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• Fuel Assembly bowing is the lateral change in the FA shapes in the reactor core. • The change in the distance between FAs caused by the bowing leads that affects power distribution. • This perturbation is time-dependent because the boric acid concentration change during the reactor operation cycle. • Neutron noise is used in the analysis and investigation of neutron perturbation time-dependent. • In this research, by using an adiabatic approximation in neutron noise theory has been made to investigate the influence of the bowing phenomenon of FAs. • The noise amplitude caused by the bowing during the cycle depends on the changes in the boric acid concentration, burnup, size, and direction of the FA bowing. Thermomechanical and radiation loads in the reactor core cause lateral changes in the shape of Fuel Assemblies (FAs). This phenomenon, known as FA bowing, has significant implications for reactor safety and operation. Bowing results in alterations to the spacing between FAs, which in turn leads to neutron flux perturbations affecting power distribution. These perturbations are time-dependent due to variations in boric acid concentration during the cycle. Neutron noise theory is employed to analyze these time-dependent neutron perturbations. In this research, it is investigated the bowing effect during the cycle in a VVER-1000 reactor using the neutron noise adiabatic approximation. It was initially considered a VVER-1000 reactor core to verify the calculations algorithm without bowed FAs. Results had acceptable accuracy in power, initial boric acid, and cycle length. Subsequently, it was explored various modes of FA bowing as neutron noise sources the results revealed that the power noise caused by bowing varies throughout the cycle, as expected. The noise amplitude during the cycle depends on changes in boric acid concentration, burnup, size, and the direction of bowing. [ABSTRACT FROM AUTHOR]

In this study, optimized preparation, quality control, cell assessments and biostribution of 188Re-HYNIC-PSMA in normal rats and tumor bearing mice are studied. Optimized conditions for radiolabeling were evaluated and radiochemical purity (> 99%) was investigated using ITLC and HPLC methods. 188Re-HYNIC-PSMA was stable both in PBS (4 °C) and in human serum (37 °C) even after 48 h. The results showed the complex was cleared from the blood very fast via urinary track. This new radiolabeled compound has a high potential to be considered as an agent for the treatment of patients with PSMA expressing tumors however more biological studies are still needed. [ABSTRACT FROM AUTHOR]

• Timely and correct identification of NPP transients could either prevent an accident or mitigate the consequences of an accident. • A hybrid deep learning technique is proposed for the online identification of nuclear power plant transients. • A CNN-LSTM network was used for the transient identification, which transformed the input matrices into vectors using a CNN and then LSTM layers predicted the probability of each transient by sequence processing. • The results were shown that the proposed technique performs better than common deep learning techniques in terms of accuracy, identification time, and computational cost. Identification of NPP transients plays an important role in the prevention of accidents and mitigation of their consequences. NPP parameters may follow different patterns during each transient. So the transients can be identified by monitoring the operating parameters. It has been shown in several studies that data-driven methods, especially deep learning approaches, have a desirable performance in NPP transient identification. A hybrid deep learning technique is proposed in the present paper, in which transient identification is done using a CNN-LSTM neural network. The training data set is taken from a VVER-1000 full-scope simulator and the most important operating parameters are determined by feature selection techniques. According to the results, the proposed technique has identified the NPP transients in a short time, with high accuracy, and with a reasonable computational cost. The effective performance of the technique makes it possible to use it as a practical tool for online transient identification. [ABSTRACT FROM AUTHOR]

A new scheme of single-phase transformerless grid-connected inverter is presented in this article. By employing the series–parallel switching conversion of the integrated switched-capacitor module in a packed unit, some attractive features for the proposed inverter can be obtained. Such promising features can be quantified as more than 97% overall efficiency for a wide range of output power, two-time boosting ability within a single-stage operation and almost total suppression of the leakage current through using the common grounding technique. In this case, a peak current controller strategy is used in order to trigger the gate of power switches and control both the active and reactive powers. Therefore, a tightly controlled current with an appropriate quality can be injected to the grid from a single source renewable energy resource. Operating procedure, design consideration, comparison study, and a 513 W built prototype results are also given to prove the correctness and feasibility of the proposed circuit. [ABSTRACT FROM AUTHOR]

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 new fundamental structure of a single-phase transformer-less grid connected multilevel inverter based on a switched-capacitor structure is presented in this study. By employing the series-parallel switching conversion of the integrated switched-capacitor module in a packed unit, attractive features for the proposed inverter can be obtained such as high efficiency and boosting ability within a single stage operation. Also, using a common grounding technique provides an additional advantage of reducing the leakage current. Moreover, the presented structure generates a multilevel waveform at the output voltage terminals which reduces the harmonics in the system. A peak current controller is utilized for triggering the gate of the power switches and controlling both the active and reactive powers. This results in a tightly controlled current with an appropriate quality that can be injected to the grid using a single source renewable energy resource. Operating procedures, design considerations, comparison studies and test results of a 620 W prototype are also presented to validate the accuracy and feasibility of the proposed multilevel inverter. [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]

Summary: This article proposes a new Zeta‐based switched‐capacitor (SC) dc‐dc converter, which has many advantages such as increased voltage gain, decreased duty‐cycle, lower voltage stress on components such as its capacitors and input switch, and increased output power over traditional dc‐dc converter structures. In traditional converters such as Zeta converter, there is only one coupling capacitor, which works as a medium for transferring the power between input and the output. However, in the proposed Zeta‐based converter, there are multiple coupling capacitors, which are used based on dc‐dc SC converter principles. By using these switched coupling capacitors, the mentioned advantages are obtained for the proposed structure, which in turn make this converter more applicable for industrial applications. The analysis has been validated by comprehensive and precise comparisons and experimental results. [ABSTRACT FROM AUTHOR]

This paper presents an interleaved converter that benefits the coupled inductor and built-in transformer voltage multiplier cell (VMC). Compared with the other converters with only a built-in transformer or only a coupled inductor, the combination of these techniques gives an extra degree of freedom to increase the voltage gain. The VMC is composed of the windings of the built-in transformer and coupled inductors, capacitors, and diodes. The voltage stress of MOSFETs is clamped at low values and can be controlled via the turns ratio of the built-in transformer and coupled inductor that increases the design flexibility. Moreover, the energy of the leakage inductances, is recycled to the clamp capacitors which avoids high voltage spikes across MOSFETs. In addition, the current falling rate of the diodes is controlled by the leakage inductances, and the reverse current recovery problem is alleviated. Meanwhile, due to the interleaved structure of the proposed converter, the input current ripple is minimized and the current stress of the power devices is decreased. All of these factors improve the efficiency of the proposed converter in high-current and high-voltage applications. The principle operation and steady-state analysis is given to explore the advantages of the proposed converter. Finally, a 1.3-kW prototype with 50–600 V voltage conversion is built to demonstrate the effectiveness of the proposed converter. [ABSTRACT FROM AUTHOR]

Abstract Analysis of gamma-ray spectra is an important step for identification and quantification of radionuclides in a sample. In this paper a new gamma-ray spectra analysis algorithm based on Particle Swarm Optimization (PSO) is developed to identify different isotopes of a mixed gamma-ray source and determine their fractional abundances. PSO is an iterative algorithm that imitates the social behaviors observed in nature to solve complex optimization problems. The PSO method is used for complex fitting to the response of a 3 × 3 inch NaI (Tl) scintillation detector and the fitting process is controlled by a test for significance of abundance and computation of Theil coefficient. To test the developed algorithm, a number of experimentally measured gamma-ray spectra related to a mixed gamma-ray source including different combinations of 60 Co , 137 Cs , 22 Na , 152 Eu and 241 Am isotopes are analyzed using information of whole spectrum. The performance of the developed PSO algorithm is compared to the multiple linear regression (MLR) method as well. The results of the developed PSO algorithm show a better match with the real fractional abundances than that of MLR method. Highlights • Introducing a novel method for gamma-ray spectrum analysis. • Whole spectrum analysis using particle swarm optimization (PSO). • Calculation of fractional abundances of isotopes of mixed gamma-ray sources. • Comparing results of the PSO method with those of multiple linear regression (MLR). • Better results of the PSO method in comparison with MLR method. [ABSTRACT FROM AUTHOR]

In this paper, the neutron noise based on transport theory and diffusion noise theory using Green’s function technique is calculated. As the neutron noise is used for core diagnostic, surveillance and monitoring, calculation of neutron noise precisely can play an important role in monitoring and safety. We compare the accuracy of Green’s function based on transport and diffusion theory in order to survey the differences between these theories. In this study some deviation between results obtained two theories are observed, and the impact of dimensions, cross sections and frequency on the results investigated. [ABSTRACT FROM AUTHOR]

In the present study, the development of a neutron noise simulator, DYN-ACNEM, using the Average Current Nodal Expansion Method (ACNEM) in 2-G, 2-D hexagonal geometries is reported. In first stage, the static neutron calculation is performed. The neutron/adjoint flux distribution and corresponding eigen-values are calculated using the algorithm developed based on power iteration method by considering the coarse meshes. The results of the static calculation are validated against the well-known IAEA-2D benchmark problem. In the second stage, the dynamic calculation is performed in the frequency domain in which the dimension of the variable space of the noise equations is lower than the time dependent equations. Induced neutron/adjoint noise distribution due to the neutron noise source of type absorber of variable strength is calculated as well. Two different methods are used to validate the neutron noise calculation. The Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (ILOFAIP) as an important neutron noise source is investigated using ACNEM in this study. [ABSTRACT FROM AUTHOR]

• To determine the most important input parameters for NPP transient identification, the present paper has proposed a hybrid feature selection method. • Several filter methods are used to generate appropriate feature subsets by calculating the score of each feature. • A Long Short-Term Memory (LSTM) deep neural network is selected as the classifier to determine the best feature subset. • The Neighbourhood Components Analysis filter method has selected the best subset of features. • The effect of the number of used features on the accuracy of the model was investigated. Accidents that occur at NPPs must be correctly identified so quickly that mitigation actions can be taken in a timely manner. Depending on the type of transient, the operating parameters follow different patterns and it might be possible to identify the transient by monitoring these parameters. Due to the large number of parameters of an NPP, it is necessary to determine the parameters that play a vital role in transient identification. Data-driven methods have shown effective performance for NPP transient identification. To determine the most important input parameters for NPP transient identification, the present paper has utilized a hybrid feature selection method, in which feature subsets are created using several filter methods and then the best feature subset is determined by comparing the training results of a deep Long Short-Term Memory network. According to the results, the Neighbourhood Components Analysis method has selected the best subset of features. [ABSTRACT FROM AUTHOR]

• Model development to calculate the bowing effects of the FAs on power in VVER-1000. • The model is developed using the DRAGON and PARCS codes. • To verify the model, the bowed fuel geometry was simulated by MCNPX-2.7 code. • Power asymmetry depends on boric acid concentration, bowing size and direction. The Lateral deformation of Fuel Assembly (FA) under the operational conditions of the reactor cores is called FA bowing. This phenomenon is caused by factors such as thermal and hydraulic loads on FAs in the reactor core. It can lead to disturbances in the movement of control rods inside of FAs, cross-contact of FAs in refueling, and also changes in power distribution. Changing the distance between the fuels along the assemblies due to bowing, leads to non-uniform distribution of water (coolant) around the FAs and results in neutronic perturbation. In this research, by developing a calculation model for the bowed FAs of the VVER-1000 reactor, based on the distribution of water around FA at different heights. Macroscopic cross-sections are calculated by using the DRAGON cell calculation code and then the effect of bowing on the power distribution is calculated by using the PARCS core calculation code. To verify the model, the bowed FA geometry was simulated in MCNPX-2.7 Monte Carlo code. Results of DRAGON and MCNPX show the relative difference in all macroscopic cross-sections is less than 5%, except for scattering cross-sections (Σs,2), which is about 9%. Comparing the results with each other shows that the model to simulate the bowing effect of VVER-1000 reactor fuel has acceptable accuracy. Besides, for the central FA with C-shape bowing, the relative differences between the results of PARCS and MCNPX for thermal and fast flux are respectively less than 2% and 4%. Also, the effect of FA bowing in different states has been investigated. At last, results show the magnitude of power asymmetry depends on the size of the deformation of the FAs, bowing direction, and boric acid concentration. In addition, the effect of FAs bowing on power distribution and its resulting asymmetry is different during the cycle according to the concentration of boric acid. [ABSTRACT FROM AUTHOR]

This paper presents an interleaved nonisolated dc–dc converter with high-voltage gain and zero-voltage switching (ZVS) performance. Both coupled inductor and voltage multiplier cell techniques are used to increase the voltage gain. The ZVS circuit is composed of an active clamp which is in series with the output filter capacitors. This will give rise to further extension of the voltage gain. Applying the interleaving technique at the input of the converter, the ripple of the input current is reduced. Due to the leakage inductances of coupled inductors, the diodes are turned-off under zero-current switching condition. Hence, the reverse current recovery problem is alleviated. The steady-state analysis of the proposed converter is also presented. Finally, a 900-V to 415-W laboratory prototype is implemented to validate the performance of the proposed converter. [ABSTRACT FROM PUBLISHER]

A novel method is introduced which allows the higher order perturbative solution of a linear operator with a time variant coefficient. This method employs a form of raising and lowering operators which generate higher and lower order harmonics from a given harmonic. The analysis has been applied to the point kinetics equations with a monotone oscillatory reactivity to place bounds on the relative error of the linearization method frequently employed in the power reactor noise techniques. As a result, the maximum permissible reactivity amplitude for a given reactor as a function of frequency has been obtained such that regular power reactor noise methods remain accurate enough. Three benchmarks in subcritical and critical configurations show the accuracy of this method. [ABSTRACT FROM AUTHOR]

In the present study, the neutron noise, i.e. the stationary fluctuation of the neutron flux around its mean value is calculated based on the 2G, 3D neutron diffusion theory. To this end, the static neutron calculation is performed at the first stage. The spatial discretization of the neutron diffusion equation is performed based on linear approximation of Galerkin Finite Element Method (GFEM) using unstructured tetrahedron elements. Using power iteration method, neutron flux and corresponding eigen-value are obtained. The results are then benchmarked against the valid results for VVER-1000 (3D) benchmark problem. In the second stage, the neutron noise equation is solved using GFEM and Green’s function method for the absorber of variable strength noise source. Two procedures are used to validate the performed neutron noise calculation. The calculated neutron noise distributions are displayed in the different axial layers in the reactor core and its variation in axial direction is investigated. The main novelty of the present paper is the solution of the neutron noise equations in the three dimensional geometries using unstructured tetrahedron elements. [ABSTRACT FROM AUTHOR]

In this study, a new type of cascaded modular multilevel inverters (CMMLIs) is presented which is able to produce a considerable number of output voltage levels with a reasonable number of components. Accordingly, each series stage of the proposed CMMLI is comprised of two same basic units that are connected with each other through two unidirectional power switches without aiming any of the full H-bridge cells. In addition, since the potentiality for generating a higher number of output voltage levels in CMMLIs hinges on the magnitude of the dc voltage sources used in each series unit, in the rest of this paper, four different algorithms for determining an appropriate value for the dc sources' magnitude are also presented. In the following, a comprehensive topological analysis between some CMMLI structures reported in the literature and proposed structure along with several simulation and experimental results will be also given to validate the lucrative benefits and viability of the proposed topology. [ABSTRACT FROM AUTHOR]

Mesh point positions in Finite Difference Method (FDM) of discretization for the neutron diffusion equation can remarkably affect the averaged neutron fluxes as well as the effective multiplication factor. In this study, by aid of improving the mesh point positions, an enhanced finite difference scheme for the neutron diffusion equation is proposed based on the neutron importance function. In order to determine the neutron importance function, the adjoint (backward) neutron diffusion calculations are performed in the same procedure as for the forward calculations. Considering the neutron importance function, the mesh points can be improved through the entire reactor core. Accordingly, in regions with greater neutron importance, density of mesh elements is higher than that in regions with less importance. The forward calculations are then performed for both of the uniform and improved non-uniform mesh point distributions and the results (the neutron fluxes along with the corresponding eigenvalues) for the two cases are compared with each other. The results are benchmarked against the reference values (with fine meshes) for Kang and Rod Bundle BWR benchmark problems. These benchmark cases revealed that the improved non-uniform mesh point distribution is highly efficient. [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 aim of this paper is to present a new structure for switched-capacitor multilevel inverters (SCMLIs) which can generate a great number of voltage levels with optimum number of components for both symmetric and asymmetric values of dc-voltage sources. The proposed topology consists of a new switched-capacitor dc/dc converter (SCC) that has boost ability and can charge capacitors as self-balancing by using the proposed binary asymmetrical algorithm and series–parallel conversion of power supply. The proposed SCC unit is used in new configuration as a submultilevel inverter (SMLI) and then, these proposed SMLIs are cascaded together and create a new cascaded multilevel inverter (MLI) topology that is able to increase the number of output voltage levels remarkably without using any full H-bridge cell and also can pass the reverse current for inductive loads. In this case, two half-bridge modules besides two additional switches are employed in each of SMLI units instead of using a full H-bridge cell that contribute to reduce the number of involved components in the current path, value of blocked voltage, the variety of isolated dc-voltage sources, and as a result, the overall cost by less number of switches in comparison with other presented topologies. The validity of the proposed SCMLI has been carried out by several simulation and experimental results. [ABSTRACT FROM PUBLISHER]

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]

In normal operation conditions of nuclear power plants, the distribution of primary coolant between fuel channels would be considered almost uniform. When different number of Reactor Circulation Pumps (RCPs) are switched off, known as an abnormal condition, this uniform distribution is disturbed and different conditions occur for each channel depending on its position in the core. In this research, the normal and abnormal condition (with one or two tripped RCPs) for a VVER-1000/446 is investigated. For evaluation of the core neutronic calculations and thermal power distribution, USNRC's PARCS system code is employed. Then a thermal-hydraulics module was developed for performing the T/H calculation of the core zone. The input velocity of each channel in abnormal condition was calculated based on developed CFD model in downcomer and lower plenum of Reactor Pressure Vessel (RPV) by ANSYS-CFX. The results show that, in normal operation, the hot channel is related to the central fuel assembly of the reactor core with the highest relative power equal to 1.29 and total power of 23.74 MW. In this case, the minimum inlet velocity, the maximum coolant outlet velocity, and the maximum fuel temperature are 5.6 (m/s), 330.96 (°C), and 1345.8 (°C), respectively. In the cases of operation with one and two tripped RCPs, the hot channel is related to the fuel assemblies with the lowest inlet velocities. The lowest velocities are 0.32 m/s, 0.24 m/s, and 0.22 m/s respectively for the condition with one tripped pump, two tripped pumps placed oppositely, and two tripped pumps placed contiguously. The hot channel numbers in these cases are 158, 102, and 90, respectively. In these channel, the condition of outlet flow would be superheated, but the fuel temperature (1006.3, 1050.9, and 987.9) do not reach the maximum allowable margin. The study confirms the necessity of the coolant distribution consideration in OLCs as well as events that may disturb the symmetry of the coolant flow. It also showed that the lateral fuel assemblies are more at risk in this situation because of a significant reduction in coolant flow. Likewise, the investigations proved the safe continuation of the operation in PLOFA conditions with the preventive algorithm of the emergency protection system and without any need for immediate mitigation actions or operator intervention. [Display omitted] [ABSTRACT FROM AUTHOR]

Neutron noise induced by propagating thermal-hydraulic disturbances (propagation noise for short) in pressurized water reactors is investigated in this paper. A closed-loop neutron-kinetic/thermal-hydraulic noise simulator (named NOISIM) has been developed, with the capability of modeling the propagation noise in both Western-type and VVER-type pressurized water reactors. The neutron-kinetic/thermal-hydraulic noise equations are on the basis of the first-order perturbation theory. The spatial discretization among the neutron-kinetic noise equations is based on the box-scheme finite difference method (BSFDM) for rectangular-z, triangular-z and hexagonal-z geometries. Furthermore, the finite volume method (FVM) has been used for the discretization of the thermal-hydraulic noise equations, which comprises the fluctuations of all the coolant parameters, as well as the radial distribution of the temperature fluctuations in the fuel, gap and cladding. Based on the discretized governing equations and the proposed solving procedure, the closed-loop noise calculations have been performed (using NOISIM) for the case study of a typical VVER-1000 reactor core and the numerical results are presented and analyzed for different scenarios. The noise sources include the inlet coolant temperature and velocity fluctuations, in addition to the power density noises. Moreover, the space- and frequency-dependence of the propagation noise are studied in this paper. [ABSTRACT FROM AUTHOR]

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304–309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137 Cs source is also compared with the experimental data. [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]

Analysis of thermal–hydraulics fluctuations in pressurized water reactors (e.g., local and global temperature or density fluctuations, as well as primary and charging pumps fluctuations) has various applications in calculation or measurement of the core dynamical parameters (temperature or density reactivity coefficients) in addition to thermal–hydraulics surveillance and diagnostics. In this paper, the thermal–hydraulics fluctuations in PWRs are investigated. At first, the single-phase thermal–hydraulics noise equations (in the frequency domain) are originally derived, without any simplifying assumptions. The fluctuations of all the coolant parameters, as well as the radial distribution of the temperature fluctuations in the fuel, gap and cladding are taken into account. Then, the derived governing equations are discretized using the finite volume method (FVM). Based on the discretized equations and the proposed algorithm of solving, a single heated channel noise calculation code (SHC-Noise) is developed, by which the steady-state and fluctuating parameters of PWR fuel assemblies can be calculated. The noise sources include the inlet coolant temperature and velocity fluctuations, in addition to the power density noises. The developed SHC-Noise code is benchmarked in different cases and scenarios. Furthermore, to show the effects of the power feedbacks, the closed-loop calculations are performed by means of the point kinetics noise theory. Both the space- and frequency-dependence of the temperature fluctuations are analyzed in this work. [ABSTRACT FROM AUTHOR]

A space dependent model for the second moment of a stochastic field has been presented. The derivation procedure is simple and avoids non-physical assumptions for completeness. Moreover, a program has been developed which obtains the solutions to this equation in the three dimensional space. Two benchmarks, namely the fissile cube and the subcritical slab, show the accuracy of the results obtained using this method. [ABSTRACT FROM AUTHOR]

Neutron noise induced by propagating disturbances in VVER-type reactor core is addressed in this paper. The spatial discretization of the governing equations is based on the box-scheme finite difference method for triangular- z geometry. Using the derived equations, a 3-D 2-group neutron noise simulator (called TRIDYN-3) is developed for hexagonal-structured reactor core, by which the discrete form of both the forward and adjoint reactor dynamic transfer functions (in the frequency domain) can be calculated. In addition, both types of noise sources, namely point-like and traveling perturbations, can be modeled by TRIDYN-3. The results are then benchmarked in different cases. Considering the noise source as propagating perturbations of the macroscopic absorption cross sections, the induced neutron noise is calculated throughout the reactor core. For the first time, adjoint approach is applied and examined for modeling moving noise sources. Moreover, the space- and frequency-dependence of the propagation noise are investigated in this paper. [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: The present paper consists of two separate sections. In the first section, the neutron noise source is reconstructed using Artificial Neural Network (ANN) in a typical VVER-1000 reactor core. In the first stage of this section, the neutron noise calculations are performed based on Galerkin Finite Element Method (GFEM). To this end, two types of noise sources including absorber of variable strength and vibrating absorber are considered. As the results of noise calculations, the neutron noise is obtained in the location of detectors. In the second stage, the multilayer perception neural network is developed for reconstruction of the noise source. Complex neutron noises (real and imaginary parts) in the location of detectors are considered as the inputs of ANN. The developed artificial neural network consists of two hidden layers of type hyperbolic tangent sigmoid transfer function and a linear transfer function in the output layer. Noise source characteristics including strength, frequency and the location (X and Y coordinates) are identified with high accuracy. The developed hybrid method which comprises scanning method and multilayer perception neural network is employed for reconstruction of two coincidence noise sources. The number, type and location of noise source are exactly reconstructed using the hybrid method. The strength and frequency of noise source(s) are also identified with high accuracy using the developed method. A sensitivity analysis of the reconstructed noise source to some ANN parameters like the number of hidden layers, neurons in each hidden layer and the applied transfer functions is performed. Variation of accuracy of reconstructed noise source versus number of detectors and their arrangement in the reactor core are investigated as well. In the second section of present work, neutron flux variations (neutron noise) due to absorber of variable strength and vibrating absorber noise sources are studied in both frequency and time domains. Time dependency of neutron noise was obtained using inverse Fourier transform. [Copyright &y& Elsevier]

Highlights: [•] The properties of Poisson distribution are exploited. [•] An Ito stochastic differential equation is obtained for a general stochastic system. [•] The formulation is applied to the point reactor. [•] Numerical results show the simplicity and accuracy of this method. [Copyright &y& Elsevier]

Highlights: [•] Scaling factors are calculated for the beginning of cycle (BOC) of VVER-1000 reactor using the direct and adjoint methods. [•] Space and energy dependent power density in core region is obtained from MCNP code. [•] Linearity assumption of relative detector response for small CBM is investigated. [•] It is shown that the direct method result is reliable for determining the CBM. [Copyright &y& Elsevier]

Highlights: [•] Lagrange method is proposed for on-line reactivity calculation in nuclear reactors. [•] The need for nuclear power history or the Laplace transform is vanished. [•] The three- and five-point formulas are presented and examined in different benchmark cases. [•] Computational time-steps of up to 1s lead to highly reliable reactivity calculations. [•] The main advantage of the proposed approach is its stability and convergence in large time-step calculations. [Copyright &y& Elsevier]

Highlights: [•] A sensitivity analysis of unfolding methods to different parameters is done. [•] ISM is proposed to reconstruct the noise source of type vibrating absorber. [•] Four algorithms are used to solve the system of equations in the inversion method. [•] The neutron noise due to a new defined noise source (ILFAIP) is calculated. [•] We find that the scanning method is a reliable algorithm for source reconstruction. [Copyright &y& Elsevier]

Abstract: Various methods for solving the forward/adjoint equation in hexagonal and rectangular geometries are known in the literatures. In this paper, the solution of multigroup forward/adjoint equation using Finite Element Method (FEM) for hexagonal and rectangular reactor cores is reported. The spatial discretization of equations is based on Galerkin FEM (GFEM) using unstructured triangle elements. Calculations are performed for both linear and quadratic approximations of the shape function; based on which results are compared. Using power iteration method for the forward and adjoint calculations, the forward and adjoint fluxes with the corresponding eigenvalues are obtained. The results are then benchmarked against the valid results for IAEA-2D, BIBLIS-2D and IAEA-PWR benchmark problems. Convergence rate of GFEM in linear and quadratic approximations of the shape function are calculated and results are quantitatively compared. A sensitivity analysis of the calculations to the number and arrangement of elements has been performed. [Copyright &y& Elsevier]

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: In the present study, the neutron noise, i.e. the stationary fluctuation of the neutron flux around its mean value, is calculated in 2-group forward and adjoint diffusion theory for both hexagonal and rectangular reactor cores. To this end, the static neutron calculation is performed at the first stage. The spatial discretization of equations is based on linear approximation of Galerkin Finite Element Method (GFEM) using unstructured triangle elements. Using power iteration method, forward and adjoint fluxes with the corresponding eigenvalues are obtained. The results are then benchmarked against the valid results for BIBLIS-2D and IAEA-2D benchmark problems and DONJON computer code. The dynamic calculations are performed in the frequency domain which leads to reducing the dimension of the variable space of the noise equations. The forward/adjoint noise in two energy group is obtained by assuming the neutron noise source as an absorber of variable strength type. The neutron noise induced by a vibrating absorber type of noise source is also obtained from the calculated adjoint Green''s function. Comparison of the calculated neutron noise at zero frequency with the results of static calculation, and utilizing the results of adjoint noise calculations are two different procedures to validate the neutron noise calculations. [Copyright &y& Elsevier]

Abstract: An improvement to the Direct Discrete Method (DDM), also known as the Cell Method, has been discussed. The improvement is based on a duality theorem between the primal and dual complexes. Also, the analog counterpart of the Integral operator has been derived in this paper. The multi-group neutron diffusion is then derived, directly in a discrete algebraic form, according to this procedure. A numerical example has shown that this method would yield a high order of convergence (approximately 4.6) if its parameters are adjusted suitably. Finally, the method is applied to the 2D IAEA benchmark problem, and has shown to yield accurate solutions with a reasonably low number of unknowns. [Copyright &y& Elsevier]

Abstract: The most general basis for approximating the transport equation has been studied. The application of the Gram–Schmidt procedure has been shown to unify the complete class of functions (polynomial type or non-polynomial type), applicable to this equation. The completeness of the series of functions is proved. A generalized version of the Fick''s law is introduced. It is shown that the spectrum of the transport equation obtained by this method agrees with the conventional methods of obtaining the spectrum. [Copyright &y& Elsevier]

Abstract: In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C# language to find an order of the fuel loading pattern of a VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process launches by considering an initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. The results on a typical VVER-1000 reactor reveal that the method could reach to a pattern with an allowed radial power peaking factor and increases the cycle length 1.1days, as well. [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: Effective delayed neutron fraction β eff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetics parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20°C) and hot (fuel, clad and coolant temperature 65, 49 and 44°C, respectively) states using MTR_PC computer code. The ratio of (β eff ) i /(β eff )core plays an important role in reactivity accident analysis codes. This parameter and its contribution to effective delayed neutron fraction from each nucleus have been calculated in cold and hot reactor states. Uncertainty of effective delayed neutron fraction is evaluated in terms of following four quantities; basic delayed neutron constants, delayed neutron spectra, energy dependence of delayed neutron yield (ν d ) and fission cross-section of 235U and 238U. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. The calculated and measured values are in good agreement. Relative Percent Errors (RPEs) are 2.8% for β eff and 5.7% for Λ in the cold state. [ABSTRACT FROM AUTHOR]

Abstract: In this paper, the development of a neutron noise simulator for hexagonal-structured reactor cores using both the forward and the adjoint methods is reported. The spatial discretisation of both 2-D 2-group static and dynamic equations is based on a developed box-scheme finite difference method for hexagonal mesh boxes. Using the power iteration method for the static calculations, the 2-group neutron flux and its adjoint with the corresponding eigenvalues are obtained by the developed static simulator. The results are then benchmarked against the well-known CITATION computer code. The dynamic calculations are performed in the frequency domain which leads to discarding of the time discretisation. Then, the developed 2-D 2-group neutron noise simulator calculates both the discretised forward and the adjoint reactor transfer function between a point source and its induced neutron noise, by assuming the neutron noise source as an “absorber of variable strength” type. The neutron noise induced by a “vibrating absorber” type of noise source may also be modeled using the calculated transfer function. The viability of the simulator is verified for different benchmark cases. [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]

The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss'' points. The accuracy of the method was examined with a well-known IAEA two-dimensional benchmark problem. The numerical results demonstrate the effectiveness of the new method. [Copyright &y& Elsevier]

• Transport Green function data is used for localization of a noise source. • Comparative study between inverse Green's function based on SN transport and diffusion is performed. • The impact of more accurate Green's function on the validation of the source localization is investigated. • The sensitivity of the localization results to the position and the number of detectors is investigated. In an attempt to explore the significance of transport theory in neutron noise, localization of a noise source by Green's function based on transport theory is investigated. There are considerable differences between Green's functions based on diffusion and transport, such as small dimensions, near edges, high heterogeneity medium and high-frequency source of perturbation. These differences are expected to significantly impact unfolding, reconstruction, and identification of the neutron noise source. Improvement in noise source unfolding methods is essential in terms of safety aspects and reactor performance enhancement. Since gaining the ability to monitor nuclear reactor based on noise diagnostics principle using few numbers of detectors is a challenging problem, more accurate methods of calculation of neutron noise can help reduce the detectors in number and sensitivity to location. This study revealed that inverse Green's function based on diffusion and transport is fundamentally different in concept. The inverse Green's function based on diffusion theory requires applying complicated reconstruction methods to gain more accurate results. This means that a large number of detectors must be used for noise source reconstruction. In contrast, in the case of inverse Green's function based on transport theory, each mesh's shape function is strongly correlated with the adjacent mesh. As a result, simple methods can be used to reconstruct the noise source. Therefore, the noise source can be identified and reconstructed with a smaller number of detectors in practical terms. [ABSTRACT FROM AUTHOR]