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Transport packages for spent fuel have to meet the International Atomic Energy Agency requirements for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. In this paper, the mechanical behaviour of high burn-up spent fuel assemblies under transport conditions is analysed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods (fine cracks or complete breakage), which can cause release of gas, volatiles, fuel particles or fragments have to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods as well as between 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 pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context some practical approaches based on experiences of BAM Federal Institute for Material Research and Testing within safety assessment of packages for transport of spent fuel are discussed.
Methodological aspects for finite element modeling of lid systems for type B(U) transport packages
(2012)
The regulatory compliance of the containment system is of essential importance for the assessment
process of Type B(U) transport packages. The requirements of the IAEA safety standards for
transport conditions implies high loading on the containment system. The integrity of the
containment system has to be ensured in mechanical and thermal tests.
The containment system of German SNF and HLW transport packages usually includes bolted lids
with metal gaskets. The finite element (FE) method is recommended for the analysis of lid systems
according to the guideline BAM-GGR 012 for assessment of bolted lid and trunnion systems. FE
analyses provide more accurate and detailed information about loading and deformation of such
kind of structures. The results allow the strength assessment of the lid and bolts as well as the
evaluation of relative displacements between the lid and the cask body in the area of the gasket
groove.
This paper discusses aspects concerning FE simulation of lid systems for Type B(U) packages for
the transport of SNF and HLW. The work is based on the experiences of BAM Federal Institute for
Materials Research and Testing as the German competent authority for the mechanical design
assessment of such kind of packages. The issues considered include modeling strategies, analysis
techniques and the interpretation of results. A particular focus of this paper is on the evaluation of
the results with regard to FE accuracy, influence of the FE contact formulation and FE modeling
techniques to take the metallic gasket into account.
Packages for the transport of radioactive material are generally equipped with specific
structures (basket) to support the radioactive content in defined position. The safety function
of the basket depends on the kind of transported inventory. In case of transport cask for spent
fuel, the basket design has to ensure the subcriticality of the fissile material in all conditions
of transport in particular. Therefore the evaluation of structural integrity and neutron
absorption capability of the basket is an important part of complete safety analysis. Sufficient
heat transfer to maintain fuel assembly and cask temperature within allowable limits has to be
verified as well. Corrosion resistance is an additional requirement on basket materials owing
to contact with water during loading and unloading operations.
Computational and experimental methods or their combination along with additional material
and component tests can be used to analyse the mechanical and thermal basket behaviour
under transport conditions defined in IAEA regulations. By deciding between the analysis
methods, the design features (including material selection concept) as well as specific safety
function should be accounted.
In approval procedures of transport packages for radioactive materials, the competent
authority mechanical and thermal safety assessment is carried out in Germany by BAM. Some
questions of safety evaluation of basket designs are discussed in this paper based on the BAM
experience within approval procedures. The paper focuses primarily on the mechanical
behaviour of baskets with regard to the assumptions that have to be used in the criticality
safety demonstration. The state of the art methodologies for computational basket stress and
deformation analysis as well as for interpretation of drop tests results are presented.