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Elemental analysis can be performed using obtained gamma-ray spectrum of the sample under study. In this work, simple Multi-Layer Perceptron (MLP) neural network models are proposed for analyzing a gamma-ray emitting sample using whole information of its obtained gamma-ray spectrum. Elemental analysis is performed in two fields of study using 3 × 3 inch NaI(Tl) detectors: Radio-Isotope Identification (RIID) and Prompt Gamma Neutron Activation Analysis (PGNAA). The gamma-ray point sources are used for an empirical study in RIID field, while a Monte Carlo simulation study is considered for determining chlorine and water content of crude oil using combination of PGNAA technique and a MLP model. According to the obtained results of both empirical and simulation studies, the proposed ANN models are appropriate for elemental analysis using whole gamma-ray spectral information of sample under study. [ABSTRACT FROM AUTHOR]

• Geant4 simulation of proton induced prompt gamma spectra of most abundance elements of human tissues. • Single peak analysis is applied on prompt gamma spectra for determining mass of elements in irradiated phantoms. • A new approach is proposed to deal with similarly found gamma-ray peaks related to multi elements. • The estimated mass of elements are in good agreement with predefined values. • The effect of estimated mass errors on calculated proton range is small. Online elemental analysis of irradiated tissues is an important challenge in proton therapy for estimating the depth of the Bragg peak and monitoring the healing process. In the present study, the prompt gamma (PG) spectra of the most abundant elements of human tissues were obtained for two different incident protons of 30 MeV and 120 MeV using the Geant4 toolkit. Train phantoms were defined to study the elemental analysis of the tissue from PG spectra. It was found that the intensities of 7.12 and 3.91 MeV PG depend on the mass of 16O and 20Ca elements for both proton energies, respectively. Furthermore, multi-variable linear functions were presented for overlapping peaks like 4.44 and 2.31 MeV PG lines. The results show that the relative errors of estimated masses relative to predefined values are small and the corresponding errors in the proton range are less than 4.5% for test phantoms. [ABSTRACT FROM AUTHOR]

• A novel sequential algorithm is developed for analyzing γ-ray spectra. • Whole spectral information are used to identify and quantify isotopes of a sample. • A sequence of spectra are produced for estimating a measured spectrum appropriately. • Both empirical and simulation studies are performed for testing the algorithm. • The identification procedure of developed algorithm is superior to the NNLS. A novel approach for analyzing complex gamma-ray spectra using a sequential algorithm is introduced. The developed Sequential Gamma-ray Spectrum Deconvolution (SGSD) algorithm produces a sequence of spectra converging to the best estimation of output spectrum of a gamma-ray detector. In each point of sequence, an isotope of unknown gamma-ray source is identified and the respective response of the detector to unknown source is reconstructed. Effectiveness of the developed algorithm is demonstrated by two empirical and simulation studies. In the case of empirical study, a number of recorded gamma-ray spectra related to a mixed gamma-ray source including different combinations of 5 isotopes (Co-60, Cs-137, Na-22, Eu-152 and Am-241) are analyzed using whole information of spectra. Furthermore, a number of simulated gamma-ray spectra related to a mixed gamma-ray source including different combinations of 30 isotopes are analyzed in simulation study. Both man-made and natural radioisotopes like Ba-133, Co-60, Ir-192, Cs-137, K-40, Th-232 series, U-238 series, Ac-227 series, etc. are used for Monte Carlo simulations. The numerical results of the SGSD algorithm are compared with those of the conventional Non-Negative Least Squares (NNLS) algorithm. Based on the results, the identification procedure of the SGSD algorithm has a remarkable superiority over the NNLS algorithm. [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]

Radiation transport techniques which are used in radiation detection systems comprise one of two categories namely probabilistic and deterministic. However, probabilistic methods are typically used in pulse height distribution simulation by recreating the behavior of each individual particle, the deterministic approach, which approximates the macroscopic behavior of particles by solution of Boltzmann transport equation, is being developed because of its potential advantages in computational efficiency for complex radiation detection problems. In current work linear transport equation is solved using two methods including collided components of the scalar flux algorithm which is applied by iterating on the scattering source and ANISN deterministic computer code. This approach is presented in one dimension with anisotropic scattering orders up to P8 and angular quadrature orders up to S16. Also, multi-group gamma cross-section library required for this numerical transport simulation is generated in a discrete appropriate form. Finally, photon pulse height distributions are indirectly calculated by deterministic methods that approvingly compare with those from Monte Carlo based codes namely MCNPX and FLUKA. [ABSTRACT FROM AUTHOR]

Many of the physical macroscopic quantities could be explained as the result of a collection of microscopic particles which act independent of each-other, in a linear fashion. Since the physical laws of the interaction of these particles with their surrounding medium are non-deterministic, one could think of these particles as the generators of a linear stochastic field. In this paper, we have introduced a derivation which has yielded an equation for the spatiotemporal correlations in such a field. The derivation is simple and extendable to include the behavior of many physical particles. A simple numerical algorithm has been devised to solve the obtained initial value integrodifferential equation and a computer program has been developed accordingly. The temporal correlations in four benchmark problems, i.e. a three dimensional one group fissile cube, two multi-region one group slabs and a two-group multi-region slab, have been obtained and compared with direct Monte Carlo simulations which show the accuracy of the method. Finally, these benchmarks show some possible applications of this formulation, e.g. generalizing the Rossi-alpha formula and extraction of the average neutron velocity in a steady-state reactor core. [ABSTRACT FROM AUTHOR]

Evaluation of multigroup constants in reactor calculations depends on several parameters, the Dancoff factor amid them is used for calculation of the resonance integral as well as flux depression in the resonance region in the heterogeneous systems. This paper focuses on the computer program (MCDAN-3D) developed for calculation of the multigroup black and gray Dancoff factor in three dimensional geometry based on Monte Carlo and escape probability methods. The developed program is capable to calculate the Dancoff factor for an arbitrary arrangement of fuel rods with different cylindrical fuel dimensions and control rods with various lengths inserted in the reactor core. The initiative calculates the black and gray Dancoff factor versus generated neutron flux in cosine and constant shapes in axial fuel direction. The effects of clad and moderator are followed by studying of Dancoff factor’s sensitivity with variation of fuel arrangements and neutron’s energy group for CANDU37 and VVER1000 fuel assemblies. MCDAN-3D outcomes poses excellent agreement with the MCNPX code. The calculated Dancoff factors are then used for cell criticality calculations by the WIMS code. [ABSTRACT FROM AUTHOR]

The evaluation of multigroup constants in reactor calculations depends on several parameters. One of these parameters is the Dancoff factor which is used for calculating the resonance integral and flux depression in the resonance region in heterogeneous systems. In the current paper, a computer program (MCDAN-3D) is developed for calculating three dimensional black and gray Dancoff coefficients, based on Monte Carlo, escape probability and neutron free flight methods. The developed program is capable to calculate the Dancoff factor for an arbitrary arrangement of fuel and moderator pebbles. Moreover this program can simulate fuels with homogeneous and heterogeneous compositions. It might generate the position of Triso particles in fuel pebbles randomly as well. It could calculate the black and gray Dancoff coefficients since fuel region might have different cross sections. Finally, the effects of clad and moderator are considered and the sensitivity of Dancoff factor with fuels arrangement variation, number of TRISO particles and neutron energy has been studied. [ABSTRACT FROM AUTHOR]

Abstract: In pebble bed reactors, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in these reactors consist of TRISO particles, which are embedded in a graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneous nature of this type of reactor. In this paper, a pebble bed reactor, the HTR-10, is used to demonstrate a treatment of this double heterogeneity using the MCNP5 Monte Carlo code and MATLAB programming. In this technique, TRISO particles are modeled in a pebble using the expanded FILL and LATTICE features of MCNP5. MATLAB is used to generate random numbers which represent the location of pebbles in the core. Centers of pebbles are generated stochastically and uniformly and then transferred into the MCNP5 input file as the centers of spherical surfaces. In this model, there is no approximation to the actual geometry. In other words, the double heterogeneous nature is preserved while truncating neither the pebbles in the core nor the particles in the pebble matrix. Finally, to validate the model, benchmark problems of IAEA are used. Very good agreement with experimental results is observed. [Copyright &y& Elsevier]

Abstract: Noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) have been simulated by MCNP computer code to calculate the prompt neutron decay constant (α 0), effective delayed neutron fraction (βeff ) and neutron generation time (Λ) in a subcritical condition for the first operating core configuration of Tehran Research Reactor (TRR). The reactor core is considered to be in zero power (reactor power is less than 1W) in the entire simulation process. The effect of some key parameters such as detector efficiency, detector position and its dead time on the results of simulation has been discussed as well. The results of proposed method in the current study are validated against both the experimental data and the results of MTR_PC computer code. [Copyright &y& Elsevier]

Abstract: Several parameters, one of which is the Dancoff Factor (DF), are used to calculate the resonance integral (RI) and reduced flux in the resonance region of heterogeneous systems as well as to accurately determine the group constants for criticality calculations. This paper is a report on the development of a program to calculate the DF correction factor using Monte Carlo method and collision probability definition in three-dimensional (3-D) geometries and with multi energy groups. Hence, the DF for any arbitrary arrangement of cylindrical and slab fuel elements is hereby calculated. The fuel elements are monitored and kept at equal levels, though different material compositions and formations are allowed rendering the materials either black or partially transparent. A separate investigation is carried out as to the effects of extension to 3-D geometry, energy group divisions, clad, coolant and moderator. The program is applied to calculate DF for slab fuels of a pool-type research reactor (PRR) containing 19 slab fuels and for cylindrical fuel element of CANFLEX fuel bundle with 43 cylindrical fuels elements. All calculations are performed in 3-D geometry and for six energy groups. The viability as well as the feasibility of the program is verified using the WIMSD computer code for the obtained 3-D and six-group DF for CANFLEX fuel bundle. [Copyright &y& Elsevier]

In this paper, a simpler approach compared to the existing approaches is developed to analyze nuclear reactor dynamics based on the explicit Monte Carlo method. A new population control method is also introduced to prevent neutron population growth and consequent computer memory shortages, which also increases simulation speed. The scheme is applied for time-dependent particle tracking in three-dimensional arbitrary geometries in the presence of feedbacks through a code named MCSP-Explicit. Changes in material density, as well as geometry dimensions, are also considered during simulation. MCSP-Explicit can be run with either continuous or multigroup data libraries, and it is further boosted by parallel processing to speed up simulations. A number of benchmark problems are studied at the end to evaluate the performance of the proposed approach in various situations. [ABSTRACT FROM AUTHOR]

One of the well-known online monitoring techniques used for quality control of bulk samples is Prompt Gamma Neutron Activation Analysis (PGNAA). PGNAA suffers from the so-called matrix effect problems such as density, thickness and moisture content of the sample under study. In this work, an Artificial Neural Network (ANN) model is introduced to deal with these effects. The required spectra for training and testing the proposed ANN model are obtained by Monte Carlo simulation of the gamma-ray spectra recorded in a PGNAA online analyzer system used in cement factories. The gamma-ray spectra related to given set of density, thickness and moisture content are corrected channel-to-channel using the proposed ANN model. The corrected spectra are then compared against the standard spectra obtained for predefined standard density, thickness and moisture content. The improvement in Theil coefficient is larger than 70% for the tested spectra. • The PGNAA online monitoring technique suffers from the so-called matrix effect problems. • A commercial PGNAA online analyzer used in cement factories is simulated to investigate the matrix effects and correct them. • Artificial neural network is used for correcting the matrix effects in the Monte Carlo simulated spectra of analyzer. • Investigated matrix effects are density, thickness and moisture content of cement sample. • Quantitative and qualitative results indicate that the proposed ANN method is accurate. [ABSTRACT FROM AUTHOR]

• Development of a Computer code for safety analysis of a nuclear reactor. • The Code works based on nodal discretization using MCNPX code. • The restrictions of previous methods in transient analysis have been removed. • The Code is capable to simulate the systems with time-varying geometry and cross sections. • Independency of the method of the code to the node size. This paper proposes a new method for solving the time-dependent neutron transport equation based on nodal discretization using the MCNPX code. Most valid nodal codes are based on the diffusion theory with differences in approximating the leakage term until now. However, the Monte Carlo (MC) method is able to estimate transport parameters without approximations usual in diffusion method. Therefore, improving the nodal approach via the MC techniques can substantially reduce the errors caused by diffusion approximations. In the proposed method, the reactor core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates and leakage ratio are estimated using MCNPX. They are then employed within the time-dependent neutron transport equation for each node independently to compute the neutron population. Based on this approach, a time-dependent code namely MCNP-NOD (MCNPX code based on a NODal discretization) was developed for solving time-dependent transport equation in an arbitrary geometry considering feed backs. The MCNP-NOD is able to simulate multi-group processes using appropriate libraries. Several test problems are examined to evaluate the method. [ABSTRACT FROM AUTHOR]

• The Code works based on Monte Carlo method using MCNPX code. • The restrictions of previous methods in transient analysis have been removed. • The Code is able to use either continuous or multi-group energy cross section libraries. • The Code is capable to simulate the systems with time-varying geometry and cross sections. • The Code is applied the effect of temperature on the cross section and density. Transient analysis which is vital in safety analysis requires a reliable calculation method. Most valid tools use diffusion theory with many approximations by now. However, the Monte Carlo method inherently overcomes these approximations and accurately calculates the parameters of a reactor. In this paper, a new time-dependent transport approach is described to simulate the nuclear reactor dynamic correctly using the MCNPX code. In this approach the fundamental parameters of a nuclear reactor like multiplication factor (K eff) and mean generation time (t G) are calculated using MCNPX code. They are then employed in the formulas to compute neutron population, proportional to K eff , during a generation time as well as precursors are decayed. Based on the approach, a dynamic Monte Carlo code namely DMCNP (Dynamical MCNPX code) is developed for a time-dependent simulation of particle tracking in an arbitrary geometry, considering the thermal-hydraulic feedbacks. The effects of temperature on cross section and density are applied at each time steps. Several test problems such as TWIGL, LMW, LRA and C5G7 are examined to assess the performance of the method. [ABSTRACT FROM AUTHOR]

A highly reliable tool for transient simulation is vital in the safety analysis of a nuclear reactor. Despite this fact most tools still use diffusion theory and point-kinetics that involve many approximation such as discretization in space, energy, angle and time. However, Monte Carlo method inherently overcomes these restrictions and provides an appropriate foundation to accurately calculate the parameters of a reactor. In this paper fundamental parameters like multiplication factor (K eff) and mean generation time (t G) are calculated using Monte Carlo method and then employed in transient analysis for computing the neutron population, proportional to K eff , during a generation time considering precursors decay. Based on this approach, a dynamic Monte Carlo code named MCSP (Monte Carlo dynamic Simulation of Particles tracking) is developed for both the steady state and time-dependent simulation of particle tracking in an arbitrary 3D geometry. MCSP is able to use either continuous or multi-group energy cross section libraries. To speed up the simulation, the MCSP was empowered with parallel processing as well. Several test problems such as C5G7, LMW and TWIGL are examined to assess the performance of the method. • Development of a 3D Computer code for safety analysis of a nuclear reactor. • The Code works based on Monte Carlo method. • The restrictions of previous methods in transient analysis have been removed. • The Code is able to use either continuous or multi-group energy cross section libraries. • The Code is capable to simulate the systems with time-varying geometry and cross sections. [ABSTRACT FROM AUTHOR]

Abstract: The time-dependent diffusion of neutrons in a spheroid as a function of the focal distance has been studied. The solution is based on an orthogonal basis and an extrapolation distanced related boundary condition for the spheroidal geometry. It has been shown that spheres and disks are two limiting cases for the spheroids, for which there is a smooth transition for the systems properties between these two limits. Furthermore, it is demonstrated that a slight deformation from a sphere does not affect the fundamental mode properties, to the first order. The calculations for both multiplying and non-multiplying media have been undertaken, showing good agreement with direct Monte Carlo simulations. [Copyright &y& Elsevier]

• Development of a computer code for safety analysis of a nuclear reactor in the presence of xenon oscillation. • The Code works based on nodal discretization using a MC code named MCSP. • The restrictions of previous methods in transient analysis have been removed. • The Code is capable to simulate the systems with time-varying geometry and cross sections. • Independency of the method of the code to the node size. In this paper a probabilistic methodology based on core nodalization is proposed to estimate the core power in the presence of xenon oscillation. A time-dependent Monte Carlo neutron transport code named MCSP-NOD is developed for dynamic analysis in arbitrary 3D geometries to simulate xenon oscillations as well as sub-critical condition with feedbacks. The new code is based on the approach adopted in MCNP-NOD which was previously introduced as a tool for core transient analysis using the MCNPX platform. As before, the core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates, leakage ratio are estimated using the MC techniques. However, as a new option the concentration of iodine and xenon are also estimated which enables us to predict the oscillatory behavior of reactor power following poison oscillation. These quantities are then employed within the time-dependent neutron transport equation for each node independently to compute the neutron population. Simulations prove the robustness of the method. [ABSTRACT FROM AUTHOR]