Preliminary Investigation on Convective Heat Transfer of Rod Bundle Flow With Spacer Grid for Nuclear Fuel Assembly Application

Abstract

The convective heat transfer of rod bundle flow with spacer grid was investigated preliminarily for nuclear reactor core application. As the test fluid, the water was used. To simulate the nuclear fuel assembly, 4×4 rod bundle with P/D (=pitch between rods/rod diameter) of ∼1.35 was prepared together with a spacer grid with twist-mixing vane. A single heated section with five thermocouples embedded in the surface along the circumferential direction was installed around the center subchannel. The measurements of wall temperatures were carried out upstream and downstream of spacer grid. For the rod bundle flow at the inlet of spacer grid (i.e., upstream of spacer grid), the wall temperatures at the gap and subchannel centers exhibited the higher and lower, respectively, which was because in the subchannel center, the axial flow velocity became higher, as compared with the gap center. On the other hand, downstream of spacer grid, the rod wall toward the tip of twist-mixing vane showed the lowest temperature in the measurements along the circumferential direction of rod wall. Near the twist-mixing vane, the averaged wall temperature was observed to be remarkably low, which implies that the twist-mixing vane is an effective tool to enhance the convective heat transfer performance. However, along the axial flow direction behind the spacer grid, the averaged wall temperatures became to increase, and the enhancement of convective heat transfer performance by mixing vane faded away.