Monte Carlo Computer Code Research Articles

Volume fraction detection in multiphase systems using neutron activation analysis and artificial neural network

This study presents an application of an Artificial Neural Network (ANN) to detect fluids in an annular flow regime using Prompt-Gamma Neutron Activation Analysis (PGNAA). The ANN was trained using gamma-ray spectra resulting from neutron interactions with chemical elements found in fluids typical of multiphase flow in oil exploration. These spectra were generated through mathematical simulation using the MCNP6 Monte Carlo computer code to model nuclear particle transport. A241Am-Be polyenergetic neutron source was simulated for these calculations. Several combinations of fluid fractions were developed to create a dataset used for both training and evaluation of the ANN. The ANN demonstrated robust generalization capabilities by accurately predicting the volume fraction of the three investigated fluids (saltwater, oil, and gas), even for cases not included in the training phase. The combination of ANN and PGNAA proved effective for analyzing multiphase systems, with over 92% of all showing errors of less than 5%.

Experiments on Criticality and Reactivity Worths in the FCA-XXII-1 Assembly Simulating Highly Enriched MOX Fueled Tight Lattice LWR Cores

A series of integral experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency, simulating light water reactor cores with a tight lattice cell of highly enriched mixed-oxide (MOX) fuel containing >15% fissile plutonium (Pu). The three experimental configurations of the FCA-XXII-1 assembly were constructed using foamed polystyrene with different void fractions (45%, 65%, and 95%) to clarify the prediction accuracy of neutronics calculation codes and nuclear data libraries among various neutron spectra. The hydrogen-to–nuclear fuel atomic ratio varied from 0.1 to 0.8. The nuclear characteristics measured in the experiments were criticality (keff), moderator void reactivity worths, and sample reactivity worths using boron carbide (20%, 60%, and 90% 10B enrichment) and Pu (92%, 81%, and 75% fissile Pu ratio). Preliminary analyses on experiments were conducted using a deterministic calculation code system conventionally used for fast reactors and the Japanese evaluated nuclear data library of JENDL-4.0. The calculated keff values overestimated the experiments beyond the experimental uncertainties. However, most reactivity worth calculations agreed well with the experimental values. Even beyond the experimental uncertainties, discrepancies between the calculation and the experiment were <13%. Specifically in the reactivity worth analyses of the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations for criticality and large reactivity worths were performed with the Monte Carlo calculation code MVP3 by modeling the experimental configurations in detail, confirming that the deterministic calculations closely agreed with the reference values.

The Study for Standardization of Dose Evaluation Method in Cone-beam CT

This study aimed to compare the dose evaluation methods by constructing simulation models using the Monte Carlo calculation code and propose an evaluation method for cone beam CT (CBCT) that ensures accuracy and practicality. The Particle and Heavy Ion Transport code System (PHITS) ver. 3.26 was used as the Monte Carlo calculation code. CBCT doses were measured by CB dose index (CBDI) and American Association of Physicists in Medicine task group 111 (TG111) methods. The CBDI was compared with the equilibrium doses obtained by the TG111 method. Although CBDI was lower than equilibrium doses obtained by the TG111 method, its practicality was ensured because it can be measured using the dosimeter and phantom that are commonly used. In contrast, the TG111 method guarantees accuracy, but it is difficult to prepare a long phantom to obtain the equilibrium dose. The TG111 method with a phantom length of 15 cm underestimated the equilibrium dose by 20% compared to that with a phantom length of 45 cm that satisfies the dose equilibrium. Therefore, the equilibrium dose obtained by the TG111 method with a phantom length of 15 cm is multiplied by 1.20 to obtain the equilibrium dose equivalent to that with a phantom length of 45 cm. This study has proposed the dose evaluation method that combines guarantees accuracy and practicality in CBCT.

Graphical Visualisation of the MCNP Mesh Tally File

This paper presents an updated version of the MTV3D (Mesh Tally Visualization in 3D) program with Windows graphical user interface for visualization of the MCNP-based mesh tally files. Several improvements over the previous program version are addressing better figure export functionality ("Save as" option), switching between linear and logarithmic values on axes, dynamic figure scaling in active window, inversion of relative errors from “max” to “min” values, etc. MCNP is a well known and widely used general purpose Monte Carlo computer code for neutron, photon and electron transport simulation through arbitrary three-dimensional configurations. An important feature of the MCNP code is a graphical display of the simulation model using auxiliary program, such as X-window server, which is useful for geometry error-checking during model setup and visualisation of Monte Carlo results from a mesh tally file (i.e. meshtal file) over a structured xyz mesh. Such inspection of the model is useful for the end user, providing an insight of the Monte Carlo convergence process in a phase space and effectiveness of the selected variance reduction parameters in shielding calculations. Basic features and functionalities of the updated MTV3D program are presented on some selected hybrid-shielding problems involving ADVANTG3.0.3 and MCNP6.1.1b codes.

Analysis of the Neutron Radiographic Image Quality and Beam Intensity Produced by a Compact Multichannel Collimator

We investigate the image quality and beam intensity of thermal neutron radiography after replacing a standard single-channel neutron collimator with a compact array of microcollimators. In this study, the MCNP6 Monte Carlo computer code was used to simulate a 2 × 2-cm-area isotropic thermal neutron source, which then was collimated by an array of micron-sized neutron collimators that measured 29.8 μm in diameter and with lengths that varied from 0.6 to 3 mm. These microcollimators were spaced 30 μm apart and assembled into a 2 × 2-cm array. The image quality of the neutron beams produced by the resulting collimator arrays was assessed by imaging the edge of a very thin (~0.01-mm) gadolinium foil to obtain the image Modulation Transfer Function (MTF). The MCNP6 resulting flux map from each simulation then was converted into a grayscale .tiff image and the image’s resulting MTF obtained using the ImageJ computer program with the imaging beam geometric unsharpness, which is a limiting factor in the image resolution determined at the 10% value of the MTF curve. In this study, we found that a 2 × 2× 0.298-cm microcollimator, corresponding to a length-to–hole diameter ratio of 100:1 and a collimator length of 2.98 mm produced a beam with a geometric unsharpness of 32 μm. Compared to a standard single-channel collimator with a 2 × 2-cm aperture, the single-channel collimator would need to be 660 cm long to produce an equivalent geometric sharpness. Yet because of its shorter length, the imaging beam intensity from our 2.98-mm-thick collimator array was approximately 50 times greater than that of an equivalent single-channel collimator.

Assessment of Cystamine's Radioprotective/Antioxidant Ability under High-Dose-Rate Irradiation: A Monte Carlo Multi-Track Chemistry Simulation Study.

(1) Background: cystamine and its reduced form cysteamine have radioprotective/antioxidant effects in vivo. In this study, we use an in vitro model system to examine the behavior of cystamine towards the reactive primary species produced during the radiolysis of the Fricke dosimeter under high dose-rate irradiation conditions. (2) Methods: our approach was to use the familiar radiolytic oxidation of ferrous to ferric ions as an indicator of the radioprotective/antioxidant capacity of cystamine. A Monte Carlo computer code was used to simulate the multi-track radiation-induced chemistry of aerated and deaerated Fricke-cystamine solutions as a function of dose rate while covering a large range of cystamine concentrations. (3) Results: our simulations revealed that cystamine provides better protection at pulsed dose rates compared to conventional, low-dose-rate irradiations. Furthermore, our simulations confirmed the radical-capturing ability of cystamine, clearly indicating the strong antioxidant profile of this compound. (4) Conclusion: assuming that these findings can be transposable to cells and tissues at physiological pH, it is suggested that combining cystamine with FLASH-RT could be a promising approach to further enhance the therapeutic ratio of cancer cure.

Electrical deterioration of 4H-SiC MOS capacitors due to bulk and interface traps induced by proton irradiation

In this study, the displacement damage dose (DDD) effects of proton irradiation on bulk crystal and interfacial properties of 4H-SiC MOS capacitors (MOSCs) were investigated. The Monte Carlo calculation code named SRIM (Stopping and Range of Ions in Matter) was first used to predict the depth profiles and damage distributions of protons. The 4H-SiC MOSCs were irradiated with 60, 100, 150 keV protons to fluences ranging from 5 × 1011 to 5 × 1013 cm−2. Capacitance-voltage (C-V) curves of the MOSCs were then measured to characterize the variation of the electrically active defects in 4H-SiC MOSCs. The results revealed that the non-ionization energy loss of incident protons leads to the accumulation of the primary defects in bulk SiC crystal as well as the increase of the interface traps between SiO2 and SiC crystal. A distinctive bump phenomenon was found in the weak inversion region of the C-V curve and it became more prominent with the increase of proton fluence, providing evidence for the existence of bulk traps. Furthermore, the incident proton energy is also a crucial factor in interface deterioration, suggesting that more interface traps would be formed when the proton damage peak is closer to the SiO2/SiC interface.

Modeling of betavoltaic cells GaN using a Monte Carlo calculation

In this study, we use a Monte Carlo calculation code to simulate the concentration of electron-hole pairs generated of each point in the solid targets under a bombardment of Ni63 source for betavoltaic cells; this model is reported to be an accurate representation of electron interaction.From this simulation we can obtain the distribution of electron-hole pairs generated in GaN/GaN junction as a function of the depth, this distribution allowed us to find the concentrations of minority carriers excess depending on the thickness, which can represent as function and inject into the continuity equations for determine the diffusion current and then the characteristics of betavoltaic chosen. The model has been tested for Ni-63/GaN/GaN structure, with energy 17 KeV.

Simulations of X-ray spectra, half value layer, and mean energy from mammography using EGSnrc Monte Carlo and SpekPy

IntroductionThis study presents the simulation results of X-ray spectra, half value layers (HVLs), and mean energies (Emean) of two mammography units using EGSnrc Monte Carlo (MC) and SpekPy computer codes. MethodsThe spectra caused by different combinations of targets/filters at various tube voltages (kVps) of two mammography units were simulated using two different computer codes. The EGSnrc MC simulated data of spectra and Emean were compared with those obtained from SpekPy. The simulated values of two units’ HVLs obtained from two computer codes were compared with those from physical measurements from mammography machines used in clinical practice. ResultsThe maximum discrepancies in Emean simulated from two codes were less than 4.1% and 1.5% for the target/filter combination of W/Rh and Mo/Mo, respectively. The HVLs of the SpekPy were well matched to the physical measurements. The percentage differences were within 3.3% and 6.8% for two units, respectively. The EGSnrc MC simulated values of HVLs show the percentage differences within 8.9% and 7.0% with those from physical measurements for two units, respectively. All methods of HVLs determination comply with the requirements of IAEA Human Health Series No.17. ConclusionsThe HVLs, Emean, spectra varied depending on the target/filter combinations and composition of the mammography tubes. The simulation results verify that the HVLs evaluation using the EGSnrc MC and SpekPy can be validated for mammography standard beam qualities and provide prediction almost immediately compared with physical experiments. Implications for practiceEGSnrc MC and SpekPy have been considered powerful toolkits to simulate typical X-ray tubes used in mammography due to the good agreement between the calculation of Emean, physical measurements and simulated HVLs.

Monte Carlo Codes for Neutron Buildup Factors

The point-kernel method is a widely used practical tool for gamma-ray shielding calculations. However, application of that method for neutron transport simulations is very limited. The accuracy of the method strongly depends on the accuracy of buildup factors used in the calculations. Buildup factors are usually obtained using appropriate computer codes, either based on discrete ordinates transport method or Monte Carlo approach. Since these codes put strong demands on computer resources, they are applied on a limited number of shielding configurations and an attempt is made to use these results and formulate an empirical expression for buildup factors estimation. Due to high physical complexity of neutron transport through shielding material it is very hard to perform parameterisation in order to establish adequate empirical formula. Existing formulas are very limited and are usually applicable to a narrow neutron energy range for few commonly used shielding materials, mostly in monolayer configuration. Recently, a new approach has been proposed for determination of gamma ray buildup factors for mono-layer, as well as multi-layer shielding configurations covering a wide gamma ray energy range. The new regression model is based on support vector machines learning technique, which has theoretical background in statistical learning theory. Development of named regression model required a large number of experimental data obtained by Monte Carlo computer code. More than 7000 Monte Carlo runs were required. Due to physical complexity neutron transport is likely to require even more experimental data in order to generate a model of reasonable accuracy. Therefore, the choice of appropriate Monte Carlo code is a very important question. One has to take into account the accuracy as well as the time required for input preparation and running the code. What also has to be considered is the possibility of the code to be incorporated in an algorithm for automated generation of experimental data. In this paper three Monte Carlo codes are analysed, namely SCALE4.4 code package (SAS3 sequence), SCALE6.0 code package (MAVRIC sequence), and MCNP5. Two simple experimental setups based on a point isotropic source in spherical and slab-like shield are modelled, and the codes are examined on previously mentioned issues. The comparison results show that each one of the examined codes has potential to be used for neutron buildup factor model generation. However, some aspects of their utilization require further analysis prior to final selection.

An Intelligent Optimization Method for Preliminary Design of Lead-Bismuth Reactor Core Based on Kriging Surrogate Model

To meet the numerous application demands of lead-bismuth reactors, different design optimization tasks need to be conducted on these reactors based on the existing reactor core solutions. However, the design optimization of lead-bismuth reactors is a challenging task because it is a complex, multi-dimensional, and nonlinear constrained problem. To resolve these issues and improve the efficiency of design optimization, a new method, called the KSM-OLHS-SEUMRE method, based on the Kriging surrogate model (KSM), orthogonal Latin hypercube sampling (OLHS), and space exploration and unimodal region elimination (SEUMRE) algorithm is proposed in this study. Based on this method, a design optimization program of lead-bismuth reactors (DOPPLER-K) is developed, which realizes functions like sample point generation, optimization analysis, pre-post processing of reactor calculation, coupling of the Reactor Monte Carlo (RMC) calculation code and the Steady-state Thermal-hydraulic Analysis Code (STAC). Further, taking lead-bismuth reactors SPALLER-4 and URANUS as prototypes, the proposed intelligent optimization method for preliminary design of lead-bismuth reactor core is verified. The results show that this method can rapidly and accurately find the target scheme satisfying the optimization conditions, and it is three orders of magnitude faster than pure Monte Carlo calculation. Compared with the initial core scheme of URANUS, the optimization rates of fuel loading, total core mass, active zone volume, and total core volume are reduced by 10.8, 11.5, 18.1, and 17.1%, respectively. These results validate the feasibility and efficacy of the proposed method for design optimization of lead-bismuth reactor core.

Neutronic Analysis of the European Sodium Fast Reactor: Part II—Burnup Results

Abstract In the framework of the Horizon 2020 project ESFR-SMART (2017–2021), the European sodium fast reactor (ESFR) core was updated through a safety-related modification and optimization of the core design from the earlier FP7 CP-ESFR project (2009–2013). This study is dedicated to neutronic analyses of the improved ESFR core design. The conducted work is reported in two parts. Part I deals with the evaluation of the safety-related neutronic parameters of the fresh beginning-of-life (BOL) core carried out by eight organizations using both continuous energy Monte Carlo and deterministic computer codes. In addition to the neutronics characterization of the core, a special emphasis was put on the calibration and verification of the computational tools involved in the analyses. Part II is devoted to once-through and realistic batchwise burnup calculations aiming at the establishing of the equilibrium core state, which will later serve as a basis for detailed safety analyses.

Neutronic Analysis of the European Sodium Fast Reactor: Part I—Fresh Core Results

Abstract In the framework of the Horizon 2020 project ESFR-SMART (2017-2021), the European Sodium Fast Reactor (ESFR) core was updated through a safety-related modification and optimization of the core design from the earlier FP7 CP-ESFR project (2009-2013). This study is dedicated to neutronic analyses of the improved ESFR core design. The conducted work is reported in two parts. Part I deals with the evaluation of the safety-related neutronic parameters of the fresh Beginning-of-Life (BOL) core carried out by 8 organizations using both continuous energy Monte Carlo and deterministic computer codes. In addition to the neutronics characterization of the core, a special emphasis was put on the calibration and verification of the computational tools involved in the analyses. Part II is devoted to once-through and realistic batch-wise burnup calculations aiming at the establishing of the equilibrium core state, which will later serve as a basis for detailed safety analyses.

Critical analysis of light-water moderation UO2 and uranium-plutonium mixed oxide lattices by continuous-energy Monte Carlo calculation

ABSTRACT Critical analysis of 120 light-water moderation lattices composed of 2.6 wt% UO2 fuel rods or 3.0 wt%-PuO2 uranium-plutonium mixed oxide (MOX) fuel rods was performed using a continuous-energy Monte Carlo calculation code to validate the JENDL-4.0 nuclear data library. As a result, the average neutron multiplication factor (keff ) of the 35 UO2 lattices (hydrogen-to-heavy-metal atomic ratio (H/HM): 4.3–8.7) was 0.99930 with the uncertainty of 0.00005 (1 σ), and that of the 85 MOX lattices (H/HM: 12.0–27.6) was 0.99856 with the uncertainty of 0.00006 (1 σ). The keff values of the UO2 lattices partly showed a decreasing trend with H/HM, while those of the MOX lattices an increasing trend. The keff values of the MOX lattices presented an increasing trend with the measurement dates. Assuming that the trend was relating to the uncertainties in the neutron capture cross-section of 241Am, a simulation to modify the cross-section was performed by increasing the atomic number densities of 241Am in the MOX fuel. It indicated that the neutron-flux-average capture cross-sections of 241Am for the MOX lattices should be increased by 10% to 20% from those calculated by the cross-section in JENDL-4.0.

THE CONCENTRATION OF ELECTRON-HOLE PAIRS GENERATED BY THE INTERACTION OF ELECTRON WITH MATTER, CASE OF CdS USING MONTE CARLO SIMULATION

In this study, we use a Monte Carlo Calculation code to simulate the stopping range and the concentration of electron-hole pairs generated of each point in the solid targets under a bombardment, those electrons and holes generating a low electrical power source for many specialized applications. The model has been tested for CdS structure, under a bombardment of one electron with energy 50 - 150 KeV.

Continuous-energy Monte Carlo calculation analysis of light-water moderation critical experiments with aqueous solution samples of B, Rh, Cs, Nd, Sm, Eu, Gd, and Er

ABSTRACT In order to validate the neutron absorption cross sections in the JENDL-4.0 nuclear data library, critical analysis was performed using a continuous-energy Monte Carlo calculation code for the light-water moderation UO2 cores loaded with the aqueous solution samples of eight elements (B, Rh, Cs, Nd, Sm, Eu, Gd, and Er). The concentrations of the elements in the solution samples were changed in the experiments, and the criticality water levels were measured. The calculated keff values of the critical cores for the different concentrations of each element were plotted against the reactivity worth of the solution samples calculated based on the differences in the critical water levels. They were fitted by an approximate linear curve to obtain the slope which was regarded as a negative value of a fractional bias of the absorption cross section in the thermal and epi-thermal neutron energy regions. The obtained fractional biases indicated that the absorption cross sections of Sm, Eu, and Gd agreed with the actual values with the uncertainties of about 3%, and those of B, Rh, and Er overestimated them by 6 3%, 5 3%, and 6 4%, respectively.

Multi-objective optimization strategies for radiation shielding performance of BZBB glasses using Bi2O3: A FLUKA Monte Carlo code calculations

Highly efficient gamma shielding of BZBB glasses with various Bi2O3 concentrations in the xBi2O3-30B2O3-(65-x)ZnO-5BaO, (5 ≤ x ≤ 25 mol%) were evaluated for their radiation shielding properties. The μm results have been evaluated via the NISTXCOM database and simulated via FLUKA code. The simulation values by Monte Carlo code (FLUKA) have been shown to converge and correlate with XCOM values. Shielding properties like HVL, MFP, Zeff, EBF, and EABF values have been computed. The values observed that μm and Zeff increase with increasing Bi2O3. 47.09, 48.91, 50.73, 52.54 and 54.36 (cm2/g) are the μm values for BZBB1, BZBB2, BZBB3, BZBB4 and BZBB5 glasses at 15 keV. Also, the μm values decrease with increasing photon energy for all glass samples. Moreover, BZBB5 sample had the lowest HVL, MFP, EBF, and EABF values. That indicates to the addition of Bi2O3 improves the radiation shielding properties of BZBB glasses. The obtained results were compared with the most commonly shielding materials such as lead and concretes. It was concluded that the improved BZBB glasses with the addition of Bi2O3 showed excellent shielding properties comparing with shielding materials. These results could be highly beneficial for fields such as medical treatment facilities.

Neutronic performance of a thorium based mixed oxide fuel in a burner sodium-cooled fast reactor

The paper discusses the feasibility of using the Thorium Mixed-Oxide fuel with transuranic isotopes in a medium-sized Sodium-Cooled Fast Reactor core dedicated for actinide incineration with a capacity of 1000MWth. The core was designed based on the Advanced Burner Reactor 1000MWth studied in OECD/NEA SFR Benchmark. The SERPENT 2.1.31 Monte Carlo computer code with ENDF library was used to perform static criticality, isotopic depletion and safety parameters calculations for five consecutive fuel cycles. The key core parameters, eigenvalue, effective delayed neutron fraction, control rod worth, sodium void, doppler reactivity coefficients and isotopic inventories were investigated at the Beginning-of-Cycle (BOC) and End-of-Cycle (EOC). The results were compared with the medium MOX burner and large thorium core. Thanks to the internal U-233 breeding, the excess reactivity remained positive up to 700 days in comparison to 400 days for the reference core. The core has the ability to operate long campaigns, and at the same time, it is the efficient transuranic burner, with the conversion ratio ∼1. The penalty of the improved cycle was reduced Doppler effect and effective delayed neutron fraction. In comparison to the MOX reference core, sodium void effect was reduced.

Applying the continuous-energy Monte Carlo calculation code, MVP3, to analysis of kinetic parameters measured for light-water moderated UO2 and MOX cores of the TCA and EOLE critical facilities

ABSTRACT In order to evaluate the accuracies of the kinetics parameters predicted with the continuous-energy Monte Carlo calculation code, MVP3, core analysis was performed for light-water moderated UO2 and mixed oxide (MOX) fuel cores for which the ratios of effective delayed-neutron fraction (βeff ) to prompt-neutron life time (ℓ) – represented hereafter by βeff /ℓ – and βeff values were measured. The results obtained with the JENDL-4.0-based neutron library showed that (the calculation values (C)/the measurement values (E)−1) ranged from −1.0% to 4.0% for the βeff /ℓ values with an average of 1.0% and a standard deviation of 1.1% for the 17 cores (the UO2, and UO2-MOX mixed cores) tested in the Tank-Type Critical Assembly (TCA). With respect to the βeff of one UO2 core in TCA, C/E − 1 was 1.2%. For the βeff values of the UO2 and MOX cores tested in the EOLE critical facility, C/E − 1 was −1.5% for the former and −4.6% for the latter. The calculated βeff value of the MOX core using the JEFF-3.2-based neutron library was larger by 4% than that calculated using the JENDL-4.0-based neutron library and showed better agreement to the measurement.

Contribution of cadmium to the total amount of positron creation in a reactor-based slow positron beamline

In the slow positron beamline at the Kyoto University Research Reactor (KUR), positron creation was enhanced by increasing the gamma-ray intensity at the positron source via the reaction of 113Cd(n,γ)114Cd. To achieve this, a cadmium (Cd) cap was attached to the positron source, surrounding it, and thus, without intentional cooling, the temperature was able to reach near the melting point of Cd via nuclear heating. In this study, the degree to which the Cd cap contributes to the quantity of positron creation was estimated by using the Monte Carlo calculation code PHITS (Particle and Heavy Ion Transport code System), which simulates radiation transportation and interaction with matter. As a result, the number of positrons created was found to become 2.0±0.1 times higher by using the Cd cap at the KUR slow positron beamline.The use of the Cd cap was confirmed to be significantly effective for enhancing positron creation.