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Metal Degradation in Angra Plants
José Eduardo Maneschy
Technical Meeting on
Degradation of Primary Components of Pressurized Water Cooled Nuclear Power PlantsCurrent Issues and Future Challenges
Vienna, Austria5 – 8 November 2013
IAEA and EC-JRC
Introduction
Angra 1Power: 657 MWStart of Operation: 1985Westinghouse
Angra 2Power: 1350 MWStart of Operation: 2001Areva (Siemens/KWU)
Introduction
Angra 1Power: 657 MWStart of Operation: 1985Westinghouse
Angra plants up to 2013
• Angra 1 – 28 years in operation (~15 EFPY)
• Angra 2 – 13 years in operation (~10 EFPY)
Degradation
Plants are aging
Introduction
History of degradation (typical for U.S. nuclear industry):
Safety systems:
1980 – Steam generator (tubes)
1990 – Reactor vessel (head penetrations)
2000 – Reactor vessel, pressurizer, steam generator (nozzles safe-end welds)
Non-safety components:
1990 – Low pressure turbine (disks)
Stress corrosion cracking (SCC)
Introduction
Confirmed degradation due to stress corrosion cracking in Angra 1:
• Steam generator tubes
- Detected in 1980s
SG replaced in 2009
• Low pressure turbine disks
- Detected in 1990s
LP 1 rotor replaced in 2006
LP 2 rotor replaced in 2013
Introduction
Potential degradation due to stress corrosion cracking in Angra 1:
• Pressurizer nozzles (safe-end welds made of alloy 600)
- Non detected
Mitigated by Weld Overlay (WOL) in 2010
• Reactor vessel head (penetrations made of alloy 600)
- Non detected
Upper head replaced in 2013
• Reactor vessel nozzles (safe-end welds made of alloy 600)
- Non detected
To be mitigated by mechanical stress improvement process (MSIP)
Introduction
Objective:
Show how the stress corrosion cracking is managed in Angra
• Pressurizer (nozzles safe-end weld)
• Reactor vessel (upper head penetrations)
• Low pressure turbine (disks)
ASME has considered fatigue in the design phase since the beginning of 1970s. SCC was not included because it was believed that ASME material was not susceptible to this form of degradation.
Pressurizer nozzles safe-end weldsDegradation is in the dissimilar metal welding, which is the welding of two metals with different properties. The original option for the nuclear industry was alloy 600 weld material (specification alloy 82/182).
Illustration from EPRI
The alloy 82/182 is susceptible to primary water stress corrosion cracking (PWSCC)
Mitigative process to avoid or eliminate the PWSCC in 82/182 weld is Weld Overlay. The WOL uses a resistant material deposited in theexisting weld, which is specified to induce a compressive residual stress in the inner portion of the weld. The total stress is lower than the threshold to PWSCC.
SS Pipe SS Field weld SS Safe-end alloy 82 alloy182 (buttering) CS Nozzle
WOL Volume Alloy 52M
Illustration from Structural Integrity
Pressurizer nozzles safe-end welds
Residual stress pre and pos WOL (axial stress):
Results from Structural Integrity
pre WOL Stress at inner portion of the weld:
32 to 48 ksi
pos WOL Stress at inner portion of the weld:
-64 to -28 ksi
Stress analysis results
Pressurizer nozzles safe-end welds
Degradation is in the dissimilar metal weld and in the base metal (PWSCC)
Control rod drive mechanism (CRDM)
Reactor vessel head insulation
Reactor vessel head (carbon steel portion)
Reactor vessel head (stainless steel cladding layer)
Typical PWR Upper Head
Reactor vessel upper head penetrations
External head surface
Head (low alloy steel)
Stainless steel cladding
Weld alloy 82/182
Surface in contact with PWR ambientWeld alloy 82/182
Typical penetration
Penetration (alloy 600)
Reactor vessel upper head penetrations
• Longitudinal cracks in base metal (France, 1991)
• Circumferential cracks in weld metal (USA, 2001, Oconne 3), Davis-Besse, North Anna 2, Cristal River 3 etc.
Huge program of inspection and analysis was implemented
PWSCC detected in the weld
Repair
PWSCC detected in the base metal
Leave it in service
Inspect based on fracture mechanics
Reactor vessel upper head penetrations
Allowable (0,75t)
– If the inspection interval is one year, only cracks below 0,3t can remain in service.
– Leakage occurs after approximately three years of crack initiation.
Reactor vessel upper head penetrations
One year
• Since early 1970s SCC is a typical degradation in low pressure (LP) turbine.
• Cracks are detected in the disks, and located close to keyway area. Although fatigue is also present, the failure is controlled by SCC.
• In some cases, after five years it is possible to detect cracks 10 mm in depth. After ten years, crack depth can reach 20 mm and length 100 mm.
• Main reason to SCC: high stress in the keyway area, aggressive environment, and susceptible material.
• Aggressive environment: impurities (chloride and oxygen), humidity and temperature.
Low pressure turbine disks
Because steam may arrives with some humidity in the center of the LP, in general, disks 1 and 2 are more susceptibles to SCC.
Low pressure turbine disks
Degradation is in the turbine disk in the keyway area
Blade Keyway area
Generator sideValve side
Steam
It is difficult to predict when SCC will initiate. When cracks are detected, the issue is to determine the time to propagate until the critical size is reached.
Turbine disk under stress corrosion cracking
Steam
Keyway
Transversal section
Axial view
Disk
Bore
shrunk-on disks on rotor
Disk material: 3.5 CrNiMoV
Low pressure turbine disks
Low pressure turbine - Disk 1
02.0004.0006.0008.000
10.00012.00014.00016.000
0 2 4 6 8 10 12 14 16 18 20
Crack depth (mm)
Tim
e fo
r re
-ins
pect
ion
(h)
Typical solution to turbine disk under SCC:
Apply fracture mechanics to determine time interval for re-inspection. Crack manual is prepared to allow a quick evaluation of the flaw during the outage.
Low pressure turbine disks
Conclusions
– Presented typical degradation in nuclear industry and how the stress corrosion cracking issue in the alloy 600/82/182 is managed in Angra.
– Fatigue has been considered in the components design since the early 1970s. However, SCC was not included (it was believed the ASME material was not susceptible). Several failures have shown the SCC is the concern.
– Solutions are components dependent.Primary components nozzles – mitigation (WOL or MSIP)RV penetrations – replacementSG tubes – replacementTurbine disks – replacement
– Fracture mechanics allows assessing the life of a cracked component due to stress corrosion cracking. Crack manual is a good practice to allow a quick assess of a crack detected during an inspection. This will avoid plant outage extension due to necessity to conduct stress and fracture mechanic analyses.
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