Ion irradiation effects on Cr-coated zircaloy-4 surface wettability and pool boiling critical heat flux.

Journal Article · Sat Aug 01 00:00:00 EDT 2020 · Nuclear Engineering and Design

DOI:https://doi.org/10.1016/j.nucengdes.2020.110581· OSTI ID:1667395

Ali, Amir  ^[1]; Kim, Hyun-Gil ^[2]; Hattar, Khalid Mikhiel ^[3]; Briggs, Samuel ^[2]; Park, Dong Jun ^[2]; Park, Jung Hwan ^[2]; Lee, Youho ^[4]

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1. Univ. of New Mexico, Albuquerque, NM (United States)
2. Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)
3. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
4. Seoul National Univ. (Korea, Republic of)

The concept of coating the currently used nuclear fuel cladding (zirconium-based alloy, typically Zircaloy-4 or Zirc-4) with an oxidation preventive layer is a progressing Accident tolerant Fuel (ATF) candidate alloys. The coated Zirc-4-based alloys could be a solution to suppress undesirable fast reaction kinetics with high-temperature steam. Zirc-4 has been the most preferred cladding material in pressurized water reactors (PWRs). Chromium (Cr) based alloys as a coating material provides excellent corrosion protection and good strength and wear resistance. In this paper, we present the surface wettability measurements and pool boiling Critical Heat Flux (CHF) for Cr-coated Zirc-4 claddings pre- and post-exposure to an ion irradiation environment. The wettability measurements, including static contact angle (contact angle, θ) and average surface roughness (surface roughness, Ra), are introduced for samples of different coating thicknesses (5–30 μm thick). The coatings fabricated by the cold spray of Cr-Al particles to 10 mm × 10 mm × 1.95 mm Zirc-4 substrates. Post fabrication, a Pilgering (cold rolling) process, was applied to finalize the coating thickness and resulted in a significant reduction in surface roughness of initially fabricated rough surfaces. The process produced three distinguished samples 5-μm unpolished (as machined), 5-μm, and 30-μm polished (cold rolled). The measurements are presented for the three surfaces and bare Zirc-4 as a baseline surface. The contact angle analyses were implemented in theoretical models from the literature to predict pool boiling CHF. Pool boiling experiments were conducted to measure the pool boiling CHF values and compare them to the predicted values. Scanning Electron Microscope (SEM) images and Energy Dispersive X-ray Spectroscopy (EDS) analysis was performed to characterize the surfaces for better understanding and interpreting the results. The SEM images showed localized surface damage due to ion irradiation. No recognized change in the measured surface roughness due to ion irradiation. The contact angles of irradiated Cr-coated surfaces are consistently higher (10°) than pre-irradiated surfaces. Decreasing the Cr-coating layer thickness resulted in lower contact angle pre- and post- ion irradiation. The predicted pool boiling CHF using the Kandlikar model is in good agreement with the experimentally measured CHF values within ±12% for all samples.

• Research Organization: Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA); USDOE Office of Science (SC); USDOE Office of Nuclear Energy (NE). Nuclear Energy University Program (NEUP)
• Grant/Contract Number: AC04-94AL85000; NA-0003525; 17-12688
• OSTI ID: 1667395
• Alternate ID(s): OSTI ID: 1693662
• Report Number(s): SAND-2020-8581J; 690039
• Journal Information: Nuclear Engineering and Design, Vol. 362; ISSN 0029-5493
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE