McNp Simulation

This paper deals with the processes involved in the generation of reliable experimental<br> data for the validation of computer simulations. In the field of neutron noise, the analysis<br> of results is based on spectral features of the detector signals in the frequency domain.<br> Neutron noise simulators also produce estimates that are subjected to studies in the same<br> domain. The validation process of such simulations begins with the generation of reliable<br> experimental data. In this work, we analyze results from two neutron noise experimental<br> campaigns. The focus is placed upon de comparison of results obtained by different data<br> acquisition systems (DAQs) that were used to record the data in parallel. The goal is to<br> verify whether results obtained by the different DAQs are consistent, and thus reliable.<br> The neutron noise-dedicated experiments were carried out in the AKR-2 reactor at the<br> Technis...

In this study, utilizing the relationships between the energy of atomic/nuclear particles released from underground piezoelectric rocks and the elastic energy stored in these rocks, we introduced some methods to estimate the time/energy of incoming earthquakes in aseismic regions by measuring the energy of radiated particles. Since piezoelectric granite rocks make up approximately 60% of the Earth's crust, the increase in the energy of the detected particles in a certain period of time can be considered as an important precursor for the impending shallow earthquake. This analysis holds significant promise for enhancing the earthquake time and energy estimation methodologies. The detection of radiated particles from piezoelectric rocks can be achieved by utilizing detectors placed either on the surface or inside deep wells that are drilled near active faults. However, it is essential to note that the presented methodologies are approximations, as they rely on constant parameters for the piezoelectric material and presuppose that earthquakes occur within a piezoelectric block.

The dose measurement using Al 2 O 3 OSL dosimeter (OSLD) was carried out at low and high energy x-ray. The dose at low energy x-ray was measured at 40, 71 and 125 kVp x-ray energies. The dose ar high energy x-ray was measured at 6 and 10 MV x-ray energies measured at the depth of maximum dose (Z max). The results were compared to that in ionization chamber and LiF: Mg,Ti thermoluminescent dosimeters (TLD100). The results showed that the dose of OSLD were less in agreement to ionization chamber compared to that in TLD100. The dose of OSLD however was in good agreement to that in ionization chamber at high energy x-ray. The dose measured using OSLD were found to be more consistence at high energy x-ray shown by the standard deviation of the readings. The measurement of x 2 showed that the readings of OSLD were close to that in ionization chamber with values of 2.21 and 4.63 for 6 and 10 MV respectively. The results indicated that OSLD is more suitable for dose measurement at high energy x-ray.

In thin low-Z media irradiated by photon energies of several tens of keV, the presence of a high-Z additive can result in manifest locally modified secondary electron dose. Present study analyses the photoelectron dose enhancement resulting from nanometre thickness gold (atomic number Z = 78) coated on commercial borosilicate (B 2 O 3) glass microscope cover-slips. Two thicknesses of B 2 O 3 cover-slip have been utilized, 0.13 ± 0.02 mm and 1.00 ± 0.01 mm, with single-sided Au coatings of 20, 40, 60, 80 and 100 nm. An additional uncoated glass slide has been kept as a comparator. The samples have been exposed to X-rays generated at kVp potentials, delivering a fixed dose of 2 Gy. Dose enhancement resulting from the 1.00 mm glass has been observed to bẽ 1.32 × that of the 0.13 mm thickness glass. The elemental composition of the samples has been obtained via Electron Dispersive X-ray (EDX), elemental content differences between the two thicknesses of glass leading to a difference in effective atomic number of less than 0.3%. The influence on photon yield of the gold coating and variations in elemental content has been modelled using Monte Carlo simulation, allowing comparison with the measured values of enhanced TL yield.

In thin low-Z media irradiated by photon energies of several tens of keV, the presence of a high-Z additive can result in manifest locally modified secondary electron dose. Present study analyses the photoelectron dose enhancement resulting from nanometre thickness gold (atomic number Z = 78) coated on commercial borosilicate (B 2 O 3) glass microscope cover-slips. Two thicknesses of B 2 O 3 cover-slip have been utilized, 0.13 ± 0.02 mm and 1.00 ± 0.01 mm, with single-sided Au coatings of 20, 40, 60, 80 and 100 nm. An additional uncoated glass slide has been kept as a comparator. The samples have been exposed to X-rays generated at kVp potentials, delivering a fixed dose of 2 Gy. Dose enhancement resulting from the 1.00 mm glass has been observed to bẽ 1.32 × that of the 0.13 mm thickness glass. The elemental composition of the samples has been obtained via Electron Dispersive X-ray (EDX), elemental content differences between the two thicknesses of glass leading to a difference in effective atomic number of less than 0.3%. The influence on photon yield of the gold coating and variations in elemental content has been modelled using Monte Carlo simulation, allowing comparison with the measured values of enhanced TL yield.

A new fuel management pattern has been applied to the RSG-GAS Research reactor operation. The new fuel management pattern is done so that the replacement of fuel in every operating cycle is more efficient. The new fuel replacement pattern required various safety analyzes, among others related to the fuel fraction and radioactivity inventory of the RSG-GAS. Both of those are the main elements in dose acceptance analysis for workers and communities around the reactor when the reactor operates both in normal or in abnormal conditions. Fuel burn-up and inventory analysis was performed using ORIGEN2 computer program. Inputs for ORIGEN2 are the fuel mass, the time required for one operating cycle as well as the peak power factor in each fuel in the specified fuel management pattern. The result for the 97 th Core configuration (T97) is that the average burn-up fraction of each cycle is 6.79%, and the maximus fuel fraction is 52.36% for Fuel Elemet and 56.52% for Contro Elemet. It is also obtained that the largest and dangerous human inventory activity at the end of cycle (EOC) for Iodine radionuclide group is I-131 for 5.18E+04 Ci, and for the alkali metal radionuclide group Cs-137 of 7.65E+03 Ci.

Monte Carlo simulations provide accurate results for the neutronic response of the system under consideration if modeling is performed appropriately since it has great influence on the results. Sensitivity analysis of modeling approaches for geometry and fissile material composition distributions in the reactor core was performed by taking ITU TRIGA Mark II Research Reactor into consideration. The method of defining fuel element positions in the core by using circular or hexagonal lattice was considered as one case and three different methods of lumping material compositions in the fuel elements was considered as another case since these approaches are used by deterministic codes hence the accuracy of deterministic codes were also investigated. The validation study showed that both MCNP and Serpent Monte Carlo codes resulted in good agreement with the experimental data. It was observed that the handling of fuel composition in different ways did not influence the results significantl...

The transfer function is a basic characteristic of every nuclear reactor. It describes how a perturbation at a given place and time influences the neutron flux. In case of a known perturbation, the determination of characteristic reactor parameters is possible. The present paper shows an experimental method to determine the gain of the zero-power reactor transfer function (ZPTF) of the AKR-2 reactor at TU Dresden and the comparison to the theoretical shape of the ZPTF derived from kinetic parameters simulated with MCNP. For the experiments, a high-precision linear motor axis is used to insert an oscillating perturbation acting at frequencies smaller than the lower bound of the plateau region of the ZPTF. For higher frequencies, a rotating absorber is used. This device emulates an absorber of variable strength. The reactor response is detected with a He-3 counter. The data evaluation shows good agreement between measured and corresponding theoretical values of the gain of the ZPTF.

Am-241 sample was irradiated in spallation neutrons produced in ADS setup QUINTA at the JINR in Dubna. The energy was 660 MeV in the proton beam. The incineration study method was based on gamma-ray spectrometry. During the analysis of the spectra, several fission products were identified. Fission product activities yielded the number of fissions. Nevertheless, the lines are assumed to belong to the neutron capture product covered by parasitic Np-238 decay lines. The Np-238 lines as a result of neutron capture by Np-237 made impossible to determine the number of captures in Am-241.

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

In this work, we propose to analyze the dosimetric properties of the glucose Electron Paramagnetic Resonance (EPR) system. This study consists of irradiating samples of glucose by megavoltage X photons and electrons delivered by a particle accelerator used in radiotherapy. After irradiation, the samples were analyzed by EPR. The measurements obtained on samples placed at different depth in a phantom of PolyMethyl Methacrylate "PMMA" have firstly identified the depths corresponding to the maximum dose deposition for both particles and samples used. Analysis by EPR showed that the threshold measurable dose for this dosimetric system is about 2 Gy. The variation of curves obtained from the peak-to-peak EPR intensity, or the results of the double integration of the spectrum as a function of the irradiation dose, are linear with a correlation coefficient exceeding 0,999. Glucose seems slightly more sensitive to megavoltage X photons than megavoltage electrons. The results reveal interesting potential dosimetry of the glucose/EPR system in the field of radiation therapy.

In this work, we propose to analyze the dosimetric properties of the glucose Electron Paramagnetic Resonance (EPR) system. This study consists of irradiating samples of glucose by megavoltage X photons and electrons delivered by a particle accelerator used in radiotherapy. After irradiation, the samples were analyzed by EPR. The measurements obtained on samples placed at different depth in a phantom of PolyMethyl Methacrylate "PMMA" have firstly identified the depths corresponding to the maximum dose deposition for both particles and samples used. Analysis by EPR showed that the threshold measurable dose for this dosimetric system is about 2 Gy. The variation of curves obtained from the peak-to-peak EPR intensity, or the results of the double integration of the spectrum as a function of the irradiation dose, are linear with a correlation coefficient exceeding 0,999. Glucose seems slightly more sensitive to megavoltage X photons than megavoltage electrons. The results reveal interesting potential dosimetry of the glucose/EPR system in the field of radiation therapy.

The higher the transparency of any civilian nuclear power program, the higher the chance of attracting and securing the longterm foreign nuclear cooperation. For newcomer states, securing nuclear cooperation is essential for successful deployment and implementation of nuclear power program. Complying with an acceptable types of safeguards commitment/protocols plays a major role in increasing transparency. The determination of transparency primarily relies on the presence of sensitive nuclear isotopes-as defined under IAEA's safeguards-with a nuclear facility. Kingdom of Saudi Arabia (KSA) is considering the deployment of civilian nuclear power program with a projected nuclear capacity ~18 gigawatt-electric (GWe) by 2032-40. The goals of this paper were the quantification of the sensitive nuclear isotopes (primarily plutonium) that will be produced within the prospective KSA nuclear facilities up to 2040 and the estimation of the uranium fuel requirements. Two scenarios were analyzed. Scenario-I: two reactors are operational started by 2022 and one reactors are added each year subsequently until the intended 11 reactors are deployed. Scenario-II is like Scenario-I, but only one reactor is added each two years subsequent to the deployment of the first 2 reactors in 2022. Simulation of EPR operation was performed from beginning of life to equilibrium cycle using Monte Carlo N-Particle (MCNP6) code. A 2-year cycle length was assumed. The proposed KSA civilian nuclear power program would require 5766 and 4585 tonnes of cumulative uranium by 2040 for Scenario-I and Scenario-II respectively. The discharged fuel (assuming full power at 90% capacity factor) would contain 17.6 and 13 tonnes of cumulative 239 Pu along with 21.4 and 15.9 tonnes of cumulative total plutonium by 2040 for Scenario-I and Scenario-II respectively. A primary concern related to transparency is the ability and readiness of KSA to handle these quantities of special nuclear material under internationally acceptable safeguards protocols as the planed nuclear power program expands. It's recommended that KSA have AP in place well before 2040 to enable IAEA draw the broader safeguards conclusion which definitely will raise the confidence of the international community.

Original CT slice deck of a Roman tomato .…………………………... 3.3 A CT image of a Roman tomato, showing artifact (tapes) …………... 3.4 A processed CT image of a Roman tomato ..………………………….. 3.5 All processed CT slices of a Roman tomato ..………………………… 3.6 Isosurfaces of a chicken carcass from CT data ……………………….. 3.7 Cutaway of a chicken carcass from CT data ………………………….. 3.8 Flow diagram of electron transport (Bielajew, 2001) ………………… 3.9 Beam scanning configuration at 10 MeV linac ……………………….. 3.10 Reference and body-attached coordinate system (Fu, Gonzales, & Lee, 1987) ……..

This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a ''VVER-1000 LEU Assembly Computational Benchmark'', at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

Am-241 sample was irradiated in spallation neutrons produced in ADS setup QUINTA at the JINR in Dubna. The energy was 660 MeV in the proton beam. The incineration study method was based on gamma-ray spectrometry. During the analysis of the spectra, several fission products were identified. Fission product activities yielded the number of fissions. Nevertheless, the lines are assumed to belong to the neutron capture product covered by parasitic Np-238 decay lines. The Np-238 lines as a result of neutron capture by Np-237 made impossible to determine the number of captures in Am-241.

) is a project on development and practical use of a universal computer code for simulation of particle transport (neutrons, photons, electrons, positrons) in three-dimensional systems by means of the Monte Carlo method. This paper provides the information on the current state of the project. The developed libraries of constants are briefly described, and the potentialities of the MCU-5 package modules and the executable codes compiled from them are characterized. Examples of important problems of reactor physics solved with the code are presented.

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

 This paper aims to compare three Monte Carlo (MC) burnup based codes, i.e. MCNP6, Monteburns and Serpent on a future prototype reactor, named ALLEGRO, based on Gas cooled Fast Reactor (GFR) technology. GFR reactors are one of the proposed Generation-IV fast reactors; ALLEGRO facility is scheduled to be built in Europe as a GFR demonstrator, so its deepened simulation can help in its future development. The present study follows other researches already performed and aims to exhibit the different approaches in burnup calculations applied to a gas cooled fast reactors, i.e. this paper would like to show and to compare some results concerning nuclear parameters as keff and flux spectra, as well as the mass inventories versus burnup for some nuclides evaluated with different Monte Carlo codes. From obtained results, it seems to exist some differences in evaluation of nuclear parameters, mainly in effective multiplication factor and in mass inventories. The remaining differences are mai...

The delayed neutron emission of fissile isotopes after irradiation can be described by the six-groups theory. Due to the different half-lives and yields of the fission fragments, the intensity-time curve of delayed neutrons has an individual shape for different fissile isotopes. According to this characteristic, a new rapid analysis method has been developed to determine the concentration of different isotopes in a mixture, such as 235 U and 239 Pu, without any chemical separation. The intensity-time curve of delayed neutrons is measured by using the multichannel scanning (MCS) method with five 3 He-detectors installed on a fast transfer system at the research reactor Munich (FRM). This method was validated by about 200 measurements with different isotope contents even below the microgram region.

Two Np-237 samples and one Pu-239 were irradiated in spallation neutrons produced in ADS setup QUINTA. The accelerated beam consisted of protons of energy 660 MeV. The method was based on gamma-ray spectrometry measurement. During analysis of the spectra several fission products and one actinide were identified. Fission product activities gave the number of fissions. The actinide (Np-238), a result of neutron capture by Np-237 gave the number of captures. In a similar manner the number of fissions in Pu-239 was determined. The Pu-240, a product of neutron capture by Pu-239, activity was impossible to measure.

Two Np-237 samples and one Pu-239 were irradiated in spallation neutrons produced in ADS setup QUINTA. The accelerated beam consisted of protons of energy 660 MeV. The method was based on gamma-ray spectrometry measurement. During analysis of the spectra several fission products and one actinide were identified. Fission product activities gave the number of fissions. The actinide (Np-238), a result of neutron capture by Np-237 gave the number of captures. In a similar manner the number of fissions in Pu-239 was determined. The Pu-240, a product of neutron capture by Pu-239, activity was impossible to measure.

This study aims to verifying the nuclear criticality criteria in spent fuel repository. The repository is an open pool type that is used temporary to store the spent fuel elements removed from the core of 22 MW power of open pool research reactor. The repository could receive up to 800 of spent fuel elements of material testing reactor (MTR) type. Arrangement configuration of spent fuel elements in the repository must verify the criticality criteria during the storage that mean that the multiplication factor should not exceed 0.85. Two models were developed in this study using MCNP5 code to simulate the spent fuel elements configuration including; the basket model, and the storage pool model (contains the maximum permissible capacity). The result shows that the multiplication factors would be less than the permissible limit for the criticality criteria of the storage.

ABSTRACT Copyright 2020 by Shdeed TA. This is an open-access article distributed under Creative Commons Attribution 4.0 International License (CC BY 4.0), which allows copying, redistribute, remix, transform, and reproduce in any medium or format, even commercially, provided the original work is properly cited. 46 Original Research | Volume 4 | Issue 2| cc Purpose To read the digital imaging and communications in medicine (DICOM) images of brain and extract intensity values and build a three dimensional model for Monte Carlo n-particle transport (MCNP) code input file in purpose to study the average particle flux and deposited energy of X-Ray photons resulting at 120 kVp and 1 mAs (form point source) as function to DICOM pixel numbers in the brain tissues for a 29-year-old female patient using MCNP code and Matlab program to read the DICOM images. Method The matrix laboratory (MATLAB) program was used to read the DICOM images and extract the intensity values in each pixel of the DICOM image corresponding to certain slice of the brain. These color levels are characteristic of different tissue, and have been relied upon to create the specific material in each volume element in MCNP input file. Results The percentage of tissues with computed tomography (CT) numbers less than 100 (except air between source and patient) equals to 87.11% and the sum of deposited energy over these regions equals to 1.77×10 10 , but the tissues with CT numbers greater than 100 equals to 12.89% and absorbed dose in these tissues reach to 9.12×10 9. Also the X-ray photon flux can be reach to 6.30×10 11 at surface of skin, and drops to 10 9 when passing the bones of the skull and skin where photons deposit most of their energy. Conclusion Values of the deposited energy at surface of skin are high, so it is always necessary to be cautious when performing the examination to obtain acceptable images from the first time and without having to repeat the imaging again for the same case unless there are necessities for it.

The aim of this study was to compare the energy dependence of Radio-Photoluminescent Glass Dosimeter (RPLGD) with thermoluminescent LiF:Mg,Ti (TLD-100) and Optically stimulated luminescent (Al2O3:C) dosimeters. In this work, a Monte Carlo simulation with MCNP5 was carried out to estimate the energy responses of these dosimeters. The following common clinical radiation qualities were applied: 50-300 kVp X-rays, 4 – 20 MeV electron beam and 60Co γ rays as the reference. The current study is compared to previous measurements and MC simulation in literature. For kilovoltage photon beams, the dosimeters displayed an increasing response with decreasing energy with a significant over response factor at 70 kVp of RPLGD and Al2O3:C about 4.39 and 3.40 respectively, and a maximum of about 1.40 for TLD-100. For electron beams were found to be with a maximum of 5.1% and 3.5% for RPLGD and TLD-100 dosimeters respectively, for OSLD a maximum of 6% in 6-20 MeV and 11% at 4 MeV. This paper has clearly confirmed that all three dosimeters should be used with an energy compensation filter in low energy X-ray (<250 kV). All studied dosimeters are recommended for clinical and experimental use particularly in radiotherapy.

This paper aims to compare three Monte Carlo (MC) burnup based codes, i.e. MCNP6, Monteburns and Serpent on a future prototype reactor, named ALLEGRO, based on Gas cooled Fast Reactor (GFR) technology. GFR reactors are one of the proposed Generation-IV fast reactors; ALLEGRO facility is scheduled to be built in Europe as a GFR demonstrator, so its deepened simulation can help in its future development. The present study follows other researches already performed and aims to exhibit the different approaches in burnup calculations applied to a gas cooled fast reactors, i.e. this paper would like to show and to compare some results concerning nuclear parameters as keff and flux spectra, as well as the mass inventories versus burnup for some nuclides evaluated with different Monte Carlo codes. From obtained results, it seems to exist some differences in evaluation of nuclear parameters, mainly in effective multiplication factor and in mass inventories. The remaining differences are mainly related to calculation time: indeed between the fastest, that is SERPENT, and the slowest, that are MCNP6 and MONTEBURNS, the differences are about one order of magnitude. As far as precision is concerned, it was considered the standard only for effective multiplication factor and it seems that all codes are in good agreement.

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.