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A comparative study has been performed with the latest evaluated nuclear data libraries JEFF-3.3 and ENDF/B-VIII.0. The study has been conducted through the benchmark calculations for 120 criticality problems and the TRIGA Mark II research reactor with the libraries processed using NJOY21 for MCNPX Monte Carlo transport code. The criticality benchmarks assemblies, taken from the ICSBEP benchmark, cover Uranium (highly enriched uranium, intermediate enriched uranium, low enriched uranium, and 233 U) and Plutonium fuel systems in a various metal forms, and under a various spectral conditions. The Moroccan TRIGA Mark II research reactor calculation is used to look into the predictive capability of those nuclear data libraries and then to compare the accuracy of the predicted results with the experimental data published elsewhere. Actually, the purpose of this study is to investigate some neutronic and kinetic parameters of those benchmarks for both libraries. The former consist of effective multiplication factor, heat distribution, neutron flux distribution, effective delayed neutron fraction (β eff), prompt removal lifetime (τ r) and the mean neutron generation time (Λ). The results show that the calculated effective multiplication factor, heat distribution, neutron flux distribution, and the kinetic parameters are in good agreement with references. However, it is found that the computed values are strongly depending on the nuclear data set used in calculations. • Processing of the latest continuous energy neutron data JEFF-3.3 and ENDF/B-VIII.0 libraries. • NJOY21 processing code is used. • Benchmarking JEFF-3.3 and ENDF/B-VIII.0 libraries with critical benchmarks and the TRIGA Mark II research reactor. • Monte Carlo calculation with MCNPX code is used. • Good performance with the JEFF-3.3 and ENDF/B-VIII.0 libraries and excellent agreement with published results. [ABSTRACT FROM AUTHOR]

Radiation based gauges have been widely utilized as a nondestructive and robust tool for measuring the thickness of metal sheets in industry. The typical radiation thickness meter can just work accurately when the composition of the material is fixed during the measurement process. In conditions that material composition may differ substantially from the nominal composition, such as manufacturing rolled metals factories, the thickness measurements would be along with errors. The purpose of the present research is resolving the problem of measuring the thickness of metal sheets with various alloys. The aluminum is investigated in this work as a case study but the procedure can be applied for other types of metals. As the first step, the performance of various arrangements of two main detection techniques, named dual energy and dual modality, were investigated using MCNPX code to obtain optimum technique and arrangement. The simulation results indicated that a binary combination of 241Am-60Co isotopes as the source and one transmission detector in dual energy technique is the most appropriate choice. After then, an experimental setup based on the obtained optimal technique from simulation investigations was established. The aluminum sheets with 4 alloy types of 1050, 3105, 5052 and 6061 and thicknesses in the range of 0.2–4 cm with a step of 0.2 cm were tested and the obtained data were implemented for testing and training the artificial neural network (ANN). The proposed methodology could predict the thickness of aluminum sheet independent of its alloy type with an error of less than 0.04 cm in experiments. • Performance of 2 main detection techniques, named dual energy and dual modality, were investigated using MCNPX code. • The simulation indicated that a binary combination of 241Am-60Co in dual energy technique is the most appropriate choice. • An experimental setup based on the obtained optimal technique from simulations was established. • The ANN model could predict the thickness of aluminum sheet independent of its alloy type with an error of less than 0.04 cm. [ABSTRACT FROM AUTHOR]

99Mo is an important radioisotope and mainly produced using uranium fission reaction in a nuclear reactor. Investigation for probable improvements, especially on target geometry and in-core location of target is the main goal of this research. This is for producing more efficient of 99Mo in a typical Material Testing Reactors (MTRs). A parametric investigation is done focused on the target characteristics such as geometry, location, material, density, heat flux (power density) and also usability in Tehran Research Reactor (TRR) as a MTR case study. Stochastic code MCNPX 2.6.0 along with CFD code ANSYS are used to perform neutronic and thermal-hydraulic analyses. A target with plate type design is specified and proposed as a final and most favorable design. Taking into account the safety criteria, the production yield, the chemical process and radioactive waste, it is demonstrated that the new target design meets the key design requirements without compromising the reactor safety. This research results indicate that the new target gives rise to a higher production of 99Mo using less amount of initial material causing to a reduction in nuclear waste and process difficulties. • An investigation is performed for producing more efficient of 99Mo in Tehran research reactor. • Stochastic code MCNPX along with CFD code ANSYS are used to perform the calculations. • The proposed design satisfied all the neutronic, safety and operational constraints. • The new target gives rise to a higher production of 99Mo enough for the local demand. • The proposed design causing a reduction in nuclear waste, process and operating difficulties. [ABSTRACT FROM AUTHOR]

Abstract Studies of radiation interactions with tissue equivalent material find importance in efforts that seek to avoid unjustifiable dose to patients, also in ensuring quality control of for instance nuclear medicine imaging equipment. Use of the Monte Carlo (MC) simulation tool in such characterization processes allows for the avoidance of costly experiments involving transmitted X- and γ-ray spectrometry. Present work investigates MC simulations of γ-ray transmission through tissue equivalent solid phantoms. Use has been made of a range of radionuclide gamma ray sources, 99mTc, 131I, 137Cs, 60Co (offering photons in the energy range from a few keV up to low MeV), popularly applied in medicine and in some cases for gauging in industry, obtaining the transmission spectra following their interaction with various phantom materials and thicknesses. In validation of the model, the simulated values of mass attenuation coefficients (μ/ρ) for different phantom materials and thicknesses were found to be in good agreement with reference values (NIST, 2004) to within 1.1% for all material compositions. For all of the primary photon energies and medium thicknesses of interest herein, results show that multiple scattering peaks are generally located at energies lower than 100 keV, although for the larger phantom thicknesses it is more difficult to distinguish single, double and multiple scattering in the gamma spectra. Transmitted photon spectra investigated for water, soft tissue, breast, brain and lung tissue slab phantoms are demonstrated to be practically independent of the phantom material, while a significant difference is observed for the spectra transmitted through bone that was proved to be due to the density effect and not material composition. Highlights • Transmitted gamma spectra from popular radionuclides was simulated using MCNPX code. • Different tissue equivalent phantom materials at various thicknesses were simulated. • Resulted spectra were entirely analyzed based on photons interaction. • Multiple scattering peaks were observed at energies lower than 100 keV. • Transmitted spectra through bone showed significant difference compared to other tissues. [ABSTRACT FROM AUTHOR]

Abstract The radiation effects of loss of coolant accident on the reactor operators are critical issues. In the present study, computational codes were used to investigate gamma and neutron dose rates inside the 5 MW pool-type Tehran Research Reactor containment. The carried out calculations showed, when the remaining coolant over the core is decreasing less than 1 m, the containment should be evacuated. In addition, 24 h after LOCA, the containment gamma dose rates drop to one tenth of their initial values immediately after LOCA. [ABSTRACT FROM AUTHOR]

Abstract The computer simulation of radiation processing in an industrial panoramic gamma facility generates a number of complex scientific and technical problems; for example, the product is manually loaded onto turntables and rotates during the entire irradiation and the Monte Carlo MCNP three-dimensional geometry simulation is static. This causes the first problem; for this it is necessary to introduce additional approaches to describe the real irradiation process. It is important to properly represent the irradiation process and the specific conditions of the irradiator to correctly reproduce the measured results and guarantee also through the use of the model developed the most accurate dose map formations in an irradiated product, improving the service quality provided to the customer. The objective of this study is to apply the proposed source model of a panoramic gamma irradiator on dose distributions in containers of different geometric shape of an irradiated product with use of the Monte Carlo code. The influence was studied with use of a new homogenized annular source model. The validation of the MCNPX model of the GammaBeam 127 source was performed by measurements. The validation of the proposed model was based on comparative evaluation of the MCNPX absorbed-dose rates (Gy/h) in the irradiated product between the real source model and the new model. The comparative results for the MCNPX absorbed-dose rate showed that the new source model provides a better description of the irradiation process in a panoramic gamma irradiator. This indicates that the physical and mathematical MCNPX models developed are reliable and correct, and the new source model adequately reproduces the observed dose distributions in the different geometric shapes of the irradiated product. Highlights • Gamma radiation generates a number of complex technical problems. • Dose distributions in an irradiated product using the Monte Carlo code. • Validation of the MCNPX model of the GammaBeam 127 source was performed with measurements. • Errors lower than 9% were found between MCNPX values and measured values. • Nonuniform distribution of isodose inside the product depending on the shape. [ABSTRACT FROM AUTHOR]

Abstract Currently, lung cancer is one of the most lethal types of cancer (IARC, 2012), the pathology being detected in advanced stage, when the tumor has considerable volume because the disease in most cases asymptomatic in the early stages (INCA, 2016). Dosimetry analysis of healthy organs under real conditions is not feasible. Therefore, computational simulations are used to aid in dose verification in organs of patients submitted to radiotherapy. The goal of this study was to calculate the equivalent dose, due to photons, in the surrounding of healthy organs of patients submitted to radiotherapy for lung cancer, through computational modeling. The simulation was performed using the MCNPX code (MNCPX, 2006), Rex and Regina phantoms (ICRP 110, 2009), radiotherapy room, Siemens Oncor Expression accelerator operating at 6 MV and treatment protocol adopted at the INCA (National Cancer Institute – Brazil). The results obtained, considering the dose due to photons for both phantoms indicate that organs located inside the thoracic cavity received higher dose, being the bronchi, heart and esophagus more affected, due to their anatomical positioning. Clinical data describe the development of bronchiolitis, esophagitis and cardiomyopathies with decreased cardiopulmonary function as one of the major effects of lung cancer treatment. In the Regina phantom, the second largest dose was in the region of the breasts with 615.73 mSv/Gy, while in the Rex the dose was 514.06 mSv/Gy, event related to the difference of anatomical structure of the organ. A qualitative analysis was performed between the dose deposition profile of the treatment planning system and the simulated treatment through the tmesh command and a similar profile of dose distribution was verified throughout the patient's body. Highlights • Dose of photons during lung cancer treatment. • Was simulated treatment for lung cancer with Rex and Regina phantoms inside a radiotherapy room with head of the Linac Siemens operating 6 MV. • In both phantoms organs located inside the thoracic cavity received higher dose due to photons, being the bronchi, heart and esophagus more affected. • It was analyzed of dose through mesh tally command in the simulation and was compared with images of the planning system. • Images of the simulation and planning system presented same energy distribution profile and present "pyramidal" shape. [ABSTRACT FROM AUTHOR]

Abstract Zinc-65 has been of great interest in medical, biomedical, agricultural, and industrial applications due to its suitable half-life and decay properties. The 65Zn was produced via neutron irradiation on natural zinc oxide and natural zinc oxide nanoparticles targets in Tehran Research Reactor (TRR) at a thermal neutron flux of 4.5 × 1013 n cm−2 s−1 for 30 min. The excitation function of 64Zn(n,γ)65Zn reaction was calculated via the TALYS-1.8 code. The MCNPX code was used to calculate the thermal neutron distribution. The 65Zn theoretical production yield was estimated using calculated cross sections and the calculated thermal neutron distribution. The obtained experimental data and simulated value of production yield for 65Zn were in reasonable agreement. Highlights • Experimental and theoretical production yield of 65Zn from the natZnO target was investigated. • Experimental and theoretical production yield of 65ZnNPs from the natZnNPs target was investigated. • The nuclear codes of MCNPX and TALYS were used to calculate the neutron distribution function. [ABSTRACT FROM AUTHOR]

Abstract Self-absorption correction factors are fundamental in spectroscopy to correct the efficiency of the samples detection whose density is different from the radioactive standard. Mathematical simulations have been widespread as a tool to facilitate the procedure of correction factors calculation. In this paper, LabSOCS was used to calculate the self-absorption correction factor for some geometries and the values found were compared to those obtained in MCNP and experimental values. The percentage deviations found for the self-absorption correction factor calculated by LabSOCS were below 1.6% when compared to experimental values. Deviations were below 1.9% in the curve extrapolation of the experimental procedure found in literature. Results obtained show that the deviations increase proportionally to the difference between the density values of the radioactive standard and the sample. High percentage deviations were also noticed in simulations whose samples had high densities, complex geometries and low energy gamma-rays. Highlights • Great density differences between the sample and the radioactive standard should not be used. • It is noticed that percentage deviations increase with an increase in density and with an energy decrease. • Geometry complexity is a factor which increases the percentage differences of the self-absorption correction factor. • Must be careful when reproducing the same chemical compositions of the experimental procedure in a computational simulation. [ABSTRACT FROM AUTHOR]

Radiation protection is crucial for the safe utilization of ionizing radiation and minimizing the harmful effect upon exposure, hence some standards have been defined by some relevant organizations for the safe uses of radiation. One of the parameters relevant to the calculation of gamma ray shielding is the half-value layer (HVL), which is normally calculated using the knowledge of linear attenuation coefficient (μ). In this research, an attempt has been made to directly calculate HVL without the knowledge of μ via Monte Carlo simulation technique. For this purpose, in the Monte Carlo N‐Particle eXtended (MCNPX) code, F1, F5 and Mesh Popul sequences tallies were defined and the optimal structure for the least measurement error was introduced. The MCNPX calculated values showed reasonable agreement with the experimental findings. According to the obtained results, it is suggested that in order to reduce the error of HVL calculations, in exchange for the MCNPX code, the values of the R parameter and the radiation angle of the source should be considered according to the calculations introduced in this plan. Because the results show that by considering the measurement error between 6 and 20%, the code output can be cited in different energy ranges. • Study uses MCNPX to investigate HVL calculation directly. • MCNPX and experiment show agreement. • Different energy ranges can be cited by considering the measurement error between 6 and 20%. [ABSTRACT FROM AUTHOR]

Neutrons as a one of the by-products of high-energy photons in radiotherapy increase the patient's risks and could cause secondary cancers. Therefore, a new corrected simplified model for linac head was introduced to calculate equivalent (H) and absorbed (D) doses in different tissues/organs of a phantom model. The photoneutron spectrum calculated with this model was in agreement with the spectrum obtained by using a detailed linac´s head model having all components. Besides, an anthropomorphic phantom was irradiated under different gantry angles of the corrected simplified model aiming to emulate 15, 18, 20, and 25 MV Siemens linacs. The results indicated that tissues which were within the treatment field received more dose than others. Furthermore, tissues which were in the vicinity of each other and the same depth in the phantom nearly received the same doses. Finally, fatal secondary cancer risk was also studied. [ABSTRACT FROM AUTHOR]

In this paper, a preliminary neutronic design study of an accelerator driven subcritical system for Jordan Subcritical Assembly (JSA) is presented. The conceptual design of coupling the JSA core with proton accelerator and spallation target is investigated, and its feasibility as a neutron source for nuclear research, and possibly for target irradiation and isotope production evaluated. 3D MCNPX model of the JSA reactor, the accelerator beam, and the Pb target was developed, based on actual reactor parameters. MCNPX calculations were carried out to estimate the absolute radial and axial neutron flux in the reactor, and to calculate the multiplication factor K eff and heat generated in the reactor. Numerical results showed an enormous increase in the neutron flux, by seven orders of magnitude, compared to the current JSA core design using Pu-Be source. In this research the results obtained are discussed and compared with those of the JSA, and do confirm the feasibility of utilizing the JSA as a viable nuclear research facility with adequate neutron flux. [ABSTRACT FROM AUTHOR]

The use of gamma irradiation technologies generates a number of complex scientific and technical problems; for example, the target is manually loaded onto turntables and is rotated during the entire irradiation process and the MCNPX three-dimensional geometry simulation is kept static. For this, it is necessary to introduce additional approaches. In this paper, two new methodologies are proposed for the simulation of irradiation process in panoramic gamma irradiator. The study was performed at the gamma irradiation facility at the Nuclear Technology Development Center of the National Nuclear Energy Commission, Brazil. The source can be reproduced with a homogenized geometry. Validation of MCNPX calculations of gamma doses were performed by thorough comparison with the experimental measurements. The contribution of this proposed source models has opened new lines of research. The results of this study showed that the proposed source models effectively represent the irradiation process. [ABSTRACT FROM AUTHOR]

67 Cu radioisotope is a beta particle-emitting nuclide used in radioimmunotherapy (RIT) as well as for imaging, tracer kinetic studies and dosimetry. 67 Cu can be produced by bombarding nat Zn with deuterons. In this study, the physical yields of 67 Cu via nat Zn(d,x) 67 Cu reaction channel as well as via subreactions of 68 Zn(d,2pn) 67 Cu, 67 Zn(d,2p) 67 Cu, 70 Zn(d,2p3n) 67 Cu, 68 Zn(d,x) 67 Ni(T 1/2 =21 s)→ 67 Cu and 70 Zn(d,x) 67 Ni(T 1/2 =21 s)→ 67 Cu in the nat Zn target have been calculated by using the MCNPX-2.6, TALYS-1.8 and SRIM codes. Also, the total cross sections for production of 67 Cu from nat Zn(d,x) 67 Cu reaction channel in the energy range of 15–45 MeV have been estimated by TALYS code. The best reaction to produce 67 Cu radionuclide in a carrier free form was chosen with deuteron energy around 30 MeV on 70 Zn thick target. Good agreement between the calculated results and the experimental values shows that the employed methods can be used for prediction and production estimation in cyclotron. [ABSTRACT FROM AUTHOR]

In the present work the effect of tungsten oxide (WO 3 ) nanoparticles on mass attenauation coefficients of concrete has been investigated by using MCNPX (version 2.4.0). The validation of generated MCNPX simulation geometry has been provided by comparing the results with standard XCOM data for mass attenuation coefficients of concrete. A very good agreement between XCOM and MCNPX have been obtained. The validated geometry has been used for definition of nano-WO 3 and micro-WO 3 into concrete sample. The mass attenuation coefficients of pure concrete and WO 3 added concrete with micro-sized and nano-sized have been compared. It was observed that shielding properties of concrete doped with WO 3 increased. The results of mass attenauation coefficients also showed that the concrete doped with nano-WO 3 significanlty improve shielding properties than micro-WO 3 . It can be concluded that addition of nano-sized particles can be considered as another mechanism to reduce radiation dose. [ABSTRACT FROM AUTHOR]

Combination of two or three dissimilar scintillator materials as a radiation detector has found major role in environmental radiation monitoring. In this paper, a three-layer Phoswich detector including BC-400, YAG, and CsI was designed to efficiently discriminate gamma-ray in the beta events up to 3.2 MeV using a simple rise-time discrimination method. MCNPX Monte Carlo code was used to obtain interaction probability of beta and gamma-rays as well as optimum thicknesses of the layers in the designing process. The optical transport of the system was simulated by GEANT4. In this regard, the pulses from simultaneous beta-gamma emitter sources were detected and discriminated based on pulse's rise-time so that the minimum number of gamma-ray contaminating events was observed in the beta spectrum. The results showed that using the proposed configuration and the method, output pulses with a rise-time shorter than 9 ns have been successfully detected as a beta particle while those with rising time longer than 15 ns have been identified as gamma-ray events. Overall results revealed that using the proposed system, an individual spectrum of beta particles or gamma-rays can be recorded from a simultaneous beta-gamma emitter source that minimizes contribution of the other radiation. • A Phoswich detector was designed for simultaneous beta-gamma radiation. • Scintillator materials used in the detector structure should have different decay time constant. • GEANT4 simulation was used for scintillation lights transport. • A pulse discrimination method based on the pulse rise time has been proposed for pulse processing. • Beta-rays were successfully discriminated against gamma-rays by the proposed detector and associated pulse processing method. [ABSTRACT FROM AUTHOR]

Regardless the motivation terrorism is the most important risk for the national security in many countries. Attacks with explosives are the most common method used by terrorists. Therefore several procedures to detect explosives are utilized; among these methods are the use of neutrons and photons. In this study the Monte Carlo method an explosive detection system using a 241 AmBe neutron source was designed. In the design light water, paraffin, polyethylene, and graphite were used as moderators. In the work the explosive RDX was used and the induced gamma rays due to neutron capture in the explosive was estimated using NaI(Tl) and HPGe detectors. When light water is used as moderator and HPGe as the detector the system has the best performance allowing distinguishing between the explosive and urea. For the final design the Ambient dose equivalent for neutrons and photons were estimated along the radial and axial axis. [ABSTRACT FROM AUTHOR]

The response of a scintillation neutron detector of ZnS(Ag) with 10 B was calculated, using the MCNPX Monte Carlo Code. The detector consists of four panels of polymethyl methacrylate (PMMA) and five thin layers of ~0.017 cm thick 10 B+ZnS(Ag) in contact with the PMMA. The response was calculated for the bare detector and with different thicknesses of High Density Polyethylene, HDPE, moderator for 29 monoenergetic sources as well as 241 AmBe and 252 Cf neutrons sources. In these calculations the reaction rate 10 B(n, α) 7 Li and the neutron fluence in the sensitive area of the detector 10 B+ZnS(Ag) was estimated. Measurements were made at the Neutron Measurements Laboratory, Universidad Politécnica de Madrid, LMN-UPM, to quantify the detections in counts per second in response to a 252 Cf neutron source separated 200 cm. The MCNPX computations were compared with measurements to estimate the efficiency of ZnS(Ag) for detecting the α that is created in the 10 B(n, α) 7 Li reaction. After validating new models with different geometries it will be possible to improve the detector response trying to achieve a sensitivity of 2.5 cps-ng 252 Cf comparable with the response requirements for 3 He detectors installed in the Radiation Portal Monitors, RPMs. This type of detector can be considered an alternative to the 3 He detectors for detection of Special Nuclear Material, SNM. [ABSTRACT FROM AUTHOR]

Intensity Modulated Radiation Therapy (IMRT) is an advanced treatment technique, widely used in external radiotherapy. This paper presents the SOFT-RT which allows the simulation of an entire IMRT treatment protocol. The SOFT-RT performs a full three-dimensional renderization of a set of patient images, including the definitions of region of interest with organs in risk (OIR), and the target tumor volume and margins (PTV). Thus, a more accurate analysis and planning can be performed, taking into account the features and orientation of the radiation beams. The exposed tissues as well as the amount of absorbed dose is depicted in healthy and/or cancerous tissues. As conclusion, SOFT-RT can predict dose on the PTV accurately, preserving the surrounding healthy tissues. SOFT-RT is coupled with SISCODES code. The SISCODES code is firstly applied to segment the set of CT or MRI patient images in distinct tissues pointing out its respective density and chemical compositions. Later, the voxel model is export to the SOFT-RT IMRT planning module in which a full treatment planning is created. All geometrical parameters are sent to the general-purpose Monte Carlo transport code-MCNP-to simulate the interaction of each incident beam towards to the PTV avoiding OIR. Computational simulations is running on MCNPx. The normalized dose results are exported to the SOFT-RT output-module, in which the three-dimensional model visualization is shown in a transparent glass procedure adopting gray scale for the dependence on the mass density of the correlated tissue; while, a color scale to depict dose values in a superimpose protocol. [ABSTRACT FROM AUTHOR]

The Laboratory of Internal Dosimetry of the Center for Development of Nuclear Technology (LDI/CDTN) is responsible for routine internal monitoring of occupationally exposed individuals. The determination of photon emitting radionuclides in the human body requires calibration of the detector in specific counting geometries. The calibration process uses physical phantoms containing certified activities of the radionuclides of interest. The objective of this work was to obtain calibration efficiency curves of the Whole Body Counter in operation at the LDI/CDTN using a BOMAB physical phantom and Monte Carlo simulations. [ABSTRACT FROM AUTHOR]

This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2 MW, with new configurations based on new ICs, different safety-related thermal–hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. [ABSTRACT FROM AUTHOR]

The European Pressurized Water Reactor (EPR) and Hualong One Pressurized Water Reactor (HPR) are two of the reactors under consideration by the Ghana Nuclear Power Programme. Radiotoxicity analysis of Spent Nuclear Fuel (SNF) assembly was carried out with these commercial Pressurized Water Reactor (PWR) nuclear power technology as case study. This will help determine which one is less radiotoxic on the environment between the two reactor technologies, in the long run. Burnup depletion calculation for the Uranium Oxide (UOX) fuel of these reactor technologies was simulated, using Monte Carlo Neutron Particle Extended (MCNPX), a code used in nuclear fuel management analysis, being a well validated code and also due to its versatile nuclei reactions cross section library. Determination of radiotoxicity for EPR and HPR SNF is the main objective of this study. The radiotoxicity was achieved taking into consideration the radioactive decay rate of the radionuclides and the Dose Factor of each radionuclide present in the SNF using the International Commission on Radiological Protection (ICRP) compendium of Dose Factors due to ingestion. The radiotoxicity for the two reactor's SNF were compared. The initial radiotoxicity for HPR SNF was higher in the duration below one hundred years but at about a hundred years and above, the radiotoxicity was higher for EPR SNF. The radiotoxicity was tremendously reduced for the reprocessed spent UOX fuel (with the Pu and U extracted) to be used as mixed oxide (MOX) fuel. The main finding is that Pu isotopes are the major contributors to the radiotoxicity of the SNF for the two reactors systems due to their very high radioactivity, long half-lifes and high dose factors as compared to other actinides and fission products present in the SNF. • The EPR and HPR fuel assemblies are both designed to have a square lattice containing 265 fuel pins. • It is observed that the initial high radiotoxicity of the SNF is from the fisssion products. • Thorium exhibited high radiotoxicity for the SNF of the two reactors considered. • The radiotoxicity of the Th-UOX fuel cycle being considered as fuel for future similar reactors be further investigated. [ABSTRACT FROM AUTHOR]

We have designed a silicon-photomultiplier-based remote radiation-sensing module consisting of a master port (displaying radiation information) and a slave port (detects radiation, transmits to master). The master port merges radiation and dose values and displays them. Counting detection efficiency and radiation response simulated using MCNPX were used to calibrate the module. We performed radioactive source tests ( 137 Cs, 22 Na, 60 Co, 55 Fe) and compared experimental and simulation results. Remote detection capability was demonstrated and the detection accuracy was determined. Applications abound in the radioactivity industry. [ABSTRACT FROM AUTHOR]

Simulation, design and fabrication of a sealing enclosure is carried out for a NaI(Tl) 2″×2″ detector, to be used as in situ gamma radioactivity measurement system in marine environment. Effect of sealing enclosure on performance of the system in laboratory and marine environment (distinct tank with 10 m 3 volume) were studied using point sources. The marine volumetric efficiency for radiation with 1461 keV energy (from 40 K ) is measured with KCl volumetric liquid source diluted in distinct tank. The experimental and simulated efficiency values agreed well. Marine volumetric efficiency calibration curve is calculated for 60 keV to 1461 keV energy with Monte Carlo method. This curve indicates that efficiency increasing rapidly up to 140.5 keV but then drops exponentially. [ABSTRACT FROM AUTHOR]

The feasibility of using a laser-accelerated proton beam to produce a neutron source, via (p,n) reaction, for Boron Neutron Capture Therapy (BNCT) applications has been studied by MCNPX Monte Carlo code. After optimization of the target material and its thickness, a Beam Shaping Assembly (BSA) has been designed and optimized to provide appropriate neutron beam according to the recommended criteria by International Atomic Energy Agency. It was found that the considered laser-accelerated proton beam can provide epithermal neutron flux of ∼2×10 6 n /cm 2 shot. To achieve an appropriate epithermal neutron flux for BNCT treatment, the laser must operate at repetition rates of 1 kHz, which is rather ambitious at this moment. But it can be used in some BNCT researches field such as biological research. [ABSTRACT FROM AUTHOR]

In this work we have studied the feasibility of photonuclear production of 47 Sc from 48 Ti via 48 Ti(γ,p) 47 Sc reaction. Photon flux distribution for electron beams of different energies incident on tungsten converter was calculated using MCNPX radiation transport code. 47 Sc production rate dependence on electron beam energy was found and 47 Sc yields were estimated. It was shown that irradiating a natural Ti target results in numerous scandium isotopes which can reduce the specific activity of 47 Sc. Irradiating enriched 48 Ti targets with a 22 MeV 1 mA beam will result in hundreds of MBq/g activity of 47 Sc and no other isotopes of scandium. Decreasing the size of the target will result in much higher average photon flux through the target and tens of GBq/g levels of specific activity of 47 Sc. Increasing the beam energy will also result in higher yields, but as soon as the electron energy exceeds the 48 Ti(γ,np) 46 Sc reaction threshold, 46 Sc starts being produced and its fraction in total scandium atoms grows as beam energy increases. The results of the simulations were benchmarked by irradiating natural titanium foil with 22 MeV electron beam incident on the tungsten converter. Measured 47 Sc activities were found to be in very good agreement with the predictions. [ABSTRACT FROM AUTHOR]

The MCNPX code was used to determine the efficiency of a N-type HPGe detector after two decades of operation. Accounting for the roundedness of the crystal`s front edges and an inhomogeneous description of the detector's dead layers were shown to achieve better agreement between measurements and simulation efficiency determination. The calculations were experimentally verified using point sources in the energy range from 50 keV to 1400 keV, and an overall uncertainty less than 2% was achieved. In order to use the detector for different matrices and geometries in radioactivity, the suggested model was validated by changing the counting geometry and by using multi-gamma disc sources. The introduced simulation approach permitted the revaluation of the performance of an HPGe detector in comparison of its initial condition, which is a useful tool for precise determination of the thickness of the inhomogeneous dead layer. [ABSTRACT FROM AUTHOR]

A prototype x-ray system based on a liquid-metal-jet anode was evaluated within the framework of the LABSYNC project. The generated spectrum was measured using a CZT-based spectrometer and was compared with spectra simulated by three Monte Carlo codes: MCNPX, PENELOPE and EGS5. Notable differences in the simulated spectra were found. These are mainly attributable to differences in the models adopted for the electron-impact ionization cross section. The simulation that more closely reproduces the experimentally measured spectrum was provided by PENELOPE. [ABSTRACT FROM AUTHOR]

Abstract: A novel free release measurement facility (FRMF) was developed within the joint research project “Metrology for Radioactive Waste Management” of the European Metrology Research Programme. Before and during FRMF design and construction, Monte Carlo calculations with MCNPX and PENELOPE codes were used to optimize the thickness of the shielding, the dimensions of the container, and the shape of detector collimators. Validation of the numerical models of the FRMF detectors and the results of the optimization are discussed in the paper. [Copyright &y& Elsevier]

This study discusses a computer simulation for the equivalent ambient dose due to photons, H*(10)p, and neutrons, H*(10)n, in the patient's plane undergoing radiation therapy. A standard radiotherapy room with an additional shielding made by one lead or steel tenth-value layer was considered. A Varian 2100/2300 C/D linear accelerator head operating at 18 MV was modeled. Jaw openings of 5 cm × 5 cm, 10 cm × 10 cm, 20 cm × 20 cm, and 30 cm × 30 cm, as well as the multileaf collimator under eight different angles of gantry inclination, were also modeled. The use of steel in the shield generates a slightly raised average value of H*(10)p (0.527%) compared to when using lead. This finding can be interpreted as that the use of lead or steel coating makes no difference to the additional shield calculations when only photons are considered. When considering the contribution to H*(10)n, there is a significant difference (11.7% increase) for using lead compared to steel shielding. • Simulation of H*(10)p on the patient plane allows assessing of additional shielding. • No difference for lead or steel additional shielding concerning photons. • Lead increases H*(10) in comparison to steel. [ABSTRACT FROM AUTHOR]

Abstract: We investigate the ability of high spatial resolution (∼120μm) Ge-doped SiO2 TL dosimeters to measure photoelectron dose enhancement resulting from the use of a moderate to high-Z target (an iodinated contrast media) irradiated by 90kVp X-rays. We imagine its application in a novel radiation synovectomy technique, modelled by a phantom containing a reservoir of I2 molecules at the interface of which the doped silica dosimeters are located. Measurements outside of the iodine photoelectron range are provided for using a stepped-design that allows insertion of the fibres within the phantom. Monte Carlo simulation (MCNPX) is used for verification. At the phantom medium I2-interface additional photoelectron generation is observed, ∼60% above that in the absence of the I2, simulations providing agreement to within 3%. Percentage depth doses measured away from the iodine contrast medium reservoir are bounded by published PDDs at 80kVp and 100kVp. [Copyright &y& Elsevier]

Abstract: Some of the radionuclides used for efficiency calibration of a HPGe spectrometer are subject to coincidence-summing (CS) and account must be taken of the phenomenon to obtain quantitative results when counting samples to determine their activity. We have used MCNPX simulations, which do not take CS into account, to obtain peak intensities that were compared to those observed experimentally. The loss or gain of a measured peak intensity relative to the simulated peak is attributed to CS. CS correction factors are compared with those of ETNA and GESPECOR. Application to a test sample prepared with known radionuclides gave values close to the published activities. [Copyright &y& Elsevier]

Abstract: The new manganese bath facility of the Laboratoire National Henri Becquerel has been modeled by using three Monte-Carlo codes: MCNPX, GEANT4, and FLUKA, in order to determine the correction factors needed in the neutron source calibration process. The most realistic source geometry has been determined, and the most reliable cross sections library has been chosen. The models were compared, and discrepancies between the codes have been pointed out. Potential causes of deviations between results were assessed and discussed using additional models. Finally, an experimental process is proposed to validate the accuracy of the different codes and their abilities in simulating the neutron capture by the manganese bath. [Copyright &y& Elsevier]

Abstract: Germanium metal targets encapsulated in Nb shells were irradiated in a proton beam. Proton and secondary neutron beam fluences as well as radionuclide activity formation were modeled using MCNPX in combination with CINDER90. Targets were chemically processed using distillation and anion exchange. Good agreement between the measured radiochemical yields and MCNPX/CINDER90 estimates was observed. A target of pentavalent 73,74As radioarsenic for neutron activation studies was prepared. [Copyright &y& Elsevier]

Abstract: Some remarkable advances have been made in the last years on the SPES-BNCT project of the Istituto Nazionale di Fisica Nucleare (INFN) towards the development of the accelerator-driven thermal neutron beam facility at the Legnaro National Laboratories (LNL), aimed at the BNCT experimental treatment of extended skin melanoma. The compact neutron source will be produced via the 9Be(p,xn) reactions using the 5MeV, 30mA beam driven by the RFQ accelerator, whose modules construction has been recently completed, into a thick beryllium target prototype already available. The Beam Shaping Assembly (BSA) final modeling, using both neutron converter and the new, detailed, Be(p,xn) neutron yield spectra at 5MeV energy recently measured at the CN Van de Graaff accelerator at LNL, is summarized here. [Copyright &y& Elsevier]

Abstract: The response of PADC was calculated by Monte Carlo simulation using MCNPX code. H, C and O recoils as well as nuclear reactions have been taken into account. The calculated results show important contribution to detection efficiency of PADC due to H, C and O recoils. The irradiated detectors with 241Am–Be source and 241Am–Be softened with a spherical polyethylene moderator were electrochemically etched. The paper presents the results of the experimental and Monte Carlo simulations, which are in a good agreement. [ABSTRACT FROM AUTHOR]

Abstract: A procedure to optimize the geometrical model of an n-type detector is described. Sixteen lines from seven point sources (241Am, 133Ba, 22Na, 60Co, 57Co, 137Cs and 152Eu) placed at three different source-to-detector distances (10, 20 and 30cm) were used to calibrate a low-background gamma spectrometer between 26 and 1408keV. Direct Monte Carlo techniques using the MCNPX 2.6 and GEANT 4 9.2 codes, and a semi-empirical procedure were performed to obtain theoretical efficiency curves. Since discrepancies were found between experimental and calculated data using the manufacturer parameters of the detector, a detail study of the crystal dimensions and the geometrical configuration is carried out. The relative deviation with experimental data decreases from a mean value of 18–4%, after the parameters were optimized. [Copyright &y& Elsevier]

Abstract: As a first step into the development of a neutron dose monitor, a stilbene detector is used to test a procedure applicable to other organic scintillators allowing for neutron–photon discrimination. The pulses are measured by numerical acquisition and their amplitude and decay time are calculated by software. The discrimination is performed in an amplitude-decay time plot, and separate amplitude spectra are built. These spectra are unfolded MAXED, using the detector response matrices calculated with MCNPX. [Copyright &y& Elsevier]

Abstract: In December 2003 the irradiator of the Portuguese 60Co irradiation facility, UTR, was replenished. Monte Carlo simulation studies were used to characterize the facility. Comparison of simulation results with experimental measurements were useful in estimating where the source elements were positioned and what the total activity was. For this purpose the MCNPX code was used. From the experimental point of view ionization chambers and PMMA dosemeters were used. The results indicate that the geometry of the irradiator used for simulation studies shows a shift on the vertical dimension relative to the actual geometry of approximately 8cm, but with reduced influence on the irradiation process. Considering the uncertainties of the experimental and simulated approaches the activity of the irradiator stated by supplier has been confirmed. [Copyright &y& Elsevier]

This study focusses on the risk of potential exposure to radiation for personnel driving a truck as well as illegal individuals being transported in cargo containers. Inspection facilities usually use a high energy linear accelerator (linac) in order to inspect the cargo. Since this type of equipment has associated health risks due to potential unwanted exposure, the occupational and public dose limits should be calculated in order to develop safer work conditions. This work used a computation model running the code MCNPX to simulate a typical cargo inspection facility which used a linac operating at 4.5 MeV. Two scenarios were considered: (1) exposure of the driver to the primary beam due to a potential failure of the safety sensors; and (2) dose received by an illegal individual being transported inside the cargo container. The results show a dose of 0.8514 mSv per scan for the driver exposed to the primary X-ray beam, and 0.1997 mSv per scan for an individual transported in the cargo box. In conclusion, both the individual and the driver received a dose below the acceptable limit considered safe for an individual (1 mSv/year). However, that was the value of one scan; in a case in which multiple scans would be performed, the dose limit can be quickly exceeded. In that case, the limit would be exceeded by the driver faster than by the individual in the cargo. • Focus on the risk of potential exposure to radiation for personnel driving a truck. • MCNPX computation model simulates a typical cargo inspection facility. • Results suggest that the system is safe even in an abnormal situation. [ABSTRACT FROM AUTHOR]

The fast neutron and gamma ray attenuation capability of the most common thermoplastic polymers used in nuclear applications has been evaluated theoretically. Monte Carlo simulation has been used to compute the gamma-ray energy absorption buildup factor in the energy range 0.015–15 MeV at penetration depths up to 40 MFP. The results of MCNPX calculations have been validated against the results derived from the Geometric Progression fitting method. To evaluate neutron attenuation performance of the polymers, the fast neutron removal cross-section has been determined using theoretical database. Despite the superior ability of polysulfone and poly (ether sulfone) in gamma ray attenuation, high-density polyethylene has been found to have the best fast neutron removal ability among all. The detailed insights into the fast neutron and gamma ray shielding properties of selected polymers in the present work might have great potential applications in nuclear systems. • The shielding properties for thermoplastic polymers were studied in mixed neutron-gamma radiation field. • EABF have been calculated using MCNPX codes and validated based on the values obtained using G.P. fitting method. • The values of macroscopic removal cross-sections for neutrons were determined. • Measured data are useful in nuclear medicine application. [ABSTRACT FROM AUTHOR]

The accuracy of an out-of-field dose from an Elekta Synergy accelerator calculated using the X-ray Voxel Monte Carlo (XVMC) dose algorithm in the Monaco treatment planning system (TPS) for both low-energy (6 MV) and high-energy (15 MV) photons at cardiac implantable electronic device (CIED) depths was investigated through a comparison between MCNPX simulated out-of-field doses and measured out-of-field doses using three high spatial and sensitive active detectors. In addition, total neutron equivalent dose and fluence at CIED depths of a 15-MV dose from an Elekta Synergy accelerator were calculated, and the corresponding CIED relative neutron damage was quantified. The results showed that for 6-MV photons, the XVMC dose algorithm in Monaco underestimated out-of-field doses in all off-axis distances (average errors: −17% at distances X < 10 cm from the field edge and −31% at distances between 10 < X ≤ 16 cm from the field edge), with an increasing magnitude of underestimation for high-energy (15 MV) photons (up to 11%). According to the results, an out-of-field photon dose at a shallower CIED depth of 1 cm was associated with greater statistical uncertainty in the dose estimate compared to a CIED depth of 2 cm and clinical depth of 10 cm. Our results showed that the relative neutron damage at a CIED depth range for 15 MV photon is 36% less than that reported for 18 MV photon in the literature. • A MC model of the Elekta Synergy head to assess out-of-field doses for both low and high energy photons. • The accuracy of out-of-field dose calculations by XVMC dose algorithm in Monaco. • An out-of-field dose analysis of microdiamond, PinPoint 3D, Semiflex 3D. • Neutron equivalent dose and fluence at CIED depths. • The CIED relative neutron damage of a 15 MV photon beam was quantified. [ABSTRACT FROM AUTHOR]

Current fission-based methods of 99Mo production require single use uranium targets which leads to spent uranium waste. This waste could be reduced if a target is developed that does not require dissolution so that it can be reused for multiple production runs. MCNP6.2 was used to model reusable targets of 20% and 1% enrichment for activity produced, target efficiency and burnup. The 1% enriched target was found to be much more efficient but had a lower activity produced compared to the 20% enriched target. The ideal target design for 99Mo production that optimises efficiency and reusability and reduces the self-shielding effect of UO 2 was found to be a target that is made from 1% enriched UO 2 with density as high as allowable for sufficient yields, efficient 99Mo extraction and having an irradiation time of 5 days, with the target able to be re-irradiated and re-processed 2–4 times. • Current fission based manufacturing of 99Mo using uranium targets is based on target destruction. • Target destruction leads to increased radioactive waste volumes. • MCNP6 modelling of uranium targets show that higher density targets have more self-shielding effects and are less efficient. • Low density uranium targets with 5 days irradiation maximise 99Mo output and minimising radioactive waste. [ABSTRACT FROM AUTHOR]

A thermal neutron flux measurement tool with two perpendicular sets of plastic scintillator arrays was designed and simulated (Ghal-Eh and Green, 2016) with the MCNPX code (Version 2.6.0, with ENDF/B-VII cross section library (ENDF, 2011)). The proposed system aimed to provide a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1 H(n t h , γ) 2 D reactions. In the present work, using Monte Carlo code FLUKA and its scintillation light transport capability, several important upgrades were carried out to include the light transport modeling in the response of plastic scintillators, analyze the cross-talk phenomenon, optimize the system geometry, and also provide a new approach in thermal neutron image reconstruction. The results showed that the last two cases played a significant role in improving the longitudinal profile of thermal neutron flux. • Thermal neutron flux imaging based on the detection of 2.22 MeV prompt gamma-rays. • Image reconstructed from 17 scintillator responses using appropriate algorithm. • Optimizing collimator holes and using B 4 C as neutron absorber led to desired image. • Cross-talk between detectors is not significant due to the energy range definition. • Response uniformity achieved in a polished scintillator with extra isolated length. [ABSTRACT FROM AUTHOR]

Abstract: This work presents a methodology for digital radiography simulation for industrial applications using the MCNPX radiography tally. In order to perform the simulation, the energy-dependent response of a BaFBr imaging plate detector was modeled and introduced in the MCNPX radiography tally input. In addition, a post-processing program was used to convert the MCNPX radiography tally output into 16-bit digital images. Simulated and experimental images of a steel pipe containing corrosion alveoli and stress corrosion cracking were compared, and the results showed good agreement between both images. [Copyright &y& Elsevier]

In the previous study, we found the feasibility of a cyclotron-based BNCT using the Ta(p,n) neutrons at 90° bombarded by 50 MeV protons, and the iron, AlF3, Al and 6LiF moderators by simulations using the MCNPX code. In order to validate the simulations to realize the cyclotron-based BNCT, we measured the epithermal neutron energy spectrum passing through the moderators with our new spectrometer consisting of a 3He gas counter covered with a silicon rubber loaded with natB and polyethylene moderator and the depth distribution of the reaction rates of 197Au(n,γ)198Au in an acrylic phantom set behind the rear surface of the moderators. The measured results were compared with the calculations using the MCNPX code. We obtained the good agreement between the calculations and measurements within ∼10% for the neutron energy spectra and within ∼20% for the depth distribution of the reaction rates of 197Au(n,γ)198Au in the phantom. The comparison clarified a good accuracy of the calculation of the neutron energy spectrum passing through the moderator and the thermalization in a phantom. These experimental results will be a good benchmark data to evaluate the accuracy of the calculation code. [Copyright &y& Elsevier]

The purpose of this research is the investigation of the 90Zr(n,p)90Y reaction as an alternative method to the traditional fission product based on the 90Sr/90Y generator. The fast neutrons necessary to activate 90Zr are generated through (p,xn) reactions during 33 MeV proton irradiation of natural tungsten or other targets. Since 90Y is a pure beta emitter, the gamma-rays from the 90Zr(n,2n)89Zr reaction were used to quantify the neutron flux incident on the 90Zr sample. Using the simulation code MCNPX (Monte Carlo N-Particle System), the angular and energy distributions of neutrons incident on the Zr target were calculated. Based on the MCNPX neutron flux predictions, the 90Y activity was determined. [Copyright &y& Elsevier]

In the oil production industry, water is used as a fluid injected into the well to raise the oil when the well is depressurized. Water thus produced presents variations in the concentrations of dissolved salts, as there is a mixture of different types of water, related to its origin (such as connate water, sea water). Because it is reused in oil production, water needs to be monitored to maintain the standard suitable for its use as it can be hypersaline, contributing to the encrustation of pipes and contamination of underground water reservoirs. In this study, a noninvasive method was developed to determine the salt concentration in seawater. The method uses a detection system that contains a NaI(Tl) detector, a241Am source, and a sample holder to measure the mass attenuation coefficient of saltwater samples. For validation, the same setup was also simulated using the MCNPX code. Saltwater samples with different concentrations of NaCl and KBr were used as a proxy for seawater. The mass attenuation coefficients for the simulation exhibited the smallest relative errors (up to 6.2%), and the experimental ones exhibited the highest relative errors (up to 25%) when compared with theoretical values. • Using a low energy radioactive source, a noninvasive method was designed to measure the salt concentrations (salinity) in seawater. • To validate the method, five saline samples were prepared and titrated to yield precise densities. The mass attenuation coefficients of the samples were measured in experiments. • The experiments were also simulated using the Monte Carlo method, and the results were compared to the XCOM reference values. • The experimental results closely follow the linear behavior predicted by the Monte Carlo simulations. [ABSTRACT FROM AUTHOR]

Measuring the wax deposition inside pipelines is one of the critical parameters in the oil, gas and petrochemical industries to control the flow through the pipelines. This paper presents a novel method using artificial neural networks to measure the thickness of the wax. This method was based on counting the backscattered gamma-ray from different thicknesses of the wax inside the pipes with different diameters. For this purpose, the system was simulated by MCNPX code and the designed setup was optimized. The main analyses were based on the simulation results but the verification was performed using a real experimental setup. The results showed a good agreement between the simulation results and the experimental data with a root mean square error less than 1%. Response of the detector was simulated for a standard industrial nominal pipe ranged from 2 to 4.5 inches and for radiation sources 137Cs and 60Co. Using these data, a multilayer perceptron for different energy sources was trained. The best prediction of the wax thickness was obtained for the case of using two radiation sources, simultaneously. The output of the trained neural network showed that the proposed method is capable of measuring the wax thickness inside the pipe with a good accuracy. • A new intelligent system was developed for detecting wax in different sized pipes. • The system works based on backscattered γ-rays technique and ANN. • Training of the ANN was done using experimentally verified MCNPX simulation results. • The proposed method can predict the wax thickness with good accuracy and a RMSE about 0.2%. • The best response was obtained when 137Cs and 60Co were employed, simultaneously. [ABSTRACT FROM AUTHOR]

In this paper, a nuclear gauge was simulated using MCNPX and FLUKA codes. This device consists of a fluid-containing vessel, three BC400 rod plastic scintillator and a60Co gamma source. Simulation studies show that the changes in the count of each of the three detectors and the sum of their counts decrease with increasing vessel water height. The simulation results were compared with the measurement results. Comparison results show that the mean count difference between MCNPX code and experimental results is about 3% lower than FLUKA code, but the computation time using FLUKA code is approximately 2.8 times less than the MCNPX code. • A nuclear gauge system was simulated fully using multi-purpose Monte Carlo codes MCNPX and FLUKA. • The performance level of MCNPX and FLUKA codes have been determined in simulating the nuclear gauge system. • The comparisons show very good agreement between the simulation results and the experimental results. • This study shows the accuracy of simulation in calibration and measurement of the liquid surfaces in a nuclear gauge. [ABSTRACT FROM AUTHOR]