Monte Carlo MCNPX Research Articles

Design and simulation of a blanket module with high efficiency cooling system of tokamak focused on DEMO reactor

Abstract In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% 6Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687 °C . As regards the tolerable temperature of beryllium (650 °C ), the design of blanket module is done in the way that beryllium temperature never exceed 600 °C . The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.

Biological polymeric shielding design for an X-ray laboratory using Monte Carlo codes.

Photon irradiation facilities are often shielded using lead despite its toxicity and high cost. In this study, three Monte Carlo codes, EGS5, MCNPX, and Geant4, were utilized to investigate the efficiency of a relatively new polymeric base compound (CnH2n), as a radiation shielding material for photons with energies below 150keV. The proposed compound with the densities of 6 and 8gcm-3 were doped with the weight percentages of 8.0 and 15.0% gadolinium. The probabilities of photoelectric effect and Compton scattering were relatively equal at low photon energies, thus the shielding design was optimized using three Monte Carlo codes for the conformity of calculation results. Consequently, 8% Gd-doped polymer with thickness less than 2cm and density of 6gcm-3 was adequate for X-ray room shielding to attenuate more than 95% of the 150-keV incident photons. An average dose rate reduction of 88% can be achieved to ensure safety of the radiation area.

Evaluation of subcritical multiplication parameters in MYRRHA-FASTEF accelerator driven system reactor

In the present study, the fast spectrum transmutation experimental facility (FASTEF) core that considered for the MYRRHA reactor is modelled using Monte Carlo MCNPX code. The effect of changing the type of material and radius of the cylindrical target source as well as the energy of the proton beam on the final production of neutrons and the subcritical multiplication of the system are evaluated. Subcritical models of the investigated reactor have been numerically investigated as well. six target materials; uranium (U), lead-bismuth eutectic (LBE), tungsten (W), lead (Pb), bismuth (Bi), and copper (Cu) are used with varying target radii from 3.5 to 20 cm. The beam energy is varied from 0.2 to 2.0GeV. The present study depends on numerical calculations of the subcritical multiplication factors and the efficiency of the external source using MCNPX code. The obtained results revealed that the favourable target material, radius, and beam energy can be precisely determined.

Monte carlo study of organ doses and related risk for cancer in Tanzania from scattered photons in cervical radiation treatment involving Co-60 source

PurposeThe present work aimed to evaluate organ doses and related risk for cancer from external beam radiation treatment (EBRT) and high-dose-rate (HDR) brachytherapy (BT) involving Co-60 source for patients with cervical carcinoma in Tanzania based on Monte Carlo methods and to evaluate the secondary cancer risks in their lifetime. MethodsEBRT and HDR-BR were modelled by using the MCNPX Monte Carlo (MC) code. The MC simulations were performed by using validated models and isocentric irradiation of an adult female computational phantom. The organ doses and cancer risks estimates were obtained. ResultsThe highest absorbed doses of 6.98 × 10−2 and 5.74 × 10−2 Sv/Gy were recorded in the bladder for BT and EBRT. The higher risk was found for colon at 1.06 × 10−3 in the HDR-BT and 9.75 × 10−5 in the EBRT per 100,000 population at exposure age of 35 years than in the other organs. The risk magnitude decreased with increasing age at exposure. In general, the secondary cancer risks in all sites considered from EBRT and HDR-BR for cervical cancer patient were lower than the baseline risks. ConclusionsThe chances of developing secondary cancer take years following radiation therapy are extremely low, but the results of present study can support to establish a future database on secondary cancer risks involving radiation therapy in patients with cervical cancer by using HDR-BR and EBRT with Co-60 source in Tanzania and other developing countries.

Neutron field study of p(24) + Be source reaction using the multi-foil activation technique

The p(24)+Be neutron producing reaction was investigated at the Nuclear Physics Institute of the Czech Academy of Sciences using the 24 MeV proton beam on thick beryllium target. Neutron field in close target-to-sample distance was determined employing the activation foils method; set of ten activation foils (Al, Au, Fe, Ni, Y, Ti, In, Co, Nb, Lu) was utilized. Neutron spectrum reconstruction from reaction rates obtained by means of gamma-ray spectroscopy was performed using the modified version of SAND-II unfolding code and relevant activation cross-sections from the EAF-2010 nuclear database. Determined neutron energy spectra were validated against the Monte Carlo MCNPX predictions. Developed p(24)+Be neutron field up to 22 MeV represents a useful tool for validation of cross-sections in energy range relevant to fusion neutrons (≥12 MeV), intensive irradiation experiments, and radiation hardness study of materials.

Assessment of the power deposition on the MEGAPIE spallation target using the GEANT4 toolkit

This work aims at evaluating the reliability of the GEANT4 (GEometry ANd Tracking 4) Monte Carlo (MC) toolkit in calculating the power deposition on the Megawatt Pilot Experiment (MEGAPIE), the first liquid–metal spallation target worldwide. A new choice of codes to study and optimize this target is provided. The evaluation of the GEANT4 toolkit is carried out in comparison with the MCNPX and FLUKA MC codes. The MEGAPIE is an international project led by the Paul Scherrer Institute in Switzerland. It aims to demonstrate the safe operation of an intense neutron source to power the next generation of nuclear reactors, accelerator-driven systems (ADSs). In this study, we used the GEANT4 MC toolkit to calculate the power deposited by fast protons on the MEGAPIE target. The calculation focuses on several structures and regions. The predictions of our calculations were compared and discussed with that of the MCNPX and FLUKA codes, adopted by the MEGAPIE project. The comparison shows that there is a very good agreement between our results and those of the reference codes.

A Monte Carlo evaluation of dose distribution of commercial treatment planning systems in heterogeneous media.

Calculations from a treatment planning system (TPS) in heterogeneous regions may present significant inaccuracies due to loss of electronic equilibrium. The purpose of this study is to evaluate and quantify the differences of dose distributions computed by some of the newest dose calculation algorithms, including collapsed cone convolution (CCC), fast Fourier transform (FFT) convolution, and superposition convolution, in heterogeneity of the lung. A 6-MV Siemens Primus linear accelerator was simulated by MCNPX Monte Carlo (MC) code, and the results of percentage depth dose (PDD) and dose profile values were compared with measured data. The ISOgray TPS was used and PDDs of CCC, FFT, and superposition convolution algorithms were compared with the results obtained by MCNPX code. CT2MCNP software was used to convert the computed tomography images of the lung tissue to MC input files, and dose distributions from the three algorithms were compared to MC method. For PDD curves in buildup region, the maximum underdosage of ISOgray TPS was at the surface (19%) and comes in closer agreement when depth increases (average 7.08%). Dose differences (DD) between different algorithms and MC were typically 4.81% (range: 1.95% to 7.30%), -1.55% (range: -5.14% to 5.26%) and 4.96% (range: 2.00% to 7.4%) in the lung for the CCC, FFT, and superposition algorithms, respectively. The difference between monitor units and maximum dose calculated using the three algorithms were 0.5% and 1.61%, respectively. The maximum DD of 7% was observed between MC and TPS results. Significant differences were found when the calculation algorithms were compared with MC method in lung tissue, and this difference is not negligible. It is recommended to use of MC-based TPS for the treatment fields including lung tissue.

TRITIUM BREEDING CALCULATION IN A STELLARATOR BLANKET

In current blanket projects, tritium breeding ratio (TBR), the ratio of tritium production rate to the neutron production rate, is low (1.1…1.2). The MCNPX Monte-Carlo code has been used to model the neutron kinetics and to look for a principal possibility of increase TBR within a stellarator blanket limited space.

Impact of Photon Spectra on the Sensitivity of Polymer Gel Dosimetry by X-Ray Computed Tomography

Introduction: The purpose of the current study was to investigate the effect of X-ray spectra on the sensitivity of a polymer gel dosimeter imaged with a conventional computed tomography (CT) scanner. Material and Methods: The whole process of CT imaging of an irradiated polymer gel was simulated by MCNPX Monte Carlo (MC) code. The imaging of polyacrylamide gel was accomplished by means of a conventional X-ray CT scan machine for different X-ray spectra, including mono-energetic beams and the spectra generated after passing through physical filters, including copper and tin. The MC-scored photon fluence inside simulated detectors was used to reconstruct the axial CT images by MATLAB software. The resultant images were used to derive the dose calibration curve of the gel for different spectra, based on which the highest sensitivity was selected. Results: Among the calculated gel sensitivities for different beam spectra, the highest increase in average sensitivity was obtained as 23% for the 140 kVp spectrum with copper filter and copper+tin filter. However, the sensitivity of mono-energetic beams showed no considerable variation with the increase of energy from 30 to 140 keV. Conclusion: As the findings indicated, the optimization of photon spectra by means of a physical filter could increase the sensitivity of polymer gels in gel dosimetry using CT imaging.

Comparison between EGSnrc and MCNPX for X-ray target in 6 MV photon beam

The Monte Carlo codes, EGSnrc and MCNPX, have been specifically developed to study the particles transport in linear accelerator used in radiotherapy. A simple X-ray target 6 MV photon beam geometry was chosen to highlight similarities and differences of particles transport algorithms implemented in these Monte Carlo codes. The purpose of this study was to compare the particles transport in X-ray target 6 MV photon beam using EGSnrc and MCNPX codes. This target has 0.2484 cm thickness consisting of 0.0899 cm Tungsten and 0.1575 Copper. The global cut-off energies used in the simulations for electron and photon were 500 KeV and 10 KeV, respectively. Monte Carlo simulations were performed for 6.4 MeV monoenergetic beam. The parameters used in this simulation were using the DEFAULT setting. The particles fluence, spectral distribution after target and simulation time were analysed in this simulation. The results showed that the spectrum from EGSnrc and MCNPX was not well-matched. There was under 5% difference, especially in the build-up region. Meanwhile, the EGSnrc was 2 to 3 times more efficient than MCNPX code to achieve certain statistical precision. This happened because the differences of transport algorithm and approximation in these MC codes. EGSnrc just follow photons and electrons during simulation, while MCNPX can simulate transport 34 types of particles, including neutrons. It also causes the simulation time difference between these two codes. MCNPX simulation took longer than the EGSnrc. However, in some cases, differences in results as well as simulation time between 2 codes can be ignored.

Determination of internal absorbed dose from radiation of111In-Rituximab in body

Introduction: There are many ways to treat cancer and pain relief. One of the methods is radiopharmaceutical therapy. Indium 111-rituximab contains the radioisotope indium-111 and rituximab. The Indium-111, after decay, emits gamma rays with energies of 245 KeV and 171 KeV, and beta-radiation of 26 KeV. In this study, the absorbed dose of 111In-Rituximab in human body was calculated using mice data and MCNPX simulation code for the organs of heart, lung, liver, spleen, kidney, stomach, intestine and pancreas. Materials and Methods: Male phantom was used for this simulation. The source and target organs were defined in code and absorbed dose obtained for per particle with error of 5 percent. The source organs were determined as volume source in SDEF card of MCNP code. Simulations were done in two steps for gamma and beta sources and the final absorbed dose of each organ were multiplied by the amount of cumulated activity of 1MBq. Results: The results showed that the absorbed dose due to injection of 111In-Rituximab for organs of spleen and stomach are 2.2252E-02, 5.5979E-03 respectively. so these organs are at risk in treating cancer with this radionuclide. The errors for were less than 5 percent. Conclusion: Monte Carlo method and MCNPX simulation code are important methods for calculation of the absorbed dose after the distribution of rradionuclides in the body. Also the results show that two organs spleen and stomach are at risk in Indium 111-rituximab therapy.

Calculation of Neutron Fluence and Energy Release in WWER-1000 Structural Components Using Monte Carlo Method

The paper describes the methodology for the calculation of neutron fluence and energy release in WWER-1000 reactor cavity and baffle using the Monte Carlo MCNPX code. It formulates an approach to the simulation of a 3-D neutron source and conditions for the transport of neutrons and photons in the core. The research presents preliminary results of neutron fluence and energy release calculation in WWER-1000 reactor cavity and baffle.

Dosimetric verification of small fields in the lung using lung-equivalent polymer gel and Monte Carlo simulation.

The main purpose of this study was evaluate a polymer-gel-dosimeter (PGD) for three-dimensional verification of dose distributions in the lung that is called lung-equivalent gel (LEG) and then to compare its result with Monte Carlo (MC) method. In the present study, to achieve a lung density for PGD, gel is beaten until foam is obtained, and then sodium dodecyl sulfate is added as a surfactant to increase the surface tension of the gel. The foam gel was irradiated with 1 cm × 1 cm field size in the 6 MV photon beams of ONCOR SIEMENS LINAC, along the central axis of the gel. The LEG was then scanned on a 1.5 Tesla magnetic resonance imaging scanner after irradiation using a multiple-spin echo sequence. Least-square fitting the pixel values from 32 consecutive images using a single exponential decay function derived the R2 relaxation rates. Moreover, 6 and 18 MV photon beams of ONCOR SIEMENS LINAC are simulated using MCNPX MC Code. The MC model is used to calculate the depth dose water and low-density water resembling the soft tissue and lung, respectively. Percentages of dose reduction in the lung region relative to homogeneous phantom for 6 MV photon beam were 44.6%, 39%, 13%, and 7% for 0.5 cm × 0.5 cm, 1 cm × 1 cm, 2 cm × 2 cm, and 3 cm × 3 cm fields, respectively. For 18 MV photon beam, the results were found to be 82%, 69%, 46%, and 25.8% for the same field sizes, respectively. Preliminary results show good agreement between depth dose measured with the LEG and the depth dose calculated using MCNP code. Our study showed that the dose reduction with small fields in the lung was very high. Thus, inaccurate prediction of absorbed dose inside the lung and also lung/soft-tissue interfaces with small photon beams may lead to critical consequences for treatment outcome.

NEUTRON FIELD MEASUREMENT OF P(35)+Be SOURCE USING THE MULTI-FOIL ACTIVATION METHOD.

Neutron field from the p+Be interaction was investigated at the NPI CAS for a proton beam energy of 35 MeV and thick beryllium target. Broad neutron spectra at close source-to-sample distances were determined using the multi-foil activation technique. Two large sets of dosimetry foils containing the Ni, Co, Au, In, Ti, Al, Y, Lu, Nb and Fe were irradiated at a distance of 74 mm at direct neutron beam axis and at a distance of 34 mm from beam axis. Supporting Monte-Carlo MCNPX calculations of the irradiation system were performed as well. From measured reaction rates, the neutron energy spectra at both positions were reconstructed employing the modified version of the SAND-II unfolding code and activation cross-section data from the EAF-2010 library. At the position of irradiated samples, the total fast neutron flux reaches the value up to 1010 cm-2 s-1, and the neutron field is utilizable for radiation hardness study and integral benchmark experiments within the International Fusion Material Irradiation Facility (IFMIF) program.

Monte Carlo Simulation of Siemens Primus plus Linac for 6 and 18 MV Photon Beams.

The aim of the present study is to simulate 6 MV and 18 MV photon beam energies of a Siemens Primus Plus medical linear accelerator (Linac) and to verify the simulation by comparing the results with the measured data. The main components of the head of Siemens Primus Plus linac were simulated using MCNPX Monte Carlo (MC) code. To verify the results, experimental data of percentage depth dose (PDD) and beam dose profile for 5 × 5 cm2, 10 × 10 cm2 and 20 × 20 cm2 field sizes were measured and compared with simulation results. Moreover, gamma function was used to compare the measurement and simulation data. The results show a good agreement, within 1%, was observed between the data calculated by the simulations and those obtained by measurement for 6 MV photon beam, while it was within 2% for 18 MV photon beam, except in the build-up region for both beams. Gamma index values were less than unity in most data points for all the mentioned energies and fields. To calculate the dose in the phantom, cells were selected in different modes, one of the modes due to the lack of dose gradient and overlapping, produced better results than others produce. There was good settlement between measured and MC simulation values in this research. The simulation programs can be used for photon modes of Siemens Primus Plus linac in conditions in which it is not possible to perform experimental measurements.

Evaluation of an alternative shielding materials for F-127 transport package

Lead is used as radiation shielding material for the Nordion's F-127 source shipping container is used for transport and storage of the GammaBeam –127's cobalt-60 source of the Nuclear Technology Development Center (CDTN) located in Belo Horizonte, Brazil. As an alternative, Th, Tl and WC have been evaluated as radiation shielding material. The goal is to check their behavior regarding shielding and dosing. Monte Carlo MCNPX code is used for the simulations. In the MCNPX calculation was used one cylinder as exclusion surface instead one sphere. Validation of MCNPX gamma doses calculations was carried out through comparison with experimental measurements. The results show that tungsten carbide WC is better shielding material for γ-ray than lead shielding.

Validation of MCNPX with Experimental Results of Mass Attenuation Coefficients for Cement, Gypsum and Mixture

Background: Shielding properties of compound or mixture is presented in terms of mass attenuation coefficients using Monte Carlo simulation. Mass attenuation coefficients of cement, gypsum and the mixture of gypsum and PbCO3 has been investigated using monte carlo MCNPX. Materials and Methods: The mass attenuation coefficients of cement, gypsum and the mixture of gypsum and PbCO3 were calculated for photon energies 365.5, 661.6, 1,173.2, and 1,332.5 keV energies. Results and Discussion: The simulated values of mass attenuation coefficients were compared avaialable experimental results, theoretical values by XCOM and found good comparability of the results. Conclusion: Standard simulation geometry used in the present investigation would be very useful for various types of sample for shielding and dosimetry applications.

NEUTRON RESPONSE FUNCTION OF BONNER SPHERE SPECTROMETER WITH 6LiI(Eu) DETECTOR

Neutron Response Function of Bonner Sphere Spectrometer With 6LiI(Eu) Detector. The detector response function was needed to measure the neutron fluence based on the count rates from Bonner Sphere Spectrometer (BSS). The determination of response function of a BSS with 6LiI(Eu) detector has been performed using Monte Carlo MCNPX code. This calculation was performed for BSS using scintillation detector of 4 mm × 4 mm 6LiI(Eu) which is placed at the center of a set of polyethylene spheres i.e bare, 2", 3", 5", 8", 10", and 12" diameters. The BSS response functions were obtained for neutron energy of 1x10-9 MeV - 1x102 MeV in 111 energy bins and each value has an uncertainty less or equal to 2 %. The response function were compared with two response functions reported in the literature i.e IAEA document in Technical Reports Series 403 (TRS-403) and the calculation from Vega-Carrillo, et al. Also validated with measurement 252Cf neutron spectra, that shown the simulated BSS spectra were quite close to the experimental measured with a differrence of 3%.

X-ray sterilization of insects and microorganisms for cultural heritage applications

The APAM (Development of Particle Accelerators and Medical Applications) Laboratory of the ENEA Frascati Research Center is engaged in the preservation of cultural heritage as part of the COBRA (Sviluppo e diffusione di metodi, tecnologie e strumenti avanzati per la COnservazione dei Beni culturali, basati sull’applicazione di Radiazioni e di tecnologie Abilitanti) project addressed to the transfer of innovative technologies and methodologies from research to small and medium enterprises involved in the restorative measures. This work aims to demonstrate the effectiveness of ionizing radiation on the disinfection of biodegraded art objects. The conventional methods for the disinfestation of works of art, using chemicals toxic to humans and environment, might cause some damage to the treated material even on micrometric scale (i. e. either cellulose degradation). Ionizing radiations interact with the infesting biological material causing an irreversible DNA degradation. For this reason, they are certainly suitable for removal treatments of both macro organisms and bacterial colonies. A 4.8MeV electron linear accelerator, normally dedicated to the characterization of dose detectors and radiographies, has been employed to produce Bremsstrahlung X-rays through a lead converter. The spectral fluence of the radiation source has been calculated using the Monte Carlo MCNPX code.The dosimetric characterization of the radiation field has been made using radiochromic films sensitive in the dose range of our interest (from 50 to 500Gy) calibrated with a Markus ionization chamber. The irradiation of the artifact prototypes are made within a lead shielded room at a variable distance from the X-rays source. Samples subjected to irradiation consist of a soil bacterium, Agrobacterium rhizogenes, and an insect, Stegobium paniceum, that are found as wall paintings invasive coloniser and as a pest of books, wood works and paintings, respectively. Tests of irradiation have been performed on pest organisms as well as on woods mock-ups to evaluate potential damage to the material during the sterilization. The growing capacity of the treated bacterial cells re-cultured at the end of the treatment was evaluated on the bacterial sample and resulted to strongly inhibit cell growth during post-irradiation incubation, so that after incubation periods at 28°C, no significant cell growth was observed. The induced levels of insect mortality and sterility vs absorbed dose and operative conditions have been also evaluated, demonstrating the induction of full sterility since the lower dose and 40% mortality by two days after the higher dose treatment. The experiments proved the ability to efficaciously treat objects of cultural heritage with X-rays in order to prevent the increase of the biodeterioration without damaging the materials: in fact, mechanical tests on both irradiated and not irradiated woods have demonstrated the absence of any induced degradation after the radiation exposition.

Mamografi Çekimlerinde Absorbe Doz Hesaplamaları için MCNPX Monte Carlo Kodunun Değerlendirilmesi

Introduction: This study aimed to investigate capabities of MCNPX monte carlo code for calculations of average absorbed dose in a breast phantom during mammography examination. Also, the effect of tube voltage and breast thickness on absorbed dose was determined by using Monte carlo method. Materials & Methods: In this study, the average absorbed dose values were investigated by using MCNPX (version 2.4.0) Monte Carlo code. Validation of modeled simulation setup has been obtained by calculating the HVL values of Al attenuator material for different tube voltage (26 to 32 kV ). Obtained HVL values were found to be comparable with experimental results and available Monte Carlo results. Thus, validated simulation setup has been used for investigation of average absorbed dose in a %50-%50 grandular-adipose breast phantom. Results: Our calculated results was consistent with the previous studies and our MC model was validated. The results showed that changing tube voltage from 26 to 32 kV leads to about 4 times increase in breast dose. While increasing breast thick ness from 2 to 5 cm results in 1.6 times higer dose to breast. Conclusion: It can concluded that the breast recieved dose depends strongly beam quality and breast thickness. Our modeling using MCNPX can be used future dosimteric studies concering breast dose investigations in mamograpgy examinations.