Critical heat fluxes to water in a vertical annulus in a natural circulation system

The influences of sub-cooling and pressure on the values of the critical heat flux in a natural circulation system were investigated. The system used could simulate an experimental loop in a nuclear reactor in case of failure of forced circulation cooling. This work is distinguished from others on the same subject in that a short heating element of large diameter and a big water container forming part of the circulation system were used. The critical heat flux (in the range of 103 to 264 W/cm2), the critical power density (in the range of 183 to 468 kW/l), the inlet sub-cooling (in the range of 49 to 111 deg C) and the pressure (in the range of 0.97 to 4 ata) were measured. The data obtained provided the basis for the determination by the least squares method of the empirical correlation: qsub(BO)sup('') = 2.190ΔTsub(sub,i)sup(0.952)psup(0.183) [W/cm2]. This correlation made it possible to dtermine the safety factors for the operation of the loop. Boiling and natural circulation boiling are discussed briefly and the system used for determining the critical heat flux to water in a natural circulation is described