Testing covariance matrices of uncertainties in the BNAB data system
The growing interest in evaluation of computational uncertainties in the main neutron-physical characteristics led to the development of nuclear data covariance libraries. Taking into account the recent achievements in this area, the covariance matrices of uncertainties in the BNAB-93 group constant library needed to be verified which could be done by using additional information from current neutron data libraries.
The present work is focused on testing covariance matrices of uncertainties which were derived from expert evaluation of the current version of the BNAB group constants library based on evaluated data files RUSFOND-2010. The testing process involved comparing and analyzing computational uncertainties of average cross-sections for typical fast reactor spectra from the BNAB covariance data library with similar data from other current libraries (like ENDF/B, JENDL, COMMARA and others).
In addition, uncertainties of important neutron-physics characteristics have been calculated for numerous test models of advanced liquid-metal-cooled fast reactors based on COMMARA and BNAB covariance data libraries. The main sources of uncertainties have been analyzed.
- Abagjan L.P., Bazazjanc N.O., Nikolaev M.N., Tsibulya A.M. Gruppovye konstanty dlja rascheta reaktorov i zashhity [Group constants for reactor and shielding calculations]. Moscow, Energoizdat Publ. 1981. 232 p. (in Russian).
- G. Palmiotti, M. Salvatores, Proposal for nuclear data covariance matrix, JEFDOC 1063 Rev.1, January 20, 2005
- Khomyakov Yu.S., Eliseev V.A., Malysheva I.V., Matveev V.I., Tsibulya A.M. Optimization of Parameters of MOX Fuel Core of Sodium Cooled Large Size Fast Reactor. Proceedings of Global 2009, Paris, France, September 6-11, 2009.
- Matveev V.I., Chebeskov A.N., Cherny V.A.,.Krivitski I.Yu, Kirushin A.I., Belov S.B., Vasiljev B.A.. Studies, Development and Justification of Core with Zero Sodium-Void Reactivity Effect of the BN-800 Reactor. In Proc. of the International topical meeting on Sodium Cooled Fast Reactor Safety. Obninsk, Russia, October 3-7, 1994.
- BN-600 Hybrid Core Benchmark Analysis, IAEA TecDoc- 1623, Vienna, IAEA, February 2010.
- Smirnov V., Orlov V. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method. Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications Palais des Papes, Avignon, France, September 12-15, 2005, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2005).
- Zrodnikov A.V., Chitaykin V.I., Toshinsky G.I., Grigoriev O.G., Dragunov U.G., Stepanov V.S., Klimov N.N., Kopytov I.I., Krushelnitsky V.N., Grudakov A.A..”SVBR-75⁄100 — Lead-bismuth cooled small power modular fast reactor for multi-purpose usage”, In Proc. of the International conference “Innovative Technologies for Nuclear Fuel Cycles and Nuclear Power”, Vienna, 23–26 June 2003, IAEA-CSP-24, 2004, p.371.
- Romanova N.V., Juhnov B.M., Mamedov T.S., Komponovochnye reshenija konturov teplootvoda RU MBIR s natrievym teplonositelem [Layout solutions of heat sink lines with sodium-cooled MBIR reactor assembly]. Moscow, OAO «NIKIET», 2011.
- Jerdev G.M. SKALA – The Computing System for an Estimation of Nuclear and Radiation Safety, M&C-2005/ Avignon, France, September 12–15 2005.
- Blyskavka A.A., Jerdev G.M., Manturov G.N., Nikolaev M.N., Raskach K.F, Tsibulya A.M.. Use of the SKALA Code Package for Computing Criticality and its Uncertainty. Proc. of International Conference on Nuclear Safety (ICNC’07), St. Petersburg, Russia, May 28 – June 1, 2007.
- M. Herman, P. Oblo~inskэ, C.M. Mattoon, M. Pigni, S. Hoblit, S.F. Mughabghab, A. Sonzogni, P. Talou, M.B. Chadwick, G.M. Hale, A.C. Kahler, T. Kawano, R.C. Little, P.G. Young, COMMARA-2.0 Neutron Cross Section Covariance Library, BNL- 94830-2011, 2011.