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.

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.

Long-term thermal aging performance of welding structures in primary coolant pipes for nuclear power plants

2020 ◽  
Vol 149 ◽  
pp. 107793
Author(s):  
Minyu Fan ◽  
Mingya Chen ◽  
Min Yu ◽  
Wenqing Jia ◽  
Yuanfei Li ◽  
...  
Keyword(s):  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Primary Coolant ◽  
Aging Performance

Regulatory Perspectives on Fitness for Service Assessments of CANDU Pressure Boundary Components

2017 ◽  
Author(s):  
John C. Jin ◽  
Blair Carroll
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Management Program ◽  
Steam Generators ◽  
Operating Life ◽  
Term Operation ◽  
Pressure Tubes ◽  
Extended Operation ◽  
Condition Assessments

Major pressure boundary components such as pressure tubes, feeder pipes and steam generators at some Canadian CANDU nuclear power plants are entering periods of extended operation beyond their initially assumed operating life. The Canadian Nuclear Safety Commission (CNSC) has approved their long term operations based on the assurance of fitness for service of those components which was composed of condition assessments and aging management programs carried out and implemented by the Canadian licensees. The condition assessments were conducted to demonstrate that components would be within their design basis for the period of intended long term operation and the aging management program was implemented to ensure that the conditions of the components would be maintained as evaluated in the condition assessments and to monitor if new degradation mechanisms would become active. This paper presents the CNSC regulatory practice adopted in the course of technical reviews of fitness for service assessments of major pressure boundary components conducted by Canadian nuclear licensees to demonstrate the safe long term operation of major components can be achieved.

Age-Related Degradation of Steam Generator Internals Based on Industry Responses to Generic Letter 97-06

2002 ◽  
Author(s):  
M. Subudhi ◽  
E. J. Sullivan
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Long Term Effects ◽  
Steam Generators ◽  
Pressurized Water ◽  
Age Related ◽  
The U.S

This paper presents the results of an aging assessment of the nuclear power industry’s responses to NRC Generic Letter 97-06 on the degradation of steam generator internals experienced at Electricite de France (EdF) plants in France and at a United States pressurized water reactor (PWR). Westinghouse (W), Combustion Engineering (CE), and Babcock & Wilcox (B & W) steam generator models, currently in service at U.S. nuclear power plants, potentially could experience degradation similar to that found at EdF plants and the U.S. plant. The steam generators in many of the U.S. PWRs have been replaced with steam generators with improved designs and materials. These replacement steam generators have been manufactured in the U.S. and abroad. During this assessment, each of the three owners groups (W, CE, and B&W) identified for its steam generator models all the potential internal components that are vulnerable to degradation while in service. Each owners group developed inspection and monitoring guidance and recommendations for its particular steam generator models. The Nuclear Energy Institute incorporated in NEI 97-06, “Steam Generator Program Guidelines,” a requirement to monitor secondary side steam generator components if their failure could prevent the steam generator from fulfilling its intended safety-related function. Licensees indicated that they implemented or planned to implement, as appropriate for their steam generators, their owners group recommendations to address the long-term effects of the potential degradation mechanisms associated with the steam generator internals.

National Crisis and Resilience Planning – How to Measure Huge and Compound Disaster that Causes National Crisis –

2016 ◽  
Vol 11 (5) ◽  
pp. 911-925
Author(s):  
Itsuki Nakabayashi ◽  
◽  
Keyword(s):  
Disaster Management ◽  
Nuclear Power ◽  
Nankai Trough ◽  
The Other ◽  
Addition Compound ◽  
National Crisis ◽  
Power Stations ◽  
Inland Earthquake ◽  
Compound Disaster

In the last two decades, three great earthquakes have occurred in Japan: the Hanshin-Awaji earthquake of 1995, the Mid-Niigata earthquake of 2004, and the East Japan Earthquake of 2011. After the East Japan earthquake, a devastating tsunami caused significant casualties and home destruction. More than 18,500 people were killed and more than 121,000 homes were destroyed. In addition, the tsunami destroyed nuclear power stations, which resulted in a severe crisis not previously experienced in Japan.On the other hand, earthquake disasters on a huge scale have been announced to occur as probability of about 70% in the next three decades. One such earthquake is Tokyo inland earthquake that destroys 610,000 homes and kills 23,000 people, and the other is the Nankai Trough earthquake that destroys 2,380,000 homes and kills 320,000 people. In addition, compound disasters where one disaster merges with another disaster may cause damage on a mega scale in this century.In order to address these mega disasters, it is very important to make efforts to reduce damage in the pre-disaster period. According to local plans for national resilience, each municipality must make efforts to reduce level of damage which is able to response trough a Business Continuity Plan (BCP). In addition, each municipality must implement long-term urban projects with a vision toward reconstruction after a mega disaster trough a pre-disaster recovery and reconstruction plan. It is necessary to make revolutionary efforts rather than standard disaster management efforts to reduce damages in the pre-disaster period.

Thermal Aging of Dissimilar-Metal Weld Joints for Reactor Pressurized Vessels at Elevated Temperature

2016 ◽  
Vol 857 ◽  
pp. 271-275
Author(s):  
Won Sik Kong ◽  
Chung Seok Kim
Keyword(s):  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Welding Process ◽  
Primary System ◽  
Microstructural Changes ◽  
Dissimilar Metal ◽  
Welding Consumable ◽  
Arc Welding Process

The purpose of this study is to investigate the thermal aging of dissimilar metal welds for reactor pressurized vessels in the primary system of nuclear power plants. The influences of long-term aging of dissimilar-metal welds on microstructural and mechanical characteristics have been studied qualitatively and quantitatively. The dissimilar-metal welds composed of SA 508 Cl.3 low alloy steel and AISI 316L stainless steel are prepared after buttering alloy 82 on the SA 508 side by the gas tungsten arc welding process using Inconel 82 welding consumable. The test specimens are heat-treated at 600°C for 10000 hours at each predetermined aging time to simulate the degraded microstructure of dissimilar-metal welds subjected to high temperature and pressure. The long-term aging tests are interrupted at various stages to obtain the different level of degraded specimens. The microstructural changes in base metals and weld metal have been evaluated by the optical and electron microscope in relation with twins, grains, precipitates, and phase transformation. The residual stress and mechanical softening were also discussed in terms of microstructural changes during long-term aging.

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.

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