Coolant Tech. of Water-Cooled Reactors Vol 2 [Corrosion in

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Extra resources for Coolant Tech. of Water-Cooled Reactors Vol 2 [Corrosion in Pri Coolant Sys] (IAEA TECDOC-667v2)

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ELECTROCHEMICAL TEST NO MAXIMUM BREAKING STRESS (MPl/klil TIME TO FRACTURE IM 1 602/87 4 188 1 870/1262 25« Ultimate Yield Strength (MPa/Ksi) Fluence 21 3xl0 2 n/cm IxlO18 n/cm2 Test No. Sample Fluence (n/cm, E>1 MeV) Tensile Strength (MPa/Ksi) 859/125 505/73 820/118 195/28 Strain Rate (min ) Elongation (Î) Test Time (h) 18 75 Max. Stress (KPa/Ksi) Fracture Mode (SEM) HC-1 3*1021 4xlO~6 191 710/103 1GSCC (100Ï) HC-2 IxlO18 4xlO~6 408 372/54 Ductile HC-3 21 IxlCf4 25 717/104 IGSCC (402) 4xl

5% The use of carbon steel for feeders, headers, system piping and vessel shells is an inherent feature of PHWRs. 8 at 298 K) and low dissolved oxygen concentration in the primary coolant. Proper hotconditioning of the PHT system surfaces prior to reactor operation, optimum water chemistry control during operation, reduced cobalt content in the system constructional materials and improved system design have been the main contributors towards the achievement of the efficient radiation field control.

Das, "Effect of impurity elements on corrosion behaviour of Zircaloy in high temperature steam". Proc. of a Tech. Comm. Meeting organised by IAEA on 'External Cladding Corrosion in Hater Power Reactors', Cadarache, France. PP 94-98 (Oct. 1985), IAEA, Vienna (1986). 6. Lister, "Review of in-reactor Zircaloy corrosion and crud deposition experience at AECL", EPRI HP-1254 (1979). 7. Sheppard, "Zirconium in Nuclear Industry", ASTM STP 824, p 519 (1982). 8. , "Oxide thickness effects on Zircaloy oxidation and hydriding under irradiation", BNWL Kept.

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