Specific features of corrosion damage to heat-transfer tubes of steam generators used at nuclear power stations equipped with VVER-1000 reactors

2009 ◽  
Vol 56 (7) ◽  
pp. 612-616 ◽  
Author(s):  
D. S. Nemytov ◽  
V. F. Tyapkov
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Corrosion Damage ◽  
Steam Generators ◽  
Power Stations

Construction Aspects of the Doel 1 and 2 SGR Projects

2009 ◽  
Author(s):  
Bert Kroes ◽  
Edmond Gobert ◽  
Xavier Delhaye ◽  
Peter Devolder ◽  
Michel Sonville
Keyword(s):  
Nuclear Power ◽  
Steam Generators ◽  
Primary Coolant ◽  
Worker Exposure ◽  
Power Stations ◽  
Exchange Operation ◽  
Radiation Worker ◽  
Opening And Closing ◽  
Leak Before Break ◽  
Specific Construction

The Doel 1 and 2 PWR Nuclear Power Stations are the oldest commercially operating units in Belgium and the last to replace their two Steam Generators. The Doel 2 Steam Generators were replaced in 2004 and those of Doel 1 will be replaced late 2009. The replacement poses a particular challenge as these are the only stations in Belgium requiring the creation of primary and secondary containment opening for the SG exchange operation. Other construction challenges result from the a-typical SG support configuration which dates from the period well before the more or less standardized support configuration as used for later PWR units. The current paper discusses the construction approaches selected to facilitate the exchange operation and to minimize the outage duration and radiation worker exposure. The main particularities of the construction effort concern the secondary containment opening and closing using a structural formwork assembly, the use of containment platforms hanging inside the primary containment allowing for parallel primary and secondary containment reconstruction and the de-activation of some of the primary coolant piping and SG restraints following the licensing acceptance of the Leak Before Break concept for the primary piping. The specific construction options that made the Doel 2 replacement a success will be presented in this paper.

Electroslag Strip Cladding of Steam Generators With Alloy 690

2006 ◽  
Author(s):  
M. Consonni ◽  
F. Maggioni ◽  
F. Brioschi
Keyword(s):  
Nuclear Power ◽  
Filler Metal ◽  
Hot Cracking ◽  
Test Results ◽  
Steam Generators ◽  
High Productivity ◽  
Cladding Layer ◽  
Power Stations ◽  
Alloy 690 ◽  
Submerged Arc

The present paper details the results of electroslag cladding and tube-to-tubesheet welding qualification tests conducted by Ansaldo-Camozzi ESC with Alloy 690 (Alloy 52 filler metal) on steel for nuclear power stations’ steam generators shell, tubesheet and head; the possibility of submerged arc cladding on first layer was also considered. Test results, in terms of chemical analysis, mechanical properties and microstructure are reproducible and confidently applicable to production cladding and show that electroslag process can be used for Alloy 52 cladding with exceptionally stable and regular operation and high productivity. The application of submerged arc cladding process to the first layer leads to a higher base metal dilution, which should be avoided. Moreover, though the heat affected zone is deeper with electroslag cladding, in both cases no coarsened grain zone is found due to recrystallisation effect of second cladding layer. Finally, the application of electroslag process to cladding of Alloy 52 with modified chemical composition, was proved to be highly beneficial as it strongly reduces hot cracking sensitivity, which is typical of submerged arc cladded Alloy 52, both during tube-to-tubesheet welding and first re-welding.

scholarly journals Local corrosion of austenitic steels and alloys for heat exchanger tubes of nuclear power stations steam generators

2019 ◽  
Vol 75 (2) ◽  
pp. 227-241 ◽  
Author(s):  
L. A. Pisarevskii ◽  
A. B. Korostelev ◽  
A. A. Lipatov ◽  
G. A. Filippov ◽  
T. Yu. Kin
Keyword(s):  
Thermal Aging ◽  
Nuclear Power ◽  
Austenitic Steels ◽  
Local Corrosion ◽  
Nickel Steel ◽  
Water Type ◽  
Low Carbon ◽  
Steam Generators ◽  
Power Stations

Elaboration of modern domestic structural materials with increased corrosion resistance in contact with advanced heatcarriers of future reactor plants is an important problem at development of innovation projects of nuclear power engineering. Heatexchanging tubes are the critical components, which influence on the safety and reliability of steam generators operation. Corrosion properties of non-stabilized nitrogen-containing corrosion resistant steels of austenite class after cold deformation, thermal treatment and long-term thermal aging studied. It was shown, that silicon introducing into chrome-nickel steel, alloyed by nitrogen and molybdenum, results in increasing of its resistance against local kinds of corrosion and equated it on resistance against intercrystallite and pitting corrosion with particularly low-carbon steels and alloys. But the experimental 03Х18Н13С2АМ2ВФБР-Ш low carbon micro-alloyed steel, proposed for operation at a heat-carrier temperature of 450–500 о С, in the first version had a tendency to a decrease of resistance against local corrosion and impact resistance after long-term thermal aging at temperatures of 360 о С and higher. At present specifying of technological parameters of production and balanced alloying element content takes place, which prevents heat exchanging tubes properties degradation. Steel 03Х17Н13С2АМ2 which has higher resistance against local corrosion and strength comparing with 316LN and 08Х18Н10Т grades, can be taken as a candidate material for production of heat-exchanging tubes of steam generators of nuclear power stations having power reactors of water-water type. The new 03Х17Н9АС2 steel, resistant against inter-crystallite corrosion in high-oxidizing media, was proposed for tests of its operation under conditions of contact with lead heat-carriers instead of 10Х15Н9С3Б1-Ш (ЭП 302-Ш) steel.

scholarly journals Steam reheater with helical tube bundle for wet steam turbine

2020 ◽  
Vol 178 ◽  
pp. 01069
Author(s):  
Mikle Egorov ◽  
Alexander Ivanov ◽  
Ivan Kovalenko ◽  
Irina Krectunova ◽  
Nadezhda Litvinova ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Tube Bundle ◽  
Pure Water ◽  
Work Conditions ◽  
Water Steam ◽  
Power Stations ◽  
Heating Pipes ◽  
Moisture Separator ◽  
Steam Heat

Since steam heat exchangers, used at steam cycle of Russian nuclear power stations, were designed while the knowledge about the separation and the heat exchange processes was limited, deviations between its empirical and theoretical characteristics occur. This limitation also determined application of heating pipes with simple straight shape rather than curved. This study explores a steam heat exchanger with helical heating pipes. It was shown that the model may work stably within the range of parameters, simulating work conditions of the moisture separator and steam reheater at Leningrad nuclear power plant. The experiment included processing of pure water steam as well as mixture of steam and nitrogen. It was obtained a relationship between empirical the heat transfer coefficient and the steam mass flow rate. It was noted that presence of incondensable gas does not affect significantly the heat transfer from the coils, processing high pressure steam.

French PWR Nuclear Power Plants Steam Generators and Reactor Pressure Vessel Heads Aging Management Criteria and Replacement Strategy

2004 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Life Time ◽  
Reactor Pressure Vessel ◽  
Steam Generators ◽  
Life Management ◽  
Power Stations ◽  
Reactor Pressure

The approach used by the French utility, concerning the Aging Management system of the Steam Generators (SG) and Reactor Pressure Vessel Heads, applied on 58 PWR NPPs, involves the verification of the integrity of the component and the Life Management of each plant to guarantee in the first step the design life management and in the second step to prepare long term life time in operation, taking into account the degradation of Alloy 600 material and the replacement of these materials by components made with Alloy 690. The financial stakes associated with maintaining the lifetime of nuclear power stations are very high; thus, if their lifetime is shortened by about ten years, dismantling and renewal would be brought forward which would increase their costs by several tens of billions of Euros. The main objectives are: • to maintain current operating performances (safety, availability, costs, security, environment) in the long term, and possibly improve on some aspects; • wherever possible, to operate the units throughout their design lifetime, 40 years, and even more if possible. This paper shows the program to follow the aging evaluation with application of specific criteria for SG and for Vessel Heads, and the replacement of the Steam Generators and Vessel Heads at the best period. The strategy of Steam Generators Replacement are developed and Vessel Head program of monitoring and replacement are detailed.

scholarly journals Influence of steam heat transfer tube damage characteristics on burst pressure

2021 ◽  
Vol 233 ◽  
pp. 01072
Author(s):  
Zhizhen Peng ◽  
Shangkun Ren
Keyword(s):  
Heat Transfer ◽  
Nuclear Power ◽  
Residual Life ◽  
Corrosion Damage ◽  
Damage Parameter ◽  
Burst Pressure ◽  
Size Change ◽  
Heat Transfer Tube ◽  
Small Depth ◽  
Transfer Tube

The corrosion damage and leakage of heat transfer tube of steam generator were important factors of nuclear power accident, which is closely related to the national nuclear power production safety and people's life and property. The influence of the size change of cylindrical defect on burst pressure of inconel690 alloy heat transfer tube was studied by finite element simulation and experimental measurement. According to the simulation and test data, the relationship between burst pressure and damage parameter is established. The results show that for cylindrical defects with small depth, there is an extreme point of cylinder damage diameter b which makes the damage maximum, which makes the burst pressure at a minimum. The research results can provide reference for accurate evaluation of residual life of heat transfer tubes.

Consideration of field experience in developing new projects of steam generators for nuclear power stations equipped with VVER reactors

2006 ◽  
Vol 53 (1) ◽  
pp. 37-42 ◽  
Author(s):  
N. B. Trunov ◽  
V. V. Denisov ◽  
S. A. Kharchenko ◽  
B. I. Lukasevich
Keyword(s):  
Nuclear Power ◽  
Field Experience ◽  
Steam Generators ◽  
Power Stations
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