Phenomenon of (irradiation assisted) stress corrosion...

Phenomenon of (irradiation assisted) stress corrosion cracking for g

internals of PWR & BWR systems

R d k N t & L i i D b b iRadek Novotny & Luigi DebarberisInstitute for Energy (IE)Petten, The NetherlandsPetten, The Netherlands

http://www.jrc.ec.europa.eu

T i t A il 2009T i t A il 2009

1

Trieste, April 2009Trieste, April 2009

CONTENTCONTENTA i- Ageing

- Corrosion- SCC – PW SCC- Radiation effects- IASCC- CONCLUSIONS

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- Effort to plan life management of ageing NPPsf- Reliability of in-core in-vessel structural materials

- Control of the degradation of the structural materials- Guiding replacement campaignsg p p g- Rising issues for alloys in LWR conditions:

- Corrosion and

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- Stress Corrosion Cracking (SCC PW SCC)- IASCC

Industry’s Top Ten R&D Priorities (from MRP-205)1. Inspection & Evaluation (I&E) Guidelines: Reactor Internals2. NDE Technology: Dissimilar Metal (DM) Butt Welds3. PWSCC Mitigation: Environmental Controls4. I&E Guidelines: Bottom Mounted Nozzles5. Vibration Fatigue: Small Bore Piping6. Environmental Fatigue Issues: Press. Boundary Components7. NDE Qualification Program: Ni-Alloy Penetrations8. NDE Accessibility: Reactor Internalsy9. PWSCC Mitigation: Stress Improvement (SI) of Butt Welds10.Thermal & Irradiation Embrittlement: Synergistic Effects on CASS & SS Welds - Internals

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Essential service water system pipe degradation

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Stress Corrosion CrackingStress Corrosion CrackingAttachment welds: SCC significant at welded pad/bracket locations in the vessel shell

Nozzles: SCC of nozzles is a significant issue

Closure studs

Penetrations: SCC of CRD stub tubes (high residual stresses in sensitized weld material)

Safe ends: Observed at several plants SCC is a potentially significantSafe ends: Observed at several plants. SCC is a potentially significant degradation mechanism for safe ends

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Water Stress Corrosion CrackingWater Stress Corrosion CrackingCorrosion Water Stress Corrosion Cracking

Fracture Mechanics

Stress Corrosion Cracking

Fracture Mechanics

Corrosion

StressStressStress

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Material

Main Degradation Mechanisms

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LoviisaLoviisa

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PWSCC Experience in VesselPWSCC Experience in Vessel and Other Components

Plants with leaks:Plants with leaks:V.C. Summer – axial, reactor vessel nozzle (2000)Tsurga 2 – axial, pressurizer valve nozzle (2003)

Plants with cracks/indications:Ri h l 3 4 i l t l l (2000)Ringhals 3, 4 – axial, reactor vessel nozzle (2000)V.C. Summer – circ. y axial, reactor vessel nozzles (2000)Tihange 2 – axial, pressurizer nozzle (2003)C l t Cliff 2 (2005) DC C k 1 (2005) C l t Cliff 1Calvert Cliffs 2 (2005), DC. Cook 1 (2005), Calvert Cliffs 1 (2006), Davis Besse (2006).....

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Corrosion & radiation- Effects of irradiation on materials are well investigated

- Defects due to irradiation M h i l i l i i h d- Mechanical properties evolution with dose

- For LWRs, passive materials used - Corrosion enhanced by water radiolysis

- oxidant species (OH-, H2O2, etc.)- together with reducing species (H, etc)

n− [y , −K IC- influence free corrosion potential

- Radiation affects also semi-conductive properties of the oxides, particularly the

H 2O→H+OH

2 OH →H2O2Fe+2

p p , p ybehaviour of the passive layer

11Cr gb ↓

Radiolysis of Water by n- and γ Radiation

- The integrity of fuel elements (zircaloy) may also be affected- Localised corrosion phenomena are also affected by the free corrosion

Water Stress Corrosion CrackingWater Stress Corrosion CrackingCorrosion

potential changes, particularly SCC phenomena

Water Stress Corrosion Cracking

Fracture Mechanics


Fracture Mechanics

Corrosion

StressStressStress

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Material

+ Radiation


+ Radiation


Fracture MechanicsIASCC


Fracture Mechanics

Corrosion

IASCC

StressStressStress

Radiation

Radiation creepRadiolysis

Radiation inducedRadiation creepRadiolysis

Radiation induced

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Radiation corrosionRadiation inducedsegregation

MaterialRadiation corrosion

Radiation inducedsegregation

Present generation

LWR BWR PWR VVERLWR BWR, PWR, VVERHW CANDUGCR MAGNOX AGRGCR MAGNOX, AGRLM FRLM FR

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BWR PWRBWR - PWR

- Interactions corrosion behaviour and radiolysis - Slightly diversified for the two main LWRs

B ili t t (BWR) d- Boiling water reactors (BWR) and - Pressurised water reactor (PWR)

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Water Chemistry Conditions in BWR and PWRWater Chemistry Conditions in BWR and PWR

BWR - PWR

BWRs: stainless steels (304 type) mainly used for core components- exposed to water often up to 288 °Cp p- SCC is controlled by hydrogen conditioning (HWC: H water chemistry)

- free corrosion potential (ECP) at low values

PWRPWRs- primary water chemistry based on B (added as boric acid) - for neutronic purposes with Li addition (added as LiOH)- increase pH to limit the general corrosion and activation of componentsincrease pH to limit the general corrosion and activation of components - large hydrogen concentration used to limit radiolysis effects

- 25 to 35 ml of hydrogen per Kg water!

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Boric Acid CorrosionBoric acid leakage is a consequence of Alloy 600 cracking

This leakage can lead to boric acid corrosion of low-alloy steel

Davis-Besse, March 2002

Order EA-03-009.Order EA 03 009. Inspection requirements according to the parameter EDY

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Davis-Besse

Davis-Besse

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Cabeza vasija Davis-Besse

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Plants with Replaced RPV Upper Head - USAPlants with Replaced RPV Upper Head - USA

25 plants replaced RPV upper heads

24 with Alloy 690 penetrations

1 with Alloy 600 (Davis-Besse)

13 of the 21 remaining high13 of the 21 remaining high

and moderate susceptibility plants

PWSCC. Vessel Head

The Alloy 600/82/182/ has been changed by the Alloy 690/52/152 in y g y ythe penetrations of the new vessel heads

The inspection practices vary from country to country, reducing the i ti i t l f th l h d ith All 600

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inspection intervals for the vessel heads with Alloy 600

Stress Corrosion CrackingBWR internals susceptible to two forms of SCC:BWR internals susceptible to two forms of SCC:

Intergranular stress corrosion cracking (IGSCC)Irradiation assisted stress corrosion cracking (IASCC)g ( )

Degradation via IGSCC is potentially significant. Programmes to effectively manage this degradation g y g gmechanisms are required

IASCC is a concern in BWR core internal t h ti f th h d dcomponents such as a portion of the core shroud and

the top guide.

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Degradation Incidents of RPVIs Safety RelevantDegradation Incidents of RPVIs Safety Relevant

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Miti ti T h l i f SCCMitigation Technologies for SCCWater chemistry control or surface treatment

Hydrogen water chemistryNoble Metal Chemical Application (NMCA)Deposition of noble metals by plasma spraySurface melting/Solution annealing

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+ Radiation


+ Radiation




Fracture Mechanics

Corrosion

IASCC

StressStressStress

Radiation



Radiation induced

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Some Effects of IrradiationSome Effects of Irradiation

Loviisa core basketLoviisa core basketVisual and ultrasonic inspection of all 312 locking bolts (Tecnatom)

Removal of defective locking bolts (Westinghouse)Removal of defective locking bolts (Westinghouse)

Assembly of the new fixing system (Westinghouse)

Internals TV-inspection of the core basket (Tecnatom)

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+ Radiation


+ Radiation




Fracture Mechanics

Corrosion

IASCC

StressStressStress

Radiation



Radiation induced

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Strong influence of radiation

H 2O→H+OH

n

F+2

− [y , −K IC

2 OH →H2O2Fe+2

C ↓Cr gb ↓

n− [y , −K IC

H 2O→H+OH

2 OH →H2O2Fe+2

↓Cr gb ↓

Note: ~15 dpa = 1022 n/cm2 E ≥ 1 MeV (for PWR and BWR neutron spectra) 7 dpa = 1022 n/cm2 E ≥ 0 1 MeV (for PWR and BWR neutron spectra)~7 dpa = 1022 n/cm2 E ≥ 0.1 MeV (for PWR and BWR neutron spectra)

Radiation Induced Segregation - RIS

+ Radiation


+ Radiation




Fracture Mechanics

Corrosion

IASCC

StressStressStress

Radiation



Radiation induced

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- Terminology used to describe cracking of materials Exposed to nuclear reactor coolant and ionizing radiation

IASCC – Irradiation Assisted Stress Corrosion Cracking- Exposed to nuclear reactor coolant and ionizing radiation- Like all Stress Corrosion Cracking phenomena it requires critical combinations of applied stress or strain, environmental chemistry & metallurgical structure to occur

Major factors influencing Environmentally Assisted Cracking (EAC)

IASCC added feature to EAC:- atomic displacements- neutron irradiation significantly alters metallurgical microstructureionizing (α β and γ) radiation modify the environmental chemistry- ionizing (α,β and γ) radiation modify the environmental chemistry

Effects of irradiation on SCC:- primary defects

defects segregation- defects segregation - dislocation interaction- grain boundaries- localized stress and strain - environment- stress relaxation by irradiation- creep (beneficial factor for IASCC)

IASCCIASCC• Radiation• Stress• Stress• Time• Temperature• Temperature• Environment

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Strong influence of radiation

H 2O→H+OH

n

F+2

− [y , −K IC

2 OH →H2O2Fe+2

C ↓Cr gb ↓

SCC - RPV internals

1180.000 0.035 0.070 0.105 0.140 0.175 0.210

Displacement (mm)

3400.035 0.070 0.105 0.140

Displacement (mm)Displacement (mm) Displacement (mm)118

0.000 0.035 0.070 0.105 0.140 0.175 0.210

Displacement (mm)

3400.035 0.070 0.105 0.140

Displacement (mm)Displacement (mm) Displacement (mm)

116

117

118

CRACK GROWTH

320

340

CRACK GROWTH

(µV)

(µV)116

117

118

CRACK GROWTH

320

340

CRACK GROWTH

(µV)

(µV)

114

115

300DC

PD (

DC

PD (

114

115

300DC

PD (

DC

PD (

800000 1200000 1600000 2000000 2400000112

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Time (s)900000 1200000 1500000 1800000

280

Time (s)

Sample: 10 x 10 x 55 mm3 Sample: 3 x 4 x 27 mm3

Time (s) Ti ( )800000 1200000 1600000 2000000 2400000

112

113

Time (s)900000 1200000 1500000 1800000

280

Time (s)

Sample: 10 x 10 x 55 mm3 Sample: 3 x 4 x 27 mm3

Time (s) Ti ( )Time (s) Time (s)Time (s) Time (s)

Testing Environmentally Assisted Cracking (EAC) of Reactor Materials using

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g y g gPneumatic Servo Controlled Fracture Mechanics (PSCFM), sub. International Journal of Pressure Vessel and Piping, 2006R. Novotny, F. Sevini, L. Debarberis, S. Petr, M.Kytka,

CONTENTCONTENTA i- Ageing

- CorrosionWater Stress Corrosion Cracking



Fracture Mechanics

Corrosion

IASCC

- SCC- Radiation effects Stress

Radiation

Stress

- IASCC- CONCLUSIONS Radiation creepRadiolysis

Radiation





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Phenomenon of (irradiation assisted) stress corrosion cracking for g

internals of PWR & BWR systems

R d k N t & L i i D b b iRadek Novotny & Luigi DebarberisInstitute for Energy (IE)Petten, The NetherlandsPetten, The Netherlands

http://www.jrc.ec.europa.eu

T i t A il 2009T i t A il 2009

48

Trieste, April 2009Trieste, April 2009

Phenomenon of (irradiation assisted) stress corrosion...

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