VITRIFIED HLW AND SPENT FUEL-MANAGEMENT

The paper covers four topics: management of vitrified waste, management of spent fuel, final disposal, and the repository safety assessment. At present, German spent fuel is reprocessed abroad and vitrified high-level radioactive waste will be returned for disposal. Interim storage of this waste in Germany will be necessary until the planned repository at Gorleben becomes available. Two interim storage facilities have been built. Additionally, about 60 m(3) of HLLW (HAWC) produced at the 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 being developed to technical maturity. A pilot conditioning and encapsulation plant is under construction at Gorleben, and repository-related demonstration tests are being performed. Layout and optimization studies for a common repository for reprocessing waste and spent fuel are underway, and a safeguards plan for spent fuel disposal has been developed. Results from these activities will be available early enough to be incorporated into the repository licensing procedure. The Gorleben salt dome has been selected for the construction and operation of a repository for all types of radioactive waste, especially heat generating, such as vitrified waste and spent fuel elements. Radioactive waste with negligible heat generation will be emplaced in disposal rooms, heat-generating waste in vertical boreholes or in drifts. Definitive decisions concerning the disposal plan, however, can be made only on the basis of the results of an underground investigation. The current conceptual design of the repository is based on modeling assumptions, since only aboveground investigations of the repository site have been completed. Underground exploration began in 1986 with the sinking of two shafts. Proof of the safety of the repository for licensing requires a long-term safety analysis. To ensure long-term safety (individual limits 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 assessed and the respective dose rates calculated. Experimental research is being carried out to characterize and understand the long-term physico-chemical and geochemical behavior of the waste forms in the near field 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.