The presence of fission and intense radioactivity due to fission fragment decay and to neutron activation of clad, structure, moderator, coolant, etc, causes intense heating. This heat must be converted to a useful form of energy as effectively as possible. The amount of produced heat and eventually power, depends on the magnitude of the heat transfer which can take place in the core without damaging the structure material or the fuel elements. In order to ensure safe conversion of heat to power, the thermal analysis (temperature distribution and its heat transfer rate) of the Reactor core must be performed. Thermal analysis of the core is as important as neutronic considerations and in fact is an integral part of Reactor design. In this paper temperature distribution and subchannel effect of a fuel rod are studied using the COBRA III- C code. Steady state, single-phase flow using cosine heat flux is assumed and thermal analysis is done for both with and without gridspacer cases. The obtained thermal analysis results compared well with the results of the analytical and the experimental data, in without gridspacer case, and the results obtained from the FLUENT code, for both cases. The good agreement observed in both cases.

The influence of fluid flow and heat transfer study within a Pressurized Water Reactor is of significant importance for the demonstration of the Reactor performance and its safety in both steady and transient states. In the present paper, the three dimensional flow distributions in the downcomer and the lower plenum of the Reactor were calculated with the computational fluid dynamics (CFD) codes. Calculations were performed for the VVER-1000, V446 Reactor at the Bushehr Nuclear power plant, Iran. Although CFD codes provide an effective tool for the calculation of flow distribution in the Reactor pressure vessel (RPV), computer capacity puts restrictions on the capacity of CFD calculations. Consequently, simplified models had to be used in simulating the RPV. Nevertheless, our investigation shows a reasonable agreement between the numerical results and the measured data.

Stability monitoring of Nuclear Reactors has been extensively investigated in the last few decades. However, the identification methods and proposed intelligent control systems are not appropriate for real cases [1]. In this study stability analyses of Nuclear Reactors have been done by using a dynamical system approach. We elaborated the relation between Nuclear fission and spatial chaos. The relation among neutron effective multiplication coefficient k, concentration rate of uranium and chemical shim, indicates that the Nuclear self-sustaining chain fission and Nuclear fusion have literally spatial nonlinear dynamical behavior. Stability boundaries are obtained in several three and two-dimensional parameter spaces. We have found the absolutely stable area of Nuclear Reactor, which is markedly different from the formerly mentioned areas.

Present paper deals with the phenomenon known as self-sustained Nuclear fission wave that is set up with establishment of the U-Pu cycle in a fast Nuclear Reactor. The safety arised from self-sustained state of the wave and also 50%the possibility of initiation and evolution of the wave, the system comprised of neutron diffusion equation coupled with burn-up and kinetic equations have been considered. The linearizationmethod used to solve this systemnot only separates diffusion subsystem from reaction one but also simplifies greately dealing with several real hundred days of simulation. Space and time discretizations of the linearized diffusion subsystem were performed by 3D finite element and crank-nicolson methods, respectively. The convergence of equation was carried out using the iterations based on the optimized BiCGStab (L) method. Rung-Kutta (5) method was used to solve the first order differential equations of the reaction subsystem since the concentrations of elements in the reaction subsystem changed slowly. Using the element by element FEMand BiCGStab (L) used for solving the systemdropped the required memory down to 1/10. Considering the long period of the simulation, 3D case of the problem and heavy computations, the OpenMp method was used to parallelize the code written in FORTRAN 95.

The present work investigates an appropriate way to calculate the 1700 atomic density changes in the Reactor operations. To automate this procedure, a computer program has been designed by C#. This program suggests a way to solve this problem which is based on the solution system of differential equations (Bitman) that it is designed according to Runge-Kutta Fehlberg method. The designed software is based on the high ability to calculate the material depletion with constant flux and constant power condition. The software inputs included, Reactor power, computation time, initial and final time, determine of Taylor series order in calculation time depended flux, determination of time unite, specifyingmaterial composition of the Reactor core at initial condition consist of light radioactive material, heavy and fission products, determining the order in the accuracy of calculations, applying the decay constants library, cross section database, the amount of generated thermal energy by various material decay, determining the type of calculations at point of view constant flux or constant power.Finally, the atomic density of light, heavy materials and fission products at various times of Reactor operation were calculated with high accuracy as the out puts of this program. At last, it is worth to say that we proposed a new approach for the use the Runge-Kutta Fehlberg method to compute atomic density changes of material composition of the Reactor core which lead us to achievement a high ability tool to solve the above problem.