Monte Carlo MCNP Research Articles

Activity and Dose Rate Calculations for Joint European Torus Outer Long-Term Irradiation Station during Tritium and Second Deuterium Tritium Experiment Campaigns

The Joint European Torus (JET) is playing an important role in preparing for the operation of the future world’s largest tokomak, ITER. In this respect, the tritium campaign (C40) and second deuterium–tritium experiment (DTE2, C41) took place in the JET during the years 2021 and 2022. In this work, a corresponding irradiation scenario was utilized for the activation calculations of eight material foils located at the JET outer long-term irradiation station (OLTIS). Neutron-induced activities and dose rates at a 30 cm distance after shutdown at specified cooling intervals were calculated with the FISPACT-II code, employing the EAF-2010 nuclear and TENDL-2021 data libraries. The Monte Carlo MCNP6.2 particle transport code equipped with the FENDL-3.1d nuclear data library was used for the calculation of the neutron flux densities.

The fast neutron fluence and the activation monitor activity calculations using fine multigroup and continuous nuclear data

The fast neutron fluence and the activation monitor activities are routinely calculated with TORT deterministic code and BUGLE-B7 nuclear data library with 47 broad energy groups. The objective of the paper is to analyse options to improve reactor dosimetry transport calculations. There are two paths to improve reactor dosimetry calculations. Increasing geometry, angular and energy mesh size is applicable for TORT code while using newer nuclear data libraries is relevant for both deterministic and Monte Carlo codes. Two new calculation options (improved TORT and Monte Carlo MCNP6) were compared with the standard TORT calculation for VVER-440 Dukovany Unit 3 Cycle 31. The fast neutron fluence with 0.5 MeV threshold as well as activity of Fe. Ni, Ti, Cu. Mn and Nb monitors were evaluated. Standard TORT calculations were improved from S16P3 to S30P3 with three times finer axial mesh size. 120° core symmetry r-ϑ mesh size with 0.5° step and fine multigroup libraries VUAMIN-B7 with 199 neutron energy groups and ENDF/B-VH.1 with 200 neutron energy groups. Both ENDFB and IRDFF activation cross sections were used. The drawback of expanded mesh size is raised calculation runtime since TORT deterministic code is not parallelized and one calculation can require multiple weeks of CPU time. An alternative option of using MCNP6 Monte Carlo code with continuous ENDF/B-VH.1 nuclear data with detailed 3-D geometry and pin-wise effective neutron source prepared by MOBY-DICK diffusion code reactor analysis wras explored. It was found that using finer mesh size affects reactor dosimetry’ tallies less than the choice of nuclear data library. BUGLE-B7 and VTTAMIN-B7 produce results typically within 1% difference. ENDF/B-VH.1 calculations with 200 neutron energy’ groups with TORT code are even in better agreement with MCNP6 calculations with continuous nuclear data libraries. The largest differences of around 2% were observed between VTTAMIN-B7 library based onENDFB-VH.O nuclear data and ENDF/B-VH.1 library. Nuclear data library’ has larger impact on the results with up to 7 % difference between all 0.5 MeV fast neutron fluence calculations. The largest intact of nuclear data was observed for Mn(n.2n) monitor.

Experimental and Monte Carlo study of energy response of BEO-based OSL detectors within photon energy range up to 15 MeV.

Response of personal dosemeters to high energy photon radiation is of great interest nowadays due to a spread of new radiation technologies and the expansion of occupational exposure domains. ICRU95 publication has expanded the range of relevant photon energies upwards, setting new horizons for individual monitoring. Beryllium oxide (BeO) material is increasingly popular due to its excellent optically stimulated luminescence (OSL) properties, simple readout and reasonable energy response in the low energy (below 100keV) range. The study considers energy dependence of OSL response at higher photon energies. Energy deposition of monoenergetic photons with energy up to 15MeV in the BeO chips of various thickness was modeled with Monte Carlo MCNP 6.2 code. Benchmark experiments were conducted at LINAC with high voltage of 6, 10 and 15 MV resulting in respective incident photon spectra. The findings of this study add knowledge regarding behavior of BeO personal dosemeters in the photon fields within the energy range above 3MeV.

Characterization of the Radiation Contamination in an 18MV Linac Treatment Room Made of some Nanoparticles Mixture

Purpose: Shielding against radiation in radiotherapy and radiology requires deep knowledge of radiation physics and shielding design methods. The application of nanoparticles in the photon and neutron dose moderation has been proven in the literature.
 Materials and Methods: Effective neutron mass removal cross-section (ΣR/ρ) was for the ordinary concrete doped with 50nm in diameter nanosphers mixture of Fe2O3(5%), WO3(5%), B4H (5%), Pb2O3(5%) was estimated with MCNP5 Monte Carlo (MC) code and N-XCOM computational program. An 18MV linac room made of the nanoparticles dopped ordinarily simulated. Additionally, the room was considered with three legs in the maze and photoneutron and capture γ-ray Dose Equivalent (DE) were estimated at the modeled rooms maze.
 Results: Total ΣR/ρ with energies 100 keV-2000keV was estimated using MC and N-XCOM as 0.02802-0.02687 cm2/gand 0.02810- 0.02687 cm2/g, respectively. Total ΣR/ρ of the neutron for the pure ordinary concrete was estimated by MC code and N-XCOM with energies 100 keV-2000keV as 0.02802-0.02687 cm2/g and 0.02810- 0.02687 cm2/g. Borated Polyethylene (BPE) and Lead required thickness calculated as 7.43×10-06 mm and 4.73×10-06 mm for the capture γ-ray shielding.

Results and lessons learned from the Generation IV SCWR-FQT comprehensive Monte Carlo computational benchmark

A joint European Canadian Chinese development of a supercritical water-cooled small modular reactor technology has been in progress since September 2020 in the framework of a Horizon 2020 project called ECC-SMART. A specific work package has been dedicated to studying the design- and safety-related neutronic parameters and reactor physics behavior of the SCW-SMR to support the pre-conceptual design process. Three Monte Carlo codes, viz., MCNP, OpenMC, and Serpent, were selected for pre-conceptual design applications and code-to-code comparison within the Gen-IV SCWR-FQT reactor physics computational benchmark. The effective multiplication factor, the axial power distribution within the fuel, the axial three-group neutron flux distribution, and the spatial distribution of the energy deposition due to neutron and photon interactions were determined. In this paper, results and lessons learned from this study are presented, and useful considerations are summarized to provide guidance in obtaining consistent results among the three Monte Carlo codes.

Virtual Reality System for Monte Carlo Transport Simulation Codes Using Web Technology and Commodity Devices

A virtual reality (VR) system for Monte Carlo (MC) transport simulation codes, MCNP and Particle and Heavy Ion Transport Code System (PHITS), was developed. This system is aimed to provide a more reliable confirmation of the simulation geometry through an intuitive user interface. The VR system is implemented by extending the Geometry and CROSS Section VIEWer (Gxsview) code to run in a JavaScript environment, and it takes advantage of web technology and does not require dedicated hardware or geometry information supplied by computer-aided design software. Therefore, only WebXR-compatible head-mounted displays and browsers are required on the user side. In the VR space, visualization of the calculation geometry, cell selection, and cell grabbing operations is possible with a handheld controller. Further enhancements will be made in the future, such as performing shielding calculations with the cell changed by grabbing operations.

Assessment of Photon Beam Skyshine Dose Equivalent Rate of a 4 MeV Radiation Therapy Bunker Using Analytical and Monte Carlo Methods

Purpose: Skyshine radiation dose equivalent dose rate is known as scattered radiation by the room above air to points at the ground level points outside the Linear Accelerator (LINAC) room. Our aim was to estimate skyshine around the LINAC-based radiotherapy by a 4MV LINAC photon beam.
 Materials and Methods: Monte Carlo (MC) MCNP code calculation was conducted to skyshine at the control room, 60Co treatment room, physics, and simulator rooms. National Council on Radiation Protection and Measurements (NCRP) 151 was also used and it reported analytical formulation methods for photon beam calculation. A Flattering Filter (FF) equipped and Flattening Filter-Free (FFF) LINACs photon beams were derived and differences and agreements were discussed.
 Results: The results showed high skyshine for FF equipped relative to FFF LINACs. This effect may be attributed to photon beam hardening by FF in the LINAC head and higher transmission through the ceiling shield and more presence of photons on the roof above the air. NCRP 151 method results were higher than MC simulated photon beam skyshine dose equivalent dose rate and it may be the cause of the inflexible analytical method in contrast to MC simulation. Finally, FFF and FF-equipped LINACs result in skyshine compared and they compared to NCRP 151 report. MC simulation performed reasonably in estimation in different conditions.
 Conclusion: Our results showed that FF-equipped skyshine is higher than FFF LINAC and NCRP 151 is an inflexible method that does not take some effective parameters into account and calculates skyshine higher.

Simulation of a NaI(Tl) detector model using OpenMC modified on code source and validation using experimental and simulated results

In order to expand and examine the capabilities of the OpenMC Monte Carlo code to characterize the response function of the gamma radiation detectors (since OpenMC does not support this simulation), this paper presents the methodology used to modify the OpenMC code source and to characterize the response function of a NaI(Tl) scintillation detector. In addition, it includes a characterization of the response function for the 1.5” × 2” NaI(Tl) scintillator detector using various Monte Carlo codes. Finally, a comparison between the experimental data and the results obtained with OpenMC modified on code source, MCNP, and Geant4 models is presented. The detector efficiency and energy resolution were calculated over the photon energy ranging of 59–662 keV by using a precise model, which was developed by using the previous codes. The validation of the OpenMC model has been accomplished by a comparison of the results with the Monte Carlo Geant4 and MCNP models results, and experimental data, which shows a good agreement. This validation confirms that the modified OpenMC model can be successfully implanted in the modeling of the response functions of the NaI(Tl) scintillation detectors.

Sensitivity analysis of gamma-ray shielding characteristics to the TiO2 concentration in the Bi2O3–ZnO–Pb3O4–Al2O3 glass sample based on the Monte Carlo method

In this paper, the MCNP Monte Carlo (MC) code, was adopted for evaluation the properties of protection against gamma-rays of the glass system with the composition of (55-x)Bi2O3–15Pb3O4–20Al2O3–10ZnO- xTiO2 with specific concentrations of (x = 0, 5, 10, 15, 20, 25, 30 and 35 mol percent). For this purpose, several parameters related to photon attenuation such as the, Half-Value Layer (HVL), Tenth-Value Layer (TVL), mean free range (MFP), mass attenuation coefficient (μm), effective atomic number (Zeff) and buildup factor (BF) was investigated at the different energy levels between 0.1 and 1.5 MeV. To verify the results of the simulation, the data obtained from the simulation were compared with the experimental data available in the XCOM database. The NIST-XCOM database and the results of the simulation in the present study were in good agreement. The percentage deviation (PD) between the XCOM database and the MCNP program for calculating the μm was less than 0.59% in most cases. The results show that compared to conventional protective materials such as concrete and lead, the new combination shows more effective attenuation parameters. The glass with the highest concentration of TiO2 has the most favorable state in terms of density compared to the investigated protective materials. The close agreement of μm measurements with the XCOM database of this research shows that MC modeling in this research is a valuable and valid method.

Nuclear Analyses for the Assessment of the Loads on the ITER Radial Neutron Camera In-Port System and Evaluation of Its Measurement Performances

The radial neutron camera (RNC) is a key ITER diagnostic system designed to measure the uncollided 14- and 2.5-MeV neutrons from deuterium–tritium (DT) and deuterium–deuterium (DD) fusion reactions, through an array of detectors covering a full poloidal plasma section along collimated lines of sight (LoS). Its main objective is the assessment of the neutron emissivity/ <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$\alpha $ </tex-math></inline-formula> source profile and the total neutron source strength, providing spatially resolved measurements of several parameters needed for fusion power estimation, plasma control, and plasma physics studies. The present RNC layout is composed of two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module (DSM) of ITER Equatorial Port 1 (EP01): the ex-port subsystem and the in-port one. The ex-port subsystem, devoted to the plasma core coverage, extends from the Port Interspace to the Bioshield Plug: it consists of a massive shielding unit hosting two sets of collimators lying on different toroidal planes, leading to a total of 16 interleaved LoS. The in-port system consists of a cassette, integrated inside the port plug DSM, containing two detectors per each of the six LoS looking at the plasma edges. The in-port system must guarantee the required measurement performances in critical operating conditions in terms of high radiation levels, given its proximity to the plasma neutron source. This article presents an updated neutronic analysis based on the latest design of the in-port system and port plug. It has been performed by means of the Monte Carlo MCNP code and provides nuclear loads on the in-port RNC during normal operating conditions (NOC) and inputs for the measurement performance analysis.

Monte Carlo modelling of pixel clusters in Timepix detectors using the MCNP code

The track structure of the signal measured by the semiconductor pixel detector Timepix3 was modelled in the Monte Carlo MCNP® code. A detailed model at the pixel-level (256 × 256 pixels, 55 × 55 µm2 pixel size) was developed and used to generate and store clusters of adjacent hit pixels observed in the measured data because of particle energy deposition path, charge sharing, and drift processes. An analytical model of charge sharing effect and the detector energy resolution was applied to the simulated data. The method will help the user sort the measured clusters and distinguish radiation components of mixed fields by determining the response of Timepix3 detector to particular particle types, energies, and incidence angles that cannot be measured separately.

Suggested two layers container for shielding the low and intermediate activity gamma-ray sources

Due to the advantages of using gamma-ray radiation sources in various fields, the average equivalent dose received due to the exposure to these sources increases. The wrong handling or direct exposure to these sources give raises the equivalent dose for the receiver. Thus, a container made up of two layers (metal and glass layers) was suggested to decrease the recorded equivalent dose received from the gamma-ray radioactive sources to an acceptable limit. The Monte Carlo MCNP simulation code was utilized to predict the equivalent dose around the container's outer surface and at 100 cm from the center of the outer container's faces. For example, the simulation study illustrated that the equivalent dose rate received at 100 cm from an unprotected 137Cs source with an activity of 3.7E+8 Bq is 26.30 μSv/h. This equivalent dose rate from an unprotected source is reduced to 22.75 μSv/h using the inner layer of the stainless steel with a thickness of 0.5 cm. After that, the equivalent dose rate reduced from 22.75 to 7.60 μSv/h, raising the WO3. Simultaneously, the equivalent dose rate from an unprotected Co-60 source reaches 50.51 μSv/h. It is reduced to 23.56 μSv/h by using a container consisting of an inner layer of stainless steel with a thickness of 0.5 cm and an outer glass layer of BaWP7 glass sample with WO3 content of 60 mol % and thickness of 5 cm.The present study showed the suggested container's ability in reducing the absorbed dose and equivalent dose rate for intermediate activity gamma-ray sources used in laboratories and nuclear medicine.

DETERMINATION OF THE NEUTRON CONTAMINATION DURING BRAIN RADIOTHERAPY USING A MODERATED-BORON TRIFLUORIDE DETECTOR AND THE MCNP MONTE CARLO CODE.

This study aimed to determine the neutron dose equivalent to the thyroid gland and eye lens in brain tumor radiation therapy with 15- and 18-MV three-dimensional conformal methods (3D-CRT). A Monte Carlo simulation was performed using the Monte Carlo N-particle transport code to calculate neutron fluence and ambient dose equivalent (H*(10)). Afterward, these parameters were measured using a model NRD roentgen equivalent in man (REM) neutron detector (Thermo Electron Corporation, USA) equipped with Eberline's ASP-2e rate meter. Finally, the organ neutron dose equivalent was obtained by applying depth corrections to the measured ambient dose equivalent at the distance of the organ center from the central beam axis. The ratio of the out-of-field photon dose equivalent, measured previously, to the neutron dose equivalent in the eye lens was high due to its proximity to the radiation field. In contrast, this ratio remained unexpectedly high in the thyroid gland that is far from the central beam axis (about 15cm). The calculated neutron parameters agreed with the measurements. The present study findings indicate that external field photon dose is the main source of thyroid gland biological effects in radiotherapy of brain tumors. In addition, it is appropriate to apply the model NRD REM neutron detector for measuring neutron contamination from high-energy linear accelerators inside and outside the treatment field.

Determination of the coincidence summing factor by Monte Carlo method using MCNP–CP and PENNUC codes

In gamma spectrometry, the quantitative information about the composition of elements present in the experimental samples can be obtained from the efficiency calibration curve, the radioactivity, and the gamma emission intensity of the radioactive. The efficiency of the detector is often affected by the coincidence summing effect. The correction of the coincidence summing factors can be done through the software based on the efficiency transfer method or Monte Carlo simulation. In this study, two Monte Carlo simulation software, MCNP‒CP and PENNUC, were applied to calculate the coincidence summing factors of four radioactive 22Na, 60Co, 133Ba, and 134Cs by the simulation method. The calculated models from the benchmark with two types of detectors of n-type and p-type, using different source geometries such as point source, water, soil, and filter samples, were used. This study aimed to validate the spectral response as well as to compare the coincidence summing factors between MCNP‒CP and PENNUC. The results showed that the coincidence summing factors calculated by two simulation software gave a good agreement with an average relative deviation of less than 2%. The coincidence summing factors calculated by two simulation programs were verified by the statistical T-test with a linear correlation value of rxy = 0.995.

Monte Carlo Simulations of scintillation detectors for the use as dosemeters

The energy dependent conversion factors from measured photon pulse height spectra to ambient dose equivalent (H*(10)) of three different detectors based on the scintillation materials CeBr_3, LaBr3 and SrI2 were simulated with the three Monte Carlo Codes, GEANT4, MCNP and PENELOPE, respectively. The performance of the detection systems, using the calculated conversion factors (fitted to response functions), is demonstrated for different scenarios ranging from laboratory conditions (i.e. a precisely defined radiation field in a calibration facility) to a free-field irradiation scenario produced by the natural radioactivity at a large lawn. With very simple geometric models of the detector systems, the response function of scintillation based dosemeters can be derived from the aforementioned Monte Carlo simulations with an accuracy of the measured doses in H*(10) of ±10% in an photon energy range from 100 keV to 3000 keV, as being typical for natural environmental gamma radiation.

Tabulation of organ dose conversion factors for terrestrial radioactivity monitoring program

Terrestrial radioactivity monitoring of 238U and 232Th series, and 40K in soil is an essential practice for radioactivity and radiation measurement of a place. In conventional practice, only basic data can be in-situ measured using a survey instrument, for example radioactivity concentration in soil and ambient dose equivalent rate. For other physical quantities, for example organ absorbed dose and organ equivalent dose, the measurement is impossible to be performed and can only be computed using Monte Carlo radiation transport simulations. In the past, most of the works only focused on calculating air-kerma-to-effective dose conversion factors. However, the information on organ dose conversion factors is scarcely documented and reported. This study was conducted to calculate organ absorbed and tissue-weighted equivalent dose conversion factors as a result of exposure from terrestrial gamma radiation. Series of organ dose conversion factors is produced based on computations from Monte Carlo MCNP5 simulations using modelled gamma irradiation geometry and established adult MIRD phantom. The study found out that most of the radiation exposed organs absorb energy at comparable rates, except for dense and superficial tissues i.e., skeleton and skin, which indicated slightly higher values. The good agreement between this work and previous studies demonstrated that our gamma irradiation geometry and modelling of gamma radiation sources are adequate. Therefore, the proposed organ dose conversion factors from this study are reasonably acceptable for dose estimation in environmental radioactivity monitoring practices.

Dosimetry investigation of a prototype of 169Yb seed brachytherapy for use in circular stapler

This study aims to investigate dosimetry parameters for the new design of 169Yb seed in the form of a surgical staple for circular staplers commonly used in the abdominal incision and the esophageal and gastric surgery, which facilitates the precise placement. This seed includes a titanium tube with the inner diameter and outer diameter 0.68 mm and 2.2 mm, respectively, and length of 0.8 mm. Both sides of the tube are closed by titanium wires with the thickness of 0.65 mm by the laser. Natural ytterbium oxide is used after the thermal neutron activation; it is necessary for cooling time of 40 days. The dosimetry parameters were calculated based on the TG-43U1 using Monte Carlo MCNP5 code. The experimental dosimetry was performed by EBT3 radiochromic film to determine 2D dosimetry at near distance of the source and validate the MC code. The dose rate constant of MC calculation was obtained at 1.39cGyh-1U-1 ± 4% with the difference of 5% compared to another study. The dose distribution was symmetrical along the Z-axis and Y-axis (around the seed) and there was a uniform activity inside the tube. The distinction of dose rate was not noticeable at the 90 and 270 degrees on the Z-axis, which indicated a slight effect on staple legs in the matter of delivery dose. However, to understand dose distribution and introduce this source in a pre-clinical study, 3D dosimetry as well as further studying the heterogeneous function is required.

Development, design, and testing of a microwave-driven compact rotating-target D-D fast neutron generator for imaging applications

A compact D-D fast neutron generator system was developed with an emphasis on imaging applications. It is equipped with an electron cyclotron resonance microwave ion source, an electric field method electron suppression electrode, and a rotating beam target. The design, mechanical assembling, and initial performance of the system is presented in this work. Electrostatics and charged particle tracing simulations were employed to dimension components in the vacuum system and estimate the size of the ion beam spot on the target surface. Stable operation of the neutron generator was possible with accelerating potentials of up to −120 kV and an estimated averaged ion beam current of around 0.5 mA. At these conditions, the neutron yield was estimated to 6 ⋅ 107 s−1 based on a combination of detailed Monte Carlo MCNP6 models and the ambient dose rate reading of an LB6411 neutron probe. The neutron emitting spot size was compared to the ion beam spot size and experimentally determined with an attenuating edge technique to be between 1.7 and 2.6 mm. With these neutron yield and emitting spot size values, the system compares favorably with commercially available neutron generators, and this context is also presented.

THE APPLICABILITY RANGE OF THE MODIFIED SOURCE MULTIPLICATION (MSM) METHOD TESTED IN THE FAST VENUS-F REACTOR

The MSM method is an experimental technique for determination of reactivity of a sub-critical reactor. It consists of one dynamic measurement followed by two static measurements, which use an extraneous neutron source. For the data analysis, the core averaged kinetic parameters need to be calculated as well as a spatially-dependent correction factor that corrects for the point kinetics approximation. In order to test the range for which the method is valid and to demonstrate the reliability of the correction factor calculations in a fast reactor, a dedicated experimental campaign was performed in the fast lead-bismuth VENUS-F reactor. The reactivity of a dozen of sub-critical configurations was measured with the MSM method using ten 235U fission chambers. The detectors were located at various distances from the active zone and from the extraneous neutron source, leading to a large range of values of a correction factor (calculated with the Monte Carlo MCNP5 code) used in the data analysis.

Application of luminescent detectors in quality assurance of clinical electron beam

Thermoluminescent dosimeters (TLD) are widely applied in quality assurance of external radiotherapy beam. The aim of this study is to check the possible use of radiophotoluminescent glass dosimeters (RPLGD), and optically stimulated luminescent dosimeters (OSLD) in quality assurance of clinical electron beam (used in radiotherapy treatments). For this purpose, we give in this paper, an estimation and a comparison of dosimetric parameters, characterizing dose profile and percentage depth dose (PDD) curves, for electron beam using the different studied detectors. The Monte Carlo MCNP5 simulation code was utilized to calculate the distribution of deposited energy along X-axis and PDD curves in a water phantom using RPLGD (FD-7glass), TLD (LiF: Mg, Ti) and OSLD (Al2O3:C) for electron beam energies of 9, 12, 16 and 20 MeV. Our calculations show that the profiles for different electron energies are almost identical for all used materials: we found similar average penumbra and full width at half maximum (FWHM). For flatness and symmetry, our results show a good correspondence with the recommended values where the flatness and symmetry calculated in all dosimeters materials are less than 1.7%. We also found concordance with the literature for the parameters derived from PDD curves, such as various ranges (R100, R90, R80, R50 and Rp) and energies specification (Ep,o and E0) for different electron beams (9, 12, 16 and 20 MeV). In conclusion, this study demonstrates a good correspondence with the results of measurements and Monte Carlo simulations found in the literature. RPLGD and OSLD can be as TLD useful detectors for quality assurance (QA) verifications concerning dose profile and PDD curve for clinical electron beam. Further research is recommended to broaden the scope of the current study.