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Lithium-bearing ceramics have been proposed for tritium-breeding blankets in D-T fusion reactors.The conceptual designs for these blankets utilize a solid breeder (e.g., LLAlO* or IJ^0)* a helium purge stream for tritium recovery, a coolant (e.g., ff-O or He), and a structural material (e.g., HT-9 ferritic steel or a titanium modified stainless steel) to separate the coolant from the breeder. Tritium breeding, release rate from the breeder, inventory, and leakage to the primary coolant have been identified as areas of concern in these blanket designs.In this work, models are developed to describe the percolation of released tritium through the breeder Interconnected porosity to the purge stream, convection of tritium by the helium purge stream, and leakage or permeation of tritium through the structural material to the primary coolant system. Important parameters in the models are tritium generation rate, breeder microstructure, tritium species in the gas phase, temperatures, tritium diffusivities and permeabilities, and effectiveness of oxide barriers.Tritium desorbs from the solid breeder in molecular form as either T~, HT, T2O, HTO, or some combination of these species. The form depends to a large extent on the chemistry of the purge stream and the amount of oxygen which diffuses from the breeder to the pore-solid surface area. A pure helium purge stream favors T 2 and/or T 2 0, while addition of enough hydrogen (i.e., protium) to the helium purge to enhance the surface desorption process favors the release of HT and/or HTO.The results of in-reactor and out-c c -reactor studies are reviewed to identify the forms of released tritium as a function of purge and breeder chemistry. While the gas phase transport of tritium due to a number of mechanisms is discussed, diffusion is identified as the rate controlling mechanism in the percolation of the tritium species from the pore/solid surfaces to the purge flow. Models are developed for ordinary diffusion, transition diffusion, and Knudsen diffusion. Once the tritium species reach the helium purge streams, convection is shown to be the dominant mode of transport.The diffusion and convection models establish the tritium partial pressures throughout the breeder porosity as a function of tritium generation rate per unit volume, purge channel location and spacing, and purge flow temperature and pressure. The partial pressures of tritium are then used as input to the permeation model to predict the leakage rate* Critical parameters in the permeation model are the permeability of the structural material, effectiveness of oxide barriers (I.e., oxide impedence factor), and the ratio of unoxldlzed tritium-to-protium on the breeder side of the structural material. *Work supported by the U.S. Department of "Energy/Office" of Fusion Energy.

The Department of Energy (DOE) has established guidelines for qualifications and training of the technical experts preparing and reviewing the safety analysis reports for packaging (SARP) and transportation of radioactive materials. One of the qualifications is working knowledge of, and familiarity with the quality assurance (QA) requirements in Subpart H of Title 10 of the Code of Federal Regulations Part 71. DOE is sponsoring a course on quality assurance for radioactive material transportation packaging. The objective of this paper is to describe the salient features of the course, the purpose of which is to provide QA training and practical experience that are required to develop and implement a QA plan or prepare the QA chapter of a SARP for the design, fabrication, assembly, testing, maintenance, repair, modification, and use of the packaging. The applicable QA requirements from DOE orders, federal regulations, and NRC regulatory guides will be highlighted, along with a graded approach to selected QA elements from Subpart H of 10 CFR Part 71. The paper will also briefly discuss ASME NQA-1 for Type B and fissile material packaging, current issues resulting from the different emphasis between a compliance-based QA program (in Subpart H, 10 CFR 71) for packaging and a performance-based QA program for DOE nuclear facilities (based on 10 CFR 830, “Nuclear Safety Management”), and the final rule changes in 10 CFR 71 that became effective on October 1, 2004.

Solid breeder/structure mechanical Interaction (BSMI) during fusion reactor blanket operation Is a potential failure mode which could Halt the Lifetime of the blanket. The severity of BSMI will generally depend on the naterlals, specific blanket designs, and blanket operating conditions. Thenaooechanlcal analyses performed for « hellua-cooled blanket employIng Li 2 O/ru-9 plates Indicate that BSMI could be a serlcus concern for this blanket.

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