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Modeling Transport of Radioactive Products of Corrosion in Loops with Sodium Coolant

A phenomenological model is proposed for finding the relationship between the observed behavior of corrosion radionuclides in liquid sodium and the experimental conditions. Reactor and extra-reactor experimental data are used to find the temperature and rate dependences of the model parameters so that the model reflects a wide range of experimental conditions. Most of the experimental data were obtained using domestic fast reactors and, to a lesser extent, foreign reactors. The design code Al’fa-M, which is based on the proposed model, is described. The code was tested on the BN-600 reactor. It was shown that the proposed modeling approach makes it possible to solve the problem of calculating the removal from the core and the distribution in the tank of a power reactor of activated products of corrosion with satisfactory accuracy within the framework of the problem at hand. In addition, discrepancies between the calculations and experiments, requiring further improvements to the model are found.

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Translated from Atomnaya Énergiya, Vol. 119, No. 1, pp. 34–41, July, 2015.

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Zhilkin, A.S., Popov, E.P. Modeling Transport of Radioactive Products of Corrosion in Loops with Sodium Coolant. At Energy 119, 37–45 (2015). https://doi.org/10.1007/s10512-015-0026-2

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  • DOI: https://doi.org/10.1007/s10512-015-0026-2

Keywords

  • Corrosion Product
  • Power Reactor
  • Fuel Assembly
  • Radioactive Product
  • 54Mn Deposit