Calculation of Dose Rates at the Surface of Storage Containers for High‐Level Radioactive Waste
Abstract In several countries, the high‐level radioactive waste that will be disposed of in deep geological formations has to be retrievable for a certain time. Since 2010, retrievability is required for the operation phase of a repository also in Germany. Depending on the effort and the feasibility of remote handling, a certain exposure of the involved employees to ionising radiation is caused. The estimation of the exposure requires the knowledge of the inventory of radionuclides and the radiation field around the storage containers. This paper focuses on German concepts for final storage casks for the drift emplacement in rock salt for both spent fuel rods and high‐active waste (HAW) coquilles. Calculations of dose rates at the surfaces of the casks are presented. The calculations show that the gamma radiation of fission and activation products can be efficiently shielded by materials like cast iron or low‐alloyed steel. Ductile cast iron, however, has a positive influence on the neutron moderation because of the high carbon content. The shielding of the neutron radiation strongly depends on the quantity and position of the polyethylene (PE) rods which are used as neutron moderator. PE has unfavourable features at elevated temperatures that can be reached in a repository for high‐level waste. The simulations show that TiH2 is a promising alternative material for the shielding of neutron radiation. The current assumption that a final storage cask for HAW coquilles has the same outer dimensions as the cask for spent fuel rods will lead to unacceptable dose rates at the side wall. An increase of the diameter is necessary to provide sufficient shielding. Graphite insets in the cask interior lead to a considerably lowered dose rate.
Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague
The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined — either by direct measurements or by involving established correlations — and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague.Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany.Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as 14C, 36Cl, 79Se, and 129I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries — Belgium, France, Germany, Japan, Netherlands, and Switzerland — an international effort can be recommended.Copyright © 2013 by ASME
Dosage du molybdene 93 et du zirconium 93 dans les effluents de retraitement
L'objectif de l'etude presentee consiste a elaborer des methodes de dosage de deux radioelements a vie longue, le zirconium 93 et le molybdene 93, dans les effluents de retraitement. Le #9#3zr ayant une periode tres longue (1,5. 10#6 ans), la mesure de cet isotope s'avere plus performante par spectrometrie de masse a couplage plasma induit (icp/ms) que par les mesures radiochimiques classiques. Le #9#3mo se desintegre avec une periode radioactive de 4000 ans par capture electronique, et peut etre detecte par mesure radiochimique, en particulier par spectrometrie x. Cependant, ces radioelements ne peuvent etre mesures directement sur l'effluent et necessitent une separation chimique prealable, qui elimine les radioelements susceptibles d'interferer et ceux responsables de la radioactivite de l'effluent. La separation est effectuee a l'aide de techniques potentiellement tres selectives, utilisant des molecules extractantes specifiques de mo et de zr : les extractions par solvant, par membrane liquide emulsionnee et par chromatographie liquide sur resine impregnee par solvant. Des etudes realisees au cea/cadarache, sur une solution synthetique simulant un effluent de retraitement, ont permis de demontrer les proprietes selectives de l'-benzoinoxime, extractant de mo, et du phosphate tributylique, extractant de zr, utilises selon les trois techniques etudiees. Les conditions optimales d'extraction et de separation ont ete atteintes a l'aide d'une methodologie de plans d'experiences. Enfin, les methodes ont ete testees sur effluents reels a cogema/la hague. Le #9#3zr a pu ainsi etre mesure a hauteur de 2. 10#6 bq/l dans une solution de dissolution et de 10#7 bq/l dans une solution de produits de fission. De meme, l'activite en #9#3mo mesuree dans la solution de dissolution s'eleve a 500 bq/l. Ces methodes ont permis en outre de determiner, pour les elements mo et zr contenus dans les effluents, les compositions isotopiques, qui concordent parfaitement avec celles issues du code de calcul cesar.