It is usually recognized that a considerable amount of heating power additional to the ohmic heating will be required to reach ignition in a tokamak reactor. To this purpose, we use the method of the injection of wave. In the large tokamaks, poloidal magnetic field is ignorable versus toroidal magnetic field. Since toroidal magnetic field is very large, we can use cyclotron waves to heat plasma. But, in small tokamaks, toroidal magnetic field is very weak and this method is not efficient. Therefore, in small tokamaks, people use current drive methods to confine plasma. Lower hybrid wave have proven excellent at driving plasma current. Steady-state current drive is important for tokamak fusion reactors. A large effort has been devoted to produce Steady-state current drive and reduce product cost. Spherical tokamaks have low ratio aspect, and can confine a higher plasma pressure for a given magnetic field strength. Since in the small tokamaks, the magnitude of the poloidal magnetic field and toroidal magnetic field are in the same order, We can use poloidal magnetic field for plasma confinement. The National Spherical Torus Experiment (NSTX) is an innovative device based on the Spherical Tokamak concept, that is investigated in this thesis. This tokamak itself can produce inductive current and reduce inductive current production costs. In this thesis the results of LSC for NSTX and TFTR tokamak, is discussed. This results indicate higher efficiency of NSTX tokamak and express the fact that we can access to a higher external power with redusing the cost and construct time. Keywords: NSTX Tokamak, Lower Hybrid Waves, Current Drive , Quasilinear Diffusion Coefficient, Quasilinear Power
