Validation and Uncertainty Analysis of the Thermal Hydraulics Module of SOCRAT-BN Code on the Rod Bundle Experiment

Yulia Vinogradova; Nikolai Ryzhov; Ruslan Chalyy

doi:10.4028/www.scientific.net/AMR.1084.717

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HomeAdvanced Materials ResearchAdvanced Materials Research Vol. 1084Validation and Uncertainty Analysis of the Thermal...

Validation and Uncertainty Analysis of the Thermal Hydraulics Module of SOCRAT-BN Code on the Rod Bundle Experiment

Abstract:

SOCRAT-BN code is developed for the analysis of design and beyond design basis accidents at sodium cooled fast reactors. To simulate the behavior of the coolant in the reactor core heat transfer and friction in rod bundle geometry are required to consider. The article describes the validation of the code SOCRAT-BN on the experiment with fuel rod imitators in the triangular geometry with wire-wound taking into account experiment and some code model uncertainties.

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Periodical:

Advanced Materials Research (Volume 1084)

Pages:

717-721

DOI:

https://doi.org/10.4028/www.scientific.net/AMR.1084.717

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Online since:

January 2015

Authors:

Yulia Vinogradova, Nikolai Ryzhov, Ruslan Chalyy

Keywords:

Fast Reactor, Integral Code, SOCRAT-BN, Sodium Coolant, Uncertainty Analysis, Validation

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