Accelerated Corrosion Tests of Nuclear Reactor Pressure Vessel Materials in NaCl–H₃BO₃ Solutions

Abstract

Following the 2011 accident at the Fukushima Daiichi nuclear power station, sea water for cooling and boric acid for maintaining a noncritical condition, both corrosive liquids, were injected into nuclear pressure vessels. In order to estimate corrosive characteristics of the pressure vessels an experimental study was undertaken to provide an accelerated corrosion test on SA533B low alloy steel and Inconel 600, materials used in the construction of the pressure vessels. In a typical experiment, samples of these materials were immersed in saturated NaCl and concentrated H 3 BO 3 aqueous solutions at a temperature of 423 K. SA533B suffered little or no corrosion in saturated NaCl solution, significant corrosion in concentrated H 3 BO 3 solution and substantial corrosion in the binary saturated NaCl-concentrated H 3 BO 3 solution. Galvanic corrosion of SA533B was accelerated when Inconel 600 was also immersed in the same solution and the two samples were electrically connected either externally by a wire lead or internally by a screw made of SA533B or both. Corrosion rate in the initial stage was 0.07 mm per hour. The corrosion product on SA533B was porous and easily detachable, indicating corrosion to be progressive without producing a stable protective corrosion layer. The validity of the extreme experimental condition for accelerated corrosion tests is discussed and experimental programs for further investigation are proposed.

Low Activation-Modified High Manganese-Nitrogen Austenitic Stainless Steel for Fast Reactor Pressure Vessel Cladding

“…The purpose of the clad is to protect the carbon steel from the environment in the vessel such as high temperature and corrosion. [2][3] Austenitic stainless steel containing nickel and Inconel alloys face a large numbers of problems such as He-embritllement and long lived nuclides as a waste material. [4][5] Hence, the alloys containing nickel should be minimized or eliminated from the reactor core.…”

Corrosion Mechanism of Low-Alloyed Steel in High-Temperature Water: Effect of Additives and Time of Exposure

“…The combination of aggressive environment and mechanical effects could lead to different forms of corrosion-mechanical degradation, such as strain-induced corrosion cracking (SICC). [3][4][5][6][7][8][9][10][11][12] The effect of chloride and sulfate in the bulk coolant on the SICC rates is well documented from laboratory work, [13][14][15][16][17][18][19][20][21][22][23][24][25] even if practically no cases of corrosion-mechanical degradation are reported from plant experience.…”