Applicability of the Kr ? ko nuclear power plant core Monte Carlo model for the determination of the neutron source term

Tanja Gori canec; Ziga   Stancar; Domen Kotnik; Luka Snoj; Marjan Kromar

doi:https://doi.org/10.1016/j.net.2021.05.022

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자료유형: 학술저널

저자정보: Tanja Gori canec (Jozef Stefan Institute) Ziga Stancar (Jozef Stefan Institute) Domen Kotnik (Jozef Stefan Institute) Luka Snoj (Jozef Stefan Institute) Marjan Kromar (Jozef Stefan Institute)

저널정보: 한국원자력학회 Nuclear Engineering and Technology Nuclear Engineering and Technology 제53권 제11호

발행연도: 2021.11

수록면: 3,528 - 3,542 (15page)

DOI: https://doi.org/10.1016/j.net.2021.05.022

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A detailed geometrical model of a Kr sko reactor core was developed using a Monte Carlo neutrontransport code MCNP. The main goal of developing an MCNP core model is for it to be used in futureresearch focused on ex-core calculations. A script called McCord was developed to generate MCNP inputfor an arbitrary fuel cycle configuration from the diffusion based core design package CORD-2, takingadvantage 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 ofthe calculated power density distributions was tested by comparison to the CORD-2 calculations, whichis regularly used for the nuclear core design calculation verification of the Kr sko core. For the hot zeropower and hot full power states differences between MCNP and CORD-2 in the radial power densityprofile were <3%. When studying axial power density profiles the differences in axial offset were lessthan 2.3% for hot full power condition. To further confirm the applicability of the developed model, themeasurements with in-core neutron detectors were compared to the calculations, where differences of5% were observed

#Pressurized water reactor #Kr sko nuclear power plant #Monte Carlo neutron transport #MCNP #CORD-2 #In-core neutron detector #PACS: 24.10.L 28.20.-v

참고문헌 (30)

참고문헌 신청

M. Kromar / 2009 / Nuclear design calculations of the NPP Krsko core / Journal of Energy Technology 2:41~50

M. P. Garces / 2013 / Activation Neutronics for the Swiss Nuclear Power Plants

J.T. Goorley / 2013 / Initial MCNP6 Release Overview - MCNP6 Version 1.0

D. Cali c / 2016 / Validation and Modification of Fuel Burnup with the Use of ISOlib Library in the CORD-2 Program

T. Goricanec / 2019 / Determination of neutron flux redistribution factors for typical PWR using Monte Carlo neutron transport methods / Proceedings of 28th International Conference Nuclear Energy for New Europe

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