Atomic Energy

, Volume 124, Issue 4, pp 272–278 | Cite as

Tritium Module for Calculating the Behavior of Tritium in a Loop of a Reactor Installation with Sodium Coolant

  • O. Kh. Il’yasova
  • S. N. Nazarova
  • A. A. Sorokin

A description of the models and the results of tests of the Tritium module, which is intended for modeling the transport of tritium in the loops of fast reactors, are presented. The characteristics features of the module are alienability, functional relation with other modules of the code, and calculation of model coefficients as a function of the temperature of the coolant and surface of the loops. The module makes it possible to perform calculations of the behavior of tritium, including transport along loops, penetrability through the channel walls, run-off in a cold trap, and flow through leaks in the reactor installation.


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© Springer Science+Business Media, LLC, part of Springer Nature 2018

Authors and Affiliations

  • O. Kh. Il’yasova
    • 1
  • S. N. Nazarova
    • 1
  • A. A. Sorokin
    • 2
  1. 1.Nuclear Safety InstituteRussian Academy of Sciences (IBRAE RAN)MoscowRussia
  2. 2.Joint Institute of High TemperaturesRussian Academy of Sciences (JIHT RAS)MoscowRussia

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