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[en] In this paper, we propose a numerical methodology for the development of a method of the spectral nodal class that generates numerical solutions free from spatial truncation errors. This method, denominated Spectral Deterministic Method (SDM), is tested as a study of the solutions (spectral analysis) of neutron transport equations in the discrete ordinates (SN) formulation, in slab geometry, multigroup approximation, with linearly anisotropic scattering, considering a heterogeneous domain with fixed-source. The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values calculated for these quantities coincide with the analytic solution of the equations. These numerical results are shown and compared with the traditional fine-mesh Diamond Difference (DD) method and the coarse-mesh spectral Green's function (SGF) method to illustrate the method's accuracy and stability. The solution algorithms problem is implemented in a computer simulator made in C++ language, the same that was used to generate the results of the reference work. (author)
[en] Throughout history, energy has played a fundamental role in human's progress living. To promote nuclear power to meet the future energy needs, ten countries including Argentina, South Africa, the United States, the United Kingdom, Brazil, Japan, Switzerland, France, Canada and Korea in a global effort (Generation IV International Forum - GIF) have agreed to investigate the next generation of nuclear energy systems known as 4 generation. These reactors are expected to enter the market after 2030. Fundamental changes in the configuration of the systems and the forms of the old reactors have led to the production of new reactors, which require basic research and development, careful examination, and the construction of semi-industrial units. The capabilities of fourth-generation reactors are seawater desalination, and thermal applications in addition to the production of electricity. In 2000, the founding countries of GIF formed their first meeting to discuss the need for conduct research on the design of next-generation reactors. Subsequently, a strategy was put forward to direct the activities, and the implementation responsibility was assigned to the US Department of Energy. In this research, we investigate the neutron behavior of the advanced reactor core with lead coolant ALFRED. The purpose of the neutron calculations of the core of a reactor is to calculate the distribution of neutron flux in the center and to calculate the effective reproduction coefficient. Given the necessity of performing lattice pitch neutron calculations, it is initially required to determine the real geometry of the core, as well as the order and fuel richness, the lattice pitch the grid, the radius and height of the fuel rods, the composition and location of the fuel absorbents, the types and locations of the control rods, the fuel complex arrangement, and radial and axial peaking factor. The MCNPX code is used to perform neutron calculations.
[en] In this paper, we propose a new deterministic numerical methodology to solve the one-dimensional linearized Boltzmann equation applied to neutron shielding problems (fixed-source), using the transport equation in the discrete ordinates formulation (SN) considering the multigroup theory. This is a hybrid methodology, entitled Modified Spectral Deterministic Method (SDM-M), that derives from the Spectral Deterministic Method (SDM) and Diamond Difference (DD) methods. This modification in the SDM method aims to calculate neutron scalar fluxes with lower computational cost. Two model-problems are solved using the SDM-M, and the results are compared to the coarse-mesh methods SDM, Spectral Green's Function (SGF) and Response Matrix (RM), and the fine-mesh method DD. The numerical results were obtained in the programming language JAVA version 1.8.091. (author)
[en] In order to improve the convergence behavior of the fixed-point iteration (Picard iteration) for neutronics/thermal-hydraulics coupled problems, Anderson acceleration is implemented in a pin-wise whole core analysis code nTER/ESCOT. The fixed-point map of Anderson acceleration is established for serially coupled whole core transport code and subchannel code. The performance of Anderson acceleration is examined with single assembly problems having low and high boron concentrations. The Anderson scheme shows the comparable convergence behavior to that of the optimum fixed-point iteration with under-relaxation factor. (author)
[en] A new approach for the application of the coarse-mesh Modified Spectral Deterministic method to numerically solve the two-dimensional neutron transport equation in the discrete ordinates (𝑺𝑵) formulation is presented in this work. The method is based on within node general solution of the conventional one-dimensional 𝑺𝑵 transverse integrated equations considering constant approximations for the transverse leakage terms and obtaining the 𝑺𝑵 spatial balance equations. The discretized equations are solved by using a modified Source Iteration scheme without additional approximations since the average angular fluxes are computed analytically in each iteration. The numerical algorithm of the method presented here is algebraically simpler than other spectral nodal methods in the literature for the type of problems we have considered. Numerical results to two typical model problems are presented to test the accuracy of the offered method. (author)
[en] The BEAVRS benchmark problem was solved by a newly developed whole core transport code, nTER (Neutron Transport Evaluator for Reactor) to verify its code systems through its core follow calculations. The nTER results for control rod worth, radial detector signal, and boron letdown curve during two cycles agree well with the measured data. Therefore, it is concluded that the nTER code is well developed in the terms of the solution accuracy for the high-fidelity nuclear parameter evaluation. (author)
[en] In new generation PWR the annular fuel has been proposed as one of the options to achieve higher power density, larger safety margin and reduced electricity generation cost. In the current work, RELAP5 code is used to compare the thermal hydraulic parameters for both solid fuel and internally and externally cooled annular fuel in a core of a PWR. MCNP6 code is used to evaluate the neutronic design and basic safety parameters of the annular fuel. To accomplish this goal, RELAP5 input files for both solid and annular fuels are developed. In these files, a 13×13 array annular fuel design is used while the 17×17 standard array design is used for the solid fuel. A 100 % core power, steady state normal operation is assumed in the current investigation. Also, MCNP6 code input files for both fuels are prepared. It is found that annular fuel has lower peak fuel temperature than the standard solid fuel, which is an important advantage of the annular fuel rather than the solid one. Also, comparisons were performed for reactivity feedback coefficients of the two fuel types at BOC. Burnup calculations were performed to study the energy conversion capability of the annular fuel as well as rim effects.
[en] An accurate analysis model for transient reactor behavior is necessary to keep sufficient safety margins of nuclear power plants to prevent cliff edge effects. In this study, the direct response matrix (DRM) method is applied to the kinetic domain and the transient analysis is enabled based on the transport equation. The kinetic DRM model introduces the time delay to four sub-response matrices. The time delay can be evaluated by a Monte Carlo calculation. The model is evaluated in homogeneous and heterogeneous problems. The Doppler feedback is considered in the heterogeneous problem and the calculation results are compared with the experimental data. The calculation results indicate that the calculation step 1.0E-7 s is sufficient for the model and the model provides results in good agreement with the experimental data. It is concluded that the present model with the DRM method can be used for transient analysis. (author)
[en] This work describes the procedure carried out to perform a hybrid parallelization (MPI-OpenMP) in the AZTRAN neutron code which is programmed in Fortran 90. The code numerically solves the neutron transport equation in Cartesian geometry using discrete ordinate method. In recent years, the increase in computational resources has allowed the trend to parallelize numerical simulation codes in order to reduce execution times. The parallelization of the transport equation is achieved with the decomposition into domains of the independent variables with the idea of associating the decomposition into sub domains, which are distributed in the CPUs of the computer equipment. The sub domains are solved independently and at the end of the calculation they are communicated to update the complete domain solution. In the code, the decomposition was done in the variables spatial (MPI interface) and energy (OpenMP interface) obtaining a hybrid parallel version. The benchmark of the C5G7 MOX fuel assembly in steady state was simulated, obtaining solutions close to the reference, and highly satisfactory results were found in relation to the acceleration obtained by the hybrid parallelization. The confidence gained encourages the use of the code as a domestic tool for the analysis of nuclear reactors. (author)
[en] The effect for different types of scattering on the critical half thickness in slab geometry for one speed neutron transport equation is studied for isotropic, linearly anisotropic and quadratic anisotropic scatterings. An extensive numerical survey is carried out for the critical thickness in order to provide the effect of the different scattering types. The numerical results are obtained by P, T and U methods. The P method is the Legendre polynomial solution that is accepted as the exact result for the neutron transport theory calculations and the U and T methods are the types of Chebyshev polynomials. Critical thickness values are calculated by using Mark boundary condition. Results are compared with the literature.