EURAD - D8.1 State-of-the-art report
The first task of WP8 is dedicated to the state-of-the-art (SOTA) report and related know-how transfer and distribution. The SOTA report is the first product of the SFC research programme, which offers an overview of the status of knowledge in the field of spent nuclear fuel characterisation and assessment during the pre-disposal phase. The document aims to focus on the current safety-significant gaps and related challenges, providing a direct link to the goals of the mandated actors of EURAD. The report is expected to be used by all EURAD colleagues in their national programmes. However, this SOTA report is intended as an initial version, to be updated at the end of the WP to become the final SOTA report. The aim of the final report is to become a key reference in the field and to gain high recognition and visibility as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art.
Depending on individual back-end country strategies, being programme-based within the framework of their national strategy, spent nuclear fuel (SNF) can be destined for direct geological disposal, for reprocessing or for long-term interim storage followed by disposal. For all cases, a proper characterisation of the spent fuel is required.
A state-of-the-art review on characterisation of SNF properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) using several numerical and experimental approaches and methodologies is presented.
The ability to reliably predict spent nuclear fuel composition and SNF properties, namely radionuclide inventory and source term, is relevant for both operational and long-term safety assessment in geological disposal, as well as for disposal cost key factors, and relies on ad-hoc calculation schemes.
The calculations require a particle transport code coupled with a depletion solver. A large proportion of the available depletion codes has been reviewed in this work. However, since the results also depend on nuclear data and operational data as well as assay data, nuclear data libraries and uncertainty aspects have also been discussed. More specifically, the uncertainty of the fuel inventory can be dominated by several factors: irradiation history of the fuel, exact composition of the fresh fuel/cladding especially the level of impurities, and the large heterogeneities in the fuel design discharged from reactors, as well as modelling limitations, nuclear data libraries and reactor core characteristics such as, e.g., void fractions (in BWR's) and mechanical changes of fuel during irradiation. Therefore, the treatment of all related uncertainties (quantification and propagation) has been part of this review. The availability of experimental data enables testing and validation of codes and models in order to understand how closely the models replicate reality. The SOTA report also focuses on experimental verification techniques, i.e. Non-Destructive Analysis (NDA) methods (e.g. neutron and gamma spectrometry, gamma tomography, calorimetry, etc.) and Destructive Analysis (DA). Other non-conventional techniques (e.g. muon tomography) are also reviewed.
Furthermore, when considering operational safety cases for the surface facilities where the fuel must be encapsulated in special disposal canisters, studies and research activities are required to assess spent fuel performance as well as developing concepts for handling of consequence scenarios. Therefore, another part of the SOTA report reviews experimental campaigns developed to investigate fuel integrity. Conventional techniques for investigating the effect of hydrogen load, hydride distribution and fuel/cladding interaction, mechanical performance of the cladding and cladding integrity, deterioration of the mechanical properties of the cladding material resulting from Delayed Hydride Cracking (DHC) for high burnup are also reviewed, all with the focus on (extended) dry storage conditions.
Finally, the status of knowledge on definition and screening of accident scenarios with respect to pre-disposal activities, such as fuel packaging and transport, is provided on the basis of IAEA general assumptions.
