MCNP Simulation Research Articles

Internal dosimetry of 64Cu-DOX-loaded microcapsules by Monte Carlo method

Some of the issues regarding introducing new radiocompounds in nuclear medicine are the distribution patterns, delivered dose to different organs, diagnostic abilities and side effects. In this study, in order to assess the biodistribution of 64Cu-DOX-loaded microcapsules, rats were IV-injected with the microcapsules, and 1, 4, 14, and 24 h later, the activities of the targeted organs were measured (%ID/g). The accumulated activities were achieved by %ID/g curves, and S-factors were obtained by MCNP outputs. The MIRD formulation and Monte Carlo method were used to determine the absorbed dose in the target organs. The biodistribution data and PET-CT images showed that the lungs were where the majority of activity was seen. According to MIRD and MCNP, the maximum dose delivered in the lungs was 5.79E+01 mGy/MBq and 4.70E+01 mGy/MBq, respectively. Also, the effective dose was 1.2E+01 for MIRD and 8.31E+00 mSv/MBq for MCNP. These results indicate that 64Cu-DOX microcapsules can be considered a new radiocompound in pulmonary imaging, and MCNP simulation can be a reliable method for internal dosimetry.

Study of the effect of photon self-absorption on the detection efficiency of HPGe detector in measurements of an activated uranium sample

The photon self-absorption effect of uranium on the full-energy peak (FEP) efficiency calibration of high purity germanium (HPGe) detector was studied as a function of energy using an activated uranium sample has a thickness of 1 mm. The absolute efficiencies of the detector were obtained with the use of gamma standard sources and gamma rays of radionuclides created by irradiating the uranium sample with secondary neutrons. The measurements were carried out with the use of a Planar HPGe detector. MCNP simulation code was used to verify the experimental results. The comparison of the empirical results with the simulation results showed a good agreement. To interpret the efficiencies, the reaction rates were calculated using two methods: (1) - the use of the efficiency curve from the standard sources combined with the self-absorption correction of the uranium material and (2) - the use of the efficiency curve obtained from gamma rays in the sample.

Comparison Study of MCNP6 and GEANT4 for Neutron Transportation and Detection in 3He Long Counter

A long counter is a commonly used device for neutron-flux measurements in a wide energy domain. Although its physical structure is well defined, device’s response differs significantly depending on the nature of radiation fields. We investigated a 3He long counter under well-defined radiation fields based on experiments and Monte Carlo simulations. The responses of the long counter were evaluated in MCNP6 and GEANT4 simulations for neutrons at energies between 10 keV and 20 MeV. The simulation results were validated with experimental data obtained with the spontaneous-fission neutron of 252Cf and quasi mono-energetic neutrons between 0.2 and 1.5 MeV. The results showed that MCNP6 underestimated count rates, calculated from the pulse-height distribution of recoiled particles in the neutron-induced nuclear reaction in the 3He volume, while neutron transportation seems equivalent in both simulation tools. For a given condition, GEANT4 reproduced the experimental values well, and GEANT4 results appeared to be more realistic for detector simulations in the energy range of interest.

Bismuth oxide nanoparticles (Bi2O3 NPs) embedded into recycled- Poly(vinyl chloride) plastic sheets as a promising shielding material for gamma radiation

In this work, novel γ-ray shielding materials based on recycled poly (vinyl chloride) (r-PVC) were prepared via melt mixing and compression molding technique. The designed materials are low-cost plastic wastes, ductile and lightweight, doped with xBi2O3 NPs (x = 0.0%, 5.0%, 15.0%, 25.0% and 35.0% (w/w)) as a non-toxic nanofiller. Scanning electron microscopy (SEM) and X-ray diffraction (XRD) were used to explore the characteristics of the as-prepared r-PVC/Bi2O3 nanocomposites. The γ-ray shielding effectiveness of the r-PVC/Bi2O3 nanocomposites were evaluated using NaI(Tl) scintillation detector at different γ-ray photon energies (121, 334, 778, 963, 1112 and 1408 keV) emitted from Eu-152 standard radioactive point source. For each of the mentioned photon energies, the mass attenuation coefficients (μm) were measured and compared to the Bi2O3 NPs’ doping ratio. The results showed that as the content of Bi2O3 NPs was increased in the nanocomposite, the value of μm significantly increased. In particular, the r-PVC nanocomposite sample containing 35.0 wt% Bi2O3 NPs exhibited the highest μm (2.5301 ± 0.22 at 121 keV). The μm was also calculated theoretically, using the MCNP5 simulation code. Experimental data of μm for all the tested nanocomposites were comparable to that obtained from a theoretical calculation. Moreover, the experimentally measured linear attenuation coefficient was the first step in calculating other shielding parameters, including half-value layer (HVL), tenth value layer (TVL), and mean free path (MFP). These parameters confirmed that addition of Bi2O3 NPs resulted in a higher shielding efficiency for r-PVC. Finally, this study shows that addition of Bi2O3 NPs to r-PVC facilitates the production of lead-free and sustainable polymer nanocomposites with improved γ-ray shielding characteristics, which can be used in nuclear energy applications, as alternatives to the shielding materials currently in use.

Elaboration of X-ray shielding of highly barite-loaded polyester concrete: structure, mechanical properties, and MCNP simulation

Various research has been done in manufacturing and designing different types of radiation shielding. In this study, lead-free polymer composite panels with new special formulations have been prepared and characterized to prevent radiation leakage in medical facilities at the diagnostic energy range (60 – 120 keV). Polyester resin is used as a binder along with large amounts of mineral barium sulfate (barite) aggregate (more than 70 % by weight). SEM micrographs showed a homogeneous distribution of barite particles. Then the effects of adding barite on the physical, mechanical, and thermal properties of composite panels were investigated. The results showed that increasing the percentage of barite powder in the formulation, with the high wetting capability of polyester resin, improved physical and mechanical properties. The composite panel containing 80 wt% of barite had the highest tensile and flexure strength. Also the thermal stability of the composite panel containing 90 wt% of barite was 40 °C to 50 °C more than other samples. The samples' mass attenuation coefficient (μm) was calculated experimentally and theoretically using MCNP-5 simulation software and the XCOM program. Experimental and calculated results have been compared and were found to be in reasonable agreement. In general, with increasing barite content, the composite shielding panel became denser and μ increased while HVL and TVL values decreased. In addition, the studied shields were superior to lead by virtue of their non-toxic nature and will be a radiation shielding solution in making wall panels.

Experimental evaluation of the performance of a novel planar segmented HPGe detector for use in gamma emission tomography

The use of segmented HPGe detectors for gamma-ray tracking applications is well established. The spectroscopic capabilities of these detectors make them most suitable for such applications. For similar reasons, the use of such detectors in the tomographic measurement of irradiated nuclear fuel has also been envisioned. Especially, these detectors can facilitate faster fuel examination with excellent energy resolution. We have proposed and designed a novel planar segmented HPGe detector for use in gamma emission tomography. The design of the detector segmentation and the mode of operation is unique and offers six simultaneous detection channels for tomographic measurements. This work reports the first experimental evaluation of the performance of the detector. Important characteristics of the detector have been obtained, such as energy resolution of the segments in 1-fold (one segment) and 2-fold (two segments) modes, throughput curves, crosstalk energy corrections, and mislocalisation rate. Collimated source tests have been performed and the results have been compared with the MCNP simulations results. The obtained results are as expected and in good agreement with the simulation results, and it is estimated that using this detector can speed up the data collection by 3.2 times in comparison to an unsegmented detector of the same overall dimensions, in a tomographic application. Further improvements are foreseeable if scaling up to a larger detector with greater segmentation.

The Covariance of PFNS Results from the Chi-Nu Experiment

The prompt fission neutron spectrum (PFNS) from neutron-induced fission is a fundamental quantity for the behavior of nuclear reactors, and has been measured many times on a wide variety of nuclei and covering different ranges of incident and emitted neutron energies. However, results from past measurements are frequently called into question in modern nuclear data evaluations because of a lack of thorough experimental documentation and incomplete uncertainty analyses. The Chi-Nu experiment at Los Alamos National Laboratory was designed to produce high-precision measurements of the PFNS of major actinides over a wide range of incident and emitted neutron energies, and with the documentation and covariance analysis required to ensure that the results of this experiment maintain their impact long into the future, thereby avoiding this pitfall of past measurements. In this work we describe the Chi-Nu experiment along with summaries of the treatment of and methods developed to address two important components of the analysis of Chi-Nu data: random-coincidence backgrounds and MCNP simulations. Furthermore, we describe the first results for correlations not just between all data points collected on a single target nucleus, but also between all data points from separate Chi-Nu measurements on 235U and 239Pu. These correlations are important for accurately calculating ratios of the PFNS from one actinide to another, which are rare and can be informative for nuclear data evaluation efforts.

Evaluation of the Effectiveness of Silicon for Radiation Shielding in Mammography

Mammography is one of the most important test techniques for screening and diagnostic examinations of breast diseases. Recently, the incidence of breast cancer has increased and the demand for mammography is increasing through the cancer screening programs in each country. Therefore, research on shielding materials is actively being pursued to reduce radiation exposure during mammography. The purpose of this study is to verify and compare the GATE simulation and experiment in terms of the shielding performance under the experimental conditions of tube voltage (25, 30, 35 kVp) and silicon thickness (1, 2, 3, 4, 5, 6 mm). Through MCNP simulation, the absorbed dose reduction rate of breast tissue according to tube voltage/silicon shielding thickness is presented. The GATE simulation and experimental results were within the error range of 0.07 to 1.42 % under all conditions. In addition, through simulations and experiments, it was confirmed that silicon can offer more than 80 % shield regardless of automatic exposure control (AEC) of mammography equipment when 5 mm of silicon is used. Therefore, the results of the study can serve as useful basic data for the development of a shielding suit or shield that can reduce the exposure of other surrounding tissues due to scattered rays during mammography.

Preliminary study of a novel radiation shield for jaw in dental radiography using the high-density polyethylene/bismuth oxide nanocomposite

Dental radiography has been linked in numerous studies to an elevated risk of brain cancer, meningioma cranial tumors, and parotid gland tumors. Unfortunately, not much attention has been given to this problem. This research aims to determine and quantify the radiation shielding properties of High-Density Polyethylene/Bismuth Oxide composite (HDPE-Bi2O3) including the linear attenuation coefficient (μ), mass attenuation coefficient (μ/ρ), half-value layer (HVL) and tenth-value layer (TVL) for photons at 59.5 keV. First, the MCNP code—which showed a reasonable degree of consistency with the XCOM data—was used to simulate the variables μ, μ/ρ, HVL, and TVL that were connected to different concentrations of the composite. Then, the nanocomposite containing of 60 wt% Bi2O3 was chosen and fabricated via the casting method. There were XRD, and FESEM tests performed. Two techniques using the CsI(Tl) scintillation detector and the Geiger counter were applied to measure the shielding quantities at 59.5 keV gamma-rays of the 241Am radioactive source. Finally, results of the measurements revealed a comparatively good agreement with the values from the MCNP and XCOM simulations, with a maximum 3% discrepancy. The artifact test at the presence of the composites shield using a dental radiographic system at 60 kV have not been influenced adversely. This composite at the optimal weight fraction and thickness has potential application to be used as a radiation shield for jaw in dental radiography.

Correction: Solar and galactic 14C production rates in atmosphere using an MCNP6 simulation

Correction: Solar and galactic 14C production rates in atmosphere using an MCNP6 simulation

Solar and galactic 14C production rates in atmosphere using an MCNP6 simulation

Solar and galactic 14C production rates in atmosphere using an MCNP6 simulation

Effect of non-nominal port gaps on nuclear responses in ITER

The size of the gaps around the ITER equatorial and upper ports has associated a certain tolerance due to manufacturing and installation. This variability has an impact on the nuclear responses at magnets and port interspace maintenance locations. This effect is not usually considered as the ITER neutronic reference models represent the nominal gaps configuration. This paper aims to estimate the effect of the gap size variability on nuclear heating in the magnets during plasma operation and shut-down dose rate at the port interspace maintenance location taking advantage of a set of MCNP simulations with different gap configurations. The effect has been shown to produce increases of up to 30% in the worst-case configurations. The novel modelling approach, which was developed ad-hoc and used to manipulate the size of the port gaps via displacements of the geometry universes, is also described here. This approach led to important time savings, and it may be used in future studies with similar requirements.

Comparison of Real and Generated by MCNP Simulation Spectrums for Validation of Neutron Activation Calculations

The objective for decommissioning planning, is to obtain a radiological understanding of the involved installation. The characterization at this stage could be carried out by means of: (a) neutron activation calculations based on reactor design and neutron flux; (b) dose rate measurements; (c) in-situ gamma spectrometry; (d) sampling for determination of the scaling factors in activated and contaminated components.Neutron activation calculations contains several uncertainties. These uncertainties are based on the input data - such as material data (composition and impurities), neutron flux and energy, nuclear data libraries - and on the methodology of the process and the simulation codes.Taking into consideration all these modeling uncertainties, this work is focused on the development of a technique for validation of the calculations. A non-destructive gamma spectrometry technique by using MCNP6 simulations is under development for interpretation of the resulting gamma-ray spectra of the radionuclides in activated components. In particular, a spectrum will be produced, based on the activities of the main radionuclides in the activated component and the results of MCNP6 simulations. This spectrum will be compared with the experimental spectrum.Furthermore, the radiological characterization of activated components, which appeared with surface contamination, is essential for the decision making process during decommissioning. The cutting techniques to be followed in order to reduce the production of secondary waste and limit the doses to personnel and the selection of decontamination techniques should be based on accurate determination of the radionuclides inside the material and/ or in the surface contamination. The proposed method could also be helpful in this case. The activities inside and on the surface of the components could be determined by comparing the experimental spectrum with that produced by MCNP6 simulations, using the arisen activities from the scaling factors and the dose rate measurements.

Characteristics comparison and Monte-Carlo simulation of isotopes used in betavoltaics for MEMS application

• Betavoltaic nuclear batteries can be used to power microelectromechanical systems . • The important isotopes used: H-3, Pm-147, Ni-63. • Characteristics, fabrication routes (see supporting information) are presented. • A mcnp simulation that studied the number of particles released: Pm-147> H-3>Ni-63. • The ideal source thickness is 2.5 μm for Ni-63 and H-3, and 50 μm for Pm-147.

Gamma radiation shielding property of continuous fiber reinforced epoxy matrix composite containing functional filler using Monte Carlo simulation

• MCNP model of continuous fiber reinforced composite containing filler for gamma radiation shielding was established. • The effects of filler size and distribution, fiber type and their stacking order on μ were illustrated. • The experimental results confirmed the XCOM and MCNP simulation results. Functional filler modified fiber reinforced polymer composites are promising structural/radiation shielding materials in application fields of nuclear technology. In this work, MCNP simulating models of epoxy resin matrix composites containing shielding micro-fillers (tungsten oxide and boron carbide) and continuous fiber for gamma radiation shielding were established by Lattice and Universe features in MCNP. The effects of filler size and uniformity of dispersion, types of filler and fiber, stacking order of composite layers on liner attenuation coefficient ( μ ) and mass attenuation coefficient ( μ m ) were investigated using MCNP simulation. The corresponding shielding mechanisms were discussed. Furthermore, typical kinds of composites were prepared by hot-press process and their μ were tested. The numerical results show that μ decreases more significantly with large filler size and uneven dispersion in low-energy gamma rays. For fibers, μ is proportional to density independent of element type of fiber. Besides, stacking order has negligible effect on μ . All MCNP results are in good agreements with the calculation data from XCOM and the experimental values, which demonstrates the validity of the proposed modeling method. Thus, this simulation model can be used to design and evaluate radiation shielding composites with complex component and microstructure.

An investigation on compatibility of experimental effective atomic numbers of some Fe–Ni alloys with those predicted by MCNP

In this work, effective atomic numbers of some Fe–Ni alloys were obtained experimentally by coherent and Compton scattering intensity ratios at scattering angle of 168° by using Americium-241 gamma rays (59.54 keV) and High Purity Germanium Detector (HPGe) detector. The experimental setup consisting of Am-241 and HPGe detector was simulated using the MCNP (Version 6.2) code. The intensity ratios of coherent and Compton scattering peaks of Cr, Mn, Fe, Co, Ni, Cu and Zn elements used for calibration were plotted as a function of their atomic numbers. The effective atomic numbers of the investigated alloys were determined by the interpolation method based on the coherent to Compton scattering ratio. Coherent to Compton intensity ratios of the alloys ranged from 1098.02 to 1165.47. The effective atomic number values obtained by MCNP simulation of the alloys were between 26.423 and 27.211, the experimentally effective atomic number values were between 26.524 and 27.382, and the theoretical effective atomic number values obtained with ZXCom were between 26.710 and 27.594. It is observed that the results obtained by the simulation method are in agreement with the experimental results and the theoretical results obtained with ZXCOM.

Comparison of Monte Carlo simulations and theoretical calculations of nuclear shielding characteristics of various borate glasses including Bi, V, Fe, and Cd

The aim of this study is to evaluate both buildup factors and photon attenuation effectiveness of some borate glasses doped by Cd, Fe, V, and Bi. The mass attenuation coefficients (μm) of these glass systems have been calculated via MCNP6 and GEANT4 computer simulation codes over 0.02–10 MeV energy range and compared with the theoretical results of WinXCOM program. And then, half-value layer, (HVL), mean free path (MFP), and effective atomic number (Zeff) for the different content based glasses have been determined. Through G-P fitting process, exposure buildup factors (EBF) have been found in the energy range of 0.015–15 MeV up to 40 mfp. It can be concluded that SrBiO20 glass is alternative material in terms of photon attenuation, respectively.

Feasibility study of using the NUR research reactor for a BNCT installation and Monte Carlo optimization of a BSA

The aim of this work was devoted to study the feasibility of using the Algerian research reactor NUR for Boron Neutron Capture Therapy (BNCT) applications. To achieve this aim, an analysis of parameters and operational safety limits of the current core configuration X-1 with a removal of the thermal column was carried out. Furthermore, this study was followed by optimization calculations of the Beam Shaping Assembly (BSA) in the channel N∘5 using the Monte Carlo code MCNP5 simulations with a highlight of a new moderating material (Modified Fluental). Subsequently, a dosimetric study was performed using the MCNP5 code and the Snyder Head phantom by taking a Boron concentration of 65 ppm and 18 ppm in tumor and healthy tissue respectively. The in-phantom figures-of-merit have been calculated and provided promising results.

Monte Carlo simulations of vehicle-mounted whole-body NaI(Tl) counter

During a nuclear emergency, an internal exposure detection device with on-site and rapid measurements should be developed to assess the level of internal exposure of people in radioactively contaminated areas. Based on MCNP simulations, this work designed a shielding structure of a NaI(Tl) type vehicle-mounted whole-body counter (VM-WBC) and it has good shielding performance against the background radiation. The mass of the shielding structure was less than 800 kg, and the minimum detectable activity (MDA) of 60Co and 137Cs obtained by simulation for a 10 min measurement was less than 50 Bq. And we also study its counting efficiency calibration by simulating radiation computing phantoms with Chinese and American reference physiological characteristics. Equations that fit counting efficiency, body build index, and photon energy were also developed. The relative efficiency deviation between the equation and the MCNP simulation was less than 5%.

Experimental and Computational Verification of a New Remote Monitoring System Design for Spent Fuel Dry Cask Safeguards Using Small-Scale, Generic Diversion Scenarios

Dry casks will be a prevalent spent nuclear fuel (SNF) storage option until solutions for long-term storage or disposal are deployed. A dry cask storing 32 pressurized water reactor fuel assemblies will likely contain about 20 significant quantities of plutonium, so these structures require effective safeguards monitoring. An external remote monitoring system (RMS) is proposed to advance the current dry cask safeguards regime which relies on containment and surveillance. The objectives of this study were to assess the performance of the external RMS as a detection system and to develop a simulation approach for estimating measurements. Small-scale experiments of generic neutron source diversions mimicking SNF diversion from a dry cask were conducted and the nondetection probability was calculated for a variety of measurement times. MCNP simulations were carried out to assess the degree to which the measurement results could be predicted. A previous simulation methodology was advanced to consider uncertainty in the activity of sources being measured. The study concluded that the external RMS performs well as a neutron detection system and that MCNP simulation is a viable tool both for predicting measurements made with the external RMS and for calculating nondetection probabilities of hypothetical, generic diversion scenarios.