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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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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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Pungerčič Anže,Mascolino Valerio,Haghighat Alireza,Snoj Luka 한국원자력학회 2023 Nuclear Engineering and Technology Vol.55 No.10
The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the wellcharacterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the keff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.
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Analysis of fluctuations in ex-core neutron detector signal in Krško NPP during an earthquake
Tanja Goricanec,Andrej Kavcic,Marjan Kromar,Luka Snoj Korean Nuclear Society 2024 Nuclear Engineering and Technology Vol.56 No.2
During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%-8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ~4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ~6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.
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