Lithium aluminate is a candidate tridunbreeding material tor fusion reactor blankets. One of the concerns with using LiA102 Is tritium recovery from this material, particularly at low operating temperatures and high fluences. The data froa various tritium release experiments with y-LiA102 and related materials are reviewed and analyzed to determine under what conditions bulk diffusion is the rate-Uniting mechanism for tritium transport and what the effective bulk diffusion coefficient should be. Steady-state and transient models based on bulk diffusion are developed and used to interpret the data. Design calculations are then performed with the verified models to determine the steady-state inventory and time to reach equilibrium for a full-scale fusion blanket.

Lithium oxide (Li2O) is a candidate material for solid breeder blankets in d-t fusion reactors. Radiation damage in Li2O was investigated in IEA BEATRIX-II phase 1 irradiation tests using the Fast Flux Test Facility (FFTF). Li2O single crystal specimens with various 6Li concentrations, 6Li/(6Li + 7Li), were irradiated at about 650 K for 300 effective full power days in FFTF by fast neutrons (the fast neutron fluence) ( > 0.1 MeV): 3.9 × 1026 n/m2). After the neutron-irradiation, measurements of electron-spin resonance (ESR) and optical absorption were carried out for the specimens at room temperature. From the measurements, colloidal lithium metal was found to be formed in Li2O irradiated with fast neutrons.

To verify the performance of permeation-resistant cladding for tritium targets designed for a New Production Reactor Light-Water Reactor, a tritium test facility was designed, developed, fabricated, and certified. Testing is ongoing to verify the performance of reference-designed targets. Accurate measurements were taken of tritium permeating from barrier-coated cladding specimens immersed in high-temperature autoclaves configured to simulate reactor coolant conditions. The tritium test pressure is controlled by heating a zirconium-alloy getter, previously charged with tritium, to a temperature that corresponds to a specified test pressure. The apparatus for testing deuterium permeation was developed to calibrate nondestructive testing procedures for evaluating barrier quality and to screen defective industrial cladding. These permeation testing facilities perform parametric tests to evaluate the sensitivity of permeation to temperature, time, pressure, fabrication variables, barrier disparities, corrosion, and other factors. The experimental activities characterize the performance and material properties of target rod components as well as validate new nondestructive examination methods that measure target rod quality. The target rod components are 1) barrier-coated stainless steel cladding, 2) lithium aluminate pellets, 3) nickel-plated Zircaloy-4 getters, and 4) zirconium liners. In addition, data generated from statistical testing provide increased confidence in current analytical models that predict target rod performance during both steady state and calculated transient conditions. The test results indicate that the tritium release from a full core of NPR-LWR targets will satisfy design requirements for release of no more than 20,000 Ci of tritium to the reactor coolant, even with four failed target rods that release up to 50% of their inventory.

A 2:1 mixture of LiF and BeF2 (FLIBE), is a potential tritium breeder material for fusion reactors, in particular, the Advanced Safe Pool Immersed Reactor (ASPIRE). A limited experimental campaign was conducted in an effort to test the postulates of the ASPIRE concept: namely, that MoF6 is effective in controlling the tritium species by maintaining the TF form and that MoF6 can serve as a source to plate out Mo on surfaces, thereby making the FLIBE system compatible with the corrosive TF. It was demonstrated experimentally that successive additions of MoF6 achieved quantitative (i.e., greater than 99.7%) conversion of H2 to HF. Thus, MoF6 is effective in controlling the tritium species. The degree of conversion of H2 to HF demonstrates that HF does not attack MO to form H2. This supports the postulate that the system is compatible with Mo. Thus, if it were possible to plate out and maintain a coating of Mo on all surfaces in contact with the FLIBE system, the ASPIRE concept could work. Thermodynamic calculations also confirmed that MoF6 should be capable of quantitatively (>99.9%) converting H2 to HF. There is both experimental and theoretical evidence that a number of MoFx species are present in both the gas phase and the FLIBE solution.

The tritium recovery system for the US ITER Li{sub 2}O/Be water cooled blanket processes two separate helium purge streams to recover tritium from the Li{sub 2}O zones and the Be zones of the blanket, to process the waste products, and to recirculate the helium back to the blanket. The components are selected to minimize the tritium inventory of the recovery system, and to minimize waste products. The system is robust to either an increase in the tritium release rate or to an in-leak of water in the purge system. Three major components were used to process these streams, first, 5A molecular sieves at {minus}196{degree}C separate hydrogen from the helium, second, a solid oxide electrolysis unit is used to reduce all molecular water, and third, a palladium/silver diffuser is used to ensure that only hydrogen (H{sub 2}, HT) species reach the cryogenic distillation unit. Other units are present to recover tritium from waste products but the three major components are the basis of the blanket tritium recovery system. 32 refs.

The Tritium Systems Test Assembly (TSTA) at Los Alamos, New Mexico, is a full-scale facility dedicated to testing tritium processing for fusion reactors. Adding a breeder blanket interface (BBI) to the TSTA is being studied. The BBI is to test the processing required for the tritium output streams for the various fusion-reactor breeder blankets. In the current phase of the study, the characteristics of the output from the various blanket types are being evaluated. Defining the output stream with respect to H/T ratio, impurity content, and radionuclide content is emphasized. Assessments of the solid breeder blanket (ceramic, Li2O) and the aqueous salt solution blanket are reported. For the aqueous lithium salt solution blanket, the chemical systems are very complex, involving a number of phenomena, including radiolysis, corrosion, and electrolysis. A system of about 0.01 ITER scale may be added to the TSTA. For the solid breeder blanket, no critical issues are identified. A system of about half-ITER scale, possibly even full scale, may be added to the TSTA

A breeder blanket interface for an aqueous lithium salt blanket is defined for TSTA. High calculated radiolysis rates result in a high overpressure in the blanket and the need for a depressurizer and a recombiner system. High projected corrosion rates for stainless steel and for beryllium result in high activity levels in the blanket and the possible dissolution of the beryllium balls. The required tritium processing systems are complex, involving seven separation steps. A flow sheet is presented of the needed tritium systems. The main processing units to recover tritium from the salt solution are a flash evaporator with condenser, a water distillation unit, and a vapor phase chemical exchange (VPCE) unit. The gas product stream from the blanket has an H/T ratio of 10/sup 5/ which requires a dedicated cryodistillation unit of very high capacity. This unit has a first column with a diameter of almost one meter to decrease the H/T ratio to 10.

A breeder blanket interface for an aqueous lithium salt blanket is defined for TSTA. High calculated radiolysis rates result in a high overpressure in the blanket and the need for a depressurizer and a recombiner system. High projected corrosion rates for stainless steel and for beryllium result in high activity levels in the blanket and the possible dissolution of the beryllium balls.The required tritium processing systems are complex, involving seven separation steps. A flow sheet is presented of the needed tritium systems. The main processing units to recover tritium from the salt solution are a flash evaporator with condenser, a water distillation unit, and a vapor phase chemical exchange (VPCE) unit.The gas product stream from the blanket has an H/T ratio of 105 which requires a dedicated cryodistillation unit of very high capacity. This unit has a first column with a diameter of almost one meter to decrease the H/T ratio to 10.

The requirements of tritium technology are centered in three main areas, (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The Tritium Systems Test Assembly (TSTA) now in operation at Los Alamos National Laboratory (LANL) is dedicated to developing and demonstrating the tritium technology for fuel processing and containment. TSTA is the only fusion fuel processing facility that can operate in a continuous closed-loop mode. The tritium throughput of TSTA is 1000 g/d. However, TSTA does not have a blanket interface system. The authors have initiated a study to define a Breeder Blanket Interface (BBIO) for TSTA. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. Various methods of tritium recovery from liquid lithium were assessed: yttrium gettering, permeation windows, and molten salt extraction. The authors' evaluation concluded that the best method was molten salt extraction.

(1988). The Blanket Interface to TSTA: Requirements for Liquid Lithium Blanket Processing Systems. Fusion Technology: Vol. 14, Proceedings of the Third Topical Meeting on Tritium Technology in Fission, Fusion and Isotopic Applications (Toronto, Ontario, Canada, May 1-6, 1988), pp. 657-662.

The second experiment of the LISA series in-situ tritium release tests concentrated on lithium orthosilicate (Li4SiO4) and, to meet the requirements of the KfK pebble-bed design for a NET blanket, especially on high density (97% TD) spheres and low temperatures. He + 0.1 vol% H2 sweep gas was used in the first cycle of LISA-2. The temperatures were varied between 600 and 350°C and the effects of flow rate and neutron flux changes were studied at different temperatures. For the high-density Li4SiO4 samples and for temperatures ≧ 500°C, the influx residence times are small ( ≦ 1 h) and in reasonable agreement with, but at 350°C remarkable higher ( ≈100 h), than observed in diffusion-controlled annealing studies. There are indications, that besides diffusion, other processes are involved and get dominant at lower temperatures. Inventories determined by out-of-flux annealing are in excellent agreement with estimations based on the difference between measured and steady-state influx release.

This paper discusses phenomena that have been observed during tritium extraction from y-lithium aluminate, specifically: Increase of rate of extraction when adding hydrogen to the sweep gas formation of tritiated water in all cases permeation of tritium through gas pipes adsorption of tritiated water on gas lines. To minimize the blanket tritium inventory a flowchart is proposed whose specificity rests in the addition of hydrogen to the gas within the blanket, followed by recovery of the tritium after oxidation of hydrogen to water, electrolysis and reconcentration. This flowchart includes a provision for detritiation of the coolant which is separate from the purge gas.

The LISA1 experiment is a test of in-situ tritium release. Conducted in the SILOE reactor at CEN Grenoble, the experiment uses the same facilities as the LILA1. The experiment has six capsules, four with Li 2 SiO 3 , one with Li 4 SiO 4 and one with LiAlO 2 . Each capsule is separately purged. The tritium activity is determined by ionization chambers and scintillation counting. An important difference as compared to LILA1 is the use of zinc beds to reduce tritiated water and thereby prevent sorption of T 2 O on the lines. Irradiation began on October 25, 1985 and was continued for three 3-week-cycles. The testing included systematic variation of four parameters: temperature (450 to 730°C), neutron flux (0.8 to 2.7 × 10 17 m −2 s −1 ), sweep gas flow rate (1.8 to 7.0 1/h), and sweep gas composition (He, He+0.1% H 2 , He+0.2% 0 2 ). Preliminary results are given.

The TRIO experiment involves the irradiation of γ-LiAlO2 under well defined conditions of temperature, neutron flux, sweep gas flow, and solid configuration. Tritium in its various chemical forms is moved by a sweep gas to an analytical train that measures and identifies the composition of the effluent. The predominant form of tritium observed in the sweep gas was the noncondensable, or HT form. At the completion of the experiment less than 0.1 wppm tritium was found in the solid.

Lithium aluminate is a candidate tritium-breeding material for fusion reactor blankets. One of the concerns with using LiAlO2 is tritium recovery from this material, particularly at low operating temperatures and high fluences. The data from various tritium release experiments with γ-LiAlO2 and related materials are reviewed and analyzed to determine under what conditions bulk diffusion is the rate-limiting mechanism for tritium transport and what the effective bulk diffusion coefficient should be. Steady-state and transient models based on bulk diffusion are developed and used to interpret the data. Design calculations are then performed with the verified models to determine the steady-state inventory and time to reach equilibrium for a full-scale fusion blanket.