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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Roger Bonniaud
Nuclear Technology | Volume 34 | Number 3 | August 1977 | Pages 449-460
Technical Paper | Radioactive Waste | doi.org/10.13182/NT77-A31810
Articles are hosted by Taylor and Francis Online.
Les solutions concentrées de produits de fission, résidus du retraitement des combustibles irradiés, constituent une des sources les plus importantes des déchets produits par l’industrie nucléaire. Leur solidification par vitrification est la solution actuellement retenue par la France. Les procédés de solidifications, qui ont été développés en France sont les suivants: un procede de vitrification en pot et un procédé continu qui combine un calcinateur rotatif à un four de fusion de verre. En général, les verres utilisés en France sont à structure silicieuse, structure plus stable que la structure phosphatique. Le problème de la séparation de phases molybdiques dans les verres à forte concentration en produits de fission est resolu par l’augmentation de la concentration en B2O3 Pour le cas des solutions issues du combustible U-Mo, où la concentration en molybdène atteint 100 g ℓ−1 , l’addition d’alumine réduit la séparation de phase, conduisant à un solide à forte phase vitreuse. Des mesures des taux de lixiviation à l’eau naturelle avaient montré des résultats qui s’étendaient de 10−1 g cm−2 jour−1 à 10−6 g cm−2 jour−1 pour les compositions silico-boratées. Pour étudier le problème posé par les emetteurs α dans les verres, des blocs de verre étaient dopes avec différents emetteurs α qui, en 1 an développaient des energies equivalents aux energies libérées pendant 10 à 1000 ans de stockage des verres réels. The current approach in France to the problem of waste management of the solutions of concentrated fission products from fuel reprocessing is solidification by vitrification. The solidification processes utilized are a batch process, in-pot vitrification, and a continuous process, a rotary kiln calciner. Generally, a silicate glass that is more stable than the phosphate is used. The problem of separation of a molybdate phase at high fission product concentration is reduced by an increase in B2O3 concentration. For wastes from a U-Mo fuel that contains 100 g/ℓ of molybdenum, phase separation is reduced by addition of Al2O3. Leach rates for the borosilicate glass were studied, and rates of 10−8 to 10−6 g/cm2 per day were measured as a function of Na2O concentration. Plutonium leach rates were measured as a function of Na2O concentration. Plutonium leach rates were measured as 10−8 to 10−7 g/cm2 per day and 241Am was 8 × 10−9 g/cm2 per day after 110 days of leaching. Stability of the glasses to alpha-particle radiation damage was simulated for a storage period of 10 to 1000 yr. These samples indicated only a slight change in viscosity as a result of these simulation studies.