Station blackout accidents increasingly become the focus of research in the field of nuclear safety after Japan’s Fukushima nuclear plant accident in March 2011. Core decay heat under station blackout condition will be transferred by natural circulation occurring between core and passive heat exchanger for the nuclear plants incorporated passive safety design concept such as AP1000 or CAP1400. As a result, response of safety systems will differ in accident sequence and kind between passive safety plant and traditional plant. What is more, cooling capacity of passive heat exchanger (PHX) which takes on heat sink has significant effect on performance of natural circulation in passive safety system.
The safety need that characteristics of passive safety plant should be verified through integral experiment facility makes scaling analysis important in design or modification of experiment facility. Furthermore, scaling analysis of natural circulation phenomena under station blackout accident plays an important role in design verification, safety review verification or thermo-hydraulic program development. It not only determines the similar similarity criteria between the nuclear power plant prototype and test facility, but also provides technical basis for selecting different experiment schemes.
As a part of scaling analysis on natural circulation phenomena for station blackout, the cooling capacity of PHX in test facility should be scaled properly and reasonably with conservatism. Therefore, scaling of passive residual heat removal (PRHR) heat exchanger under station blackout accident is investigated analytically in this paper. The analytical model for natural circulation in passive heat exchanger is established based on the performance characteristics of PRHR system in passive plant. By proper hypothesis and simplification, the governing equations for PHX are normalized using steady-state solutions, initial or boundary conditions. The similarity criteria that should be preserved between PHXs in test facility and prototype are finally obtained from non-dimensionalized equations. Furthermore, the distortion analysis for PHE design is also investigated based on the similarity criteria for selected scaling factors and parameters. The safety analysis based on models of nuclear power plant prototype and test facility is conducted on transient performance of designed PHX with PHX of prototype. The results show that: heat source number is the dominant similarity criteria for PHXs design under SBO condition. Requirements of Richardson number and friction number could be satisfied by resistance adjusting on test loop. The performance of PHX designed following heat source number requirement can better represent the transient response characteristics of prototype under SBO condition.