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Your search keyword '"Košt'ál, Michal"' showing total 15 results
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15 results on '"Košt'ál, Michal"'

2. On radiation situation in vicinity of PET production cyclotron.

Author

Zmeškal, Marek, Košt'ál, Michal, Czakoj, Tomáš, Šimon, Jan, Schulc, Martin, Vadják, Šimon, Antoš, Michal, and Matěj, Zdeněk

Subjects
*CYCLOTRONS, *RADIOISOTOPES, *GAMMA rays, *NEUTRON radiography, *NUCLEAR counters
Abstract

The medical cyclotrons intended for the production of medical isotopes are relatively widespread. The most common radioisotope product of these facilities is 18F which origins from 18O(p, n)18F reaction. Due to relatively large amounts of produced nuclei, the leakage neutron and gamma flux is a big issue in radiation safety in such cyclotron labs. Due to relatively high energies of accelerated protons, high energy gammas and neutrons are produced with energies above 10 MeV with high penetration properties. Neutron and gamma spectra around cyclotron IBA Cyclone 18/9 intended for production of 18F was measured with stilbene detector. It was compared to simulated neutron spectra with various Monte Carlo codes and with spectral indices calculated from reaction rates in activation detectors. Strong disagreement in simulation of spectra in backward angles was found for MC codes Geant4 and MCNP6.2. [ABSTRACT FROM AUTHOR]

Published
2024
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12. Validation of calculational determination of 18O(p,n) secondary neutron field.

Author

Zmeškal, Marek, Košt'ál, Michal, Czakoj, Tomáš, Šimon, Jan, Zach, Václav, Lebeda, Ondřej, and Matěj, Zdeněk

Subjects
*NEUTRONS, *CYCLOTRONS, *PARTICLE accelerators, *NUCLEAR reactions, *NUCLEAR activation analysis
Abstract

Cyclotrons used to produce medical isotopes are relatively widespread. Nowadays, it is popular to place small and compact accelerators directly in hospitals. This approach simplifies handling of the produced radiopharmaceuticals, but it imposes strict radiation safety measures during production. For optimal utilization of isotope production cyclotrons, the exact knowledge of leakage neutron field is essential due to the deep penetration ability of the high energy neutrons and the accompanied secondary radiation production. Our paper presents measurement of the neutron leakage spectra for various angles from an open target assembly at the cyclotron U-120M at NPI of CAS. These spectra are compared with data obtained from a compact medical cyclotron IBA Cyclone 18/9 in UJV Rez and also with activation measurements of reactions with different threshold energies. All data were also compared with calculations made with different Monte Carlo codes using both models and data libraries. The preliminary results show significant disagreement between experiments and theoretical predictions. These findings could have implications not only to the nuclear data community but also to the production accelerators operators at the licensing stage. [ABSTRACT FROM AUTHOR]

Published
2023
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13. Measurement of various monitors reaction rate in a special core at LR-0 reactor.

Author

Košt'ál, Michal, Schulc, Martin, Šimon, Jan, Burianová, Nicola, Harutyunyan, Davit, Losa, Evžen, and Rypar, Vojtěch

Subjects
*CHEMICAL kinetics, *RADIATION dosimetry, *FLUX (Energy), *NEUTRON emission, *SPECTRAL imaging, *SPECIFIC catalysis
Abstract

Validation of the selected materials cross sections is an important task for their use as flux monitors within experimental measurements. As the reaction rate is a function of neutron spectrum and material cross section, the neutron spectrum must be well defined for the correct cross section validation. This criterion is met in the special core of the LR-0 reactor, where the reaction rates of 181 Ta(n,γ), 54 Fe(n,p), 54 Fe(n,α) and 89 Y(n,2n) reactions were measured. The 197 Au(n,γ) and 58 Ni(n,p) reactions were used as flux monitors during measurements. Results of calculation show that in case of 181 Ta(n,γ) reaction, the values reported in JEFF-3.2 provide a good agreement, while ENDF/B-VII.1 is underpredicting the experiment. In case of 54 Fe(n,p) reaction, JEFF-3.2 is in satisfactory agreement, while ENDF/B-VII.1 overpredicts the result by about 10%. The 54 Fe(n,α) shows significant discrepancies, JEFF-3.2 under predicts by about 20%, while ENDF/B-VII.1 overpredicts result by 40%. The high threshold 89 Y(n,2n) reaction is interesting from the point of view of spectral averaged cross section. It is worth noting that the resulted spectral averaged cross section, being 0.169 with 1σ uncertainty of 0.007 mb, is in very good agreement with previous result, being 0.172 ± 0.006 mb. [ABSTRACT FROM AUTHOR]

Published
2018
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15. Absolute determination of power density in the VVER-1000 mock-up on the LR-0 research reactor.

Author

Košt'ál M, Švadlenková M, and Milčák J

Subjects
Computer Simulation, Equipment Design, Equipment Failure Analysis, Energy Transfer, Models, Theoretical, Nuclear Fission, Nuclear Reactors instrumentation
Abstract

The work presents a detailed comparison of calculated and experimentally determined net peak areas of selected fission products gamma lines. The fission products were induced during a 2.5 h irradiation on the power level of 9.5 W in selected fuel pins of the VVER-1000 Mock-Up. The calculations were done with deterministic and stochastic (Monte Carlo) methods. The effects of different nuclear data libraries used for calculations are discussed as well. The Net Peak Area (NPA) may be used for the determination of fission density across the mock-up. This fission density is practically identical to power density., (Copyright © 2013 Elsevier Ltd. All rights reserved.)

Published
2013
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