VITRIFIED HLW AND SPENT FUEL-MANAGEMENT

The paper covers four topics: management of vitrified waste, managemen t of spent fuel, final disposal, and the repository safety assessment. At present, German spent fuel is reprocessed abroad and vitrified hig h-level radioactive waste will be returned for disposal. Interim stora ge of this waste in Germany will be necessary until the planned reposi tory at Gorleben becomes available. Two interim storage facilities hav e been built. Additionally, about 60 m(3) of HLLW (HAWC) produced at t he reprocessing plant in Karlsruhe prior to shutdown will be vitrified at the Pamela plant in Mol, Belgium, following plant adaptations and the installation of a new melter. Direct disposal of spent fuel is bei ng developed to technical maturity. A pilot conditioning and encapsula tion plant is under construction at Gorleben, and repository-related d emonstration tests are being performed. Layout and optimization studie s for a common repository for reprocessing waste and spent fuel are un derway, and a safeguards plan for spent fuel disposal has been develop ed. Results from these activities will be available early enough to be incorporated into the repository licensing procedure. The Gorleben sa lt dome has been selected for the construction and operation of a repo sitory for all types of radioactive waste, especially heat generating, such as vitrified waste and spent fuel elements. Radioactive waste wi th negligible heat generation will be emplaced in disposal rooms, heat -generating waste in vertical boreholes or in drifts. Definitive decis ions concerning the disposal plan, however, can be made only on the ba sis of the results of an underground investigation. The current concep tual design of the repository is based on modeling assumptions, since only aboveground investigations of the repository site have been compl eted. Underground exploration began in 1986 with the sinking of two sh afts. Proof of the safety of the repository for licensing requires a l ong-term safety analysis. To ensure long-term safety (individual limit s of 0.3 mSv/a, effective dose rate, and 0.9 mSv/a, organ dose rate), any possible release of radionuclides via the water path must be asses sed and the respective dose rates calculated. Experimental research is being carried out to characterize and understand the long-term physic o-chemical and geochemical behavior of the waste forms in the near fie ld of the repository and the aquatic chemistry of radionuclides in the near and far fields. Through this work, an experimental data base for quantitative long-term safety analyses will be acquired.