A method of calculating the leakage of first-loop coolant into the boiler water in modified steam generators in power stations containing VVÉR-1000 reactors

Atomic Energy ◽  
1997 ◽  
Vol 83 (4) ◽  
pp. 775-778
Author(s):  
I. V. Pyrkov ◽  
E. A. Ivanov ◽  
L. P. Kham'yanov
Keyword(s):  
Steam Generators ◽  
Boiler Water ◽  
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.

Section 1. Introduction

2022 ◽  
pp. 1-2
Author(s):  
Keyword(s):  
Water Chemistry ◽  
Steam Generators ◽  
Thermal Systems ◽  
Boiler Water ◽  
System Components ◽  
Steam System ◽  
Current Operating

This document has been prepared by the Water Technology Subcommittee of the ASME Research and Technology Committee on Steam and Water in Thermal Systems as a consensus of proper current operating practices for the control of feedwater and boiler water chemistry in the operation of industrial and institutional, high duty, primary fuel fired boilers. These practices are aimed at minimizing corrosion, deposition, cleaning requirements, and unscheduled outages in the steam generators and associated condensate, feedwater and steam systems for boilers, and steam system components which are currently available. This publication is an expansion and revision of the operating practice consensus documents previously issued by the Committee [1-3]. The tabulated values herein update and replace the ones previously published. Titles have been edited and clarified. The text has been reordered and modified where necessary. THE TEXT IS OF PRIME IMPORTANCE AND SHOULD BE CONSIDERED FULLY BEFORE USING THE TABULATED VALUES. One Appendix has been added to provide additional guidance.

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.

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.

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