Volume 85, Issue 6

Thermo-mechanical performance of fuel rods is a key aspect for assessing the spent fuel safety during dry storage. There is a need of assessing fuel rod behaviour codes for dry storage stage and uncertainty quantification can be a useful tool for this assessment. As a first step of these studies, sensitivity analyses can help to understand the effects and to identify the relevant parameters for the predictions. The present work studies the effect of the scenario on sensitivity analyses focused on the cladding stress at the beginning of dry storage. To make this comparison 10 design parameter’s tolerances and 7 model biases have been studied for 2 different fuel rods in terms of design and irradiation. Cladding diameter tolerance and fuel gas release and thermal conductivity model biases are the more influencing parameters independently of the scenario chosen. Overall, moderate differences between scenarios have been obtained. Results obtained in this work are specifically related to the modelling of the code FRAPCON-xt.

According to the Swiss waste management policy, spent nuclear fuel is planned to be disposed in a deep underground repository. Prior to final disposal, spent nuclear fuel is stored at reactor sites and in a centralized dry storage facility. Since the operation dates of the final repository are unknown, extended periods of interim storage have to be considered. A research program to investigate the fuel rod integrity during long-term dry storage has recently been launched at Paul Scherrer Institute. In the context of the project, fuel rod performance simulations will be carried out with the code Falcon. Until now, an extensive literature survey concerning current trends in dry storage modelling has been written and first developments of Falcon’s capabilities towards dry storage modeling have been made. In this work, Falcon has been modified by implementing a long-term cladding creep model. The original and upgraded versions of Falcon have been used to simulate a demonstration case consisting of the base irradiation, wet storage, drying and dry storage. Results obtained with both versions have shown an important difference in the cladding hoop creep which justifies implementation of the new creep model.

One of the key aspects for the safety of extended dry storage of used nuclear fuel in casks is the prediction of the fuel rod behavior. A detailed and reliable prediction has to take into account the entire life span of the fuel rod: Service in reactor with the associated radiation history, wet storage in the fuel pool, loading and drying of the cask, and finally the in-cask dry storage itself. Enhancing existing fuel performance codes to include all these steps and the associated phenomena is an ongoing task tackled by several groups. To foster the communication between these groups a benchmark was proposed at the 2019 Safety of Extended Dry Storage (SEDS) workshop. The article at hand summarizes some of the preliminary results which were presented at the 2020 SEDS workshop.

Within the scope of the pending search for a final storage site location for highly radioactive waste such as used fuel assemblies, the cladding integrity in case of a relocation of the rod from an interim storage to a final storage cask remains uncertain. In order to assess the potential reduction of the rod integrity by hydride precipitation and reorientation, a study of the long-term behaviour of hydrides in Zircaloy 2 is carried out. The focus is on simulation and analysis of the precipitation behaviour under different load conditions such as starting temperatures of 673 K, hoop stresses of up to 150 MPa and cooling rates between 4 and 1 K/h. The analytical methods comprised microscopic investigations as well as X-ray diffraction analyses and first results from imaging with X-rays.

The investigation of the long-term performance of sealing systems employed in containers of radioactive waste and spent nuclear fuel is one research focus area for division 3.4 \Safety of Storage Containers" at the Bundesanstalt für Materialforschung und -prüfung. The behaviour of metallic seals, which are employed in interim storage casks for spent nuclear fuel and high active waste, over time is of high importance for ensuring safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal performance and the main part is focussed on the time and temperature dependent behaviour. Our aim is to understand the longterm behaviour of the sealing systems for evaluation of their performance during possible extended interim storage and subsequent transportation. In this contribution an overview of the performed tests and their respective results is given. A focus lies on a comparison between seals with large torus diameter in comparison to small diameters.

A key parameter to predict the integrity of the fuel rod cladding during dry storage is the temperature. Almost all degradation mechanisms are temperature dependent. A realistic prediction of the temperature field at a given time for the loaded cask is needed to investigate predictions for long-term degradation mechanisms properly. Overly conservative temperature estimates with the purpose to assure peak cladding temperatures within technical specifications might lead to non-justified, even counterproductive conclusions. In this work, we present two modelling approaches used to calculate temperature fields of loaded casks. One approach is a detailed numerical simulation via a specific code, and the second is a semi-analytical approach focusing on a faster and feasible calculation. Steadystate and transient calculations are compared and discussed for both modelling approaches.

Tomographic Gamma Scanning (TGS) is a nondestructive analysis technique for accurately measuring the content of radioactive material in sealed containers. It has important applications in the fields of nuclear safety assurance and the measurement and analysis of radioactive waste drums. As the key of TGS technique, the quality of the emission image reconstruction algorithm will directly affect the accuracy of measurement results. In this paper, a computer simulation platform for TGS technique is established. The application of four emission image reconstruction algorithms in TGS technique is systematically studied by using 6 · 6 · 6 sample voxel model. Three representative sample media and radioactive source distribution models are established and the applicability of the four reconstruction algorithms is analyzed and evaluated in the presence and absence of detector noise interference. Computer simulation and experimental results show that ML-EM algorithm is the most applicable in the process of TGS image reconstruction.

If a radiation source is lost, it will do harm to human being, especially for the searching person. In order to find the lost radiation source easily, a detector system, including 3 detectors, was designed to search the lost c-ray radiation source autonomously in any environment. First, based on GEANT4 simulation, the radiation dose rates in 3 Geiger Muller (GM) counters were simulated at the different source-detector distances, distances between detectors and angles. Various analyses were performed by experimentally and verified the simulated detector designed system. A Mono-energetic 137Cs c-ray source with energy 662 keV and activity of 1.11 GBq was used for the observation. The simulated results were compared with the experimental dose rate values and obtained good agreements for various cases. Only based on the dose rates in three detectors, the radiation source was searched in a different location at a specific source activity and angle. The corresponding angles of deviation and detection limit were calculated for sensitivity and ability of detector designed system. The proposed detector designed system can be used to navigate radiation sources in the low level and high radiation environments.