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Ayoobian, N. ,
Khodadadi h., H. ,
Mokhtari j., ,
Abbassi y., Y. Nuclear Engineering and Design (00295493) 427
This paper analyses the thermal–hydraulic performance of a proposed 50-kW aqueous homogeneous reactor (AHR) aimed at producing Mo-99. The reactor thermal–hydraulic characteristics are analyzed using ANSYS-FLUENT software and Relap code, and the results are validated through experiments conducted by a prototypic cylindrical sector of the reactor vessel. Boundary conditions in the CFD simulation are set up based on Relap modelling, while CFD outputs are validated by experimental results achieved using the laboratory model. Then, various parameters, including coolant mass flow rate, first cycle pipe diameter, coolant outlet temperature, average fuel solution temperature, and the elevation difference between the reactor and heat exchanger, are optimized using the least-squares optimization method. Results demonstrate that the heat removal system provides sufficient cooling capacity to ensure stable operation of the proposed AHR core by preventing fuel solution overheating and, consequently, boiling (void formation) in the active fuel solution. © 2024
Journal of Instrumentation (17480221) 17(11)
Electron linear accelerator (LINAC) is a high-energy X-ray generator with extensive industrial, inspection, and medical applications. The studied LINAC uses a 2.6 MW magnetron as its RF generator derived with a Marx modulator. An electron gun of filament voltage 6 V and current 2.2 A connected to the accelerator tube is applied to establish electron current inside the drift tube. For this purpose, 9 kV and 25 kV voltages are applied to the cathode and anode of the electron gun pulse by pulse, generating dual-energy x-ray beams as outputs. An Automatic Frequency Control (AFC) system is designed for studied LINAC so that variation of output dose rate is lower than 2% and the required time to achieve the maximum output dose rate is lower than 1 s. Optimizing the Marx modulator's switching algorithm, the ripple in the output beam has been lower than 5%. The half-value layer (HVL) of steel has been measured for high and low energies equal to 6 MeV and 3 MeV, respectively. The output pulse of the electron beam and dose rate have been measured using a linear detector array (LDA) and dosimeter, respectively, with a pulse frequency of 35-400 Hz and pulse width of 2-4 μs. © 2022 IOP Publishing Ltd and Sissa Medialab.
Nuclear Engineering and Technology (2234358X) 53(1)pp. 266-272
In the pulsed electron beam generators, such as plasma focus devices and linear induction accelerators whose electron pulse width is in the range of nanosecond and less, as well as in cases where there is no direct access to electron beam, like runaway electrons in Tokamaks, measurement of the electron energy spectrum is a technical challenge. In such cases, the indirect measurement of the electron spectrum by using the bremsstrahlung radiation spectrum associated with it, is an appropriate solution. The problem with this method is that the matrix equation between the two spectrums is an ill-conditioned equation, which results in errors of the measured X-ray spectrum to be propagated with a large coefficient in the estimated electron spectrum. In this study, a method based on the neural network and the MCNP code is presented and evaluated to recover the electron spectrum from the X-ray generated by collision of the electron beam with a target. Multilayer perceptron network showed good accuracy in electron spectrum recovery, so that for the X-ray spectrum with errors of 3% and 10%, the network estimated the electron spectrum with an average standard error of 8% and 11%, on all of the energy intervals. © 2020 Korean Nuclear Society
Thermal Engineering (00406015) 68(12)pp. 922-929
Abstract: Computational Fluid Dynamics (CFD) is widely used in different industrial applications. In this research, the application of CFD in the thermo-hydraulic evaluation for a typical small modular Pressurized Water Reactor (PWR) was studied using ANSYS Fluent software. First, reactor core with different (pitch—fuel rod diameter) was simulated using MCNP code. Subsequently, axial neutron heat flux was calculated in the hottest rod. In the following, different fuel channels were simulated using ANSYS workbench and corresponding mass flow rate according to the fluid outlet temperature was computed using Fluent. Then, thermo-hydraulic parameters including pumping power, convective heat transfer coefficient and turbulent intensity were calculated. Artificial neural network (ANN) coupled with genetic algorithm (GA) was used for optimization; and pair pitch ‒ fuel rod diameter (0.012 m, 0.0072 m) was selected as the optimum value. Also, Critical Heat Flux (CHF) was computed with CFD-simulation, and compared with Tong CHF correlation and Groeneveld look-up table. A good agreement was observed between results, but CHF obtained from CFD simulation was more conservative. According to the results, Minimum Departure from Nucleate Boiling Ratio (MDNBR) was obtained as 2.12, which was compatible with its typical value. Accordingly, it could be concluded that the optimum reactor core was in the safe mode in the steady state conditions. © 2021, Pleiades Publishing, Inc.
Progress in Nuclear Energy (01491970) 121
Production of medical isotopes such as 99Mo, 90Sr, 131I and other fission products using Aqueous Homogeneous Reactors (AHRs) offers several advantages in comparison to other methods. In this method, after about one week operation, required isotopes can be directly extracted from the solution and the remaining fuel solution will be returned to the core for the next campaign. Moreover, AHRs are inherently safe and reliable due to their large negative reactivity feedbacks. This paper demonstrates technical features and conceptual scheme of the Iranian homogeneous reactor named RAHA, which is mainly used for the production of 99Mo/99mTc. Neutronic and thermal-hydraulic characteristics of the reactor were analyzed with MCNPX-2.6 and RELAP5/Mod3.3 codes, respectively. It was assumed that the reactor is fueled by Uranyl Nitrate, and is able to remove generated heat using natural circulation. The coolant circulation velocity was obtained as 6.5 cm/s. Four different types of neutron absorbers were considered as control rod material, and BORA resin with a diameter of 1.4 cm and Stainless-Steel cladding was chosen. As a new characteristic which may improve safety of the system, the possibility of reactor startup without any external neutron source for the case of fresh fueled core was assessed. Finally, effects of the possible uncertainties on some physical and material properties of the core were investigated and sensitivity of the effective multiplication factor relative to the uncertain parameters was also extracted using Spearman's formula. © 2019
Progress in Nuclear Energy (01491970) 119
The main purpose of this study was to design small modular PWR with power of 150 MWth for long life operation, based on the coupled neutronic and thermal-mechanical evaluations. In this regard, pitch and fuel rod diameter were changed from 0.9 cm to 2.5 cm and 0.6 cm–0.96 cm, respectively; and desirable pairs (pitch - fuel rod diameter) were obtained. Neutronic parameters such as excess reactivity, maximum to average ratio of the radial neutron flux and fuel burnup were calculated using MCNP code. Also, thermal-mechanical parameters of the fuel rod including fuel centerline temperature, cladding hoop stress and fuel rod internal gas pressure were computed using FRAPCON code. Independency of results from number of axial nodes in FRAPCON code was analyzed. The number of axial nodes was considered 40 to ensure the accuracy of the FRAPCON results. Neutronic and thermal-mechanical cost values for each pair were calculated and Pareto front of the desirable pairs were obtained. Based on the results, eight desirable pairs were selected. By assuming equal importance for neutronic and thermal-mechanical parameters, total cost value was defined and pair of (1.1, 0.64) was selected as the most desirable one. Finally, fuel cycle lifetime was assessed for desirable pair considering neutronic (criticality and burnup) and thermal-mechanical (cladding strain, oxidation and hydriding as well as fuel rod internal gas pressure) limits. It was observed that fuel cycle lifetime was enhanced to 1750 days (40 GWD/ton). © 2019
Progress in Nuclear Energy (01491970) 111pp. 183-194
Nonlinear systems such as nuclear reactor dynamics are associated with parametric uncertainties and stability conditions. These properties must be taken into account in controller design. In this paper, an effective and robust Proportional–Integral–Derivative (PID) controller is designed and tuned. Particle Swarm Optimization (PSO) is used as a metaheuristic algorithm to optimize PID gains. The PSO–PID controller is scheduled based on the power level control of a PWR-type reactor. According to Axial Offset (AO) power-distribution characteristics in the PWRs, the two-point kinetic model is used with a constant AO strategy. Lyapunov stability synthesis is intended to guarantee the stability condition of the system at various time intervals. Disturbance Rejection System (DRS) is designed based on a Lyapunov approach to avoid output disturbance. The control signal disturbance is properly removed by the designed DRS. Simulation results show output performance and robustness of the tuned PSO–PID controller against parametric uncertainties. © 2018
Progress in Nuclear Energy (01491970) 106pp. 44-50
In this study, the core behavior following the reactor vessel lower head rupture together with the moderator system failure in a tank type heavy water research reactor, IHWRR, is analyzed through the MELCOR code. The focus is on the safety assessment of the reactor core for Design Extension Conditions and overall assessment of safety features of the reactor for any residual risk posed by severe accidents. The inherent features of IHWRR provide a broad spectrum of scenarios where the fuel does not melt, even if primary and moderator cooling are lost. Accordingly, coincident with vessel rupture, three different scenarios are considered for moderator failure: i) loss of moderator forced circulation, calandria tube rupture in ii) the upper and iii) the lower parts of the core. The obtained results showed that the vessel rupture that coincides with the tube rupture in the lower part of the core leads to the oxidation of entire Zr cladding and failure of the calandria tubes without core melt or hydrogen explosion in the containment. © 2018
Nuclear Engineering and Technology (2234358X) 50(6)pp. 877-885
Various controllers such as proportional–integral–derivative (PID) controllers have been designed and optimized for load-following issues in nuclear reactors. To achieve high performance, gain tuning is of great importance in PID controllers. In this work, gains of a PID controller are optimized for power-level control of a typical pressurized water reactor using particle swarm optimization (PSO) algorithm. The point kinetic is used as a reactor power model. In PSO, the objective (cost) function defined by decision variables including overshoot, settling time, and stabilization time (stability condition) must be minimized (optimized). Stability condition is guaranteed by Lyapunov synthesis. The simulation results demonstrated good stability and high performance of the closed-loop PSO–PID controller to response power demand. © 2018
Annals of Nuclear Energy (03064549) 118pp. 107-121
Load following is an importance topic in the Nuclear Power Plants (NPPs). One of the conventional and simplest ways is the use of Proportional-Integral-Derivative (PID) controller. The reactor power is simulated based on the point kinetic model. PID gains of a nonlinear time-varying system (a PWR NPP) are optimized and scheduled using real-coded genetic algorithm (GA). To this end, the objective function of the decision variables, include the overshoot, settling time and stabilization time (based on the Lyapunov approach) of the system is minimized. The presented control system track demand power level change within a wide range of time. The simulation results demonstrate good stability of this method and show high performance of the optimized PID gains to adapt any changes in the output power. © 2018
Annals of Nuclear Energy (03064549) 114pp. 206-213
Transport and deposition of fission products within the primary coolant circuit during severe accident condition have significant effect on the amount of released radionuclides in the containment. In this study, a new code, STRCS, compatible with RELAP5/SCDAP code, has been developed that calculates the fission products’ transport through the piping system. The homogeneous and heterogeneous nucleation, aerosol agglomeration, aerosol and vapor deposition and aerosol resuspension are considered in this code. The developed code was validated through international standard problems STROM SR11 and PHEBUS FPT-1 experiments. The STRCS predictions are in good agreement with the ISPs published results, although the obtained results deviate from the PHEBUS FPT-1 experimental data. © 2017 Elsevier Ltd
Journal of Food Science and Technology (00221155) 54(13)pp. 4277-4283
In this study, the effect of gamma irradiation on the shelf life and properties of cucumber was investigated. These properties include weight reduction, fruit density, juice, tissue firmness, total soluble solids (TSS), total titratable acidity, chlorophyll and vitamin C, pH, marketability, flavor, frostbite and fungal effects. For this purpose, cucumbers were irradiated with dose of 2, 2.5 and 3 kGy. The exposure time was calculated by MCNP4C; the Monte Carlo particle transport code. Three types of fungi (white-Sclerotinia sclerotiorum, gray-Botrytis cinerea and olive-Cladosporium cucumerinum), were used to infect some samples. The chlorophyll and vitamin C preservation abilities were increased to about 3 and 1.4 times, respectively with irradiation treatment. Also, the shelf life was increased about 1 week, while chilling injuries is decreased. Samples’ resistance to the fungal growth was evident and the process of fungal growth on the irradiated samples was delayed up to 1 week. The best properties were obtained at the irradiation dose of 2 kGy since it had less effect on flavor, TSS and tissue firmness. © 2017, Association of Food Scientists & Technologists (India).
Journal Of Applied Clinical Medical Physics (15269914) 17(2)pp. 194-205
High-dose-rate (HDR) brachytherapy is a popular modality for treating cancers of the prostate, cervix, endometrium, breast, skin, bronchus, esophagus, and head and neck as well as soft-tissue sarcomas. Because of different source designs and licensing issues, there is a need for specific dosimetry dataset for each HDR source model. The main objective of the present work is to measure 2D relative dose distribution around a new prototype 192Ir source, referred to as IRAsource-HDR, in PMMA phantom in the framework of AAPM TG-43 and TG-55 recommendations for radial distances of 0.5cm to 4 cm. Radiochromic films (RCFs) Gafchromic EBT and HD-810 were used for measurements. The dose rate constant, A, of the source was determined to be 1.084± 4.6%, 1.129 ± 4.4%, and 1.112 ± 0.8% cGyh-1U-1 using EBT RCF, HD-810 RCF, and Monte Carlo (MC) simulation, respectively. The results obtained in this study are in good agreement with previously published data for HDR interstitial 192Ir-HDR sources with a maximum discrepancy of ± 4.5%. An acceptable agreement (within ± 2%) between MC calculations and RCFs measurements showed that HD-810 RCF dosimetry is as good as EBT RCF, within HDR brachytherapy, and justifies the use of specific data for this new source. These data could be used as a benchmark for dose calculations in the conventional brachytherapy treatment planning systems. © Creative Commons Attribution 4.0 International License.
Australasian Physical and Engineering Sciences in Medicine (18795447) 39(2)pp. 413-422
High-dose-rate (HDR) brachytherapy is a common method for cancer treatment in clinical brachytherapy. Because of the different source designs, there is a need for specific dosimetry data set for each HDR model. The purpose of this study is to obtain detailed dose rate distributions in water phantom for a first prototype HDR 192Ir brachytherapy source model, IRAsource, and compare with the other published works. In this study, Monte Carlo N-particle (MCNP version 4C) code was used to simulate the dose rate distributions around the HDR source. A full set of dosimetry parameters reported by the American Association of Physicists in Medicine Task Group No. 43U1 was evaluated. Also, the absorbed dose rate distributions in water, were obtained in an along-away look-up table. The dose rate constant, Λ, of the IRAsource was evaluated to be equal to 1.112 ± 0.005 cGy h−1 U−1. The results of dosimetry parameters are presented in tabulated and graphical formats and compared with those reported from other commercially available HDR 192Ir sources, which are in good agreement. This justifies the use of specific data sets for this new source. The results obtained in this study can be used as input data in the conventional treatment planning systems. © 2016, Australasian College of Physical Scientists and Engineers in Medicine.
Annals of Nuclear Energy (03064549) 88pp. 95-99
Maintenance planning is a critical issue for all heavy industrial sectors such as aeronautics, automobile factories and power plants. The study herein is aimed towards improvement of safety systems reliability in a nuclear power plant. Optimization of maintenance and surveillance test activities is one of the best strategies to improve the reliability of the related safety systems in this kind of plants. Maintenance programs can be formulated in terms of a multi-objective optimization where unavailability, cost and Exposure Time (ET) act as decision criteria and surveillance tests, Allowed Outage Time (AOT) and Preventive Maintenance (PM) intervals act as decision variables. In this paper, Genetic Algorithm (GA) is used to find the best series of answers in the form of Pareto front curve. Sensitivity Index (SI) is applied as a decision making tool to extract the most optimized and promising solution. The unavailability, cost and ET functions, for the most optimal solutions, were reduced by 86%, 58% and 30%, respectively. © 2015 Elsevier Ltd. All rights reserved.
Journal of Cancer Research and Therapeutics (19984138) 8(4)pp. 610-618
Aim: Simulating Many Accumulative Rutherford Trajectories Electron Photon and Neutral Transport Solver (SMARTEPANTS) is a discrete ordinates S N Boltzmann/Spencer-Lewis solver that was developed during 1988-1993 by William Filippone and his students. The code calculates particle fluxes, leakage currents as well as energy and charge deposition for coupled electron/photon in x-y-z geometries both in forward and in adjoin modes. Originally, SMARTEPANTS was designed to utilize CEPXS cross-section library for shielding calculation in satellite electronics. The aim of this study was to adapt SMARTEPANTS to use a new photon cross-section library from Evaluated Photon Data Library, 1997 version (EPDL97) for intravascular brachytherapy 125 Isimulations. Materials and Methods: A MATLAB (MathworkNatick, Massachusetts) program was written to generate an updated multigroup-Legendre cross-section from EPDL97. The new library was confirmed by simulating intravascular brachytherapy Best® Model 2301 and Intersource 125 I dosimetry parameters using SMARTEPANTS with different energy groups (g), Legendre moments (L) and discrete ordinate orders (S). Results: The dosimetry parameters for these sources were tabulated and compared with the data given by AAPM TG-43 and other reports. The computation time for producing TG-43 parameters was about 29.4 min in case of g = 20, L = 7 and S = 16. Conclusion: The good agreement between the results of this study and previous reports and high computational speed suggest that SMARTEPANTS could be extended to a real-time treatment planning system for 125 I brachytherapy treatments.
Annals of Nuclear Energy (03064549) 35(12)pp. 2313-2320
This paper describes the application of a multilayer cellular neural network (CNN) to model and solve the time dependent one-speed neutron transport equation in slab geometry. We use a neutron angular flux in terms of the Chebyshev polynomials (TN) of the first kind and then we attempt to implement the equations in an equivalent electrical circuit. We apply this equivalent circuit to analyze the TN moments equation in a uniform finite slab using Marshak type vacuum boundary condition. The validity of the CNN results is evaluated with numerical solution of the steady state TN moments equations by MATLAB. Steady state, as well as transient simulations, shows a very good comparison between the two methods. We used our CNN model to simulate space-time response of total flux and its moments for various c (where c is the mean number of secondary neutrons per collision). The complete algorithm could be implemented using very large-scale integrated circuit (VLSI) circuitry. The efficiency of the calculation method makes it useful for neutron transport calculations. © 2008 Elsevier Ltd. All rights reserved.
