Abstract
The main feature of an NPP steam generator (SG) is that in addition to steam generation, it shall reliably and constantly cool the NPP reactor’s core. The SG shall also comply with a stringent requirement of ensuring tightness of the primary and secondary circuits with respect to each other; i.e., its design shall exclude the possibility of damage to the heat-transfer and other elements loss of tightness of which entails ingress of radioactive primary coolant into the NPP steam--water circuit with the subsequent possibility of radioactive products releasing into the environment. Thus, ensuring reliable operation and design life of SG heat-transfer tubes (HTTs) is among top-priority objectives for different types of NPPs used both in the domestic nuclear power industry and abroad. Thin-walled SG HTTs constitute an important part of the primary circuit boundary, and in order to perform the function of an efficient barrier, HTTs shall not have through defects or defects generating the need to blank off them. The article describes methods used to predict blanking off heat-transfer tubes and the residual technical life of the SG tube bundle used at an NPP equipped with a water-cooled water-moderated power-generating reactor (VVER) and presents the results from calculating the permissible average value of chloride ion concentration during the operation. To assess the SG operation time to failure, a computer program has been written in the MathCad environment, using which the above-mentioned assessment can be performed within an acceptable timeframe and with sufficient accuracy. A conclusion has been drawn about the need to adopt more stringent requirements for the chloride ion concentration in the SG blowdown water to avoid the need of blanking off leaky tubes damaged as a result of stress corrosion cracking.
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