[en] Numerical solutions of one-group and one-dimensional neutron transport problems are reported for isotropic, forward, and backward scattering. Numerical solution is carried out by using two different methods, the SGF 'spectral Green's function' method and the DD 'diamond-difference' scheme, to test the accuracy of the results. Results of cell-edge scalar fluxes obtained for both methods are presented in the tables

[en] By using the one-group and one-dimensional neutron transport equation, we discussed the effect of stochasticity on the criticality size and reactivity of a slab reactor. In our analysis we used a sufficiently simple stochasticity to obtain the exact results. It is shown that the stochasticity increased the reactivity and decreased the critical thickness and our conclusions agreed with the previous conclusions

[en] P₁ approximation method is used to solve the spherical transport problems. This method is based on the analytic solution and the use of the P₁ approximation which is obtained by spherical harmonic method in the pseudo-slab equation. We also investigate diffusion coefficient in spherical geometry. Results are very accurate than classic diffusion theory

[en] The solution of one-group discrete ordinates S_N problems with linearly anisotropic scattering in x,y,z -cartesian geometry has been studied by using SGF-CN ''spectral Green's function-constant nodal'' method, developed first by De Barros and Larsen (1990-1992) for one dimensional and two dimensional x,y -cartesian geometries. The solutions of S_N transverse -integrated nodal equations in which only constant nodal volume leakage approximation is made, are obtained by a simple iterative algorithm. Finally, tabulated numerical results of average cell scalar fluxes are provided. (author)

[en] The critical slab problem has been studied in the one-speed neutron transport equation with isotropic scattering by using the first kind of Chebyshev Polynomials. The moment criticality solutions were obtained for the uniform finite slab using Mark and Marshak type vacuum boundary conditions. The results obtained by this approximation are presented in tables which also include the results obtained by the P_N method for comparison. (orig.)

[en] A general theoretical scheme for the U_N method and the evolution of a general eigenvalue spectrum are described. The applicability of U_N method to one dimensional slab geometry neutron transport problems is discussed. The eigenvalue spectrum is calculated for isotropic scattering with different values of the parameter c, the mean number of secondary neutrons per collision, known as the fundamental eigenvalue. Then the critical slab problem has been studied. The critical half thicknesses are computed for different values of c. For the solution, Mark and Marshak boundary conditions are used. Results are obtained for both, U_N and P_N approximations for comparison. (orig.)

[en] The critical slab problem of the one-speed and one dimensional neutron transport equation for an isotropic homogeneous medium was studied by using the Chebyshev polynomial approximation, i. e. U_N method. The results obtained by the U_N method for various c values, using the two type boundary conditions which are known as Mark and Marshak boundary conditions, are presented in tables. The tables also include the results obtained by the P_N method for comparison. (orig.)

No abstract available

[en] The one-speed neutron transport equation which includes isotropic forward and backward scattering has been studied using the T_N approximation. Critical thicknesses for different values of the reflection coefficient R and c values have been calculated. The study shows that the T_N method gives nearly the same results as the P_N approximation in one-dimensional geometry. The results obtained in this study are presented in tables and are compared with P_N approximation results. (orig.)

[en] Since in many cases curvilinear geometry is more appropriate than cartesian geometry for precise modeling of the complex systems for reactor calculation, we have developed the spectral Green's function (SGF) method which is employed to obtain angular and scalar flux distributions in heterogeneous sphere geometry with isotropic scattering. In this study, we showed that the neutron transport problems of homogeneous spheres could be reduced to the solution of plane geometry equation. Finally, some results are discussed and compared with those already obtained by diamond difference scheme to test the accuracy of the results. The agreement is satisfactory. SGF method is very suitable for the numerical solution of the neutron transport equation with isotropic scattering