No abstract

No abstract

No abstract

No abstract

The temperature of the high-pressure regeneration system of power-generating units with VVI~R-1000 reactors is determined by the reliability of the high-pressure heater. Such a system includes two lines, each line containing three heaters. Compartment heaters of the type BV-2000-120A with a stainless steel tube bundle were installed on the experimental fifth power-generating unit of the Novovoronezh nuclear power plant. The series units, starting with the first unit of the Southern Ukraine nuclear power plant, were outfitted with cheaper but less reliable collector heaters with a carbon steel tube bundle. Their heat-exchange surface consisted of ordinary flat spiral-tubular elements, during the operation of which it was found that the entrance sections were subjected to erosion-corrosion. The number of heater failures in the power-generating units with VVt~R-1000 reactors due to erosion-corrosion wear is 60-65% of the total number of failures. The erosion-corrosion wear of the entrance sections of the spirals in some works are called impingement corrosion [I]. When the entrance sections of the heaters are damaged by impingement corrosion, the feed water is fed along a bypass pipeline. The heaters can be switched off instantaneously (within 5 sec), as required in the case of collector failure [2], or more evenly with a decrease (sharp or continuous) temperature of the feed water entering the steam generator. The temperature oscillations arising in the process could give rise to the development of thermocyclic stresses and fatigue-crack growth in the exit collectors of the steam generators. References [3][4][5][6] give no data on the temperature level of the high-pressure regeneration system of power-generating units with VVI~R-1000 reactors. Our objective in the present paper is to analyze the effect of oscillations of the feed-water temperature tf. w, which arise when the heaters are switched off, on the degree of damage to the exit collectors of the steam generators in power-generating units with VVI~R-1000 reactors. The method of analysis consisted of the following. The following were determined over the period from startup of the unit to June-December 1992: daily temperature difference t d = tf.w(n) -tf.w (n + 1); average daily temperature difference Atav = EAta/EN, where EAt d is the sum of all daily temperature differences included in the analysis; EN is the total number of days included in the analysis; average daily temperature difference with amplitude equal to or greater than 10~ Atav hs = EAths/ENhs, where r.Aths is the sum of the daily temperature differences with amplitude equal to or greater than 10~ X;Nf is the total number of days during which the amplitude of the daily temperature differences was equal to or greater than 10~ In addition, we estimated the fraction of the days, included in the analysis, of the total number of days during which the unit was operated (EN/ENop) as well as the average frequency of appearance of daily differences with amplitude Atdhs _> 10~ (,6"hs = ~N/ENhs ). The quanti...