Ядерна фізика та енергетика Nuclear Physics and Atomic Energy   ISSN: 1818-331X (Print), 2074-0565 (Online)   Publisher: Institute for Nuclear Research of the National Academy of Sciences of Ukraine   Languages: Ukrainian, English, Russian   Periodicity: 4 times per year   Open access peer reviewed journal

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Abstract: The description of calculation scheme of fuel assembly for preparation of few-group characteristics is considered with help of Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data libraries. Serpent code is devoted for calculation of fuel assembly characteristics, burnup calculations and preparation of few-group homogenized macroscopic cross-sections. The results of verification simulations in comparison with other codes (WIMS, HELIOS, NESSEL etc.), which are used for neutron-physical analysis of VVER type fuel, are presented.

Keywords: fuel assembly, VVER-1000, Monte Carlo method, reactor calculation, Serpent, MCNP, SCALE, WIMSD5B.

1. M. Kostal, V. Rypar, V. Juricek. The criticality of VVER-1000 mock-up with different H₃BO₃ concentration. Annals of Nuclear Energy 60 (2013) 1. https://doi.org/10.1016/j.anucene.2013.04.014

2. J.R. Askew, F.J. Fayes, P.B. Kemshell. A general description of the lattice Code WIMS. Journal of British Nuclear Energy Society 5(1) (1966) 564.

3. V.V. Galchenko. Data preparation comparison analysis using difference computer codes. Part 2. Yaderni ta Radiatsiini Tekhnologiyi (Nuclear and Radiation Safety) 7(3-4) (2007) 29. (Rus) Article

4. V.O. Babenko, V.I. Gulik, V.M. Pavlovych. Modeling of Two-zone Accelerator-Driven Systems. Yaderna Fizyka ta Energetyka (Nucl. Phys. At. Energy) 13(3) (2012) 266. https://jnpae.kinr.kyiv.ua/13.3/Articles_PDF/jnpae-2012-13-0266-Babenko.pdf

5. V.I. Gulik, A.H. Tkaczyk. Cost optimization of ADS design: Comparative study of externally driven heterogeneous and homogeneous two-zone subcritical reactor systems. Nuclear Engineering and Design 270 (2014) 133. https://doi.org/10.1016/j.nucengdes.2014.01.007

6. V.I. Gulik, V.M. Pavlovych, A.H. Tkaczyk. Using SERPENT Monte Carlo and Burnup code to model Traveling Wave Reactors (TWR). Proc. of Joint Int. Conf. on Supercomputing in Nuclear Applications & Monte Carlo (SNA&MC2013) (Paris, France, 27 - 31 Oct., 2013).

7. V.M. Khotyayintsev, A.V. Aksonov, O.M. Khotyayintseva et al. Velocity Characteristic and Stability of Wave Solutions for a CANDLE Reactor with Thermal Feedback. Annals of Nuclear Energy 85 (2015) 337. https://doi.org/10.1016/j.anucene.2015.04.044

8. J.C. Wagner, D.E. Peplow, S.W. Mosher, T.M. Evans Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory. Progress in Nuclear Science and Technology 2 (2011) 808. https://doi.org/10.15669/pnst.2.808

9. W.R. Martin. Challenges and prospects for whole-core Monte Carlo analysis. Nuclear Engineering and Technology 44(2) (2012) 151. https://doi.org/10.5516/NET.01.2012.502

10. J. Briesmeister. MCNP General Monte Carlo Code N-Particle Transport Code Version 4C, LA-13709-М (1993). http://permalink.lanl.gov/object/tr?what=info:lanl-repo/lareport/LA-13709-M

11. R.A. Forster, T.N. Godfray. MCNP - a general Monte Carlo code for neutron and photon transport. Lecture Notes in Physics. Monte Carlo Methods and Applications in Neutronics 240 (1985) 33. https://doi.org/10.1007/BFb0049033

12. J. Leppanen. Development of a new Monte Carlo reactor physics code: PhD thesis (Helsinki University of Technology: VTT Publications, 2007) 640. Thesis

13. J. Leppanen, M. Pusa, T. Viitanen et al. The Serpent Monte Carlo code: Status, development and applications in 2013. Annals of Nuclear Energy 82 (2015) 142. https://doi.org/10.1016/j.anucene.2014.08.024

14. E. Fridman, E. Shwageraus. Modeling of SFR cores with Serpent-DYN3D codes sequence. Annals of Nuclear Energy 53 (2013) 354. https://doi.org/10.1016/j.anucene.2012.08.006

15. E. Fridman, J. Leppanen. On the use of the Serpent Monte Carlo code for few-group cross section generation Annals of Nuclear Energy 38 (2011) 1399. https://doi.org/10.1016/j.anucene.2011.01.032

16. http://montecarlo.vtt.fi/

17. Benchmark calculations for VVER-1000 fuel assemblies using uranium or MOX fuel: Validation report for FY 1997 (Final report). (2000) 114 p.

18. V.V. Galchenko, V.V. Solovyov, О.S. Gorodnycha. Nuclear Fuel Burn-Up Credit for Criticality Safety Justification of Spent Nuclear Fuel Storage Systems. Yaderni ta Radiatsiini Tekhnologiyi (Nuclear and Radiation Safety) 2(54) (2012) 45. (Ukr)

19. V.V. Galchenko, А.A. Mishyn. Comparative analysis of reactor cycle neutron characteristics using different WIMSD5B nuclear data libraries. Yaderni ta Radiatsiini Tekhnologiyi (Nuclear and Radiation Safety) 3(67) (2015) 8. (Ukr)