Abstract
Steady state flow boiling experiments were conducted on a technically smooth Inconel 625 tube with outer diameter 9.1 mm at inlet pressures 131, 220 and 323 kPa, inlet temperatures 62, 78 and 94 °C and approximately 400, 600 and 1000 kg/(m2.s) mass flow. Water of these parameters was entering into the vertically aligned annulus, where the uniformly heated tube was placed until the critical heat flux (CHF) appeared. The experimental data were compared to estimations of CHF by local PGT tube correlation and Groeneveld’s look-up tables for tubes. The results imply that in the region of low pressure and low mass flux, the differences between calculations and experiments are substantial (more than 50 % of CHF). The calculations further imply that look-up tables and tube correlations should be corrected to the annulus geometry. Here, the Doerffer’s approach was chosen and led to a substantial enhancement of CHF estimation. Yet, a new correlation for the region of low pressure and flow is needed.
Highlights
Current trend in cladding materials of nuclear fuel is an accident tolerant fuel which is designated to replace the current zirconium-based alloy
The key factor is the critical heat flux (CHF) phenomena which lead to a fast reduction of heat transfer coefficient with a steep rise of cladding temperature which can induce the rupture of the heated material
The flow boiling experiments were performed on the Mobile Research Critical Heat Flux Apparatus (MRCHA) which was constructed and operated by the University of Technology in Brno at Faculty of Electrical Engineering and Communication and its thorough description can be found in the article [21]
Summary
Current trend in cladding materials of nuclear fuel is an accident tolerant fuel which is designated to replace the current zirconium-based alloy. Towards making the ATF cladding material heat transfer analysis, the experimental facility with a primary database of different materials is needed. Heat transfer characteristics of cladding material focuses mainly on cooling the heated fuel tubes and since the fuel is cooled mostly by water, a thermohydraulic analysis needs to be done. In this analysis, the key factor is the critical heat flux (CHF) phenomena which lead to a fast reduction of heat transfer coefficient with a steep rise of cladding temperature which can induce the rupture of the heated material. Better understanding of the boiling process is important for new fuel design and crucial for nuclear power plant safety analyses
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