Tritium Module for Calculating the Behavior of Tritium in a Loop of a Reactor Installation with Sodium Coolant
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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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- 1.F. A. Kozlov, V. M. Poplavskii, V. V. Alekseev, et al., “Simulation of tritium mass transfer in a three-loop sodium-cooled NPP,” At. Energ., 98, No. 3, 175–182 (2005).Google Scholar
- 2.F. Franza, Tritium Transport Analysis in Advanced Sodium Fast Reactors: Thesis for Master Deg. in Nucl. Eng., Politecnico di Torino, Italy (2011).Google Scholar
- 3.L. A. Bol’shov, N. A. Mosunova, V. F. Strizhov, and O. V. Shmidt, “Next- generation computational codes for a new technological platform for nuclear power,” At. Energ., 120, No. 6, 303–312 (2016).Google Scholar
- 4.V. M. Alipchenkov, A. M. Anfimov, D. A. Afremov, et al., “Basic positions, current development status, and prospects for subsequent development of the next-generation thermohydraulic computational code HYDRA-IBRAE/LM for simulation of fast reactor installations,” Teploenergetika, No. 2, 54–64 (2016).Google Scholar
- 5.J. McGuire and T. Renner, “Control of tritium in liquid-metal cooled fast breeder reactors,” At. Energy Rev., 16, No. 4, 657–695 (1978).Google Scholar
- 7.C. Oh and E. Kim, Development and Verification of Tritium Analyses Code for a Very High Temperature Reactor, INL-09-16743 (2009).Google Scholar
- 9.H. Ohashi and S. Sherman, Tritium Movement and Accumulation in the NGNP System Interface and Hydrogen Plant, INL-07-12647 (2007).Google Scholar
- 10.L. Yang, A. Baugh, and N. L. Baldwin, Study of Tritium Permeation Through Peach Bottom Steam Generator Tubes, Prepared under Contract EY-76-C-03-0167 Project Agreement No. 56 for the San Francisco Operations Office U.S. Energy Research and Development Administration (1977).Google Scholar