PREDICTING THE SERVICE LIFE OF THE HEAT-EXCHANGE BUNDLE OF A STEAM GENERATOR AT A NUCLEAR POWER STATION WITH WATER-MODERATED WATER-COOLED REACTORS WITH RESPECT TO STRESS-CORROSION CONDITIONS.

被引：0

作者：

Sereda, E.V. ^{[1
]}

Gorbatykh, V.P. ^{[1
]}

机构：

[1] Moscow Power Inst MEI, USSR, Moscow Power Inst MEI, USSR

来源：

Thermal Engineering (English translation of Teploenergetika) | 1987年 / 31卷 / 11期

关键词：

NUCLEAR REACTORS; WATER COOLED - Heat Exchangers - STEAM GENERATORS - Heat Exchangers;

D O I：

暂无

中图分类号：

学科分类号：

摘要：

Austenitic chrome-nickel stainless steels are used in the nuclear power industry due to their combination of strength and physico-chemical properties. However, when there are certain corrosive components present in the working media these steels have a susceptibility to corrsion under stress. For the case of stress corrosion of austenitic stainless steels in hot chloride solutions a film dislocation approach to describing the mechanism of the process has been proposed by the authors for predicting the kinetics of failure of the heat exchange tubes of steam generators during the course of an operational campaign. Formulas are given for the dependence of rate of development of a corrosion crack on chloride concentration and electrochemical potential. 12Cr18Ni10T: steel was tested.

引用

页码：640 / 641

共 36 条

[21] Application of the system of automated monitoring of the residual service life for the solution of the problems of operation of nuclear power plants with water-cooled water-moderated reactors
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[22] Monitoring of mechanical properties of vessel materials of water-moderated water-cooled reactors of nuclear power plants for assurance of their safe service
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[23] GASEOUS SATURATION OF THE PRIMARY CIRCUIT COOLANT IN NUCLEAR STEAM SUPPLY-SYSTEMS OF WATER-MODERATED WATER-COOLED POWER-REACTORS
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LASTOCHKIN, AP
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[24] VIBROACOUSTIC DIAGNOSIS OF STATE OF MAIN CIRCULATION PUMPS OF FIRST LOOP IN A NUCLEAR-POWER STATION WITH WATER-MODERATED WATER-COOLED REACTORS
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THERMAL ENGINEERING, 1974, 21 (06) : 27 - 31
[25] INVESTIGATIONS OF STEAM SEPARATION IN STEAM-GENERATORS OF NUCLEAR-POWER STATIONS WITH VVER-440 WATER-MODERATED WATER-COOLED POWER-REACTORS
DMITRIEV, AI
KOZLOV, YV
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VOLKOV, AP
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THERMAL ENGINEERING, 1977, 24 (12) : 25 - 28
[26] INVESTIGATING THE RESISTANCE TO CORROSION OF VARIOUS STRUCTURAL-STEELS IN THE CONDENSATE-FEED CIRCUIT OF A NUCLEAR-POWER STATION WITH WATER-MODERATED WATER-COOLED REACTORS
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SOVETNIKOV, KV
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[27] CALCULATIONS OF CHANGES IN PARAMETERS OF PRIMARY LOOP OF A NUCLEAR-POWER STATION WITH WATER-COOLED WATER-MODERATED REACTORS IN EVENT OF PIPING BURSTS
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[28] BASIC TRENDS IN THE RESEARCH AND DEVELOPMENT OF HORIZONTAL STEAM GENERATORS FOR NUCLEAR POWER STATIONS (AES) WITH WATER-MODERATED WATER-COOLED (VVER) POWER REACTORS.
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Soviet Energy Technology (English translation of Energomashinostroenie), 1986, (05): : 49 - 53
[29] HERMETIC SEAL CONTROL OF PRIMARY CIRCUIT IN STEAM-GENERATORS AT NUCLEAR-POWER PLANTS USING WATER-MODERATED WATER-COOLED REACTORS
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[30] INVESTIGATING THE STEAM GENERATORS OF THE MAIN UNIT ON A NUCLEAR POWER STATION (AES) WITH A WATER-MODERATED WATER-COOLED VVER-1000 REACTOR.
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Soviet Energy Technology (English translation of Energomashinostroenie), 1986, (05): : 54 - 58

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