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 3Power: 1350 MWStart of Operation: 2018Areva (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

WOL in Angra 1 pressurizer safety nozzles (2010)

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.