Takayoshi Kusunoki

Effects of fluid properties on CCFL characteristics at a vertical pipe lower end

The purpose of this study is to derive a counter-current flow limitation (CCFL) correlation and evaluate its uncertainty for steam generator (SG) U-tubes in a pressurized water reactor (PWR). Experiments were conducted to evaluate effects of the liquid viscosity on CCFL characteristics using air–40 wt% or air–60 wt% glycerol water solution and saturated steam–water at atmospheric pressure with vertical pipes simulating the lower part of the SG U-tubes. The steam–water experiments confirmed that CCFL characteristics could be expressed in terms of the Wallis parameters (JG* and JL*) for the pipe diameters of D = 14, 20, and 27 mm. A CCFL correlation was derived using the ratio μG/μL of the viscosities of the gas and liquid phases, μG and μL, as a correction term representing effects of fluid properties, where JG*1/2(μG/μL)−0.07 was expressed by a cubic function of JL*1/2(μG/μL)0.1. In the correlation, the constant C indicating the value of JG*1/2(μG/μL)−0.07 at JL* = 0 was (1.04 ± 0.05), and this uncertainty of ±0.05 would cover most of the previous experimental data including the ROSA-IV/LSTF data at 1, 3, and 7 MPa.

Condensation experiments for counter-current flow limitation in an inverted U-tube

In this study, we measured counter-current flow limitation (CCFL) characteristics in an inverted U-tube (18.4 mm diameter and 1.0 m straight-part length) simulating steam generator (SG) U-tubes under conditions of steam condensation at pressures of 0.1–0.14 MPa. Differential pressure ΔP between the top of the inverted U-tube and the lower tank was measured, and the flow patterns wave estimated by comparing the waveforms of ΔP with those in air–water experiments. As a result, we classified the flow patterns under CCFL conditions into CCFL-P, CCFL-L and CCFL-T. The falling water flow rate under CCFL conditions slightly increased as the pressure increased and the cooling water temperature decreased (subcooling of cooling water increased). In the case of CCFL-L, CCFL characteristics in the inverted U-tube were between those in air–water and saturated steam–water experiments at 0.1 MPa. Furthermore, we derived a Wallis type CCFL correlation and its uncertainty from CCFL data, including previously measured data, i.e., J*1/2G + 0.88JL*1/2 = 0.76 ± 0.05.

Prediction of Countercurrent Flow Limitation and Its Uncertainty in Horizontal and Slightly Inclined Pipes

We proposed prediction methods for countercurrent flow limitation (CCFL) in horizontal and slightly inclined pipes with one-dimensional (1-D) computations and uncertainty of computed CCFL. In this study, we applied the proposed methods to a full-scale pressurizer surge line [inclination angle θ = 0.6 deg, diameter D = 300 mm, and ratio of the length to the diameter (L/D) = 63] in a specific pressurized water reactor, performed 1-D computations and three-dimensional (3-D) numerical simulations, and found that uncertainties caused by effects of the diameter and fluid properties on CCFL were small. We also applied the proposed methods to experiments for hot-leg and surge line models (θ = 0 and 0.6 deg, D = 0.03 to 0.65 m, and L/D = 4.5 to 63) to generalize them, performed 1-D computations, and found that uncertainties caused by effects of θ and L on CCFL were large due to the setting error for θ and differences among experiments. This shows that a small-scale air-water experiment with the same θ and L/D as those in an actual plant is effective to reduce the uncertainty of CCFL prediction.

Prediction of temperature and water level in a spent fuel pit during loss of all AC power supplies

A prediction method for water temperature in a spent fuel pit of a pressurized water reactor (PWR) has been developed to calculate the increase in water temperature during the shutdown of cooling systems. In this study, the prediction method was extended to calculate the water level in a spent fuel pit during loss of all AC power supplies, and predicted results were compared with measured values of spent fuel pools in the Fukushima Daiichi Nuclear Power Station. The calculations gave reasonable results, but overestimated the decreasing rate of the water level and the water temperature. This indicated that decay heat was overestimated and evaporation heat transfer from the water surface was underestimated. Results of calculations with 80% decay heat and 155% (Unit 4 pool) or 230% (Unit 2 pool) evaporation heat flux were in good agreement with measured values. The data-fitted evaporation heat fluxes agreed rather well with the evaporation heat transfer correlation proposed by Fujii et al.