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Uranyl sulfate obtained by uranium ore leaching of an industrial sample deposit of Gachin site was used for uranium separation by the solvent extraction technique. The presence of other elements in the sulfate leach liquors has a negative impact on the uranium extraction process using Alamine 336; therefore, the operating costs are increased. In this study, the separation of uranium(VI) and iron(III) by Alamine 336 and kerosene have been examined as an extractant and a diluent, respectively. For this purpose, the effects of operating parameters on the extraction process such as Alamine 336 concentration, modifier concentration, contact time, initial aqueous pH, sulfate ion concentration, temperature, and stripping agents were investigated. Also, the complete removal of iron(III) interference from the sulfate leach liquor before the extraction step has been studied with reducing agents. The equilibrium constants and stoichiometric coefficients for uranium and iron extraction with Alamine 336 in the sulfate leach liquors were calculated. Likewise, the values of the thermodynamic parameters such as Gibbs energy, enthalpy and entropy were determined to prove the exothermic and spontaneous reactions. The mentioned procedure is proposed for the uranium separation from the impurities in the sulfate ores using tertiary amine for production of purified uranium.

Magnetite CoFe2O4 nanoparticle and CoFe2O4/MWCNT nanocomposite adsorbents with high specific surface area, high adsorption capacity, and easily separable property were synthesized by the hydrothermal method for uranium and lead removal from aqueous solutions. The synthesized samples were characterized by FT-IR, XRD, SEM, and BET analysis. The estimated specific surface areas were obtained as 113.4 and 186.1 m2 g−1 for CoFe2O4 and CoFe2O4/MWCNTs, respectively. The synthesized nanostructures resulted in achieving the maximum monolayer adsorption capacity. The adsorption capacity of U(VI) and Pb(II) onto CoFe2O4 nanoparticles and CoFe2O4/MWCNT nanocomposites were obtained as 294.12 (at pH 5) and 500 mg g−1 (at pH 6) for uranium and 109.89 and 217.39 mg g−1 for lead (at pH 5), respectively, which indicated the favorable effects of decorated MWCNT on the removal of uranium and lead. Several parameters, including pH, contact time, initial U(VI) concentration and solution volume, and adsorbent concentration, which influence the adsorption capacity, were also investigated.

Reliable prediction of nuclear fuel rod behavior of nuclear power reactors constitutes a basic demand for steady-state calculations, design purposes, and fuel performance assessment. Perfect design of fuel rods as the first barrier against fission product release is very important. Simulation of fuel rod performance with a code or software is one of the fuel rod design steps. In this study, a software program called MARCODE is developed in MATLAB environment that can analyze the temperature distribution, gap conductance value, and fuel and clad displacement in both solid and annular fuel rods. With a comparison of the maximum fuel temperature, fuel average temperature, fuel surface temperature, and gap conductance in solid and annular fuel, the effects of a central hole on the fuel temperature distribution are investigated.

In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

Four different magnetic Fe3O4 nanoparticles were synthesized and characterized by solvothermal method based on different sodium salts. Sodium salts which were used to synthesize the nanoparticles were NaOAc, Na2CO3, a mixture of NaOAc and Na3Cit, and a mixture of NaOAc and Na2C2O4. The structural and optical properties of the synthesized nanoparticles were examined by XRF, XRD, SEM and FT-IR. The results estimated from XRD pattern and SEM image indicated that the second sample (Na2CO3) had the lowest average particle and crystallite size around 29 nm and 43 nm. It was also shown that the first (NaOAc) and second (Na2CO3) samples had the best FT-IR spectra, similar to the available commercial sample which was provided by Merck. At last, the prepared Fe3O4 nanoparticles were applied as sorbents to sorb uranium ions (U(VI)) from radioactive wastewater. The adsorption results showed that the highest U(VI) adsorption was obtained for the second sample in the solution with pH around 10.

A full scope thermal-neutronic couple has been developed to analyze WWER-1000 nuclear reactor core by using PARCS v2.7, COBRA-EN and WIMSD-5B. First, WIMSD-5B has been used to calculate macroscopic cross sections for each fuel assembly. Neutronic calculations have been done by PARCS v2.7 code to obtain multiplication factor and PPF distributions of all fuel assemblies in the core. COBRA-EN calculates thermal-hydraulic parameters for all the fuel assemblies. This couple can be used for steady-state and transient calculations. Neutronic accidents such as Rod Ejection Accident that its initial event is a neutronic event (control rod movement that yields to reactivity changes) and thermal-hydraulic accidents such as LOFA that its initial event is a thermal-hydraulic event (coolant flow rate decrease) can be analyzed by this couple. In this study, steady-state calculations and LOFA have been simulated by this couple in the reactor core. A new “steady-state convergence” method is introduced. Results have been compared to Bushehr Nuclear Power Plant FSAR. The results show great similarities with FSAR results; and confirm the ability of this package to simulate accident in the core; especially thermal-hydraulic accident that is out of the ability of PARCS code individually.

Nuclear safety has been one of the major issues to be studied since the inception of the nuclear industry. Establishing and maintaining core cooling and ensuring containment integrity are the two main goals that nuclear safety must guarantee. The specific type of large LOCA is DECL (Double Ended Cold Leg) break which means a total guillotine type of break in cold leg pipe and is one of the most dangerous accidents in the reactor containment. In this paper, thermal-hydraulic parameters (Temperature and Pressure) of WWER-1000 (Bushehr Nuclear Power Plant) containment in a DECL accident have been simulated by CONTAIN2.0 code and a single cell model. The Containment has been divided to 23 cells in CONTAIN code but for simplicity only one cell has been considered in modeling. The model has been programmed by MATLAB. The accident has been simulated for Short-term (0 < time

The consequences of sever reactor accident depend greatly on containment safety features and containment performance in retaining radioactive material. The specific type of large LOCA is DECL (Double Ended Cold Leg) break which means a total guillotine type of break in cold leg pipe and is one of the most dangerous accidents in the reactor containment. In this paper, thermal–hydraulic parameters (temperature and pressure) of WWER-1000 (Bushehr Nuclear Power Plant) containment in a DECL accident have been simulated by CONTAIN 2.0 code and a single cell model. The containment has been divided to 23 cells in CONTAIN code but for simplicity only one cell has been considered in modeling. The model has been programmed by MATLAB. The accident has been simulated for a short time (initial 200 s) and all of the results have been compared with Bushehr’s Nuclear Power Plant FSAR.

The rod ejection accident is defined as the postulated rupture of a control rod drive mechanism housing that results in the complete ejection of a rod cluster control assembly from the reactor core. The consequences of the mechanical failure are a rapid positive reactivity insertion and an increase in the local power peaking with high local energy deposition in the fuel assembly, accompanied by an initial pressure increase in the reactor cooling system. In this study, the REA has been simulated in a WWER-1000 reactor by using WIMSD-5B and PARCS v2.7 codes. First, macroscopic cross-sections have been calculated for various types of fuel assemblies using WIMSD-5B. Results have been fed as input to PARCS v2.7 code. Steady-state, transient and specially thermal–hydraulic feedback blocks of PARCS code have been handled in this simulation. Finally, results have been compared with Final Safety Analysis Report of WWER-1000 reactor. The results show a great similarity and confirm the ability of PARCS code in simulation of transient accidents.

In this paper a whole-core transport technique using response matrix and collision probability (CP) methods is presented for large-scale, highly-heterogeneous reactors. Integral transport method has been used to provide sufficient accuracy for response matrix formation of pin cell sized node with considerably less computational expense. Ray tracing is efficiently applied using macro-bands. For practical application, double P 2 (DP 2 ) Legendre polynomial expansion is applied to approximate interface angular flux that is used to couple nodes. The proposed method is based on a sound mathematical foundation and leads to dramatically reduced memory requirements in contrast to the conventional transport method. This method is also applied to several problems such as C5G7, containing mixed oxide (MOX) and UO 2 fuel assemblies, to show the effectiveness of the proposed method. The results clearly indicate that the method is quite promising and acceptable.

In this paper, a new practical robust water level control system for the U-tube steam generator (UTSG) using the quantitative feedback theory (QFT) is proposed. The steam generator is a nonlinear uncertain plant. However, the steam generator behaves as a linear uncertain and nonminimum phase plant at its different operating points, which makes its control a challenging problem. The control problem is to design controllers such that the closed-loop plant satisfies the robust stability, disturbance rejection, and robust tracking specifications that are derived from a desired steam generator performance. In the QFT design methodology, these specifications are satisfied by generating the plant templates, the composite bounds, and a nominal plant loop shaping procedure to satisfy these bounds. Simulation results reveal that the designed QFT water level controllers will ensure all the designers’ closed-loop specifications. Also, comparison results are provided that show the effectiveness of the robust Q...

In this paper, a robust water level control system for the horizontal steam generator (SG) using the quantitative feedback theory (QFT) method is presented. To design a robust QFT controller for the nonlinear uncertain SG, control oriented linear models are identified. Then, the nonlinear system is modeled as an uncertain linear time invariant (LTI) system. The robust designed controller is applied to the nonlinear plant model. This nonlinear model is based on a locally linear neuro-fuzzy (LLNF) model. This model is trained using the locally linear model tree (LOLIMOT) algorithm. Finally, simulation results are employed to show the effectiveness of the designed QFT level controller. It is shown that it will ensure the entire designer’s water level closed loop specifications.

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.