atw - International Journal for Nuclear Power | 02.2020

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atw Vol. 65 (2020) | Issue 2 ı February 110 REPORT Backscatter Diffraction (EBSD) as well as neutron radio graphy. Finite- Element- Modelling is used to simulate new geometries and test conditions. Regarding the delayed hydride cracking (DHC) investigations, differently shaped Zircaloy-2 cladding tubes with and without initial axial and radial cracks are prepared and undergo the described testing methods. The goal of the work is to clarify the role of cladding toughness for the DHC behavior. Elmar W. Schweitzer from Framatome GmbH, Germany, gave a lecture on the “End-of-Reactor-Life State of Spent Nuclear Fuel as Major Input for Long Term Dry Storage Fuel Integrity Assessment” from a vendor’s point of view. Framatome as a manufacturer of nuclear power plants and has been delivering fuel assemblies for operation of the plants. The behavior of nuclear fuel under irradiation up to end-of-life (EOL) is a prerequisite for evaluating the additional damage permissible during the dry storage period. Limitation of temperature and hoop stress by the present design criteria is the best way to circumvent any issues arising from long-term storage of used fuel. Nevertheless, an exact knowledge of the EOL state of the fuel rods is necessary in order to assess effects related to hoop stress and cladding strain. Also, parts from the fuel assembly structure, e.g. guide tubes, spacer grids, water channels, fuel channels etc. start to raise interest, since these structures are important for a safe repacking of the spent fuel from the storage and transport cask into a disposal cask. Mechanical properties of irradiated cladding and fuel assembly components (fast neutron fluence, corrosion state) are necessary for transport evaluation of the spent fuel. The presentation of Dimitri Papaioannou from the European Commission Joint Research Centre (JRC) in Germany was titled “Experimental Studies on the Mechanical Stability of Spent Nuclear Fuel Rods”. He presented recent experimental results from the spent fuel studies at the JRC in Karlsruhe on safety issues associated to handling and transportation of nuclear fuel rods. In the experiments, a pressurized rod segment has been subject to dynamic impact and quasi-static three-pointbending tests. The devices are installed in a hot cell. The rod segment stemmed from a PWR fuel rod with burn up 67 GWd/tHM. A high-speed camera was used to record the impact test and thereby to determine the deflection and absorbed energy. In the threepoint-bending tests, the load, pressure and displacement were recorded and plotted. Post-test examinations were carried out to characterize the released mass upon rupturing in both experiments. The final goal of these investigations is to determine criteria and conditions governing the response of spent fuel rods to an external mechanical load in accident scenarios. Uwe Zencker from the Bundesanstalt für Materialforschung und -prüfung (BAM), Germany, gave a talk about “Brittle failure of spent fuel claddings during long-term dry interim storage”. The current research project BRUZL, which translates to fracturemechanical analysis of spent fuel claddings during long-term dry interim storage, has the general aim to develop risk assessment methods for potential brittle failure under mechanical loads after extended dry storage. The project foresees ring compression tests with unirradiated cladding samples with representative hydride distribution. Additional finite-element-analysis of the ring compression tests will include fracture-mechanical calcu lations, allowing failure analysis and the identification of failure criteria dependent on hydride distribution (density, orientation, and size), properties of cladding material, mechanical load, and temperature. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi). Another new research project was introduced by Benedict Bongartz from the University of Hannover, Germany. He gave a presentation on the “ Investigation of the temporal rearrangement behavior of zirconium hydride precipitates in interim and final storage”. Within this work, the specific experimental equipment and the required process technology is set up to load cladding tubes with hydrogen contents of up to 500 wppm. After the cladding tubes have been loaded with hydrogen, a combination of cooling and mecha nical stress application is planned in order to recreate and investigate the reorientation of the hydrides in a laboratory environment. The hydride precipitation in the zirconium cladding will be investigated with classical materials science investigations such as metallography, scanning and transmission electron microscopy and X-ray diffraction. Additionally, new investigation methods such as X-ray microscopy are envisaged to obtain new three- dimensional geometric data about the precipitates. In his talk, Marc Péridis from GRS gave an update on his work about “Temperature fields in a loaded spent fuel cask”. The temperature is a key parameter during dry storage since it governs most of the claddings aging mechanisms. As both, high and low temperatures are relevant for different effects, conservative models or a limited consideration only on the hottest fuel zone are insufficient for safety studies. Considerably more, it is necessary to carry out best-estimate calculations. A generic detailed cask model, inspired by the GNS CASTOR® V/19, was set up and used to calculate the temperature propagation from the inventories to the cask body with COBRA-SFS. The comparison of the results with similar models in COCOSYS and ANSYS CFX showed good agreement. Within the recent work, ParaView was introduced as a graphic interface to visualize the COBRA-SFS results. In the future, the COBRA-SFS model is intended to be used for transient calculations. This will enable the user to describe the temperature evolution during the drying process, which has an important impact on the material properties. In the second contribution about thermal modeling, Marta Galbán Barahona from ENUSA, Spain, re ported about her work progress with the presentation entitled “Analysis in Spent Nuclear Fuel Cask Using COBRA-SFS”. In comparison to the GRS work, ENUSA used the COBRA- SFS code to simulate a storage cask of the TN-24P type. The results obtained for the helium filled TN-24P cask were compared to measured temperature data. There was a particularly good agreement in the center of the fuel assembly, where the maximum temperature is located. In the peripheral assemblies, the maximum differences in temperature values were approximately 15 °C. Recently implemented post-process scripts allowed a simpler evaluation of the data with graphics and colored maps. As a result of the scripts, parameters such as helium flux could be analyzed, where an unusual flux distribution was found. Sensitivity studies have been performed to analyze the impact on the tem peratures. It was found, that the impact of the specific flux distribution was negligible. Francisco Feria from CIEMAT, Spain, gave a talk about the “Progress on the modeling of in-clad hydrogen behavior within FRAPCON-xt”. FRAPCONxt in its base version is a fuel performance code, which has been extended to simulate fuel rods under dry storage conditions. The code has been further developed to model the inclad hydride radial reorientation as a continuation of the modelling derived on hydrogen migration/precipitation. Moreover, an uncertainty quantification method has been adapted to predict the best estimate plus the corresponding Report Workshop on the “Safety of Extended Dry Storage of Spent Nuclear Fuel” – SEDS 2019 ı Florian Rowold, Klemens Hummelsheim and Maik Stuke
atw Vol. 65 (2020) | Issue 2 ı February uncertainty. Recent results allowed the conclusion that realistic scenarios prevent the formation of radial hydrides, whereas giving credit to limiting conditions would not allow ruling out this degrading mechanism. Depending on experimental data made available, further work is foreseen to validate the modelling with more representative data based on irradiated material and to derive a technological limit concerning the cladding embrittlement due to radial hydrides formed. To further address and support thermo-mechanical activities, Felix Bold from GRS held a presentation called “Proposal of a Benchmark Describing the Thermo-Mechanical behavior During Dry Storage”, wherein he invited all interested parties, who are willing to improve their modeling experience and to share their knowledge about the extended storage of spent nuclear fuel. The goal of this benchmark is the prediction and the code-to-code comparison of the thermo-mechanical parameters such as cladding temperature, hoop stress and strain as well as the hydrogen and hydride behavior during the storage period. For the starting conditions it is planned to use fuel rod data or output of a fuel performance code capable of simulating the fuel rod state at the end of operation. This will provide rod and pellet geometry, corrosion, internal gas state and the initial hydrogen load. The transient conditions will include the change of environment from water cooling in the spent fuel pool to helium atmosphere in the cask as well as the temperature changes during reactor shut down and the drying process. The results of the benchmark will be published and presented in 2020 on the 4th GRS workshop. With his talk about the “Long-term evaluation of sealing systems for radioactive waste packages”, Matthias Jaunich provided a round-up of another important research area addressed by the BAM. The work performed for many years now, aims at understanding the long-term behavior of the sealing systems during possible extended storage and sub sequent transportation scenarios. It comprises accelerated aging tests on metallic and elastomeric seals and covers experimental investigations to get a database on the component/ material behavior. Based on the results, appropriate analytical descriptions and models were developed. For the metal seals, a linear logarithmic correlation and an extrapolation of the remaining seal force and useable resilience for up to 100 years seems possible, but the question about the confidence range has yet to be answered. The elastomer seals exhibited hardness increase and sealing force decrease during the aging test. Deriving from the tests, the researchers were able to present an approach for a lifetime prediction of the seals. With his presentation about “Radio nuclides present at inner PWR fuel rod segment Zircaloy cladding surfaces in the context of safety of extended dry storage of spent nuclear fuel”, Michel Herm from the Karlsruhe Institute of Technology (KIT) shifted the focus to another interesting issue. Beside the often-discussed hydrogen effects, the fuel rod cladding could also be affected by various other processes during reactor operation and beyond. Precipitates of fission or activation products, e.g. Cs, I, Te, and Cl, present at the fuel-cladding interface, possibly exhibit corrosive properties and thus affect the integrity of the cladding. Therefore, irradiated pressurized water reactor UO 2 and MOX fuel rod segments were prepared and examined. The composition of agglomerates found on the inner surfaces of the plenum area and in fuel-cladding interaction layers were analyzed by means of SEM-EDS/WDS, XPS, and synchrotron radiation- based techniques. In addition, the present radionuclide inventory was compared to calculated values using a MCNP/ CINDER approach. In a second contribution from the KIT, Mirco Große reported on the “ Experimental Simulation of Long- Term Dry Storage in the QUENCH Facility of KIT – Availabilities and Plans”. The QUENCH facility at KIT is dedicated for tests simulating design basis and beyond design basis accidents in light water reactors on a fuel rod bundle scale. However, this test bundle can also be used for long-term cooling experiments simulating dry storage conditions. A description of the facility and reports about the planned experiments within the framework of the collaborative research project SPIZWURZ was given. The project investigates the behavior of cladding materials under typical long-term storage conditions. The experimental bundle consists of 21 to 31 fuel rod simulators with Zircaloy-4, ZIRLO and Dx/D4 Duplex claddings. The rods are electrically heated and can be pressurized separately. The inner pressure will be up to 5 MPa. The test will start with a maximum cladding temperature of 500 °C at the hottest bundle position and the temperature will be reduced by 1 K/day during a period of 8 month. The axial and radial hydrogen distribution will be measured post-test by neutron imaging methods. Metallographic investigations will be used to determine the change in the hydride orientation. Michael Wagner from the Technical University of Dresden, Germany, closed the workshop with the last talk about the “Investigations on potential methods for the long-term monitoring of the state of fuel elements in dry storage casks: recent results”. Four non-invasive measuring methods were assessed regarding their suitability for the condition monitoring of the cask inventory by means of simulations and experiments. For this purpose, damage scenarios of the cask inventory were assumed in a CASTOR V/19. The identified scenarios based on investigations on damage mechanisms. The simulations and experiments showed that the measurement of neutron and gamma radiation fields and muon imaging have the greatest potential as monitoring methods. These will also be further investigated in a follow-up project. In principle, the acoustic methods have a high informative value, but the transfer of experimental results to real con ditions is difficult. Thermography showed a low practicality as a monitoring method due to its limited expressiveness. The 2019 SEDS workshop showed that the topic of extended storage of spent nuclear fuel with all its different aspects is continuing to draw a large interest in the scientific landscape. In fact, the efforts in terms of research projects and collaborations increased in the recent past. Especially for Germany and European countries, where very high burnup and mixed oxide fuels were used and dry storage in casks is a preferred option for the spent fuel, this is a positive sign, since not all knowledge gaps are answered yet and require further work. The annual workshop “Safety of Extended Dry Storage of Spent Nuclear Fuel” established itself as a place to address those knowledge gaps and exchange information in a broad community on a very scientific level. This year the 4th workshop will be held again as a three-day event at GRS in Garching during the first week of June 2020. Authors Florian Rowold Klemens Hummelsheim Maik Stuke Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH Bereich Stilllegung und Entsorgung Abteilung Stilllegung und Zwischenlagerung Forschungszentrum, Boltzmannstr. 14, 85748 Garching b. München 111 REPORT Report Workshop on the “Safety of Extended Dry Storage of Spent Nuclear Fuel” – SEDS 2019 ı Florian Rowold, Klemens Hummelsheim and Maik Stuke
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atw - International Journal for Nuclear Power | 02.2020

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