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      저자 : Stancar, Ziga (검색결과 2 건)
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        Applicability of the Krško nuclear power plant core Monte Carlo model for the determination of the neutron source term

        Goricanec, Tanja,Stancar, Ziga,Kotnik, Domen,Snoj, Luka,Kromar, Marjan Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.11

        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.

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        Computational design and characterization of a subcritical reactor assembly with TRIGA fuel

        Alvie Asuncion-Astronomo,Ziga Stancar,Tanja Goricanec,Luka Snoj 한국원자력학회 2019 Nuclear Engineering and Technology Vol.51 No.2

        The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly(SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offersa versatile and safe training and research facility since it can produce neutrons through nuclear fissionreaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuelconfigurations. Based on extensive neutron transport simulations an SRA configuration is proposed,comprising 44 TRIGA fuel rods arranged in a 7 7 square lattice. This configuration is found to have amaximum keff value of 0:95001±0:00009 at 4 cm pitch. The SRA is characterized by calculating the 3-dimensional neutron flux distribution and neutron spectrum. The effective delayed neutron fraction andmean neutron generation time of the system are calculated to be 748 pcm±7 pcm and 41 ms, respectively. Results obtained from this work will be the basis of the core design for the subcritical reactorfacility that will be established in the Philippines.

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