Corrosion reduction of aluminum alloys in flowing high-temperature water

Report describes a technique for reducing the corrosion rate of aluminum by adding colloidal substances in a closed-loop system. Experimental work shows that the addition of graphite and colloidal hydrated aluminum oxide significantly reduces the corrosion rate in flowing high-temperature water

Study made of corrosion resistance of stainless steel and nickel alloys in nuclear reactor superheaters

Experiments performed under conditions found in nuclear reactor superheaters determine the corrosion rate of stainless steel and nickel alloys used in them. Electropolishing was the primary surface treatment before the corrosion test. Corrosion is determined by weight loss of specimens after defilming

Corrosion of Some Reactor Materials in Dilute Phosphoric Acid

Corrosion tests in dilute phosphoric acid (pH 3.5) at elevated temperature are described for X8001 aluminum, 18-8 stainless steels, aluminized carbon steel, and Zircaloy. In a 307-day dynamic test at 18 ft/sec and 315 deg C, X8001 aluminum corroded at a rate of 1/2 mdd for the first 240 days. In subsequent exposures, the corrosion rate increased, but the total average penetration at 307 days was only 0.0005 inch. At 200 days, the total corrosion in this test was one-fiftieth that in distilled water. Static tests at 225 deg C gave corrosion rates too low to measure (<0.2 mdd). Of several different 18-8 stainless steels tested in this solution at 315 deg C, only sensitized type 316 suffered intergranular attack. General attack rates of the other samples, of the order of 1/4 mdd, were obtained for the period from 94 to 186 days. Although this is much larger than the rate in distilled water, it represents a penetration rate of only about 5 x 10/sup -//sup 5/ inch/year. Aluminized carbon steel did not suffer rapid corrosion in this solution at 315 deg C, even when large areas of the carbon steel were exposed. There was a tendency for corrosion to separate the steel and aluminum with some specimens, depending on the heat treatment. Zircaloy-2 and Zircaloy-3 corrosion were of the same order in this solution at 315 deg C as in water. (auth

CORROSION OF SOME REACTOR MATERIALS IN DILUTE PHOSPHORIC ACID

Corrosion tests in dilute phosphoric acid (pH 3.5) at elevated temperature are described for X8001 aluminum, 18-8 stainless steels, aluminized carbon steel, and Zircaloy. In a 307-day dynamic test at 18 ft/sec and 315 deg C, X8001 aluminum corroded at a rate of 1/2 mdd for the first 240 days. In subsequent exposures, the corrosion rate increased, but the total average penetration at 307 days was only 0.0005 inch. At 200 days, the total corrosion in this test was one-fiftieth that in distilled water. Static tests at 225 deg C gave corrosion rates too low to measure (<0.2 mdd). Of several different 18-8 stainless steels tested in this solution at 315 deg C, only sensitized type 316 suffered intergranular attack. General attack rates of the other samples, of the order of 1/4 mdd, were obtained for the period from 94 to 186 days. Although this is much larger than the rate in distilled water, it represents a penetration rate of only about 5 x 10/sup -//sup 5/ inch/year. Aluminized carbon steel did not suffer rapid corrosion in this solution at 315 deg C, even when large areas of the carbon steel were exposed. There was a tendency for corrosion to separate the steel and aluminum with some specimens, depending on the heat treatment. Zircaloy-2 and Zircaloy-3 corrosion were of the same order in this solution at 315 deg C as in water. (auth

CORROSION OF ALUMINUM AND ITS ALLOYS IN SUPERHEATED STEAM

The corrosion behavior of pure aluminum and some of its alloys in superheated steam was found to depend markedly on the method of starting the corrosion test. Pure aluminum samples survived only in tests that were brought to temperature and pressure very rapidly. Resistant Al-- Ni-- Fe alloys performed well only if a relatively slow starting procedure was used, suffering extensive blistering or complete disintegration in a test started rapidly. Over the range of temperature and pressure investigated, 400 to 540 deg C and 150 to 600 psig, with optimum starting conditions both pure aluminum and resistant Al-- Ni-- Fe alloy samples quickly formed a very protective oxide film. Interference colors were noted for exposures of several weeks. Samples surviving a 260-day test at 540 deg C and 600 psig had less than 1-mg/cm/sup 2/ weight gain. Nonresistant alloys disintegrated in short corrosion exposures. A penetrating attack, initiated in only a few spots, rapidly destroyed the samples. The effects of composition, dispersion of second-phsse compounds, hydrogen porosity, and pretreatments were investigated for 5.6% Ni--0.3% Fe-0.1% Ti in 540 deg C, 600-psig steam. It was concluded that porosity produced by corrosion product hydrogen was a major factor in the survival of samples. A mechanism for the rapid penetrating attack was proposed as based on observations made during the study of hydrogen porosity. Pretreatment of resistant alloy samples in dry air at 540 deg C or in high-temperature water at 350 deg C greatly reduced the amount of porosity produced by corrosion in superheated steam. (auth

Environmentally assisted cracking in light water reactors. Semiannual report, July 1998-December 1998.

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to {approx}0.3 and 0.9 x 10{sup 21} n {center_dot} cm{sup -2} (E &gt; 1 MeV) in helium at 289 C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to {approx}0.3 x 10{sup 21} n {center_dot} cm{sup -2} in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments

Stress corrosion cracking behavior of irradiated model austenitic stainless steel alloys.

Slow-strain-rate tensile tests (SSRTs) and posttest fractographic analyses by scanning electron microscopy were conducted on 16 austenitic stainless steel (SS) alloys that were irradiated at 289 C in He. After irradiation to {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} and {approx}0.9 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E &gt;1 MeV), significant heat-to-heat variations in the degree of intergranular and transgranular stress corrosion cracking (IGSCC and TGSCC) were observed. Following irradiation to a fluence of {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2}, a high-purity laboratory heat of Type 316L SS (Si {approx} 0.024 wt%) exhibited the highest susceptibility to IGSCC. The other 15 alloys exhibited negligible susceptibility to IGSCC at this low fluence. The percentage of TGSCC on the fracture surfaces of SSRT specimens of the 16 alloys at {approx}0.3 x 10{sup 21} n{center_dot}cm{sup {minus}2} (E &gt; 1 MeV) could be correlated well with N and Si concentrations; all alloys that contained &lt;0.01 wt.% N and &lt;1.0 wt. % Si were susceptible, whereas all alloys that contained &gt;0.01 wt.% N or &gt;1.0 wt.% Si were relatively resistant. High concentrations of Cr were beneficial. Alloys that contain &lt;15.5 wt.% Cr exhibited greater percentages of TGSCC and IGSCC than those alloys with {approx}18 wt.% Cr, whereas an alloy that contains &gt;21 wt.% Cr exhibited less susceptibility than the lower-Cr alloys under similar conditions

Irradiation-assisted stress corrosion cracking of austenitic stainless steels: Recent progress and new approaches

Irradiation-assisted stress corrosion cracking (IASCC) of several types of BWR field components fabricated from solution-annealed austenitic stainless steels (SSs), including a core internal weld, were investigated by means of slow-strain-rate test (SSRT), scanning electron microscopy (SEM), Auger electron spectroscopy (AES), and field-emission-gun advanced analytical electron microscopy (FEG-AAEM). Based on the results of the tests and analyses, separate effects of neutron fluence, tensile properties, alloying elements and major impurities identified in the American Society for Testing and Materials (ASTM) specifications, minor impurities, water chemistry, and fabrication-related variables were determined. The results indicate strongly that minor impurities not specified by the ASTM-specifications play important roles, probably through a complex synergism with grain-boundary Cr depletion. These impurities, typically associated with steelmaking and component fabrication processes, are very low or negligible in solubility in steels and are the same impurities that have been known to promote intergranular SCC significantly when they are present in water as ions or soluble compounds. It seems obvious that IASCC is a complex integral problem which involves many variables that are influenced strongly by not only irradiation conditions, water chemistry, and stress but also iron and steelmaking processes, fabrication of the component, and joining and welding. Therefore, for high-stress components in particular, it would be difficult to mitigate IASCC problems at high fluence based on the consideration of water chemistry alone, and other considerations based on material composition and fabrication procedure would be necessary as well

Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials