Critical Power Experiment with a Tight-Lattice 37-Rod Bundle

Abstract

Since most of critical power or CHF data have been collected in tube, annulus, or BWR geometries under BWR flow conditions, critical power data for highly tight and triangular lattice bundles under low mass velocity are indispensable for thermal-hydraulic design of Reduced-Moderation Water Reactor. Large-scale thermal-hydraulic experiments which use a basic 37-rod bundle test section (rod diameter: 13.0 mm, gap width between rods: 1.3 mm) were therefore carried out in this study within range of 2-9 MPa in pressure and 150-1,000 kg/(m 2 Ás) in mass velocity. Fundamental characteristics of boiling transition were investigated through effects of flow parameter on critical power and those of rod number. It was confirmed that the fundamental characteristics in 37-rod bundle are similar to those in 7-rod bundle and in case of the BWR geometry. The results of the transverse non-uniform power distribution test and subchannel analysis suggest that the critical power becomes higher when the transverse local quality distribution closes to uniform.

Effect of Rod Bowing on Critical Power Based on Tight-Lattice 37-Rod Bundle Experiments

“…The rod diameter of the test sections was 13.0 mm, and the gap between rods was 1.3 mm. These values were the same as those of the basic test section 15) without rod bowing. The flow channel was surrounded by a flow shroud that was made of stainless steel.…”

“…Bowing Rod Power supply system (A-F units) 15,16) Effect of Rod Bowing on Critical Power Based on Tight-Lattice 37-Rod Bundle Experiments stable condition was established, the heater power was increased step by step. The increment of the power input near the boiling transition (BT) condition was smaller than 0.2%.…”

“…The results of the above tests using the test sections with 1.0 and 1.3 mm gap widths were reported in other papers. 15,16) In this paper, we describe the experimental results of the effect of rod bowing on critical power under II. Experimental Method Figure 2 is a schematic view of the test loop mainly consisting of the test section, separator, pressurizer and pump.…”

“…The working fluid was purified water that was deionized and degassed, lowering the electric conductivity below 150 mS/m. The Table 2 Test items of large-scale thermal-hydraulic experiments 15,16) Type of test section Base case 15) Gap width effect 16) Rod bowing effect Rod diameter: 13 Fig. 1 Design of reactor, core and fuel assembly for FLWR 3) water was circulated in the test loop using the pump, and the error in the flow rate measurement was within AE0:5%.…”

Enhancement of Critical Heat Flux in Tight Rod Bundle with Wire Spacer

“…CHF behavior in tight rod bundle with grid spacer was investigated experimentally for tight 37-rods bundle using Freon-12 [5] and water [6] as a coolant. The results with Freon-12 showed that, the effects of pressure, mass flux and steam quality on CHF were similar to those of data in round tube channel [7].…”