The neutron-physical characteristics of reactor systems with a fast spectrum, sodium coolant, and uraniumplutonium fuel load have been analyzed on the basis of computational studies of the BFS-62-3A critical assembly and a BN-600 hybrid core with mixed oxide fuel. The large differences in the spectra in an expanded thermal range to 1 keV for the central and peripheral regions with uranium oxide and mixed oxide fuel show that spatially differentiated fission and absorption cross sections must be used for the main uranium and plutonium isotopes in neutron-physical calculations.
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References
M. Chadwick, P. Oblozinsky, M. Herman, et al., “ENDF/B-VII.O: next generation evaluated nuclear data library for nuclear science and technology,” Nucl. Data Sheets, 107, No. 12, 2931–3060 (2006).
K. Shibata, T. Kawano, T. Nakagawa, et al., “Japanese evaluated nuclear data library, version 3, revision-3: JENDL-3.3,” J. Nucl. Sci. Technol., 39, No. 11, 1125–1136 (2002).
BFS-62-3A Experiment: Fast Reactor Core with U and U–Pu Fuel of 17% Enrichment and Partial Stainless Steel Reflector: IRPhEP Handbook, NEA/NSC/DOE (2006).
X-5 Monte Carlo Team. MCNP – a General Monte Carlo JV-Particle Transport Code, Version 5, LA-UR-03-1987, April 2003.
ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6, BNL-NCS-44945-01/04-Rev. (2001).
R. MacFarlane, The NJOY Nuclear Data Processing System, Version 91.0, LA-12740-M, (1994).
BN-600 Hybrid Core. Benchmark Analyses: IAEA-TECDOC-1623, IAEA, Vienna (2010).
V. Tiberi, E. Ivanov, and S. Pignet, “An approach of SFR safety study through the most penalizing sodium void reactivity,” in: Proc. PHYSOR’2010, May 9–14, 2010, USA, CD-ROM No. 700356.
G. Rimpault, “The ERANOS code and data system for fast reactor neutronics analyses,” in: Proc. PHYSOR’2002, October 7–10, 2002, Seoul, Korea, CD-ROM No. 700298.
L. P. Abagyan, N. O. Bazazyants, M. N. Nikolaev, and A. M. Tsibulya, Reference Data on Group Constants for Calculating Reactors and Protection, Energoizdat, Moscow (1981).
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Translated from Atomnaya Énergiya, Vol. 109, No. 5, pp. 253–262, November, 2010.
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Mitenkova, E.F., Novikov, N.V. Particulars of neutron-physical calculations of sodium-cooled fast reactors with mixed oxide fuel. At Energy 109, 309–320 (2011). https://doi.org/10.1007/s10512-011-9361-0
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DOI: https://doi.org/10.1007/s10512-011-9361-0