United States Nuclear Regulatory Commission - Protecting People and the Environment

Boric Acid Corrosion of Light Water Reactor Pressure Vessel Materials (NUREG/CR-6875, ANL-04/08)

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Publication Information

Manuscript Completed: May 2004
Date Published: July 2005

Prepared by:
J.-H. Park, O.K. Chopra, K. Natesan, and W.J. Shack
Energy Technology Division
Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439

P. Krishnaswamy, D. Rudland, and G.M. Wilkowski
Engineering Mechanics Corporation of Columbus
Columbus, Ohio 43221

William H. Cullen, Jr., NRC Project Manager

Prepared for:
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code Y6722

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Abstract

This report presents experimental data on electrochemical potential and corrosion rates of the materials found in the reactor pressure vessel head and control rod drive mechanism (CRDM) nozzles in boric acid solutions of varying concentrations at temperatures of 95–316°C (203–600°F). Tests were conducted in (a) high-temperature, high-pressure aqueous solutions with a range of boric acid concentrations, (b) high-temperature (150–316°C) H-B-O solutions at ambient pressure, wetted and dry, and (c) low-temperature (≈95°C) saturated, aqueous, boric acid solutions. These correspond to the following situations: (a) low leakage through the nozzle and nozzle/head annulus plugged, (b) low leakage through the nozzle and nozzle/head annulus open, and (c) significant cooling due to high leakage and nozzle/head annulus open. The results indicate significant corrosion only for the low-alloy steel and no corrosion for Alloy 600 or 308 stainless steel cladding. Also, corrosion rates were significant in saturated boric acid solutions, and no material loss was observed in boric acid melts or deposits in the absence of moisture. The results are compared with the existing corrosion/wastage data in the literature.

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