Structural Integrity

1. 1
Vivekanand Kain
Head, Materials Processing & Corrosion Engineering Division
Bhabha Atomic Research Centre, Mumbai 400 085
vivkain@barc.gov.in
Structural Integrity:
Role of materials and corrosion
• Forms of corrosion:
main concerns
• Materials for Reactor
Pressure Vessel & Piping
(Radiation embrittlement)
(Boric acid wastage)
• SCC / IASCC
• Pitting
Outline:
Workshop on Structural Integrity Assessment of Nuclear Energy Assets
May 9 – 10, 2018
 Mechanism of FAC: Primary vs. Secondary circuit
 FAC and LDI (Liquid Droplet Impingement)
 FAC in primary circuit: Main features,
 FAC in secondary circuit
- Single phase FAC cases
- Dual phase FAC cases
- Stainless steels: FAC resistance
 FAC at weldment: extra care needed
 Small bore piping: Change material
 Control measures against FAC
Flaw tolerant material: Use in nuclear industry
Origin of flaws:
* From manufacturing stages
* Service induced flaws
• Hydriding in PTs
• Nodules on OD side of PTs
• Movement of bundles on PT ID side
 Crud depositing on bottom of vessels;
leading to under deposit corrosion
 Initiation of cracks/pits
 Thinning from FAC
Materials:
* Inherent properties
(Crystal structure)
* Aging behaviour
(Thermal aging;
radiation aging)

2. 2
Dominant degradation mechanisms in
carbon steel/stainless steel
in Swedish nuclear power plants (2002)
Cracking data base as a basis for risk
informed inspection by Karen Gott,
10th International conference on
Environmental degradation of
materials in nuclear power systems
– Water reactors
Worldwide Corrosion Events in LWR
(1995 – 2004) R. Kilian, Areva, 2008
Stress corrosion cracking: 39%
Flow accelerated corrosion: 33.6%
Corrosion: 18.7%
Pitting: 1.4%
Under deposit corrosion/MIC: 5.3%
Galvanic corrosion: 1.1%
H induced cracking: 1.1%
PHWR
TAPS 3&4
Primary coolant:
Temperature: 249 – 305 °C
Pressure: 10 MPa
Water Chemistry: D2O
LiOH: 1-2 ppm
pH: 9.5 – 10.5
H2: ~ 2 ppm
BWR
TAPS 1&2
Primary coolant:
Temperature: 270 - 300 °C
Pressure: 7 MPa
Water Chemistry:
DM water
pH: 7.0
H2: In case of HWC
PWR
KKNP
Primary coolant:
Temperature: 280 – 330 °C
Pressure: 20 MPa
Water Chemistry:
Li as LiOH: 2 ppm
B as H2BO3: 1200 ppm
pH: 6.9-7.4
H2: ~ 2 ppm

3. 3
Western LWRs
Mn-Mo Type
SA 302 B- Plates
VVERs
VVER-440 RPV
15X2MΦA
(Cr-Mo-V type)
Mn-Mo-Ni Type
SA 533 B - plates
SA 508 Cl3- Forgings
20 MnMoNi55- German
 16 MnD 5- French
VVER-1000 RPV
15X2MHΦA
(Cr-Mo-Ni-V Type)
Ni added to increase hardenability in thick sections
Concerns: Fatigue, Corrosion (inside SS clad),
Thermal and Irradiation embrittlement
RS Yadav, BARC
IRRADIATION DEGRADATION OF RPV STEEL
Neutron Fluence more
than 1018 n/cm2 (>1MeV)
Low Alloy
Steel
Irradiation
Embrittlement
1. Direct matrix damage (due to creation of V and I)
2. Precipitation hardening of matrix (due mainly to Cu: nano aggregates)
(Ni known to associate in and stabilize Cu clusters & micro-voids; Mn & Si similar effects)
3. Phosphorus segregation (in matrix / gb causes embrittlement)
>1 MeV
RS Yadav, BARC

4. 4
RPV pitting and CGR
LAS
ASS
Pitting corrosion at or along
the phase boundary between
the austenitic cladding and
the LAS base material of the
Japanese research reactor
If loading is:
a) constant (in form of tensile loading in service or tensile residual stresses)
failure is due to stress corrosion cracking (SCC),
b) transient as during startup/shutdown, strain-induced corrosion cracking
(SICC) becomes dominant and
c) cyclic then failure is by corrosion fatigue.
• A high corrosion potential ECP > +100 mV SHE/ high DO (>> 200 ppb) and
quasi-stagnant flow conditions.
• Cl- >> 5 ppb and SO4
2- >> 100 ppb
• A high steel S-content (> 0.020 wt. % S).
• A high susceptibility to dynamic strain ageing (DSA) in between 180 – 270 °C.
• A high hardness/yield stress level (> 350 HV, Rp > 800 MPa).
SCC favoured in laboratory simulated conditions by:
IAEA publicaion
Siefert, PSI publications
PWR RPV – concentrated boric acid at high temperature
Excessive thinning of RPV material
IAEA report
4” x 5” cavity
Down to 3/8” SS
None of 68 PWRs
in USA had any
similar boric
acid wastage
Wastage Rate:
4800 mpy
At 212 °C

5. 5
Gorman et al., Compa. G. ASME BOIL. Press. Vess. Code
IAEA Nucl. Energ. Series
SS tube
Small circumferential crack
Blunts at Low Alloy Steel Axial cracks in SS
IGSCC in PWR: SS-LAS (RPV) penetration welds
Turkey Point 4:
230 Kg of boric acid crystals
on RPV
And
Consequent severe corrosion
3 of 58 studs were so
corroded that these bolts and
nuts had to be replaced
Similar leakages from
Salem 2
Ringhals 2
Wastage rate: 25 mm/y
Boric acid wastage

6. 6
STAINLESS STEELS IN NUCLEAR INDUSTRY
SS 316NG
SS 304L
SS 304LN
SS 403M
17 – 4 PH / 13 – 8 Mo
SS 440C
SS 410
SS 321
SS 304L
SS 304L (NAG)
SS 310L (NAG)
PITTING CORROSION
Self initiating form of crevice corrosion
Occurs on open surfaces (does not require crevice geometry)
Metals/Alloys showing passive behaviour are prone to pitting corrosion
(Stainless steels, Ni and Fe based alloys, Al and alloys, Ti and alloys)
Halides/sulfides species cause pitting
Pits initiate preferentially at:
Propagation mechanism
Same as for crevices
(Autocatalytic)
Pit solution chemistry:
Low pH and high Cl- levels

7. 7
Pit Shapes: ASTM G 46
Pitting Damage
Y
X
Manifestation:
X/Y << 1
Pit chemistry: Low pH and high chloride conc.
Sensitization of Stainless Steels
• Precipitation of Cr23C6 at the grain boundary Cr depletion
• Temperature range: 500-8500C.
• At Cr concentration < 12%, passive film becomes weak and prone to
attack in aggressive environment.
Prone to IGC: Ditch structure and > 24 mpy in Practice C
> 48 mpy in Practice B

8. 8
SCC: Prerequisites
Crack initiation: Localized film
breakdown
(Cl-/halides/sulfides
At inhomogeneities, slip steps etc.)
Crack Propagation: Slip dissolution/
Film rupture
- Adsorption at crack tip
- Brittle films
-Liquid film model
- Internal oxidation model
Manifestation: Thin cracks with crack tip radius of atomic dimensions.
Branching in SS – IG/TG/Mixed
Material
Composition
Heat treatment
Microstructure
Surface-
condition
Environ-
ment
Composition
temperature
Electrode potn
Flow rate
Stress-
strain
Service stress
Fit up stress
Residual stress
Strain rate
..
.
.
.
SCC: A brittle failure at a low constant stress of an alloy exposed
to a corrosive environment
A synergistic action of corrosive environment and tensile stress
EAC
Mech
fracture
?
.
Corrosion
SCC inside the reactor: IGSCC
SCC outside the reactor / by chloride ions: TGSCC
Two different mechanisms of SCC

9. 9
Susceptibility to SCC as a function of
nickel content
SCC of stainless steels with
different nickel content in
boiling MgCl2 test
3-6 months
350 °C water; > sys
Pure water
Water + 1000 ppm Cl-
304
18-20Cr
800
20-23 Cr
690
27-31 Cr
600
14-17 Cr
Copson CurveCouriou cracking
IGSCC of non sensitized SS
after ~ 7- 8 EFPY of the reactors:
(Core Shrouds, core internals, recirculation pipelines)
SS 304L/316L
Indian BWR: No cracking observed in Core Shrouds
(3) Core shroud cracking: Retained strain in constrained weldments
(2) IASCC: * All SS show cracking after a threshold fluence
* The threshold fluence varies for materials
(In-core components)
(1) Other Mechanisms:
* LTS / Cold work enhancing/causing sensitization
(Recirculation pipelines)

10. 10
As welded
As welded +
4000C, 201 days
Stages of M23C6 growth during
grain boundary migration at T < 4000C
Hanninen et al. 1985
Corrosive Environment
• High purity water
• 288C
• Pressure ~ 8MPa
• Oxidizing species – 200 – 300 ppb
BWR Operating environment
Generation of oxidizing species due to radiolysis of water
CorrosionPotential,VSHE
Dissolved
oxidising
species of
200-300ppb
produced due
to radiolysis
sufficient to
serve as a
driving force
for IGSCC
200 ppb
300 ppb
Potential developed
Negligible
IGSCC crack
growth rate
Ford et al, 1987

11. 11
Corrosion Potential, VSHECorrosion Potential, mVSHE
Crackgrowthrate,mm/s
Crackgrowthrate,mm/s
Peter Andresen et al
Non-sensitized SS has shown
IGSCC in BWR and PWR
CGR of 10-7 mm/s – 3 mm / year
10-8 mm/s – 0.3 mm/year
10-9 mm/s – 0.03 mm/year
Irradiation Damage of Stainless Steels
Irradiation Damage in Stainless
Steels (SS) occur by neutron
irradiation, energy > 1 MeV
Neutron Irradiation results in
generation of point defects
(vacancy and self-interstitial)
far in excess of equilibrium
concentration at reactor
operating temperature
Diffusion of point defects results in
variation of microchemistry along
grain boundaries – RIS (Radiation
Induced Segregation)
Bruemmer et al

12. 12
BWR IASCC occurs in SS after RIS
induced by neutron fluence of
• 5 x 1020 n/cm2 for SS 304
• 1 x 1021 n/cm2 for SS 316
Bruemmer et al
Radiation Induced Segregation
(RIS) in high purity stainless steel
Cr(wt%)Fe(wt%)Ni(wt%)Si(wt%) Distance from grain boundary, nm
IASCC of Austenitic SS in Nuclear Reactors
.
.
.
Radiation Induced
Segregation (RIS)
.
Variables in IASCC: Materials and Water Chemistry effects
Radiolysis: Local (crack tip)
and bulk changes in chemistry
(oxidizing species) and ECP
Andresen and Ford

13. 13
Hydrogen Water Chemistry
Dissolved oxygen control required
Possible by adding: Ammonia, Hydrazine, Ammonia-Hydrazine, Hydrazine-morpholine
or Hydrogen
Hydrogen chosen: (a) No effect on pH, (b) Minimum system impact & (c) Not corrosive or toxic
Hydrogen Water Chemistry (HWC)
At 0.015 mg/L of Oxygen, IGSCC initiation difficult
At ECP < - 230 mVSHE IGSCC is controlled
Top of the core components not protected
Hydrogen required varies from plant to plant
Feed water Hydrogen (SCFM) Feed water Hydrogen conc (ppm)
mV(SHE)
V(SHE)
Jones et al
Effect of yield strength on CGR
Andresen et al

14. 14
Warm worked SS 304LN: irregular IGSCC vs
regular (smooth) crack front for sensitized SS 304LN
A clear distinction to establish if SCC dictated by RIS or radiation hardening !
Cold worked Alloy 600 in PWR
Water chemistry: Uneven crack front
Shoji et al, Corr Sci, 2008
BARC results
on SS 304LN
In BWR WC
WW 20%
Vs
Sensitized SS
S. Roychowdhury & V. Kain: Corrosion Science 2011 & JNM 2011
Stresses arising from
solidification of weld pool
Stresses arising from constraintment
of the structure preventing weld shrinkage
Region of heavy plastic deformation
leading to increase in strength and
IGSCC susceptibility
Weld pool
Welding of
thick plate
SCC at H3 Weld Line
of BWR Core Shroud
Ref. : TEPCO/GE data
IG crack
Weld pool
Actual IGSCC
failures in
BWR’s in non
sensitised SS
SCC on type 316NG Piping System
Ref. : Ulla Ehrnsten et. al;"Tenth International Conference on
Environmental Degradation of Materials in Nuclear Power Systems
Water Reactors"
IG crack
Weld pool
Residual strain as a function
of distance from weld pool
Strain(%)
Distance from fusion line (mm)
Strain levels
about 20% near
welded regions
HEAVY PLASTIC STRAIN IN WELDS MAIN CAUSE OF IGSCC IN REACTORS
REDUCE STRAIN: NARROW GAP WELDING; USE INSULATIVE COATINGS?
S. Roychowdhury & V. Kain, Acta Met 2012

15. 15
SS316 L weldments:
Peak strains formed at root and at weld fusion line
Andresen, 2013 Corrosion
After a heavy sensitisation heat treatment at 675 C, 24h
In non-sensitized, warm rolled (20%) condition
S. Roychowdhury & V. Kain: Corrosion Science 2011 & JNM 2011

16. 16
EAC Detection and Sizing:
Dye penetrant (DP) testing; often Fluorescent DP
(On surfaces from where initiated)
Ultrasonic Testing (UT)
(From the opposite surface)
measures crack depth LBB concept
Relevant provisions in codes:
“Localized thinning/metal loss”
API 579 Lower thicknesses are allowed in case of
cracks/pits vs. general (uniform) corrosion
THANK YOU