering replacements. This cation sublattice disorder is the largest retained damage ever measured in an irradiated spinel material.

Why is magnesia spinel a radiation-resistant material?

Side-by-side irradiation of stoichiometric MgAl 2 O 4 and α-Al 2 O 3 in JOYO shows that the radiation-induced microstructural evolution to exposure of ≤6 dpa proceeds by very different paths in these two materials. The large difference in dislocation loop evolution appears to account for the ease of void swelling in α-Al 2 O 3 and the strong resistance to void formation in MgAl 2 O 4 . Irradiation of MgAl 2 O 4 to much higher exposure (22-230 dpa) in FFTF confirms the details of the dislocation evolution, which involves a progressive change in Burgers vector and habit plane as interstitial loops increase in size. Constraints unique to the MgAl 2 O 4 crystal structure do not allow the formation of dislocation network structures that favor swelling

Current experimental activities for solid breeder development

Lithium-containing ceramics are among the principal materials being considered for tritium production in future fusion reactors. To ensure development of a data base adequate for evaluation of solid breeder materials, ongoing experimental studies are focused on resolving critical issues related to thermodynamic, thermophysical, and mechanical behavior; to tritium transport and release; and to material response to a neutron environment.

Dimensional stability, optical and elastic properties of MgAl2O4 spinel irradiated in FFTF to very high exposures

Stoichiometric MgAl{sub 2}O{sub 4} spinel specimens irradiated in FFTF-MOTA at temperatures between 385 and 750C to fluences ranging from 2.2 to 24.9 {times} 10{sup 22} n cm {sup {minus}}2 (E>0.1 MeV) darken significantly, but do not develop any significant loss in weight or change in dimensions. Similar behavior was observed in both single crystal and fully dense polycrystalline specimens. Measurements of elastic constants by an ultrasonic technique show that no measurable changes occur as a result of the irradiation. These and other results confirm the stability of this material for fusion application as an electrical insulator.

Composite materials for fusion applications

Ceramic matrix composites, CMCs, are being considered for advanced first wall and blanket structural applications because of their high-temperature properties, low neutron activation, low density and low coefficient of expansion coupled with good thermal conductivity and corrosion behavior. This paper presents a review and analysis of the hermetic, thermal conductivity, corrosion, crack growth and radiation damage properties of CMCs.

Caustic recycle from high-salt nuclear wastes using a ceramic-membrane salt-splitting process

Abstract An electrochemical salt-splitting process, based on sodium-ion selective ceramic membranes, is being developed to recover and recycle sodium hydroxide from high-salt radioactive tank wastes in the U.S. Department of Energy complex. The ceramic membranes are from a family of materials known as sodium (Na), super-ionic conductors (NaSICON). Two membrane compositions based on the rare-earth elements, neodymium and dysprosium, and a new proprietary material, NAS-D, have been fabricated as disks and are currently being tested with waste simulants. The membranes have been incorporated into a polyethylene scaffold for implementation into commercially available plate-and-frame electrochemical cells. A purified caustic product with a sodium hydroxide in excess of 3 M was produced from waste simulants with the Dy- and Nd-NaSICON membranes. This is the nominal concentration for onsite recycle and higher concentrations are expected. Membrane fouling was not observed, even though gibbsite {A1(OH) 3 } was precipitated in large amounts during some of the runs. Preliminary testing of the NAS-D material indicates that a sodium current density of 38 mA/cm 2 with a sodium current efficiency of approximately 90% is achievable over 1000 h of operation with an applied potential of 4.5 V.

Salt splitting using ceramic membranes

Abstract Inorganic ceramic membranes for salt splitting of radioactively contaminated sodium salt solutions are being developed for treating U.S. Department of Energy tank wastes. The process consists of electrochemical separation of sodium ions from the salt solution using sodium (Na) Super Ion Conductors (NASICON) membranes. In contrast to conventional organic-based bipolar or ion exchange membranes used in salt splitting, NaSICON membranes are resistant to gamma/beta radiation and are highly selective for sodium ions. Potential applications include 1) caustic recycle for sludge leaching, regeneration of ion exchange resins, inhibition of corrosion in carbon steel tanks, or retrieval of tank wastes; 2) pH adjustmet and reduction of competing cations to enhance cesium ion exchange processes; 3) sodium reduction in high-level waste sludges; and 4) sodium removal from acidic wastes to facilitate calcining. Initial experiments with dysprosium-based NaSICON membranes have demonstrated the feasibility of the ...

Irradiation of lithium zirconate pebble-bed in BEATRIX-II Phase II

Abstract BEATRIX-II was an in-situ tritium recovery experiment that was designed to characterize the behavior of lithium ceramics irradiated to a high burnup, and to assess their suitability for use in a fusion reactor blanket. This paper describes the results from the vented canister containing 29.47 g of lithium zirconate spheres packed in a bed 13.2 mm OD, 2.3 mm ID and 103 mm long. The enriched lithium spheres (85% 6Li) were irradiated to a burnup of 5.2% (total lithium) in a steep temperature profile −400°C edge, 1100°C center. The sweep gas was He-O.1% H2, with systematic tests using alternate compositions: He-0.01% H2 and pure He (maximum duration 8 days). Tritium recovery decreased slightly at lower H2 concentrations; for example, the buildup of inventory during a 4-day test in pure He was 0.8 Ci, approximately 6.5% of the tritium generated in the lithium zirconate during that period. The steadiness of the bed central temperature and the tritium release rate, together with low moisture release indicate good performance of the zirconate bed.

BEATRIX-II: A multinational solid breeder materials experiment

BEATRIX-II is an in situ tritium recovery experiment in the fast flux test facility (FFTF) reactor designed to characterize the feasibility of utilizing solid breeder materials at extended burnups in a fast neutron flux. Although not yet complete, the BEATRIX-II experiments have already substantiated that the solid breeder selected for ITER, Li 2 O, has good irradiation stability and tritium recovery. Temperature stability, lithium transport, dimensional stability and tritium recovery issues of Li 2 O up to 5% Li burnup were addressed in this experiment, Temperature gradients far more severe than in the ITER design, 400 to 1000°C, were found to be essentially unchanged by burnup and produced no observable instability, either from swelling or lithium vapor transport. Temperature change experiments illustrated that lithium inventories do not appear to increase as a result of irradiation to burnups of 5%.