Izvestia Vysshikh Uchebnykh Zawedeniy. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

Application of MCPN nonanalog techniques for calculations of reaction rate measurements at the BFS facilities

3/28/2016 2016 - #02 Modelling processes at nuclear facilities

Andrianova O.N. Manturov G.N. Rozhikhin Ye.V.

UDC: 621.039.51.17

The BFS fast critical assemblies of the JSC «SSC RF – Institute for Physics and Power Engineering» are a unique experimental base for measuring different reactor functionals required to clarify and justify expected evaluations of both reactor core and ex-core characteristics. For years the data obtained on BFS integral critical experiments have been widely used to test and improve neutron cross-sections libraries and transport codes used for applications on fast reactor core design studies. Critical experiments performed at the BFS facilities provide a way to carry out a large series of studies needed for refining neutron data, including, for instance, measurements of central reactivity coefficients (reactivity introduced by material samples of various sizes) which allows testing the resonance structure of neutron cross-sections, but these experiments have not often been used previously for neutron cross-section adjustments. Experiments carried out in different years on various BFS core configurations and compositions were used for testing of project values of fast reactor models, studying physical properties of fuel, structural, and other materials, that brought to light data on neutron cross-sections used for correcting and adjusting neutron data for the neutron data library and transport codes.

An extensive experimental program included criticality, measurements of central reaction rate ratios, and fission rate distributions. The results of these measurements were evaluated and accepted for use as criticality and reactor physics parameters benchmarks. Calculations of criticality, central reaction rate ratios, and fission rate distributions were performed using the MCNP Monte-Carlo code using different files of evaluated nuclear data. The calculations showed that it is necessary to use variance reduction techniques to get the desired uncertainty of reaction rate calculations. Using of different variance reduction techniques implemented in MCNP for calculations of local functionals in critical multiplicating systems with complex heterogeneous geometry is considered in the report. Calculational analysis of effectiveness of variance reduction techniques was performed by the example of calculations of the central reaction rates. Results of criticality, central reaction rate ratios, and fission rate distributions calculations with different files of evaluated nuclear data are presented.

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integral critical experiments BFS critical assemblies reaction rate ratios variance reduction methods MCNP