Browsing by Author "Tracz, Grzegorz"
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Item Angular-energy distributions of neutrons emitted from the Wendelstein stellarator – Monte Carlo simulations(Institute of Nuclear Physics Polish Academy of Sciences, 2009) Tracz, Grzegorz; Drozdowicz, KrzysztofAngular-energy distributions of neutron currents expected from the Wendelstein 7-X stellarator have been calculated by means of Monte Carlo methods using a simplified model of the stellarator. The MCNP5 code has been employed. The obtained spectra are to be used to design five sets of neutron and gamma detectors placed above this D-D fusion reactor. The counters are to provide information concerning both thermonuclear plasma diagnostics and dosimetry.Item Assumptions for the design of a neutron pinhole camera dedicated to the PF-24 device(Institute of Nuclear Physics Polish Academy of Sciences, 2014) Scholz, Marek; Bielecki, Jakub; Wójcik-Gargula, Anna; Wiącek, Urszula; Drozdowicz, Krzysztof; Igielski, Andrzej; Kulińska, Agnieszka; Tracz, Grzegorz; Woźnicka, UrszulaThe report presents main assumptions on the design of the neutron pinhole camera dedicated to the PF-24 (Plasma Focus) device. The pinhole camera will be used for the investigation of the spatial and temporal distributions of DD neutrons from the PF-24 source. It makes use of principles of the optical geometry adopted for neutron imaging. In the report the evaluation of pinhole geometrical layout has been made on the basis of principles of the geometrical optics. A further optimization of the pinhole geometry has been carried out by means of neutron transport calculations (the MCNP code). The main aim of this report is to provide information on technical solutions for the neutron pinhole.Item Budowa modeli geochemiczno-mineralogicznych na podstawie danych uzyskiwanych z zaawansowanych metod geofizyki jądrowej. Część 1(Institute of Nuclear Physics Polish Academy of Sciences, 2007) Woźnicka, Urszula; Cywicka-Jakiel, Teresa; Drabina, Andrzej; Dworak, Dominik; Tracz, Grzegorz; Drozdowicz, Krzysztof; Zorski, Tomasz; Ossowski, Andrzej; Środoń, Jan; Kawiak, Tadeusz; Bakowska, Dorota; Zielińska, Małgorzata; Zalewska, JadwigaAn extensive set of conventional and special core analysis measurements, complemented by additional chemical and mineralogical analyses was performed, forming the basis for the evaluation of the complex thin bedded gas-bearing shaly-sand formations of Miocene age in the Carpathian Foredeep region of Southern Poland. Mineralogy was established by using full elemental composition, XRD method, total surface area and cation exchange capacity (CEC). The analysis showed that the most important clay component is illite – smectite mixed layer. Statistical multivariate analysis of all data helped to set up a comprehensive general petrophysical model. For the notoriously difficult CEC information for the rock matrix we could establish a reliable correlation (corr. coefficient r2 around 0.95) between thermal neutron absorption cross section SIGMA and total natural radioactivity GR with CEC, with boron and rare earth elements the two most important SIGMA contributors in the rock matrix. This good correlation permits a continuous on-line CEC determination and therefore a reliable application of the Waxman-Smits water saturation model to properly take into account the clay mineral effects in evaluating water saturation.Item Calculations of neutron and gamma energy spectra in surroundings of the HRNS for ITER(Institute of Nuclear Physics Polish Academy of Sciences, 2016-03) Tracz, GrzegorzAn optimisation of the High Resolution Neutron Spectrometer HRNS) shield was performed. Concrete and polyethylene shields of different thicknesses were considered. Two boxes for two sets of detectors are to be built of polyethylene though the underside is planned to be made of concrete. The final external dimensions of the first box are to be 175 cm (length), 150 cm (width) and 170 cm (height) while the second box is to be 215 cm long, 150 cm wide and 200cm high. All the shield elements are to be 10 cm thick including a wall between the boxes. Next, photons and scattered neutrons spectra in the first box detectors were calculated. The results will be used as a source to estimate the background contribution on the different detectors. Finally, the neutron and photon spectra through the inner and outer surfaces of the first box were computed. The results are to be used as gamma and neutron sources in subsequent evaluations of shut-down dose rates in the port cell. A detailed MCNP input file was elaborated, based on a CAD model, which geometric part contains over 600 cells of HRNS, not counting void.Item Calculations of the neutron-induced activity in air inside the Cuboid 2 of the High-Resolution Neutron Spectrometer for ITER(Institute of Nuclear Physics Polish Academy of Sciences, 2017-01) Wójcik-Gargula, Anna; Tracz, Grzegorz; Scholz, Marek; Drozdowicz, KrzysztofThis report presents results of the calculations performed in order to predict the neutron - induced activity in air inside the second Cuboid in the ITER equatorial port cell #1 where the Time of Flight detectors of the High Resolution Neutron Spectrometer are planned to be installed. Based on the neutron spectra generated by 2.45 MeV and 14 MeV neutron sources, calculated with the MCNP code, it was possible to determine the activity of the dominant radionuclides using the FISPACT-II Transmutation - Activation Inventory Code and the EAF - 2010 nuclear data library. The results show that although long - lived isotopes such as e.g. H-3 or C-14 are produced following neutron irradiation, their activity will not be significant since they will not exceed the maximum permissible levels.Item Detection of delayed neutrons from neutron activation of fissionable substance samples. Monte Carlo modelling of response of the DET-12 device(Institute of Nuclear Physics Polish Academy of Sciences, 2013) Tracz, Grzegorz; Bieńkowska, Barbara; Drozdowicz, KrzysztofActivation of fissionable elements by neutrons has been considered as a possible diagnostics of D-D and D-T fusion plasma. Fission caused by fusion neutrons leads up to emission of secondary neutrons: prompt and delayed. Proper interpretation of the time decay of delayed neutrons enables an assessment of the parameters of the primary neutron flux inducing fission. Monte Carlo calculations have been carried out by means of the MCNP code in order to elaborate the method considered. Three nuclides: pure 235U, 238U and 232Th, and additionally sintered UO2 have been selected as possible materials for the sample to be irradiated. Four energies of neutrons irradiating samples have been chosen: thermal, fast and two high ones. Computations have been accomplished for two variants: pure physical effect and real experimental conditions of neutron registration in the DET-12 device designed and built in IFJ PAN. Decay curves have been obtained for each case. Detection efficiency of DET-12 has been also estimated.Item Monte Carlo simulations of neutron and photon radiation fields at the PF-24 plasma focus device at IFJ PAN in Krakow(Institute of Nuclear Physics Polish Academy of Sciences, 2016-04) Tracz, Grzegorz; Wiącek, Urszula; Bieńkowska, BarbaraThe medium scale PF‐24 facility was installed at the Institute of Nuclear Physics of the Polish Academy of Sciences (IFJ PAN), Poland. The MCNP model of the PF ‐ 24 device in the main hall of the laboratory was elaborated. Two variants of the plasma source emitting particles were considered: a point neutron source and a volume source with photon and neutron emission. Based on presented calculations the influence of the laboratory construction (the walls, the ceiling and the floor) on neutron and photon space distributions in the main hall were assessed. A study of fast uncolided and collided neutrons contributions to the total field of neutrons was performed. The contribution of photons emitted directly from the source and created as the result of neutron scattering was established. Neutron spectra at selected points were calculated.Item Optimization of the fission-converter and filter set-up for the boron-neutron capture therapy (BNCT)(Institute of Nuclear Physics Polish Academy of Sciences, 2003) Tracz, Grzegorz; Dąbkowski, Ludwik; Pytel, Krzysztof; Woźnicka, UrszulaThe paper presents the third step of the numerical modeling of the fission-converter-based epi-thermal neutron source designed for the Polish Boron Neutron Capture Therapy (BNCT) facility to be located at the Polish research nuclear reactor MARIA at Świerk. The optimization of the fission con-verter has been carried out again. The epithermal neutron flux has increased 240 % comparing with the variant proposed previously while the number of fuel rods was significantly reduced. The specific photon and fast neutron doses meet the requirements of the therapy. Optimization of the reflector sur-rounding the filter/moderator as well as collimator shape, length and liner has been also carried out. Configuration of the filter/moderator has remained the same. Criticality calculations show that keff of the fission converter filled with light water is below 1. The MCNP code has been used during compu-tations.Item Preliminary study of a target-moderator assembly for a linac-based neutron source(Institute of Nuclear Physics Polish Academy of Sciences, 2005) Bartalucci, Sergio; Drozdowicz, Krzysztof; Dworak, Dominik; Tracz, Grzegorz; Angelov, VladimirThe report concerns a design of a future pulsed neutron source at an electron linac. A massive target is irradiated with an electron beam and the neutrons are generated mainly by collisions of the bremsstrahlung photons. A first step of the work, related to the optimization of the target materials and geometry using numerical simulations, is presented. The Monte Carlo FLUKA and MCNP codes are used. The water-cooled tantalum target is investigated: 0.41 cm Ta slices separated with 0.15 cm H2O layers. Two different sizes of the cylindrical target are assumed: 5 or 2.5 cm in diameter. The 1 GeV and 1.5 GeV electron beams are tested. The outgoing neutron angular-energy spectra are presented. The angular space 0°÷180° is divided in 10-degree intervals. The neutron emission in the direction perpendicular to the originated electron beam has been observed in the particularly narrow (2°) interval. The FLUKA results of a comparison of the neutron currents in main directions (0° – forward, 90° – perpendicular, 180° – backward) are as follows. For the 5 cm target the distribution is quite uniform: at 1 GeV input electrons – n(180°)/n(0°) = 1.00, n(90°)/n(0°) = 1.04, and at 1.5 GeV electrons – n(180°)/n(0°) = 0.91, n(90°)/n(0°) = 1.00. For the 2.5 cm target the relative neutron current in the perpendicular direction is significantly higher: at 1 GeV electrons – n(180°)/n(0°) = 1.02, n(90°)/n(0°) = 1.53, and at 1.5 GeV electrons – n(180°)/n(0°) = 0.93, n(90°)/n(0°) = 1.45. In the cases when the FLUKA and MCNP simulation results can be compared, a high similarity of the neutron energy distributions is stated although a possible discrepancy of the values reaches 20 %. Spectra of the accompanying radiation (photons, electrons, positrons, protons, charged pions) have been also obtained. The angular distributions of photons, electrons, and positrons are strongly peaked up towards the beam direction. Their emission at 90° is significantly lower, which means a decrease of the background in this direction. The energy deposition in the target is estimated on a simplified model with no cooling system. About 63 % of energy is then stored in the space at ca. 20÷40 % target length along the initial electron beam axis.Item Technical design and operation tests of the DET-12 device for detection of delayed neutrons(Institute of Nuclear Physics Polish Academy of Sciences, 2014) Bieńkowska, Barbara; Drozdowicz, Krzysztof; Gabańska, Barbara; Igielski, Andrzej; Janik, Władysław; Kurowski, Arkadiusz; Tracz, Grzegorz; Wiącek, Urszula; Dankowski, JanA technical design of the device for detection of delayed neutrons emitted from neutron-activated fissionable material samples has been performed according to physical assumptions which were earlier elaborated. The DET-12 device was constructed. The detection system was composed, consisting of twelve 3He neutron detectors, related electronics lines, and the data acquisition and recording system. The detectors were adjusted to work in groups by three connected to one preamplifier, considering a weak intensity of emission of the delayed neutrons. Laboratory measurement tests of the device operation were made with use of an isotopic neutron source. A total efficiency of neutron detection in DET-12 was experimentally determined and a relative benchmark calculation was made by means of a Monte Carlo modelling of the neutron transport in the device from the source to detectors.Item Wtórne źródła neutronowe do generowania specyficznych strumieni neutronów(Institute of Nuclear Physics Polish Academy of Sciences, 2007) Tracz, GrzegorzThe foregoing paper presents the doctor’s thesis entitled “The secondary neutron sources for generation of particular neutron fluxes”. Two secondary neutron sources have been designed, which exploit already existing primary sources emitting neutrons of energies different from the desired ones. The first source is devoted to boron-neutron capture therapy (BNCT). The research reactor MARIA at the Institute of Atomic Energy in Świerk (Poland) is the primary source of the reactor thermal neutrons, while the secondary source should supply epithermal neutrons. The other secondary source is the pulsed source of thermal neutrons that uses fast 14 MeV neutrons from a pulsed generator at the Institute of Nuclear Physics PAN in Kraków (Poland). The physical problems to be solved in the two mentioned cases is are different. Namely, in order to devise the BNCT source the initial energy of particles ought to be increased, whilst in the other case the fast neutrons have to be moderated. Slowing down of neutrons is relatively easy since these particles lose energy when they scatter in media; the most effective moderators are the materials which contain light elements (mostly hydrogen). In order to increase the energy of neutrons from thermal to epithermal (the BNCT case) the so-called neutron converter should be exploited. It contains a fissile material, 235U. The thermal neutrons from the reactor cause fission of uranium and fast neutrons are emitted from the converter. Then fissile neutrons of energy of a few MeV are slowed down to the required epithermal energy range. The design of both secondary sources have been conducted by means of Monte Carlo simulations, which have been carried out using the MCNP code. In the case of the secondary pulsed thermal neutron source, some of the calculated results have been verified experimentally.