Analysis of a PWR core baffle considering irradiation induced creep

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The core baffle of a PWR is loaded by the pressure difference between bypass and core and by temperature profiles developing from gamma and neutron heating and heat transfer into the coolant. Strain, deformation and gaps between the sheets resulting from this load are determined considering the effect of neutron irradiation induced creep of the core baffle bolts. The finite element code ANSYS® is applied for the thermal and mechanical analyses. The FE-model comprises a complete 45° sector of the core baffle structure including the core barrel, the formers, the core baffle sheets and about 230 bolt connections with non-linear contact between the single components and the effect of friction. The complete analysis requires three major steps:1.Evaluation of the three dimensional distribution of neutron flux and gamma induced internal heating with the Monte Carlo code MCNP®. These calculations are based on pin wise power distributions at the core edge for typical loading patterns.2.Calculation of the temperature distribution in the core baffle for different operational conditions and core loading patterns, considering heat conduction in the components with internal heat sources and convectional boundary conditions (heat transfer coefficients and bulk temperature of the coolant).3.Calculation of time dependent deformation, stresses and strains taking into account weight, pressure loads, temperature fields for different load situations, prestressing, irradiation induced creep of the bolts as correlated to neutron flux. The results show the equalizing effect of redistribution of bolt loads from high flux to lower flux exposure locations in a self controlled process, keeping the mechanical and geometrical stability of the core baffle structure and leaving the gaps between sheet edges unaffected.




Altstadt, E., Kumpf, H., Weiss, F. P., Fischer, E., Nagel, G., & Sgarz, G. (2004). Analysis of a PWR core baffle considering irradiation induced creep. Annals of Nuclear Energy, 31(7), 723–736.

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