Advanced Monte Carlo Calculations using MCNP of Ex-Vessel Neutron Dosimetry for Cycle 25 of Krško Nuclear Power Plant
Authors/Creators
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1.
Jožef Stefan Institute
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2.
University of Ljubljana
Description
About
This repository includes calculation results of reaction rates for Ex-Vessel Neutron Dosimetry (EVND) program at the Krško nuclear power plant during 25th fuel cycle [1].
25th fuel cycle took place from October 25, 2010, to April 15, 2012.
Calculations were performed using Monte Carlo N-Particle transport code MCNP [2] in combination with the hybrid code ADVANTG [3] to speed up the convergance of the results outside the reactor core. Calculated reaction rates were normalized [4] to the nominal reactor power of 2 GW. Calculations were performed with the detailed geometrical model of the reactor core and detailed fixed neutron source modeled representing hot full power core state for the Krško NPP 25th fuel cycle [5], [6]. Detailed core model and fixed neutron source were included in the simplified geometrical model of the Krško NPP containment building [7].
A detailed description of the calculation procedure, calculation models, calculation results, experimental data and sensitivity studies is presented in the scientific paper:
T. Goričanec, B. Barbarič, V. Radulović, D. Čalič, L. Snoj and M. Kromar, "Advanced Monte Carlo Analysis of Ex-Vessel Neutron Dosimetry for Cycle 25 of Krško Nuclear Power Plant", Nuclear engineering and technology. [Online ed.]. Jan. 2026, vol. 58, iss. 1, [article no.] 103887, DOI: 10.1016/j.net.2025.103887.
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Additional details
Funding
- Slovenian Nuclear Safety Administration
- Izračun transporta nevtronov v zadrževalnem hramu jedrske elektrarne Krško z uporabo Monte Carlo metode C2563-24-430012
Dates
- Updated
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2025-03-17
References
- Hunter, M. A., "Ex-Vessel Neutron Dosimetry Program for Krško Cycle 25.", Westinghouse Electric Company LLC, WCAP-17734-NP.
- Goorley, T., et al., "Initial MCNP6 release overview.", Nuclear Technology, 180.3 (2012): 298-315.
- Mosher, S. W., et al., "ADVANTG An Automated Variance Reduction Parameter Generator", Rev.1, Tech. rep., Oak Ridge National Lab. (ORNL), Oak Ridge, TN, USA, 2015.
- Žerovnik, G., et al., "On normalization of fluxes and reaction rates in MCNP criticality calculations", Annals of Nuclear Energy 63 (2014) 126-128.
- Goričanec, t., et al., "Applicability of the Krško nuclear power plant core Monte Carlo model for determination of neutron source term", Nuclear Engineering and Technology 53 (11) (2021) 3528-3542.
- Goričanec, T., et al., "Determination of neutron flux redistribution factors for a typical pressurized water reactor ex-core measurements using Monte Carlo technique", Frontiers in Energy Research 11 (2023).
- Kos, B, et al., "Neutron Streaming Analysis and Shielding Determination for the Krško Nuclear Power Plant", In: Proceeedings of 25th International Conference Nuclear Energy for New Europe, Nuclear Society of Slovenia, 2016.