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Apr 19 – 22, 2022
Vienna, Austria
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FR22 starts in Vienna 19 - 22 April 2022 Online Stream:

The solution of nuclide kinetic equation for fast reactor in the OpenBPS code with options of choosing calculation method and uncertainties analysis.

Apr 20, 2022, 11:52 AM
Vienna, Austria

Vienna, Austria

ORAL Track 6. Modelling, Simulations, and Digitilization 6.1 Neutronics


Ivan Bukhtiiarov


The knowledge about fuel nuclear concentration behavior, fission product etc. both in fast reactor core and after withdrawing fuel from reactor is necessary from the point of view of nuclear safety as well as in future for fuel cycle closing up. There are a lot of methods and calculation codes in practice of burnup equation solution. Three different approaches can be highlighted, among which: a solution of nuclear kinetics equation with one of the matrix exponent method, iteration method with ability of uncertainties analysis, direct analytical solution method.
The software with open source code Open BurnuPSimulation (OpenBPS) provide for users a huge amount of means both for working with input data and choosing a solution method. Codes are being developed based on other open source instruments and approaches for solution burnup equation accessible in literature.
The Chebyshev Rational Approximation Method is implemented in the program for matrix exponent solution. An ability of using uncertainties analysis of the output nuclear concentration is provided by the iteration approach based on uncertainties of input, decay and nuclear cross-section data. The authors developed an analytical calculation method using modified Bateman functions (Analytical Solution of Burning Equations) for the accelerated solution of the burning equations. The calculation code ASBE was created on the basis of this method to calculate changes in the fuel concentrations during the operation of the BN-600 and BN-800 reactors. The speed of obtaining a solution in this code is more than an order of magnitude higher than the speed of obtaining a solution in programs of the iterative method, with high accuracy of the solution.
The program based on open source software provides a flexible and user-friendly interface with a choice of any of the listed solution methods and has unlimited potential for using at all stages of the fuel cycle from fabrication and placement into a reactor core to assessing of the activity and residual heat release during processing and preservation of used material.

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Country/Int. organization Russian Federation

Primary author


Evgeny Seleznev (NUCLEAR SAFETY INSTITUTE OF RUSSIAN ACADEMY OF SCIENCES) IURII Drobyshev (Russian Federation) Mr Sergey Karpov (All-Russian Research Institute for Nuclear Power Plants Operation)

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