Thermal Diffusion Properties of Moderating Materials H2O dnd D2O in Thermal Nuclear Reactors

Elsadig Omer FaduilDepartment of Physics, Faculty of Science, University of Kordofan, El-Obeid, 51111, SudanEbtisam Abdalla YousifDepartment of Physics, College of Science, Jazan University, P. O. Box 114, Jazan 45142, Kingdom of Saudi Arabia

Vol 10 No 6 (2026): Volume 10, Issue 6, June 2026 | Pages: 26-32

International Research Journal of Innovations in Engineering and Technology

OPEN ACCESS | Research Article | Published Date: 05-06-2026

doi Logo doi.org/10.47001/IRJIET/2026.106002

Abstract

Using techniques method of lines (MOL) and finite difference (FD) methods to solve the neutron diffusion equations (NDE). In this work, a new strategy was proposed to compute the neutron flux, concentration, and temperature properties of thermal nuclear reactors. This approach holds the potential for developing the static diffusion equation for different grades of temperature. Finite difference method was used to convert the parabolic partial differential equations into ordinary differential ones. The converted equations were generalized to consider three different geometries namely: spherical, cylindrical, and Cartesian. Moreover, MOL was used to solve the time-dependent NDE with space-time terms which describe the dynamics of one and two-groups. Results of the study showed that the thermal diffusion properties of moderating materials for H2O found that the diffusion coefficient (D), diffusion length (L) and Albedo β(∞) were 0.142 cm, 2.88 cm, and 0.82, respectively. While the same diffusion properties for heavy water (D2O) were D=0.8 cm, L=100.0 cm, and β(∞)=0.968. Meanwhile, D=0.903 cm, L=50.0 cm, and β(∞)=0.930 for Graphite (14C). The generated results are compatible with other different methods of calculations. This study recommends using the MOL to solve the NDE as it gives more accurate results and practically efficient for thermal neutron reactors.

Keywords

Neutron diffusion equations, Method of lines, Finite difference methods, Neutron flux, Thermal nuclear reactors.


Citation of this Article

Elsadig Omer Faduil, & Ebtisam Abdalla Yousif. (2026). Thermal Diffusion Properties of Moderating Materials H2O dnd D2O in Thermal Nuclear Reactors. International Research Journal of Innovations in Engineering and Technology - IRJIET, 10(6), 26-32. Article DOI https://doi.org/10.47001/IRJIET/2026.106002

References
Weston M. Stacey, Nuclear Reactor Physics, Wiley, New York, (2001).

John C. Lee, Nuclear Reactor Physics and Engineering, Wiley, New York, USA, (2020).

A.S. FakhrulIslam, Modeling Neutron Interaction Inside a 2D Reactor Using Monte Carlo Method,  Master Thesis, University of South Carolina (2019).

P. H. Thien, “Solving Neutron Transport Equation in the Reactor using the Integral Average Derivative Method”, International Journal of Science and Research (IJSR), (2013).

Collins, Benjamin, et al. Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT. Journal of Computational Physics Vol. 326 (2016).

Nahla, Abdallah A., Faisal A. Al-Malki, and Mahmoud Rokaya, Numerical techniques for the neutron diffusion equations in the nuclear reactors, Adv. Stud. Theor. Phys. Vol.  6.14 (2012).

S. Mohammed, Ahmad El-Ajou and Mazen Nairat, Analytical Solution for Multi-Energy Groups of Neutron Diffusion Equations by a Residual Power Series Method, ∑ mathematics, (2019).

R.A. El-Nabulsi, Fractal neutrons diffusion equation: uniformization of heat and fuel burn-up in nuclear reactor, Nucl. Eng. Des.Vol.380 (2021).

S. M. Virginijus, The solution of two-dimensional neutron diffusion equation with delayed neutrons, Informatica Vol. 12.2 (2001).

S. Mohammed, Solution of different geometries reflected reactors neutron diffusion equation using the homotopy perturbation method. Results in Physics Vol.12 (2019).

S. Mohammed, Developing a new approaching technique of homotopy perturbation method to solve two-group reflected cylindrical reactor. Results in Physics Vol. 12 (2019).

R.A. El-Nabulsi, Nonlocal effects to neutron diffusion equation in a nuclear reactor, J. Comp. Theor. Transp. Vol. 49 (2020).

M. Dahmani, A.M. Baudron, J.J. Lautard and L. Erradi, A 3D nodal mixed dual method for nuclear reactor kinetics with improved quasistatic model and a semi-implicit scheme to solve the precursor equations, Annals of Nuclear Energy, Vol. 28, (2001).

K. Kobayashi, A rigorous weight function for neutron kinetics equations of the quasi-static method for subcritical systems, Annals of Nuclear Energy, Vol. 32 (2005).

W. Yaqi, W. Bangerth, and J. Ragusa, Three-dimensional h-adaptivity for the multigroup neutron diffusion equations, Progress in Nuclear Energy Vol. 51.3 (2009).

M. A. Shafii, et al. Characteristics of neutron diffusion coefficient as a function of energy group in the one-dimensional multi-group diffusion equation of finite slab reactor core, Journal of Physics, Conference Series. Vol. 1869.No. 1.IOP Publishing, (2021).

L.M. Grossman and J. P. Hennart, Nodal diffusion methods for space-time neutron kinetics, Progress in Nuclear Energy, 49 (2007).

M. Rafael, et al. A nodal modal method for the neutron diffusion equation. Application to BWR instabilities analysis. Annals of Nuclear Energy Vol.29.10 (2002).

S. Pintor, D. Ginestar and G.Verdu, Time integration of the neutron diffusion equation on hexagonal geometries, Mathematical and Computer Modelling, Vol. 52 (2010).

A.E. Aboanber and Y. M. Hamada, Generalized Runge Kutta method for two- and three-dimensional spacetime diffusion equations with a variable time step, Annals of Nuclear Energy, Vol. 35 (2008).

E. Aboanber and Y. M. Hamada, Computation accuracy and efficiency of a power series analytic method for two- and three-space-dependent transient problems, Progress in Nuclear Energy, Vol. 51 (2009).

M. K. Butler and I. M. Cook, "One Dimensional Diffusion Theory," and A. Hassitt, "Diffusion Theory in Two and Three Dimensions," in H. Greenspan, C. N. Kelber, and D. Okrent, eds., Computing Methods in Reactor Physics, Gordon and Breach, New York (1968).

M. Shqair and E. R. El-Zahar, Analytical solution of neutron diffusion equation in reflected reactors using modified differential transform method. Computational Mathematics and Applications. Springer, Singapore, (2020).

A.Carreño, et al. Spatial modes for the neutron diffusion equation and their computation. Annals of Nuclear Energy Vol.110 (2017).

S. P. Sandra Dulla and P. Ravetto. On the spectrum of the multigroup diffusion equations. Progress in Nuclear Energy Vol.59 (2012).

E. F. Anley, Z. Zheng, Finite difference approximation method for a space fractional convection-diffusion equation with variable coefficients, Symmetry vol.12 (2020).

A.Bernal García, Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses (Doctoral dissertation, UniversitatPolitècnica de València), (2018).

H. Sun, W. Chen and D. Baleanu, Numerical solution of the space-time fractional diffusion equation: Alternatives to finite differences, Proceedings of the 5th Symposium on Fractional Differentiation and its Applications, (2012).

A.Kazuo. New finite element solution technique for neutron diffusion equations. Journal of Nuclear Science and Technology Vol. 17.2 (1980).

D. M. Chiara, I. Sgura, and V. Simoncini, Matrix-oriented discretization methods for reaction-diffusion PDEs: Comparisons and applications, Computers & Mathematics with Applications Vol. 79.7 (2020).

W. E. Schiesser, Numerical Method of Lines Integration of Partial Differential Equations, (Academic Press, San Diego, CA, 1991).

W. E. Schiesser, Computational Mathematics in Engineering and Applied Science, (ODEs, DAEs, and PDEs, CRC Press, Boca Raton, FL 1994).

P. Howard, Partial differential equations in Matlab 7.0. University of Maryland, College Park (MD) (2005).