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Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency. In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by BAM for the authority assessment of spent fuel behavior within the package design approval procedure of German package designs. In particular, cracks or failures in the fuel rod cladding can occur under regulatory transport conditions. These defects can cause the release of gas, volatiles, fuel particles or fragments into the package cavity and have to be considered properly in the safety analysis. Another issue is the transport of defective fuel rods. One concept is to use special canisters which can be handled like fuel assemblies. This concept requires additional assessment concerning drying, sealing and the mechanical and thermal design of such canisters. The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible. The simplified approaches to consider conservatively spent fuel behavior currently accepted by BAM are presented here.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety. The objective of this paper is to discuss the methodologies accepted by BAM for the authority assessment of spent fuel behavior within the package design approval procedure.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC-ITU and BAM have started a joint research project. In this context, JRC-ITU has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety.
The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder.
For a better understanding about the behavior of spent fuel rods, JRC and BAM have started a joint research project. In this context, JRC has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions.
Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different Transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under Transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel.
To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a segment of a spent fuel assembly are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers. Explicit dynamic finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. A beam element formulation is used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The load applied is gathered from experimental drop tests with spent fuel casks performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by
simulating the test setup of JRC and optimizing the results to fit the
experimental load deflection curve. The simulations of the fuel Assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport.
The containment systems of transport and storage casks for spent fuel and highlevelradioactive waste usually include bolted lids with metallic or elastomeric seals. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of such containment system.
Scaled cask models are often used for providing the required mechanical and thermal tests series.Leak tests have been conducted on those models.
It is also common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals.
In this paper questions of the transferability of scaled test results to the full size design of the containment system will be discussed.
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The specified transport condition imply high loading on the lids and the bolt connections of the package. The response of the lid systems on these load conditions is generally investigated by drop tests or numerically. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connection under loadings typical for drop tests.
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The specified transport condition imply high loading on the lids and the bolt connections of the package. The response of the lid systems on these load conditions is generally investigated by drop tests or numerically. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connection under loadings typical for drop tests.
Experimental and numerical investigation of prestressed bolt connections under lateral displacements
(2022)
The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results.