Construction Aspects of the Doel 1 and 2 SGR Projects

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.

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Fouling Of Nuclear Steam Generators: Fundamental Studies, Operating Experience and Remedial Measures Using Chemical Additives

AECL Nuclear Review ◽

10.12943/anr.2013.00007 ◽

2013 ◽

Vol 2 (1) ◽

pp. 61-88 ◽

Cited By ~ 7

Author(s):

C.W. Turner

Keyword(s):

Economic Performance ◽

Nuclear Power ◽

Power Plants ◽

Tube Bundle ◽

Operating Experience ◽

Feed Water ◽

Chemical Additives ◽

Steam Generators ◽

Primary Coolant ◽

And Performance

Fouling remains a potentially serious issue that if left unchecked can lead to degradation of the safety and performance of nuclear steam generators (SGs). It has been demonstrated that the majority of the corrosion product transported with the feed water to the SGs accumulates in the SG on the tube-bundle. By increasing the risk of tube failure and acting as a barrier to heat transfer, deposit on the tube bundle has the potential to impair the ability of the SG to perform its two safety-critical roles: provision of a barrier to the release of radioactivity from the reactor coolant and removal of heat from the primary coolant during power operation and under certain post accident scenarios. Thus, it is imperative to develop improved ways to mitigate SG fouling for the long-term safe, reliable and economic performance of nuclear power plants (NPPs). This paper provides an overview of our current understanding of the mechanisms by which deposit accumulates on the secondary side of the SG, how this accumulation affects SG performance and how accumulation of deposit can be mitigated using chemical additives to the secondary heat-transport system. The paper concludes with some key questions that remain to be addressed to further advance our knowledge of deposit accumulation and how it can be controlled to maintain safe, economic performance of nuclear SGs.

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Electroslag Strip Cladding of Steam Generators With Alloy 690

Volume 1: Plant Operations, Maintenance and Life Cycle; Component Reliability and Materials Issues; Codes, Standards, Licensing and Regulatory Issues; Fuel Cycle and High Level Waste Management ◽

10.1115/icone14-89434 ◽

2006 ◽

Cited By ~ 1

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.

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

Ferrous Metallurgy Bulletin of Scientific Technical and Economic Information ◽

10.32339/0135-5910-2019-2-227-241 ◽

2019 ◽

Vol 75 (2) ◽

pp. 227-241 ◽

Cited By ~ 1

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.

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

Aging Management and License Renewal ◽

10.1115/pvp2004-2964 ◽

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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