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.
Status of the Development of Final Disposal Casks and Prospects in Germany
In Germany, progress has been made in the development of a cask system for final disposal of spent fuel. The requirements for a final disposal cask and the planning requirements for a final repository are discussed, and a cask system is described that meets these requirements. Besides the description of the cask system and the planned inventory, the results of the shielding, critical-ity, and thermal calculations are described, as well as the results of the drop tests that have been performed with the cask. As a rther development, a consolidated spent-fuel canister is described.
Nondestructive Assay of Nuclear Low-Enriched Uranium Spent Fuels for Burnup Credit Application
Abstract Criticality safety analysis devoted to spent-fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent-fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as burnup credit. To be used, burnup credit involves obtaining evidence of the reactivity loss with a burnup measurement. Many nondestructive assays (NDA) based on neutron as well as on gamma-ray emissions are devoted to spent-fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously, and the link to burnup is a power link. As a result, burnup determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels, but the convenience of the gamma spectrometric methods is very dependent on the characteristics of the spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several nuclear power plants in western Europe, gives the average burnup within a 5% uncertainty and also the extremity burnup. 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (active neutron interrogation, passive neutron counting, and gamma spectrometry).