Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

Applicability of the Krško nuclear power plant core Monte Carlo model for the determination of the neutron source term

  • Received : 2021.01.12
  • Accepted : 2021.05.17
  • Published : 2021.11.25
Download PDF
⟨ Previous Next ⟩

Abstract

A detailed geometrical model of a Krško reactor core was developed using a Monte Carlo neutron transport code MCNP. The main goal of developing an MCNP core model is for it to be used in future research focused on ex-core calculations. A script called McCord was developed to generate MCNP input for an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, taking advantage of already available material and temperature data obtained in the nuclear core design process. The core model was used to calculate 3D power density profile inside the core. The applicability of the calculated power density distributions was tested by comparison to the CORD-2 calculations, which is regularly used for the nuclear core design calculation verification of the Krško core. For the hot zero power and hot full power states differences between MCNP and CORD-2 in the radial power density profile were <3%. When studying axial power density profiles the differences in axial offset were less than 2.3% for hot full power condition. To further confirm the applicability of the developed model, the measurements with in-core neutron detectors were compared to the calculations, where differences of 5% were observed.

Keywords

Acknowledgement

The authors acknowledge the project "Development of Computational Tools for the Determination of the Neutron Field in the Containment of a Pressurized Water Reactor" (L2-8163) was financially supported by the Slovenian Research Agency. The authors acknowledge the financial support from the Slovenian Research Agency (research core funding No. (P2-0073)).

References

  1. M. Kromar, A. Trkov, Nuclear design calculations of the NPP Krsko core, Journal of Energy Technology 2 (November 2009) 41-50.
  2. M.P. Garces, "Activation Neutronics for the Swiss Nuclear Power Plants", Dissertation, Diss. ETH No. 21623, ETH Zurich, 2013.
  3. J.T. Goorley, et al., Initial MCNP6 Release Overview - MCNP6 Version 1.0, 2013. LA-UR-13-22934.
  4. D. Cali c, "Validation and Modification of Fuel Burnup with the Use of ISOlib Library in the CORD-2 Program', PhD Thesis, Faculty of Civil Engineering, Transportation Engineering and Architecture, University in Maribor, 2016.
  5. T. Goricanec, et al., Determination of neutron flux redistribution factors for typical PWR using Monte Carlo neutron transport methods, Slovenia, in: Proceedings of 28th International Conference Nuclear Energy for New Europe, September 2019.
  6. T. Goricanec, et al., "Predicting ex-core detector response in a PWR with Monte Carlo neutron transport methods", 6th international conference on advancements in nuclear instrumentation measurement methods and their applications, EPJ Web Conf. 225 (2020).
  7. J.R. Askew, F.J. Fayers, P.B. Kemshell, A general description of the code WIMS, J. Br. Nucl. Energy Soc. 5 (1966) 564.
  8. A. Trkov, GNOMER - Multigroup 3-Dimensional Neutron Diffusion Nodal Code, " Institute Jozef Stefan, Ljubljana, Slovenia, IJS-DP-6688, March 1993.
  9. A. Trkov, M. Najzer, L. Skerget, "Variant of Green's function nodal method for neutron diffusion, J. Nucl. Sci. Technol. 27 (1990) 766-777. https://doi.org/10.1080/18811248.1990.9731251
  10. M. Kromar, S. Slavic, A. Trkov, "CTEMP: A Code for Thermohydraulic Calculations,", Institute Jozef Stefan, Ljubljana, Slovenia, IJS-DP-6143, February 1991.
  11. M. Kromar, B. Kurincic, Energija, [Print ed], Validation of the CORD-2 System for the NPP Krsko Nuclear Core Design Calculations, vol. 65, 2016 str. 105-115, ISSN 0013-7448, http://journalofenergy.com/specIss/2016_VOL65_No1.pdf.
  12. Z. Stancar, M. Kromar, B. Kos, L. Snoj, Construction of a Monte Carlo benchmark pressurized water reactor core model,, Slovenia, in: Proceedings of 25th International Conference Nuclear Energy for New Europe, September 2016.
  13. J.L. Conlin, W. Ji, J.C. Lee, M.R. Martin, Pseudo material construct for coupled neutronic-thermal-hydraulic analysis of VHTGR, Transactions of ANS 91 (2005).
  14. J.I. Marquez Damian, Thermal Scattering Law Interpolator, IAEA web-page, https://nds.iaea.org/TSL_LibGen/. (Accessed 30 September 2020).
  15. R. MacFarlane, et al., The NJOY nuclear data processing system, version 2012. Tech. Rep. LA-UR-12-27079, LANL, 2012.
  16. D.A. Brown, et al., ENDF/B-VIII.0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data, Nucl. Data Sheets 148 (2018) 1. https://doi.org/10.1016/j.nds.2018.02.001
  17. G. Zerovnik, M. Podvratnik, L. Snoj, On normalization of fluxes and reaction rates in MCNP criticality calculations, Ann. Nucl. Energy 63 (2014) 126-128. https://doi.org/10.1016/j.anucene.2013.07.045
  18. F.B. Brown, "Fundamentals of Monte Carlo Particle Transport", LA-UR-05-4983, Los Alamos national Laboratory, 2005.
  19. F.B. Brown, "On the Use of Shannon Entropy of the Fission Distribution for Assessing Convergence of Monte Carlo Criticality Calculation", PHYSOR-2006, ANS Topical Meeting on Reactor Physics, Canada, Vancouver, September 2006.
  20. F.B. Brown, K-effective of the world and other concerns for Monte Carlo eigenvalue calculations, Prog. Nuc. Sci. Tech 2 (2011) 738-742. https://doi.org/10.15669/pnst.2.738
  21. C.M. Perfetti, B.T. Rearden, W.J. Marshall, Diagnosing undersampling biases in Monte Carlo eigenvalue and flux tally estimates, Nucl. Sci. Eng. 185 (1) (2017) 139-158, https://doi.org/10.13182/NSE16-54.
  22. B.T. Mervin, Uncertainty underprediction in Monte Carlo eigenvalue calculations, Nucl. Sci. Eng. 173 (3) (2017) 276-292, https://doi.org/10.13182/NSE11-104.
  23. B.R. Herman, Monte Carlo and Thermal Hydraulic Coupling Using Low-Order Nonlinear Diffusion Acceleration, Doctoral Thesis, Massachussetts Institute of Technology, 2014.
  24. F. Brown, "A Review of Best Practices for Monte Carlo Criticality Calculations", LA-UR-09-03136, American Nuclear Society, 2009 Nuclear Criticality Safety Topical Meeting, Richland, WA, 2009.
  25. S.S. Gorodkov, Using core symmetry in Monte Carlo dominance ratio calculations, Nucl. Sci. Eng. 168 (2011) 242-247. https://doi.org/10.13182/NSE10-37
  26. F. Franceschini, et al., Simulation of the NPP krsko startup core with CASL core simulator, VERA-CS. Proceedings of 23rd International Conference Nuclear Energy for New Europe, Portoroz, Slovenia, 2014.
  27. F. Franceschini et al., "Simulation of the NPP Krsko cycle 2 with CASL VERA core simulator compared to the CORD2 and PARAGON2/ANC industrial code systems", In: Proceedings of 25th International Conference Nuclear Energy for New Europe, Portoroz, Slovenia.
  28. M. Copic, Thermal Calculation in the Axial Direction of a PWR Core, Institute Jozef Stefan, Ljubljana, IJS-DP-2212, November 1980.
  29. R.A. Knief, Nuclear Engineering, Taylor & Francis Ltd., USA, 1992.
  30. M. Kromar, S. Slavic, B. Zefran, The Nuclear Design and Core Management of the Krsko NPP, IJS-DP-11854, 2015.