In the Belgian radioactive waste management programme, low and intermediate level waste short lived (LLW & ILW SL) are intended to be disposed-off in a near surface disposal facility (called cAt) located in Mol and Dessel, two localities situated in the north of Belgium. The major part of this LLW & ILW SL is produced by the Belgian NPP’s (located in the localities Tihange and Doel), the Belgian nuclear research center (SCK CEN) and the institute for the production of radioisotopes (IRE). The management of radioactive waste is centralized by the Belgian waste management organization (WMO) ONDRAF-NIRAS. This organization has an acceptance system that allows an in-depth control of the compliance of the waste with several waste acceptance criteria (WAC), from the production of the waste to its interim storage in facilities located in Dessel. The WAC have evolved these last years (e.g. considering return of experience and the licensing of cAt). For instance, some new WAC have been defined and some existing WAC, becoming more stringent, revised. It is therefore necessary to assess whether waste packages which have been accepted in the past by ONDRAF-NIRAS based on the previous WAC still comply with the new WAC. A project has been started by the WMO in collaboration with FANC & Bel V (forming together the Belgian nuclear regulatory body) to assess the compliance of LLW & ILW SL accepted by ONDRAF-NIRAS in the past with the current WAC related to cAt. The project aims also at establishing a strategy for demonstrating the compliance using non-destructive and destructive tests (NDT/DT), in complement to the evaluation of the historical characterization data. Finally, the scope of the project includes also the definition of a strategy for demonstrating the compliance to WAC of waste that will be produced in the future using for instance a strengthened control programme at the source (as closed as to the waste production step).
How to cite: Sghir, R., Mertens, J., and Detilleux, V.: Control of the eligibility of waste packages to the Belgian near surface disposal for short lived LL&ILW, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-62, https://doi.org/10.5194/safend2025-62, 2025.
The former iron ore mine Konrad near Salzgitter is the approved final repository for radioactive waste with negligible heat production in Germany and should go into operation in the 2030s. The waste comes from operation and decommissioning of nuclear power plants, as well as from medicine, research and industry using radiation sources or radioactive materials. These types of waste are comparable to international low- and certain types of intermediate-level waste. The waste includes a large variety of products, materials and properties. The containers for such waste have the primary purpose of enabling handling and storing of the waste by safely enclosing the radioactive inventory and shielding of radiation. Due to the large variety of waste forms to be disposed of in the Konrad repository, different container types are specified to account for the different requirements related to each waste type.
Each container design for waste disposal in Konrad has to be approved by BGE on the basis of the “final disposal conditions” (Endlagerungsbedingungen) and the respective “product control” (Produktkontrolle). For demonstrating the container’s compliance with the regulatory framework, the waste producers or container manufacturers have to apply for design approval on the basis of a comprehensive safety assessment including all relevant reports about analytical, numerical and experimental safety demonstrations and quality assurance measures for container manufacturing and operation. Usually, BGE commissions independent experts like BAM to evaluate these documents and the design testing (e.g. drop or fire tests).
While the basic assessment principle is to verify the container design and quality to be in line with all regulatory requirements, there are often challenges associated with the container’s safety evaluation. As containers of different types have already been produced in large numbers and partially loaded with waste before the regulatory framework became effective (so-called “old” containers), they are considered in the regulations as well and safety assessments have to be provided in an equivalent way. As the requirements for the documentation of manufacturing quality were not yet known at the time these containers were manufactured, this is often challenging in practice. On the other hand, containers requiring the highest safety level (ABK II, sf) due to their nuclear content sometimes cause significant challenges and very long approval processes concerning safety demonstrations for severe accidental conditions like a 5 m drop without impact limiter onto a nearly unyielding target or a one hour 800 °C fire scenario. Furthermore, as the regulatory framework has not been updated for several decades, some requirements and the respective safety assessment methods have to be interpreted under consideration of the current state of knowledge in science and technology.
How to cite: Kömmling, A. and Völzke, H.: Experiences and perspectives for the design evaluation of Konrad containers, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-38, https://doi.org/10.5194/safend2025-38, 2025.
The waste acceptance criteria for the final repository Konrad are defined in the Konrad waste acceptance requirements (status: December 2014, SE-IB-29/08-REV-2). The requirements are derived on the basis of the results of a site-specific safety analysis. They include general requirements for waste containers, waste packages as well as specific requirements for the waste product and the activity limitations for individual radionuclides and mass limitations for non-radioactive hazardous substances.
Depending on the activity inventory of the waste packages, higher-quality packaging may be required. In this case, the owners of radioactive waste must provide evidence that a possible release of radionuclides during the postulated accidents is limited down to permissible values.
As part of the "plan approval decision for the construction and operation of the Konrad mine in Salzgitter", the underground fire of a transport vehicle was defined as design basis for the thermal accident to determine the release, which is modelled by a corresponding time-flame temperature model curve.
According to the Konrad waste acceptance requirements and the applicable implementing provisions (product control of radioactive waste, radiological aspects, SE-IB-30/08-REV-1), both experimental and computational considerations are permitted for verification procedures, provided that the equivalence of the thermal test with the postulated accident is demonstrated regarding heat input and spatio-temporal temperature profile.
In addition to the caloric properties of packaged waste products, the material composition and structure of the waste container and its installations have an influence on the heat transfer and thus on the temperature reached in the waste product. The release behaviour at a given temperature depends on the chemical/physical composition of the waste product. Covering heat penetration calculations lead to high calculated temperatures in the waste product, therefore restrictions are necessary for the properties of the waste product to be packaged. Nevertheless, there remain degrees of freedom in the packaging strategy. On the other hand, if the specific packaging variant is used as basis, more accurate heat penetration calculations can be carried out, which lead to lower temperatures. Thus a sufficiently low release of radionuclides can be demonstrated for a large number of waste products. This presentation compares different approaches of the numerical verification to suggest the waste owners to choose the most efficient packaging strategy for their waste products.
How to cite: Glindkamp, A. and Alraun, S.: Strategies for numerical verification of nuclear waste packages of compliance with the permissible activity release during a postulated damaging fire , Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-58, https://doi.org/10.5194/safend2025-58, 2025.
Nagra, the National Cooperative for the Disposal of Radioactive Waste, holds the national mandate to plan, construct and operate a deep geological repository for the Swiss radioactive waste. The current realisation plan as part of the Swiss waste management plan foresees the start of operation for disposal of low- and intermediate-level waste in 2050 and for high-level waste in 2060. Until then all Swiss nuclear power plants will be shut down and most probably already be fully decommissioned, as well.
Given the long timeline until operation of the geological repository, the challenge is two-fold. As nuclear technology has been utilized for decades, radioactive waste is continually conditioned, ideally from the beginning in a form suitable for final disposal and according to preliminary waste acceptance criteria of future facilities. In parallel, Nagra as the implementor continually refines the disposal concept and the design within the siting process and subsequent licensing steps, taking into account developing knowledge and state-of-the-art industrial solutions as well as all three key issues fundamental for the actual implementation of a geological repository: safety, acceptance and affordability. The challenge arising from the concurrent developments in the nuclear field and the repository project is met with a flexible and proven administrative framework governing the interaction between waste producers, Nagra and the regulator.
The second aspect of the challenge is posed by the long time scales between actual waste package production and characterisation until the final acceptance of the physical waste packages into the geological repository. Given past experiences and continually developing regulations, requirements and advances in technologies for waste characterisation, it seems advantageous to consider ways to ensure a characterisation and documentation as comprehensive as reasonably possible. Databases supporting traceable record-keeping, refinement, re-structuring, reviewing and consolidation of waste characteristics and associated properties are essential. One step in the Swiss administrative framework mentioned above is the pre-acceptance procedure, deemed to provide an important aspect for building and maintaining trust of stakeholders into adequate, reliable and sufficient waste characterisation. This step has been included in the administrative framework from the beginning but is only now being carried out in first pilot-procedures by Nagra. Within the pre-acceptance of individual waste packages, the completeness of declared properties and conformity to the original waste type specification are verified. In addition, all declared properties are assessed for plausibility and early assumptions regarding characteristics of the waste type refined, if necessary, based on these assessments. Within the procedures, all open issues that might potentially prevent or delay the physical acceptance of waste packages into a facility are addressed. This is only possible in close cooperation with the waste owners, who are, today, still operating the origin facilities and in many cases additional records or descriptions based on individual’s experiences and know-how can be utilised, thus improving and complementing the oftentimes rather sparse declared information on decades old waste packages. Simultaneously, lessons learnt from the pre-acceptance of old waste packages are applied to improve the waste package documentation of waste packages in production, facilitating future administrative procedures.
How to cite: Pudollek, S., Cavaliere, A., Wilhelm, S., and Hild, N.: How to stand the test of time - Waste Package Characterisation for final disposal , Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-148, https://doi.org/10.5194/safend2025-148, 2025.
Plots showing the radiotoxicity of different nuclide groups of HLW are often used to argue for Partitioning and Transmutation (P&T) as a nuclear waste management option. Apart from the limited value of using radiotoxicity to assess the impact of P&T on a repository, these representations are based on assumptions that are usually not explicitly stated. Using a model spent fuel inventory, this paper examines two aspects: the reference level against which spent fuel radiotoxicity is compared, and the dose conversion coefficients used to calculate radiotoxicity.
The reference level for radiotoxicity is usually referred to as "uranium ore". It is argued that no special precautions need to be taken below this level because uranium ore occurs naturally. However, uranium ore is found in various compositions and concentrations throughout the world. Different "reference levels" are often calculated based on different assumptions about how much uranium is actually needed for fuel. The impact of these assumptions on the conclusions drawn from the radiotoxicity plot is discussed.
In recent years, gender-specific effects of ionizing radiation have become an increasingly important area of research. Re-evaluation of data from the bombings of Hiroshima and Nagasaki has shown that women and children are disproportionately affected by the effects of ionizing radiation. The reasons for this are not fully understood. Nevertheless, for many types of radiation assessment and radiation protection analysis, the default is still to use dose conversion coefficients for males - the group least susceptible to harm from ionizing radiation. The dose conversion coefficients for children published by the International Commission on Radiological Protection (ICRP) are used to calculate the radiotoxicity of spent fuel for different age groups. Again, the focus is on how this adjustment in assumptions changes the conclusions that can be drawn from the figure.
How to cite: Vogt, Y. and Frieß, F.: Implicit Assumptions in Radiotoxicity Plots for High-Level Nuclear Waste: Reference Values and Choice of Dose Coefficients, Third interdisciplinary research symposium on the safety of nuclear disposal practices, Berlin, Germany, 17–19 Sep 2025, safeND2025-69, https://doi.org/10.5194/safend2025-69, 2025.
