Effect of PVD-coated chromium on the subcooled flow boiling performance of nuclear reactor cladding materials

Journal Article · Fri May 20 00:00:00 EDT 2022 · Applied Thermal Engineering

DOI:https://doi.org/10.1016/j.applthermaleng.2022.118670· OSTI ID:1976861

Seong, Jee Hyun ^[1]; Wang, Chi ^[2]; Phillips, Bren ^[2]; Bucci, Matteo ^[2]

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1. Massachusetts Institute of Technology (MIT), Cambridge, MA (United States); OSTI
2. Massachusetts Institute of Technology (MIT), Cambridge, MA (United States)

Here we elucidate the separate effect of a thin Cr coating deposited by physical vapor deposition (PVD) on the subcooled flow boiling performance of zircaloy-4. First, we run flow boiling experiments on prototypical zircaloy-4 surfaces mimicking the scratch pattern and surface roughness of nuclear reactor claddings. Then, we PVD-coat a 0.3 µm thick chromium layer on the same exact surface and repeat the same flow boiling investigations. All experiments are run using deionized water at atmospheric pressure, flowing on a 1 × 3 cm² rectangular cross section channel at a rate of 1000 kg/m²/s and a subcooling of 10 K. We measure the average temperature of the boiling surface at increasing surface heat fluxes, covering a wide range of heat transfer regimes, from single-phase forced convection to the boiling crisis. We also record high-speed videos of the boiling process, which we postprocess to measure bubble nucleation site density, growth time, departure diameter and frequency. The surface analysis reveals that, while the chromium coating does not seem change the surface roughness and morphology, it improves surface wettability. However, it decreases the critical heat flux. The chromium coating causes an increase of nucleation temperature, bubble departure diameter and growth time, and a reduction of the nucleation site density. The concurrence of these observations indicates that a size reduction of the nucleation sites, conformally covered by the chromium coating, may be the cause of the boiling performance deterioration. We confirm this hypothesis repeating the same analysis on a FeCrAl sample prepared and tested using the same protocol as the zircaloy-4 sample, but with a different initial surface texture.

• Research Organization: Massachusetts Institute of Technology (MIT), Cambridge, MA (United States)
• Sponsoring Organization: US Nuclear Regulatory Commission (NRC); USDOE Office of Nuclear Energy (NE)
• Grant/Contract Number: NE0008693
• OSTI ID: 1976861
• Journal Information: Applied Thermal Engineering, Journal Name: Applied Thermal Engineering Journal Issue: C Vol. 213; ISSN 1359-4311
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATION
42 ENGINEERING
FeCrAl
PVD
accident tolerant fuel
boiling
chromium
zircaloy-4