An estimation of the void fraction in a tight-lattice rod bundle was needed for the R&D of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR). For this purpose, we measured the void fraction and studied the behaviors of boiling flow. The void fraction was measured by a neutron radiography, a quick-shut-valve technique, and an electro void fraction meter. The data were taken using the 7-, 14-, 19- and 37-rod bundle test sections with the rod gap of 1.0 or 1.3 mm under from atmospheric pressure to 7.2 MPa conditions. A spacer effect test was also carried out. The following estimations were conducted: (1) a similarity of the advanced analysis codes with the 3D void fraction data, (2) the comparisons of the TRAC-BF1 code and a drift-flux model with the 1D data. Followings were made clear: (a) The void fraction becomes lower at the peripheral and higher at the rod gap part of the lower core and at the center of the subchannel of the upper core, (b) the codes calculates the similar distribution to the data, and (c) the TRAC-BF1 and the drift-flux model tends to overestimate the void fraction at the lower quality region, on the other hand at the higher quality, those methods tend to same characteristics to the data. It was confirmed that several special features were existed in the tight-lattice rod bundle but the codes were applicable.