# A. Zolfaghari's research while affiliated with Shahid Beheshti University and other places

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## Publications (89)

Thermal-hydraulic parameters of the floating reactor fluctuate constantly, which can endanger the safety features and make challenges for safety systems. The rolling motion has more effect on the thermal-hydraulic of the floating reactor between other six motions. In this study, the effect of fuel shape and heated channel shape on thermal-hydraulic...

The simplified double-spherical harmonics ($SDP_N$) approximation of the neutron transport equation is proposed. The $SDP_N$ equations are derived from the multi-group $DP_N$ equations for N=1,2,3 (equivalent to the $SP_3$, $SP_5$, and $SP_7$ equations, respectively), and are converted into the form of second order multi-group diffusion equations....

After the Fukushima accident, to increase the safety of nuclear power plants, work began on Floating Nuclear Power Plants with the construction of the power plant floats above 100 m eliminate the risk of earthquakes and tsunamis. The study investigated integration of three-fluid model and Drift Flux model, to analyze different two-phase flow regime...

In this paper, a novel spatially adaptive Mesh free approach using local nodes and radial basis functions,
based on weak-form formulation is presented to solve the even-parity neutron transport equation. A
residual based error indicator is proposed and used in an adaptive scheme for error estimation. The local
balance of neutrons is used as an indi...

High-fidelity simulation tools are required to treat isotope separation cascades. In the present work, a multi-objective optimization method based on Whale Optimization Algorithm, WOA, is developed to obtain the optimal parameters of recycling cascades for separating the binary mixture of isotopes. Detail treatment and simulation of isotope separat...

High-fidelity simulation tools are required to treat nuclear reactor cores under ocean conditions. Majority of acceptance criteria for rigorous safety analysis are based on local values within the core. For the thermal-hydraulic analysis of a core, reliable fuel assembly sub channels codes are necessary. In this paper, a Drift-Flux model is extende...

In this work, the analysis of thermal–hydraulic boiling flow in a vertical channel under static along rolling and heaving motions is investigated by using the drift-flux model. A code called DAFNO, based on the staggered finite volume technique to discretize governing equations in the drift-flux model is developed; Newton method is implemented to s...

Common family of numerical methods such as finite element and finite difference involve the discretization of domain into regular grids or meshes. Mesh generation and re-meshing process, however, for obtaining required accuracy are complicated and time-consuming. Recently, Mesh Free (MFree) methods have been developed to overcome the drawbacks of s...

Since the separative power of a single centrifuge is relatively small, to achieve adequate enrichment and throughput, a large number of centrifuges are interconnected to form a cascade. An enrichment plant usually holds thousands of centrifuges. The pattern of connections is determined by the properties of the individual machines, the required quan...

Accident tolerant fuels (ATFs) are designed to mitigate damaging interaction of hot steam and fuel clad during off-normal condition. These materials can withstand against high temperatures for longer period of time. They have lower oxidation kinetics, higher resistance to core degradation and lower hydrogen pick-up. The main purpose of this study i...

The application of continuous and discontinuous approaches of the finite element method (FEM) to the neutron transport equation (NTE) has been investigated. A comparative algorithm for analyzing the capability of various types of numerical solutions to the NTE based on variational formulation and discontinuous finite element method (DFEM) has been...

Nowadays, the scarcity of energy resources and the intensive demand of energy leads the researchers and scientists to find the efficient ways to save the energy and decrease the exergy loss. In this study, a new exergy optimization framework for a typical WWER1000 nuclear power plant (Bushehr NPP) is suggested. It is aimed to optimize the total eff...

Loading pattern optimization (LPO) is a main issue of in-core fuel management. In this work, the Grey Wolf Optimizer (GWO) algorithm based on behavior of gray wolves for hunting is introduced for the multi-objective LPO of a WWER-1000 core. Besides the GWO, the Genetic Algorithm (GA) and Gravitational Search Algorithm (GSA) have been applied and th...

In this study, thermal-hydraulic parameters of a two phase flow in different channels of a same hydraulic diameter is studied. The Drift-Flux model is adopted for two phase flow investigations where four principal conservation equations are discretize phased using the mesh staggered finite volume technique. The boiling flow parameters at the sub-co...

In this part, an implicit time dependent solution is presented for the Boltzmann transport equation discretized by the analytic coarse mesh finite difference method (ACMFD) over the spatial domain as well as the simplified P3 (SP3) for the angular variable. In the first part of this work we proposed a SP3-ACMFD approach to solve the static eigenval...

Accident Tolerant Fuels (ATFs) are fuels or fuel clads with improved features in comparison with standard UO2-Zircaloy in commercial LWRs. The ATFs could tolerate in vessel loss of coolant accidents. Iron-Chromium-Aluminum (FeCrAl) alloy and Silicon carbide (SiC) are being introduced as enhanced cladding candidates to mitigate accident consequences...

The present work proposes a solution to the static Boltzmann transport equation approximated by the simplified P3 (SP3) on angular, and the analytic coarse mesh finite difference (ACMFD) for spatial variables. Multi-group SP3-ACMFD equations in 3D rectangular geometry are solved using the GMRES solution technique. As the core time dependent analysi...

Water hammer or propagation of pressure waves generates profound forces through pipelines of industrial high pressure processes which causes structural vibration of the pipe in both radial and axial directions. To model the sudden rupture of a pipeline system the fluid–structure interaction, FSI, is taken into account by coupling the structural vib...

After the Fukushima disaster, interest in the evaluation of severe accidents in nuclear power plants and off-site consequences has significantly increased. Because experimental studies are difficult to conduct, computational methods play a substantial role in accident analysis. In this study, a severe accident in the Bushehr pressurized water react...

Boiling flows in a BWR rod bundle is studied using the Drift Flux Model solved by the Broyden method. Four constitutive equations in the drift flux model plus auxiliary lateral momentum equations are discretized via the mesh staggered finite volume method. Then, the Broyden method is invoked for solving the resulted nonlinear system of equations. A...

This survey focuses on a goal-oriented approach for the adaptive mesh refinement (AMR) of even parity neutron transport equation on the unstructured triangular or quadrilateral grids. Two a posteriori error estimates are compared for conducting the refinement process. User specified quantity of interest is set as the source in the adjoint equation...

In the present investigation, isolation and determination of a new strain of Acidithiobacillus ferridurans SBU-SH2 from an Iranian sulfur hot spring are carried out and then it is used to evaluate the performance of uranium bioleaching from a low-grade natural resource host. The efficiency of uranium recovery is studied by considering the influence...

The detailed dynamic modeling and the simulation of the rapid depressurization of PWR following leak or loss of coolant accident, LOCA, is a key element of the safety analysis in the nuclear power plants. Early in a LOCA, the blowdown at the break point causes the propagation of an acoustic wave through the primary circuit. The local pressure gaps...

The microbial-assisted dissolution of minerals from low-grade ore resources has been recently developed into interesting processes to recover valuable metals, because of limitation in high- grade ore resources.
In the present study, lab-scale column bioleaching of low-grade uranium ore by a new isolated strain of chemolithotrophic Acidithiobacillus...

Algal biomass has a great potential for cleaning metal pollutants from wastewaters. In the present study, biosorption of three metals i.e., vanadium, titanium, and uranium which are appeared in contaminated effluent during the uranium ore mines processing or in sludge resulting from pure UO2 processing are investigated by a new strain of Galdieria...

In this paper, a Drift-Flux model is presented for the analysis of the thermal-hydraulic behavior of a vertical boiling channel with various wall thermal fluxes. Detail treatment and simulation of the two-phase flow phenomena are critical to the safety analysis of nuclear power reactors. Four principal conservation equations in the Drift-Flux model...

Generally for numerical solution of the differential equations the mesh generation process is a major problem in classical mesh-based methods, especially in irregular domains. Contrary to the meshed based methods, Mesh Free (MFree) methods have been developed to avoid mesh generation. In the MFree methods a set of scattered nodes is used without re...

The α-k iteration method is a common approach for calculating the fundamental α- or time-eigenvalue. The bottleneck of the method lies in how to estimate or adjust the amount of α value in each iteration. Prolonged convergence as well as the need for a proper initial guess for the α-eigenvalue are two main deficiencies of commonly employed α adjust...

In the majority of research reactors, the core cooling is a combination of forced and Natural Circulation (NC). Typically, in normal operation the generated heat is removed by forced convection whilst the NC prepares residual decay heat removal. In this study, the focus is on evaluation of Tehran Research Reactor (TRR) core in failure of the natura...

An even parity approach for the detection of main stream channels of response flux inside the material is presented. The product of forward and adjoint flux is called the response flux which plays an important role in assessing the performance of shielding materials. Based on two distinct maximum principles, even parity forward and adjoint fluxes (...

In this paper a variational formulation for the adjoint even parity neutron transport equation (NTE) based on the generalized least squares method is adopted. The so-called PN method or expansion via Spherical Harmonics Polynomials (SHPs) is then summoned to treat the angular dependency of the equation while Finite Element Method (FEM) is invoked f...

A family of variational principles based on discontinuous finite element for solving the transport equation is considered. Furthermore in this paper the adaptive h-refinement approach based on conjoint variational formulation has been presented. The conjoint maximum principle derived, not only ensures global particle conservation for the whole syst...

A new variational approach with anisotropic scattering kernel for first order neutron transport equation based on Finite Element Method (FEM) and Double-PN (DPN) approximation has been introduced. In presented variational principle, the angular dependence of the neutron flux has been separated into two sub-ranges of the forward and backward moving...

The finite element method (FEM) is a widely-used method to solve neutron transport equation in an arbitrary domain, but in order to ensure the accuracy of solution a re-meshing process is often required and some regions of the domain need to be meshed with finer meshes and consequently the whole domain should be re-meshed again. To overcome this pr...

Accurate implementation of reflective Boundary Condition (B.C.) in the PN method for both forward and adjoint analysis in one, two and three dimensional geometries is investigated in detail. Slab, spherical and cylindrical geometries are covered for 1D space, while a comprehensive procedure for general 3D Cartesian geometry is explained which encap...

This paper focuses on a new direct simulation Monte Carlo (DSMC) code based on combinatorial geometry (CG) for simulation of any rarefied gas flow. The developed code, called DgSMC-A, has been supplied with
an improved CG modeling able to significantly optimize the particle-tracking process, resulting in a highly reduced runtime compared to the con...

Practical challenges concerning the solution of multi-dimensional even parity Neutron Transport Equation (NTE) is investigated in depth. Semi-analytic approach is provided by analytic integration of angular matrices appeared through expansion of even parity angular flux density, , via Spherical Harmonics Polynomials (SHPs). Spatial variable treatme...

In this paper, a neutron noise simulator based on the point flux nodal expansion method (PFNEM) for nuclear reactor cores with rectangular geometry is developed. The point flux scheme is a kind of nodal expansion method (NEM) which is based on keeping balance of neutrons in every node and has been proved very efficient for LWR analysis. The method...

The method of characteristics has been widely used for neutron transport calculations in lattice physics computations. This method recently has attracted a lot of attention for whole core calculation because of its flexibility in geometry definition and to highly impose parallelization. One area of whole core calculation is perturbation calculation...

In this paper, we describe the structure of a new Direct Simulation Monte Carlo (DSMC) code that takes advantage of combinatorial geometry (CG) to simulate any rarefied gas flows Medias. The developed code, called DgSMC-B, has been written in FORTRAN90 language with capability of parallel processing using OpenMP framework. The DgSMC-B is capable of...

A major group of accidents in PWRs is related to improper operation of control rods. Malfunction in control and protection system (CPS) of control rods or operator's error can be a cause of such events. This paper presents a study of the drop of one control rod accident with multipoint kinetic approach. The main aim of this work is the introduction...

In this work we presented the adaptive hp-SEM approach which is obtained from the incorporation of Spectral Element Method (SEM) and adaptive hp refinement. The SEM nodal discretization and hp adaptive grid-refinement for even-parity Boltzmann neutron transport equation creates powerful grid refinement approach with high accuracy solutions. In this...

Inspired by fireflies behavior in nature, a firefly algorithm has been developed for solving optimization problems. In this approach, each firefly movement is based on absorption of the other one. For enhancing the performance of firefly algorithm in the optimization process of nuclear reactor loading pattern optimization (LPO), we introduce a new...

Thermo-mechanical behavior of fuel rod is of great importance for safety assessment of nuclear reactors. This paper deals essentially with the mechanical description of pressurized water reactor (PWR) fuel rods under long-term burnups. The main goal of the work is generation of a numerical code for study of pellet–cladding interaction (PCI) as long...

In this paper, a new method for optimizing the fuel arrangement in a WWER-1000 reactor core during refueling cycle is presented. Finding the best configuration corresponding to the desired pattern, an enhanced PSO with a Novel Mutation operator is applied. WIMS and PARCS (Purdue Advanced Reactor Core Simulator) codes are used to calculate the neutr...

In this paper, we developed a new parallel optimization algorithm, P-PSOSA, for performing the fuel management optimization; we define two different fitness function considering the multiplication factor maximizing and power peaking factor minimizing objectives simultaneously. For this purpose, we developed a FORTRAN program in order to gain the po...

Although there have been well established transport based codes for core neutronics analysis, it is yet impractical to implement them in the real core treatment because their performance is not so great on ordinary server computers. For this reason, most of neutronics codes for core calculation are subject to two steps calculation procedure which c...

In this work, we developed a new high order nodal code for the neutronic analysis of hexagonal-z geometry using the first order accuracy of Average Current Nodal Expansion Method (ACNEM) for radial direction and the second order solution of ACNEM for axial direction. For this purpose, we prepared Average Current Nodal Expansion Code for three-dimen...

In this paper, we develop a novel optimization algorithm, Bat Algorithm (BA), in order to implement in the Loading Pattern Optimization (LPO) of nuclear reactor core. For performing the fuel management optimization, we define a fitness function considering the multiplication factor maximizing and power peaking factor minimizing objectives simultane...

The aim of this work is to develop a coarse mesh treatment strategy using adaptive polynomial, p, refinement approach for average current nodal expansion method in order to solve the neutron diffusion equation. For performing the adaptive solution process, a posteriori error estimation scheme, i.e. flux gradient has been utilized for finding the pr...

In this paper, we propose a core reloading of pressurized water reactors technique based on a hybrid Artificial Bee Colony (ABC) algorithm. Our approach integrates the merits of both ABC algorithm and Particle Swarm Optimization (PSO). The neighborhood search scheme of the algorithm is improved by location of personally encountered the most flowers...

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the fuels. In this paper, in order to...

The aim of this work is to apply the new developed optimization algorithm, Self-adaptive Global best Harmony Search (SGHS), for PWRs fuel management optimization. SGHS algorithm has some modifications in comparison with basic Harmony Search (HS) and Global-best Harmony Search (GHS) algorithms such as dynamically change of parameters. For the demons...

In this article we introduce a code called FDBACE (Fast Doppler Broadening ACE) which has been developed as a tool for MCNP users to enhance the generation of high temperature ACE data tables. In this code, we developed new broadening, thinning and unionization subroutines, implemented in FORTRAN programming language, for directly broadening the AC...

One of the main goals of reliability-centered maintenance programs is to find optimal maintenance strategies. Availability of emergency systems through preventive maintenance scheduling has an important role in Probabilistic Safety Analysis (PSA). Preventive maintenances can be replaced with reloading time maintenance and reduce unavailability of t...

The operation and fuel management of reactors are of utmost importance. Core performance analysis constitutes an important stage in core fuel management optimization. In this article, a new algorithm for core optimization on safety limitation of nuclear reactors is introduced. A Parallel Integer Coded Genetic Algorithm, PICGA, is applied to obtain...

In this paper, we develop a new optimization approach, i.e., Discrete Firefly Algorithm (DFA), in order to implement in the loading pattern optimization of nuclear reactor core. For the evaluation of proposed strategy, we use a multi-objective fitness function with two main goals in the nuclear core pattern design i.e. maximizing the core multiplic...

In this work, we developed an adaptive hp-refinement strategy for average current nodal expansion method in order to solve the neutron balance equation. A flux gradient based a posteriori estimation scheme has been utilized for searching the nodes with numerical errors. The relative Cartesian direction net leakage of nodes has been considered as an...

This work addresses applications of the classical harmony search (HS), improved harmony search (IHS) and the harmony search with differential mutation based pith adjustment (HSDM) to PWR core reloading pattern optimization problems. Proper loading pattern of fuel assemblies (FAs) depends on both neutronic and thermal–hydraulic aspects; obtaining op...

The aim of this work is to develop a coarse mesh code using various orders of average current nodal expansion method to solve the neutron balance equation implementing the proposed adopted iterative solution algorithm for reactor core calculations. Modern nodal methods have the ability to treat diffusion equation with coarse meshes which cause the...

The aim of this work is to develop a spatially adaptive coarse mesh strategy that progressively refines the nodes in appropriate regions of domain to solve the neutron balance equation by zeroth order nodal expansion method. A flux gradient based a posteriori estimation scheme has been utilized for checking the approximate solutions for various nod...

In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (Keff) in order to extract the maximum cycle energy and minimizing the power peaking f...

The objective of this work is to develop a core loading optimization technique using differential harmony search algorithm in the context of obtaining an optimal configuration of fuel assemblies in pressurized water reactors. To implement and evaluate the proposed technique, differential harmony search nodal expansion package for 2-D geometry, DHSN...

In this paper a core reloading technique using Harmony Search, HS, is presented in the context of finding an optimal configuration of fuel assemblies, FA, in pressurized water reactors. To implement and evaluate the proposed technique a Harmony Search along Nodal Expansion Code for 2-D geometry, HSNEC2D, is developed to obtain nearly optimal arrang...

The efficient operation and fuel management of PWRs are of utmost importance. Core performance analysis constitutes an essential phase in core fuel management optimization. Finding an optimum core arrangement for loading of fuel assemblies, FAs, in a nuclear core is a complex problem. In this paper, application of classical harmony search (HS) and...

Despite remarkable progress in optimization procedures, inherent complexities in nuclear reactor structure and strong interdependence among the fundamental indices namely, economic, neutronic, thermo-hydraulic and environmental effects make it necessary to evaluate the most efficient arrangement of a reactor core. In this paper a reactor core reloa...

This work presents a study on the performance comparison of zeroth order nodal expansion methods, NEM. A computer code based on zeroth order nodal expansion methods, ZONEM-3D, is developed for the steady state diffusion equation calculation using nodal expansion family methods including average current, average flux, point current and point flux in...

Thermal upscattering has considerable impact on the neutronic analysis of light water reactors. This phenomenon is more significant for reactors with advanced fuels having high plutonium content: Mixed Oxide (MOX) fuels. In this paper, a sensitivity analysis for the impact of upscattering boundary energy on the integral results of MOX and UO2 fuels...

This paper describes an improvement pertaining to the energy group structure of WIMS code in thermal lattices using Particle Swarm Optimization method.The 69 group WIMS cross section library for the specified energy structure is generated using NJOY data processing system. The integral parameters of thermal reactor lattices BAPL-UO2 and TRX are cal...

Thermal–hydraulic subchannel treatment of a typical hexagonal fuel assembly of VVER nuclear reactor core in steady-state or transient conditions needs consideration of detail geometry of fuel rods and subchannels. The COBRA-EN code could not generate the hexagonal subchannel and fuel-subchannel connections automatically and user must enter all requ...

Core performance analysis constitutes an essential phase in core fuel management optimization. The output consists mainly of the neutron flux and core power distributions which are needed for deriving the safety related thermal margins. Based on the results of the core simulation, feasible options of loading patterns and control strategies can be s...

The numerical treatment of partial differential equations with element-free discretization techniques has been attractive research area in the recent years. In this paper an Element-free Galerkin, EFG, method is applied to solve the neutron diffusion equation in X–Y geometry. The Moving Least Square (MLS), interpolation is used to construct the sha...

In this paper a core reloading technique using Artificial Bee Colony algorithm, ABC, is presented in the context of finding an optimal configuration of fuel assemblies. The proposed method can be used for in-core fuel management optimization problems in pressurized water reactors. To evaluate the proposed technique, the power flattening of a VVER-1...

Analysis of economical aspects of centrifuge-based separation shows that the bulk of the cost is proportional to the number of centrifuges in a cascade. In this paper a program which is called MAKNO is used to obtain velocity field in a centrifuge by solving popular purely axial flow in a gas centrifuge. Through using MAKNO and solving concentratio...

The aim of this work is to develop a new hybrid mutation integer for integer coded genetic algorithm, ICGA, to design the loading pattern, LP, in pressurized water reactors. Because of the huge number of possible combinations for the fuel assemblies, FAs, loading in a core and finding the optimum solution is a truly complex problem. In common genet...

A major objective in reactor design is to provide the capability to withstand a wide range of postulated events without exceeding specified safety limits. Assessment of the consequence of hypothetical loss of coolant accident (LOCA) in primary circuit is an essential element to address fulfilment of acceptance criteria. In addition, finding the pos...

The efficient operation and fuel management of PWRs are of utmost importance. Recently, genetic algorithm (GA) and particle swarm optimization (PSO) techniques have attracted considerable attention among various modern heuristic optimization techniques. GA is a powerful optimization technique, based upon the principles of natural selection and spec...

The objective of nuclear fuel management is to minimize the cost of electrical energy generation subject to operational and safety constraints. In the present work, a core reload optimization package using continuous version of particle swarm optimization, CRCPSO, which is a combinatorial and discrete one has been developed and mapped on nuclear fu...

A major objective in reactor design is to provide the capability to withstand a wide range of postulated events without exceeding specified safety limits. Assessment of the consequence of hypothetical Loss of Coolant Accident (LOCA) in primary circuit is an essential element to address fulfillment of acceptance criteria. In addition to analysis of...

## Citations

... In practical engineering, the application of pipeline systems transporting gas-liquid two-phase flow for offshore facilities is increasing gradually, such as the coolant flow pipeline in offshore floating nuclear plants (OFNPs) Behzadi et al., 2021), the oil production and transport pipelines in LNG-FPSO (Jiang et al., 2015), and fluid transportation pipelines in deep-sea mining systems (Wu et al., 2021). Under these circumstances, the gas-liquid two-phase flow is inevitably affected by the ocean conditions, such as rolling, heaving, and pitching (Chen et al., 2015;Basit et al., 2019). ...

... For optimally tune the parameters involved in the LSTM-SAE technique, an optimal parameter tuning process is done by the GWO algorithm. GWO algorithms are presented by Mirjalili et al., [22] in 2016, stimulating in the hunting nature of grey wolf [22][23][24]. ...

... The projects also focus on successful implementation of the design upgrades in the commercial nuclear power plants before 2021 (Goldner, 2012). Researchers are working on different ATF materials for prospective use in light water reactors (LWRs) (Bragg-Sitton et al., 2016;Spencer et al., 2016;Terrani, 2018;Ebrahimgol, Aghaie and Zolfaghari, 2021). Table 1 summarizes the recent research focused on the properties of ATF materials and their responses to normal and transient conditions. ...

... The α, β and δ determine the pursue position and the other wolves will update their position randomly around the pursue. This step is continued until the required condition is met [29]. The steps of the GWO algorithm are illustrated in Figure 4. ...

... To overcome such problems, the drift-flux model is derived from the two-fluid model by adding phase mass, momentum and energy balance equations where the lost physical information are substituted by an empirical slip equation. See for example [14,15,16,17,18,19,20,21,22] and references therein. This framework has shown that the drift-flux model governing equations are simpler than the two-fluid model. ...

... Further research, analysis and possible optimization of the observed steam turbine will be performed by using real assumptions and with involving various artificial intelligence and optimization methods [26][27][28][29]. LPC which has the highest loss of specific work and the lowest energy efficiency will be a baseline of any possible improvement. ...

... A harmonics-nodal collocation algorithm has also been proposed in [5] for solving timedependent simplified P N models. More recently, a static solution of the simplified P 3 approximation has been investigated for a reactor core in [21]. ...

... The properties of cladding materials affect the responses of NPPs during severe accidents. It has been demonstrated from many studies (Qiu et al., 2018;Gurgen and Shirvan, 2018;Ebrahimgol et al., 2019;Wu and Shirvan, 2020;Jin et al., 2020) that ATF claddings could improve the fuel performance of nuclear power plants in severe accidental conditions compared to conventional Zr alloys. These studies have evaluated the plant performance using the properties of conventional Zr alloys and existing cladding candidate materials such as FeCrAl and SiC. ...

... For decades, microbiological renovations were focused on (i) the search for more effective acidophilic species and strains [6][7][8] or (ii) optimization of the processing regimes [9][10][11]. Good examples of these approaches are presented in some modern research investigations where authors used (i) a new Sulfobacillus strain, (ii) a higher temperature (47 • C), and (iii) low pH (1.8) [12]. ...

... When metal bacterial complex is introduced with tetravalent uranium mineral it gets oxidised into hexavalent uranium which is soluble with acidic mixture of bacteria and hence leach liquor can be obtained for further uranium recovery process. Thus, Bio-oxidation of uranium is the key mechanism involved during the leaching process (Jalali et al., 2019). ...