Modeling and simulation of coupled nuclear heat energy deposition and transfer in the fuel assembly of the Ghana Research Reactor-1 (GHARR-1)

• Published in: Nuclear engineering and design Vol. 241; no. 12; pp. 5183 - 5188
• Main Authors: Ameyaw, Felix, Ayensu, Akwasi, Akaho, E.H.K.
• Format: Journal Article
• Language: English
• Published: Amsterdam Elsevier B.V 01.12.2011; Elsevier
• Subjects: Applied sciences; Assembly; Computer simulation; Controled nuclear fusion plants; Energy; Energy distribution; Energy use; Energy. Thermal use of fuels; Exact sciences and technology; Fission nuclear power plants; Fuels; Installations for energy generation and conversion: thermal and electrical energy; Nuclear engineering; Nuclear fuels; Nuclear power generation; Nuclear reactor components; Conduction; Energetic value; Coolant; Energy distribution; Aluminium; Spatial variation; Fuel pin; Modeling; Pressure; Steady state; Nuclear reactor; Kinetic model; Thermodynamics; Nuclear energy; Mass flow rate; Nucleate boiling; Maximum temperature; Safety; Melting point; Deposition; Thermal neutron; Research reactor; Heat transfer; Reactor core
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2011.09.013

► We model heat energy distribution without exceeding thermal limits ► We ascertain the hottest fuel rod is within design limits. ► Axial fuel rod heat energy increases until maximum. ► Radial energy profile suggest the hottest region in the core. ► We model convective heat transfer processes of the core. Monte Carlo N-Particle (MCNP) code coupled with PLTEMP/ANL code were used to model and simulate the heat transfer problems in the fuel elements assembly of the Ghana Research Reactor-1 (GHARR-1) by solving Boltzmann transport approximation to the heat conduction equation. Coupled neutron radiation-thermal codes were used to determine the spatial variations of thermal energy in the fuel channels, the heat energy distribution in the radial and axial segments of the fuel assembly and the convective heat transfer processes in the entire core of the reactor. The thermal energy at maximum reactivity load of 4 mk, reactor power of 30 kW and inlet system pressure of 101.3 kPa were found to be 8.896 × 10 −16 J for a single fuel pin, and 1.104 × 10 −15 J and 7.376 × 10 −16 J, for the radial and axial sectioning of the core respectively. Using the PLTEMP/ANL V4.0 code and given that the inlet coolant temperature was 30 °C, the maximum outlet coolant temperature was 51 °C. The energy values were obtained using the following thermodynamic parameters as maximum pressure drop of 0.7 MPa and mass flow rate of 0.4 kg/s. Neutronics point kinetics model and Safety Analysis Report used to validate the results confirmed that the heat distribution in the core did not exceed 100 °C. The heat energy profiles based on the data suggested no nucleate boiling at the simulated energies, and since the melting point of U–Al alloy fuel material is 640 °C, the reactor was considered to be inherently safe during normal or steady state operations.