The Experts below are selected from a list of 34509 Experts worldwide ranked by ideXlab platform
Abdallah A Nahla - One of the best experts on this subject based on the ideXlab platform.
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developed mathematical technique for fractional stochastic point Kinetics model in nuclear Reactor dynamics
Nuclear Science and Techniques, 2018Co-Authors: Ahmed E Aboanber, Abdallah A Nahla, Adel M EdressAbstract:Fractional stochastic Kinetics equations have proven to be valuable tools for the point Reactor Kinetics model, where the nuclear reactions are not fully described by deterministic relations. A fractional stochastic model for the point Kinetics system with multi-group of precursors, including the effect of temperature feedback, has been developed and analyzed. A major mathematical and inflexible scheme to the point Kinetics model is obtained by merging the fractional and stochastic technique. A novel split-step method including mathematical tools of the Laplace transforms, Mittage–Leffler function, eigenvalues of the coefficient matrix, and its corresponding eigenvectors have been used for the fractional stochastic matrix differential equation. The validity of the proposed technique has been demonstrated via calculations of the mean and standard deviation of neutrons and precursor populations for various reactivities: step, ramp, sinusoidal, and temperature reactivity feedback. The results of the proposed method agree well with the conventional one of the deterministic point Kinetics equations.
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numerical treatment for the point Reactor Kinetics equations using theta method eigenvalues and eigenvectors
Progress in Nuclear Energy, 2015Co-Authors: Abdallah A NahlaAbstract:Abstract The point Reactor Kinetics model is a stiff system of linear/nonlinear ordinary differential equations. In fact, the numerical solutions of this stiff model need a smaller time step intervals within various computational schemes. The aim of this work is an accurate numerical solution without need to the smaller time step intervals. Theta method is the most popular, simplest and widely used method for solving the first order ordinary differential equations. In light of this fact, theta method is treated for solving the matrix form of this model via the eigenvalues and corresponding eigenvectors of the coefficient matrix. In this work, the matrix form of the stiff point Kinetics equations with multi-group of delayed neutrons is introduced. The treatment theta method is applied to solve the stiff point Kinetics equations with six groups of delayed neutrons. The performance of the treatment theta method is evaluated in several case studies involving step, ramp, sinusoidal and pulse reactivities. The results of the treatment theta method are more accurate than the theta method comparing with the conventional methods.
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an efficient technique for the point Reactor Kinetics equations with newtonian temperature feedback effects
Annals of Nuclear Energy, 2011Co-Authors: Abdallah A NahlaAbstract:Abstract The point Reactor Kinetics equations of multi-group of delayed neutrons in the presence Newtonian temperature feedback effects are a system of stiff nonlinear ordinary differential equations which have not any exact analytical solution. The efficient technique for this nonlinear system is based on changing this nonlinear system to a linear system by the predicted value of reactivity and solving this linear system using the fundamental matrix of the homogenous linear differential equations. The nonlinear point Reactor Kinetics equations are rewritten in the matrix form. The solution of this matrix form is introduced. This solution contains the exponential function of a variable coefficient matrix. This coefficient matrix contains the unknown variable, reactivity. The predicted values of reactivity in the explicit form are determined replacing the exponential function of the coefficient matrix by two kinds, Backward Euler and Crank Nicholson, of the rational approximations. The nonlinear point Kinetics equations changed to a linear system of the homogenous differential equations. The fundamental matrix of this linear system is calculated using the eigenvalues and the corresponding eigenvectors of the coefficient matrix. Stability of the efficient technique is defined and discussed. The efficient technique is applied to the point Kinetics equations of six-groups of delayed neutrons with step, ramp, sinusoidal and the temperature feedback reactivities. The results of these efficient techniques are compared with the traditional methods.
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taylor s series method for solving the nonlinear point Kinetics equations
Nuclear Engineering and Design, 2011Co-Authors: Abdallah A NahlaAbstract:Abstract Taylor’s series method for solving the point Reactor Kinetics equations with multi-group of delayed neutrons in the presence of Newtonian temperature feedback reactivity is applied and programmed by FORTRAN. This system is the couples of the stiff nonlinear ordinary differential equations. This numerical method is based on the different order derivatives of the neutron density, the precursor concentrations of i -group of delayed neutrons and the reactivity. The r th order of derivatives are derived. The stability of Taylor’s series method is discussed. Three sets of applications: step, ramp and temperature feedback reactivities are computed. Taylor’s series method is an accurate computational technique and stable for negative step, negative ramp and temperature feedback reactivities. This method is useful than the traditional methods for solving the nonlinear point Kinetics equations.
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solution of the nonlinear point nuclear Reactor Kinetics equations
Progress in Nuclear Energy, 2010Co-Authors: Abdallah A Nahla, Elsayed M E ZayedAbstract:Abstract Analytical approximation and numerical solution of the point nuclear Reactor Kinetics equations with average one-group of delayed neutron and temperature feedback are presented. The analytical approximation is based on transforming the differential equations with respect to time to differential equations with respect to reactivity and neglecting very small term. The numerical solution is based on Taylor’s series method. These methods are applied to different types of initial reactivity and the results of these methods are compared.
Mitsuru Kambe - One of the best experts on this subject based on the ideXlab platform.
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Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M"
2020Co-Authors: Mitsuru KambeAbstract:A metal fueled modular island core sodium cooled fast breeder Reactor concept RAPID-M to improve Reactor performance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the Reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified refueling by replacing IFAs in which all the core and blanket fuel elements are comprised. In this paper, the 600 MWe RAPID-M design consists of 7 IFAs is presented. Significant Reactor mass savings and the improvement of inherent safety features are discussed. Plant dynamics analyses using the multi-point Reactor Kinetics equations to accommodate the modular core configuration demonstrated a favorable transient response in case of unprotected transient over power (UTOP)
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conceptual design of a modular island core fast breeder Reactor rapid m
Journal of Nuclear Science and Technology, 2002Co-Authors: Mitsuru KambeAbstract:A metal fueled modular island core sodium cooled fast breeder Reactor concept RAPID-M to improve Reactor performance and proliferation resistance and to accommodate various power requirements has been demonstrated. The essential feature of the RAPID-M concept is that the Reactor core consists of integrated fuel assemblies (IFAs) instead of conventional fuel subassemblies. The RAPID concept enables quick and simplified refueling by replacing IFAs in which all the core and blanket fuel elements are comprised. In this paper, the 600 MWe RAPID-M design consists of 7 IFAs is presented. Significant Reactor mass savings and the improvement of inherent safety features are discussed. Plant dynamics analyses using the multi-point Reactor Kinetics equations to accommodate the modular core configuration demonstrated a favorable transient response in case of unprotected transient over power (UTOP).
Tadashi Narabayashi - One of the best experts on this subject based on the ideXlab platform.
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sensitivity analysis for Reactor period induced by positive reactivity using one point adjoint kinetic equation
Nuclear Data Sheets, 2014Co-Authors: Go Chiba, Masashi Tsuji, Tadashi NarabayashiAbstract:Abstract In order to better predict a kinetic behavior of a nuclear fission Reactor, an improvement of the delayed neutron parameters is essential. The present paper specifies important nuclear data for a Reactor Kinetics: Fission yield and decay constant data of 86 Ge, some bromine isotopes, 94 Rb, 98 m Y and some iodine isotopes. Their importance is quantified as sensitivities with a help of the adjoint kinetic equation, and it is found that they are dependent on an inserted reactivity (or a Reactor period). Moreover, dependence of sensitivities on nuclear data files is also quantified using the latest files. Even though the currently evaluated data are used, there are large differences among different data files from a view point of the delayed neutrons.
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sensitivity and uncertainty analysis for Reactor stable period induced by positive reactivity using one point adjoint Kinetics equation
Journal of Nuclear Science and Technology, 2013Co-Authors: Go Chiba, Masashi Tsuji, Tadashi NarabayashiAbstract:In order to better predict the kinetic behavior of a nuclear fission Reactor, improvement of delayed neutron parameters is essential. Since it is required to establish a path from the microscopic nuclear data to the macroscopic delayed neutron parameters for the improvement, the present paper identifies important nuclear data for Reactor Kinetics. Sensitivities of the Reactor stable period, which describes Reactor kinetic behavior, to microscopic nuclear data such as independent fission yields, decay constants and decay branching ratios are calculated efficiently by using the adjoint Kinetics equation. Furthermore, nuclide-wise and nuclear data-wise uncertainties of the Reactor stable period are quantified using the variance data given in the nuclear data file, and the nuclear data that require further improvement are identified.The results obtained through the present study are quite helpful, and can be a driving force for further nuclear physics studies.
Yoshihiro Yamane - One of the best experts on this subject based on the ideXlab platform.
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explicit time integration scheme using krylov subspace method for Reactor Kinetics equation
Journal of Nuclear Science and Technology, 2011Co-Authors: Tomohiro Endo, Akio Yamamoto, Yoshihiro YamaneAbstract:The spatial discretization form of the space-dependent Reactor Kinetics equation is a first-order simultaneous ordinary differential equation in time. Conventional numerical methods of the space-dependent Kinetics equation, i.e., the generalized Runge-Kutta method, the implicit method (backward Euler method), and the Theta method, are based on the time difference approximation. However, the present study adopts the analytical solution of the space-dependent Kinetics equation expressed by the matrix exponential and no time difference approximation is used. In this context, our present approach is classified as an explicit method in which no iteration calculation on space and energy is necessary. The Krylov subspace method is used to evaluate the matrix exponential observed in the solution of the spatially discretized space-dependent Kinetics equation. The Krylov subspace method is implemented into a space-dependent Kinetics solver. In order to examine the effectiveness of the Krylov subspace method, the TW...
Go Chiba - One of the best experts on this subject based on the ideXlab platform.
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sensitivity analysis for Reactor period induced by positive reactivity using one point adjoint kinetic equation
Nuclear Data Sheets, 2014Co-Authors: Go Chiba, Masashi Tsuji, Tadashi NarabayashiAbstract:Abstract In order to better predict a kinetic behavior of a nuclear fission Reactor, an improvement of the delayed neutron parameters is essential. The present paper specifies important nuclear data for a Reactor Kinetics: Fission yield and decay constant data of 86 Ge, some bromine isotopes, 94 Rb, 98 m Y and some iodine isotopes. Their importance is quantified as sensitivities with a help of the adjoint kinetic equation, and it is found that they are dependent on an inserted reactivity (or a Reactor period). Moreover, dependence of sensitivities on nuclear data files is also quantified using the latest files. Even though the currently evaluated data are used, there are large differences among different data files from a view point of the delayed neutrons.
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sensitivity and uncertainty analysis for Reactor stable period induced by positive reactivity using one point adjoint Kinetics equation
Journal of Nuclear Science and Technology, 2013Co-Authors: Go Chiba, Masashi Tsuji, Tadashi NarabayashiAbstract:In order to better predict the kinetic behavior of a nuclear fission Reactor, improvement of delayed neutron parameters is essential. Since it is required to establish a path from the microscopic nuclear data to the macroscopic delayed neutron parameters for the improvement, the present paper identifies important nuclear data for Reactor Kinetics. Sensitivities of the Reactor stable period, which describes Reactor kinetic behavior, to microscopic nuclear data such as independent fission yields, decay constants and decay branching ratios are calculated efficiently by using the adjoint Kinetics equation. Furthermore, nuclide-wise and nuclear data-wise uncertainties of the Reactor stable period are quantified using the variance data given in the nuclear data file, and the nuclear data that require further improvement are identified.The results obtained through the present study are quite helpful, and can be a driving force for further nuclear physics studies.
