ML15331A178

Official Exhibit - NYS000495-00-BD01 - Slides, Irradiation Assisted Degradation of LWR Core Internal Materials: Brief Review, Presentation by Appajosula S. Rao, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Apr

ML15331A178
Person / Time
Site:	Indian Point
Issue date:	04/14/2015
From:	State of NY, Office of the Attorney General
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 27908, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15331A178 (51)

Text

Irradiation Assisted Degradation of LWR Core Irradiation Assisted Degradation of LWR Core Internal Materials: Brief Review Appajosula S. Rao Office of Nuclear Regulatory Research US Nuclear Regulatory Commission, Washington, DC, 20555 The views expressed in this presentation are not necessarily those of the U S Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission For Presentation at the Office of Nuclear Regulatory Research, Seminar on April 14th 2015.

United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of:

Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #:

Identified:

Admitted:

Withdrawn:

Rejected:

Stricken:

Other:

NYS000495-00-BD01 11/5/2015 11/5/2015 NYS000495 Submitted: June 9, 2015

Contents

• Introduction / Technical Issues B

k d

Background

Chemical Composition of Austenitic Stainless Steels Irradiation Microstructure - Density/Size of Loops with Dose Dose

• Recent NRC Research on Irradiation Microstructure Micro-chemical changes at the grain boundary of Steels versus Neutron Irradiation versus Neutron Irradiation Irradiation Hardening Slow Strain rate Test Results Crack Growth Rate (CGR) Tests

(

)

Summary 2

Introduction Technical Issues identify susceptible materials & conditions

identify susceptible materials & conditions,

determine crack growth rates to assure that the selected inspection intervals are adequate,

verify effectiveness of proposed mitigative measures i

ti t

t ti l f

di ti b ittl t i l di i ti 3

investigate potential of radiation embrittlement, including synergistic effects of thermal & neutron embrittlement

evaluate effects of void swelling, including the associated decrease in ductility

Void Swelling

Void swelling of most reactors internals is Void Swelling

Void swelling of most reactors internals is not expected to be limiting over the current licensing period licensing period

Continued research work will explore the extent of void swelling over the the extent of void swelling over the extended operating life of present reactors (i.e 54 EFPY) 4 reactors (i.e 54 EFPY)

Helium bubbles in Mono crystalline Nickel (110) Irradiated with 100 keV ions 5

y

(

)

6

=

Background===

Neutron irradiation changes the material microstructure (radiation hardening) & micro-chem. (radiation induced segregation) that leads to:

increase in yield strength & decrease in ductility

degradation of fracture toughness

increased susceptibility to IASCC

void formation, coalescence and swelling

creep relaxation

Neutron irradiation also changes the water chemistry (radiolysis);

BWR normal water chemistry results in an increase in corrosion y

potential,

Ecorr 200 mV (SHE);

PWR water contains H2 toscavenge radiolysis products,

E

800 mV (SHE);

Ecorr - 800 mV (SHE);

Extent of micro-structural & micro-chemical changes vary with irradiation temperature, neutron fluence, flux, & energy spectrum 7

Materials series with low dose exposure Help to understand the thresholds for irradiation effects:

Fracture and tearing toughness, Irradiation-assisted stress corrosion cracking (IASCC)

Irradiation assisted stress corrosion cracking (IASCC)

Materials with high dose exposure Help to understand if saturation of mechanical Help to understand if saturation of mechanical properties occurs:

Radiation-induced segregation (RIS)

Fracture toughness (FT) and tensile properties Fracture toughness (FT) and tensile properties Irradiation-assisted stress corrosion cracking (IASCC)

Help to understand the dose dependence for the :

Onset for the void swelling and the extent of the void swelling (if any at a given temperature and exposure) 8

Threshold and Saturation Slow Irradiation : No shock Threshold : Lower Boundary diti t

b ifi ff t

Slow Irradiation : No shock Time for stress induced damage recovery condition to observe a specific effect LOW DOSE / dpa Fast Irradiation : Not enough time for Saturation : Upper Boundary condition to observe a specific effect any stress recovery before damage 9

co d t o to obse ve a spec c e ect High Dose / dpa

Chemical composition of austenitic stainless steels austenitic stainless steels 10 Ref: Y, Chen et al., NUREG/CR - 7018(2010) and 6965 (2008)

Neutron irradiation of SSs leads to the formation of:

Point defect clusters (black dots)

Dislocation loops Dislocations network s Cavities

(

id f

i d/

b bbl

)

Based on experimental observations it has been (voids of vacancies and/or gas bubbles)

Precipitates Ref: O. Chopra and A. S. Rao, NUREG/CR-7027 (2010)

Based on experimental observations it has been suggested that:

Microstructural changes vary with irradiation condition, i.e., temperature, fluence, g

y p

dose rate, & spectrum, and material condition & composition Below 300oC: black spot defect clusters & faulted dislocation loops Above 300oC: large faulted loops, network dislocations, cavities/voids, & precipitates Depending on material & irradiation condition, loop density & size reach saturation 11 p

g p

y Microstructural changes correlate well with change in yield strength

Continuing increase in IASCC susceptibility at higher dose can not be explained readily

Irradiation Microstructure -

Density/Size of Loops with Dose Density/Size of Loops with Dose

N t

i di ti f SS 11 12

Neutron irradiation of SSs leads to the formation of:

Point defect clusters (black dots)

Dislocation loops 1022 1023 8

9 10 nsity (m-3)

Diameter (nm)

Loop Size Loop Density Dislocation loops Network dislocations Cavities (voids of vacancies and/or gas bubbles) 1021 5

6 7

8 Type 304 SS Loop Den Average Loop LWR Irradiation at 275-290C and/or gas bubbles)

Precipitates 1020 4

5 0

1 2

3 4

5 6

7 8

Type 316 SS Irradiation Dose (dpa)

Ref: O. Chopra and A. S. Rao, NUREG/CR-7027 (2010)

Microstructural changes vary with irradiation condition, i.e., temperature, fluence, dose rate, & spectrum, and material condition & composition Below 300oC: black spot defect clusters & faulted dislocation loops Above 300oC: large faulted loops, network dislocations, cavities/voids, & precipitates 12 Above 300 C: large faulted loops, network dislocations, cavities/voids, & precipitates Depending on material & irradiation condition, loop density & size reach saturation Microstructural changes correlate well with change in yield strength,

Continuing increase in IASCC susceptibility at higher dose can not be explained readily

Density & Size of Loops in LWR BOR 60 I di ti vs BOR-60 Irradiation Small symbols LWR LWR 275°C BOR-60 320°C y

Big symbols BOR-60 Edwards et al., also observed (Ref: 12th Intl. Conf. Env. Deg., 2005)

Loop size is similar for both LWR and BOR -60 reactor irradiations 13 Loop size is similar for both LWR and BOR -60 reactor irradiations.

Loop density is higher for BOR-60 than LWR Both temperature and thermal spectrum has effect.

Irradiation temperature 275oC versus 320oC has some effect in loop evolution initially (dose < 5 dpa).

However for dose levels > 5 dpa irradiation temperature has very little effect.

Recent NRC Research on Irradiation Microstructure - Density/Size of Loops with Dose (Y Chen et al NUREG/CR 7128 2012) with Dose (Y. Chen et al., NUREG/CR-7128, 2012) 14

15 Typical microstructure of un-irradiated solution annealed 304 stainless steel

Irradiated SA 304 SS Irradiation BOR 60 reactor Irradiation BOR -60 reactor Bright Field (BF) images Dark Field (DF) images

(

)

g

loop size distributions 16 Microstructure of irradiated solution annealed 304 SS

Irradiated SA 304 SS Irradiation BOR 60 reactor Irradiation BOR -60 reactor 17 Microstructure of irradiated solution annealed 304 SS with low sulfur

Irradiated 316 LN SS Irradiation BOR 60 reactor Irradiation BOR -60 reactor 18 Microstructure of 316 LN SS irradiated to doses of (a) 24.5 and (c) 45.0 dpa. (b) and (d) are magnified view of (a) and (c) respectively.

Irradiation BOR -60 reactor 19 Quantitative data on the average loop size and loop density and qualitative conclusion on the presence of voids and precipitates in the irradiated stainless steel samples.

Mi h

i l

h t th i

b d

f Micro-chemical changes at the grain boundary of Steels due to Neutron Irradiation (Ref: O. Chopra and A. S. Rao, NUREG/CR-7027 (2010))

20

Micro-chemical changes at the grain boundary of Steels versus N

t I

di ti Neutron Irradiation RIS results in GB depletion of Cr, Mn, Mo & enrichment of Ni, Si, P, C, B Segregation depends strongly on irradiation temperature, dose, & dose rate In LWRs RIS increases with neutron dose peaks at intermediate temp In LWRs, RIS increases with neutron dose, peaks at intermediate temp,

& increases at lower dose rates At 300°C, saturates at 5 dpa 21

Dose Dependence of Grain Boundary Cr, Ni & Si Contents for Stainless Steels Irradiated in LWRs & Fast Reactors Irradiated in LWRs & Fast Reactors PNNL fast 320°C Fujimota fast 330°C Fujimoto PWR 295 320°C Cr Ni Si Fujimoto PWR 295-320°C RIS results in GB depletion of Cr and the enrichment of Ni, Si. (also P, C, B - not shown here)

Stronger RIS in LWRs than BOR-60 (except Fujimoto HP), particularly above 5 dpa Irradiation temperature comparable differences most likely due to dose rate Data from Edwards et al., 13th Intl. Conf. Env. Degrad., P0139, 2007 Irradiation temperature comparable, differences most likely due to dose rate Edwards et al. conclude, BOR60-produced GB compositions provide reasonable representation of LWR-irradiated SSs; however, PWR-irradiated SSs contain cavities that may effect fracture properties Fujimoto et al. conclude from SSRT data, IASCC susceptibility of FBR-irradiated SSs is j

p y

significantly lower than PWR irradiated SSs - <5%IGSCC vs 50-100%IGSCC Data suggest FBR irradiations may not be prototypical of LWRs, particularly >5 dpa 22

Irradiation Hardening 600 800 MPa)

Type 304 SS Irradiated & Tested at 289C High Purity Water with 8 ppm DO 600 800 MPa)

Type 316 SS Irradiated & Tested at 289C High Purity Water with 8 ppm DO 1.35 dpa 1.35 dpa 3.0 dpa 3.0 dpa 400 ngineering Stress (M 400 600 ngineering Stress (M Unirradiated Un irradiated 0.45 dpa 0.45 dpa 0

200 0

10 20 30 40 50 60 Unirradiated 0.45 dpa 1.35 dpa 3.00 dpa En 0

200 0

10 20 30 40 50 Unirradiated 0.45 dpa 1.35 dpa 3.00 dpa En Defect structure & precipitates act as obstacles to dislocation motion that lead to Engineering Strain (%)

Engineering Strain (%)

Ref: Y, Chen et al., NUREG/CR - 7128(2012) matrix strengthening - increase in yield strength & decrease in ductility In general, cavities (voids) are strong barriers, large faulted dislocation loops are intermediate barriers, & small loops & bubbles are weak barriers 23

Effect of Irradiation Dose on Tensile Yield Strength Irradiation temperature 100-427oC Test temperature 125-427oC Yield strength can increase up to five times by 3 to 5 dpa g

p y

p Increase in yield strength follows a square root dependence on dose Yield strength of solution annealed SSs saturates between 3 & 5 dpa At higher dose(>5 dpa), drastic change in deformation mode - dislocation channeling 24

Increase in Yield Stress -

T 304 SS Type 304 SS 1000 600 800 (MPa)

Proposed Curve for BWR Core Shrouds 400 304L/LWR 304/LWR 304L/LWR 304/LWR 304/LWR 304/LWR Yield Stress Irradiation temperature 100-427oC test temperature 125-427oC 0

200 0

2 4

6 8

10 12 304L/LWR 304/LWR 304L/LWR 304/LWR 304/LWR 304/LWR 304/Fast 304/Fast 304/Fast 304L/Fast YS of solution annealed SS increases from 150-200 to 800 MPa at 3-5 dpa Yield stress of SA SSs saturates between 3 & 5 dpa; nearly all SSs show strain Neutron Dose (dpa) 25 softening at higher dose & little or no uniform elongation Proposed curve for BWR core shrouds represent lower bound of the data for Types 304 & 304L SS & their welds 25

Increase in Yield Stress Type 316 SS

- Type 316 SS Irradiation temperature 90-427oC, test temperature 100-427oC 800 1000 800 1000 400 600 316CW 316 316L 316LN ield Stress (MPa)

Fast Reactors 400 600 316CW 316L eld Stress (MPa)

LWRs 0

200 0

2 4

6 8

10 12 316L 316-20%CW 316-20%CW 316-12%CW 316LN 316L Y

0 200 0

2 4

6 8

10 12 316L 316 316L 316 316 316CW Yie YS of SA SS increases from 180-250 to 800 MPa at 3-5 dpa YS of cold worked SS increases from 500-700 to 1000 MPa at 3-5 dpa 0

2 4

6 8

10 12 Neutron Dose (dpa) 0 2

4 6

8 10 12 Neutron Dose (dpa) 26 26 p

Both Fast reactor and LWR irradiation results on the YS of materials is the same SSs irradiated >3 dpa show strain softening & little or no uniform elongation 26

SLOW STRAIN RATE SLOW STRAIN RATE TENSILE TESTING (SSRT)

Ref: Y, Chen et al., NUREG/CR - 7018(2010) and 6965 (2008)

Ref: Y, Chen et al., NUREG/CR 7018(2010) and 6965 (2008) 27

Tests performed in PWR water p

48 10 5

Dose (dpa)

Material Material Type 48 10 5

Dose (dpa)

Material Material Type

304L CW 304 304L

304L SA

304 CW 48 10 5

Type

304L CW 304 304L

304L SA

304 CW 48 10 5

Type

316LN 2 SA

HP 304L 1 SA, Low O

HP 304L 1 SA, High O

304L CW 304, 304L

316LN 2 SA

HP 304L 1 SA, Low O

HP 304L 1 SA, High O

304L CW 304, 304L

316 CW

316 SA

316LN-Ti 2 SA 316, 316L

316 CW

316 SA

316LN-Ti 2 SA 316, 316L 1

High purity 304L stainless steel with high (0.047 wt.%) and low oxygen (0.008 wt.%) content 2 Low carbon, nitrogen 316 stainless steels with and w/o Ti addition (N0 06 0 1 wt % Ti0 027 wt %)

(N0.06-0.1 wt.%, Ti0.027 wt.%)

SA - Solution Annealed, CW - Cold Worked

Irradiation and SSRT Tests Irradiated in BOR-60 (a fast flux reactor)

- Irradiation temperature ~ 320°C

- Damage rate ~ 10-6 dpa/s ( <10-7 dpa/s in Halden)

- Three doses: 5, 10, 48 dpa

SSRT strain rate: 7.4 x 10-7 s-1

Test Conditions:

Sample grip

- In PWR water DO < 10 ppb Temperature: 315°C Pressure: 1800 psig Sample A fractured sample Pressure: 1800 psig Conductivity: 20 mS/cm pH: 6.6 ECP: - 650 mV (ss)

- 700 mV (Pt)

Flow rate: 25 ml /min Flow rate: 25 ml /min 29

Type 304 and 304L SSs Irradiation BOR -60 reactor Irradiation BOR -60 reactor

• CW samples exhibit much higher yield stress and less elongation.

Microstructure of SA and CW Type 304L SS CW Type 304L SS Small dimples with some brittle areas Large dimples 304L SA, 10 dpa 304L CW, 10 dpa p

brittle area brittle area brittle area brittle area area

• SA samples possess fully ductile features while brittle features can be seen in CW samples.

Microstructure of irradiated Type 304L CW SS More brittle areas in higher dose sample and cleavage on sample surface 10-dpa 48-dpa S

ll di l

ith b ittl and cleavage on sample surface.

Small dimples with some brittle areas brittle area brittle area brittle area brittle area 10

Type 316 SS - Effect of Dose Irradiation BOR 60 reactor Irradiation BOR -60 reactor

• Yield stress is higher and elongation is lower for CW SS than that of SA SS at 48 dpa.

• For CW SS, yield stress (0.2) increases with dose from 5 to 10 dpa, and saturates above 10 dpa.

Type 316 SS Irradiated to 47 dpa b

a Irradiated to 47 dpa SA IG cracking IG cracking c

d CW c

d TG cracking TG cracking TG cracking

• Some IG features in SA samples, only TG cracking in CW samples.

• Despite a much lower elongation, the failure surface of the CW sample does not appear more susceptible to cracking than the SA sample at 48 dpa.

15 34

Type 316 SS - CW, 5, 10 d 48 d and 48 dpa 5 dpa 10 dpa 48 dpa Mixed mode cracking Mixed mode mode cracking Mixed mode Mixed mode cracking

Dimple fracture with small mixed-mode cracking areas at all dose levels.

mode cracking 35

Yield Stress - Dose effects SA versus CW SA versus CW

• The increase of yield stress by CW is not affected by irradiation beyond 10 dpa.

• The yield stress differences between SA and CW materials are consistent between 10 to 48 dpa.

p

• The yield stress seems saturate at 5-10 dpa.

36

Loss of ductility - Dose effects SA versus CW Halden SA versus CW Test in Environment Total elongation Reduction of area BOR60

Total elongation and reduction of area decreases with increasing dose up to 48 dpa.

37

SSRT Tests - Effect of S Content 1000 Type 304 SSs, SA Irr. Temp ~320oC Dose 4 8 dpa a

Irr. Temp 320oC D

4 8 d 600 800 Stress (MPa)

Dose = 4.8 dpa C9 C1 C12 Low-S Dose ~ 4.8 dpa 0

200 400 0

2 4

6 8

10 12 14 S

Strain rate = 7.4x10-7 s-1 C1 Low-S (0.003%)

Low-S High-S Strain (%)

38 IG cracking is severe in the high-S Type 304 SS, but no IG fracture in the low-S Type 304 SS High-S (0.016%)

HP Type 304L SS SA with high - O (0.008%) and low-O (0.0047%)

Low-O, RA ~ 80%

High-O, RA ~ 60%

700 HP 304L SS, SA 500 600 700 a)

High-O, 9.6 dpa HP 304L SS, SA Low-O, 9.6 dpa 9.6 dpa 200 300 400 Stress (MPa High-O, 47.5 dpa p

Low-O, 47.5 dpa 9.6 dpa 0

100 0

1 2

3 4

5 6

7 8

Strain (%)

Test temp. = 315oC Strain rate = 7.4 x 10 -7 s-1 47.5 dpa

• A load drop beyond yield is observed for all HP 304L samples, regardless of their O content.

The low O specimens are more ductile than the high O specimens 47.5 dpa

• The low-O specimens are more ductile than the high-O specimens.

• No IG cracking was observed in low-O specimens.

RA - reduction of area

500 600 700 High-O, 9.6 dpa HP 304L SS, SA 300 400 500 Stress (MPa)

High-O, 47.5 dpa Low-O, 9.6 dpa Low-O, 47.5 dpa 0

100 200 0

1 2

3 4

5 6

7 8

St i

(%)

Test temp. = 315oC Strain rate = 7.4 x 10 -7 s-1 p

Strain (%)

40

HP 304L SS - 10 dpa, low O vs high O low-O vs. high-O Low-O High-O 10 dpa, 10 dpa, RA=60%, dimples RA=82%, dimples g

p,

RA - reduction of area

HP 304L SS - 48 dpa, low O vs high O RA76%, dimples RA58%, dimples Low-O High-O low-O vs. high-O 48 dpa, 48 dpa,

Fracture morphology was unchanged with increasing dose from 10 to 48 dpa.

Dimples remain the dominant features on failure surface.

Reduction of area (RA) was similar to that of 10-dpa, ~60% for high-O, and

~80% for low-O specimens.

Crack Growth Rate (CGR) Testing 43

Crack Growth Rate (CGR) Tests

• Fatigue cyclic loading with triangle waveform at 1-2 Hz and load ratio 0.2-0.3 are used to obtained a sharp crack.

p

• Cyclic loading with saw tooth waveform of increasing load ratio and rise time.

- The obtained CGR is compared with CGR in air to evaluate environmental enhancement for each step.

- If the observed CGR is higher than CGR in air, then we continue to increase the rise time, load ratio (R) up to 1000 s rise time and R=0.5-0.7. Load ratio = (Minimum load/Maximum load)

(

)

- If the observed CGR fall back to CGR in air, we repeat the cyclic loading steps until we observe the environmental enhancement again.

S t th t

t i t

t l d

ith ith t

i di

• Set the test in a constant load with or without periodic partial unloading (PPU)

- Eg. CASS is known to be difficult to crack, especially in low-corrosion-potential environments. Therefore PPU is applied for CASS sample 44 potential environments. Therefore PPU is applied for CASS sample testing.

- PPU is done every 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> depending upon the situation (whether crack is initiated or not).

Effect of Environment on Crack Growth Rate on Crack Growth Rate Under more rapid cycling loading condition pre-cracking occurs &

crack growth will be dominated by mechanical fatigue Note:

For stress intensity Kmax 15-18 MPa m1/2 45 y

max environmental enhancement typically occurs at R 0.5 & rise time 30 s; fracture morphology changes from transgranular (TG) to intergranular(IG)

Note:

To transition TG fatigue crack to IG SCC fracture Change rise time from 30 1000 seconds & loading conditions (R) to R = 0.5 0.7.

Cyclic CGR on austenitic stainless steels in 300-335 ppb DO environment 10-7 10-6 Austenitic SS 0.45 dpa 289°C, 300-350 ppb DO 10-7 10-6 Austenitic SS 1.35 dpa 289°C, 300-350 ppb DO 10-10 10-9 10-8 GRenv (m/s) pp 10-10 10-9 10-8 GRenv (m/s) pp 10-13 10-12 10-11 10 304L SS (Spec. C3-A) 316 SS (Spec. C21-A)

CG 10-13 10-12 10-11 10 304L SS (Spec. C3-B) 316 SS (Spec. C21-B)

CG 10 3 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 CGRair (m/s) 10 13 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 CGRair (m/s)

Cyclic CGRs for austenitic SSs irradiated to (a) 0.45 dpa, (b) 1.35 dpa, and (c) 3 dpa.

At 0 45 dpa no environmental enhancement was detected for Type 304L SS

At 0.45 dpa no environmental enhancement was detected for Type 304L SS.

Moderate enhancement was observed for Type 316 SS (specimen C21-A).

With increasing dose, environmentally enhanced cracking also increases.

The difference in cyclic CGRs for different austenitic SSs tend to decrease.

At 1.35 and 3 dpa the cyclic CGRs for Type 304L and 316 SS are nearly identical.

46 y

y y

The low-carbon Type 316 SS (specimen C16-B) showed slightly lower cyclic CGRs than the normal-carbon-content Type 316 SS (specimen C21-C) at 3 dpa

Dose dependence of constant load CGR for austenitic stainless steels in BWR (NWC) environment steels in BWR (NWC) environment Constant-load CGRs versus stress intensity for austenitic SSs irradiated to (a) 0 45 dpa (b) 1 35 dpa and (c) 3 dpa 47 (a) 0.45 dpa, (b) 1.35 dpa, and (c) 3 dpa.

Dose dependence of cyclic CGRs for Type 304L and 316 SSs 10 7 10-6 Austenitic SSs (304L, 316) 289°C, 300-350 ppb DO for Type 304L and 316 SSs.

10-9 10-8 10-7 m/s) 1.35 dpa 3 dpa 10-11 10-10 10 CGRenv (m 0 45 dpa 10-13 10-12 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 0.45 dpa 1.35 dpa 3 dpa 0.45 dpa 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 CGRair (m/s)

The IASCC susceptibility is observed to increase with an increase in neutron dose 48

44 CGRI-JR48, Spec. BR-01 H

t CIR BR i 5 d c

10-8 CGR_JR-48, Spec. BR-01 7.700 7.750 28 32 36 40 ength (mm)

Pa m0.5)

Heat: CIR BR, irr. ~ 5 dpa Low-DO water, 320oC 18a Partial unloading, R=0.7, every 2 hrs K

a 10-9 m/s)

NUREG-0313 Curve Heat: CIR BR, irr. ~ 5 dpa Low-DO water, 320oC 7 550 7.600 7.650 12 16 20 24 Crack Le K (MP 18b Partial unloading, R=0.7, every 4 hrs a

10-10 mental CGR (m Curve 7.550 2390 2395 2400 2405 2410 2415 2420 2425 2430 Time (h) 10-11 Experim 287°C Constant K

~K1.3 CGRs for a 2-hour hold time are about a factor of two greater than the CGRs of a 4-hour hold time, suggesting the stepped crack growth also occurred at 10-12 5

10 15 20 25 30 35 40 unloading every 2 hrs unloading every 4 hrs constant K suggesting the stepped crack growth also occurred at lower stress intensity levels.

The ligaments are being broken to support additional crack growth and or The passive layer is broken with unloading to result in 49 Stress Intensity K (MPa m1/2) p y

g higher CGR.

Summary Summary

- Microstructural changes vary with irradiation condition, i.e., temperature, fluence dose rate & spectrum and material condition & composition fluence, dose rate, & spectrum, and material condition & composition

- Below 300oC: black spot defect clusters & faulted dislocation loops

- Above 300oC: large faulted loops, network dislocations, cavities/voids, &

precipitates p

p

- RIS results in GB depletion of Cr, Mn, Mo & enrichment of Ni, Si, P, C, B

- Segregation depends strongly on irradiation temperature, dose, & dose rate In LWRs, RIS increases with neutron dose, peaks at intermediate temp,

, p p,

& increases at lower dose rates

- Defect structure & precipitates act as obstacles to dislocation motion that lead to matrix strengthening (work hardening) - increase in yield strength &

decrease in ductility

- Yield stress is higher and elongation is lower for CW SS than that of SA SS 50

Summary Cont.

The increase of yield stress of either SA or CW is not affected by neutron irradiation beyond 10 dpa, however the total elongation and reduction of Summary Cont.

y p

g area tends to continuously decrease with increasing dose up to 48 dpa.

SA samples possess fully ductile features while brittle features are seen in CW samples.

While some inter granular (IG) cracking is observed in SA samples, predominant trans granular (TG) cracking is noticed in CW samples.

With an increase in the neutron dose, the environmentally enhanced cracking increases in both 304 and 316 steels cracking increases in both 304 and 316 steels.

IG cracking is severe in the high-S Type 304 SS, but not in the low-S 304 SS The low-O specimens are more ductile than the high-O steels.

At 0 45 dpa no environmental enhancement was detected for Type 304L SS At 0.45 dpa no environmental enhancement was detected for Type 304L SS.

Moderate enhancement was observed for Type 316 SS (specimen C21-A).

With an increase in the neutron dose, the environmentally enhanced cracking increases in both 304 and 316 steels 51 cracking increases in both 304 and 316 steels.

The IASCC susceptibility of both 304 and 316 steels increase with an increase in neutron dose up to 5 dpa.

Irradiation Assisted Degradation of LWR Core Irradiation Assisted Degradation of LWR Core Internal Materials: Brief Review Appajosula S. Rao Office of Nuclear Regulatory Research US Nuclear Regulatory Commission, Washington, DC, 20555 The views expressed in this presentation are not necessarily those of the U S Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission For Presentation at the Office of Nuclear Regulatory Research, Seminar on April 14th 2015.

United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of:

Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #:

Identified:

Admitted:

Withdrawn:

Rejected:

Stricken:

Other:

NYS000495-00-BD01 11/5/2015 11/5/2015 NYS000495 Submitted: June 9, 2015

Contents

• Introduction / Technical Issues B

k d

Background

Chemical Composition of Austenitic Stainless Steels Irradiation Microstructure - Density/Size of Loops with Dose Dose

• Recent NRC Research on Irradiation Microstructure Micro-chemical changes at the grain boundary of Steels versus Neutron Irradiation versus Neutron Irradiation Irradiation Hardening Slow Strain rate Test Results Crack Growth Rate (CGR) Tests

(

)

Summary 2

Introduction Technical Issues identify susceptible materials & conditions

identify susceptible materials & conditions,

determine crack growth rates to assure that the selected inspection intervals are adequate,

verify effectiveness of proposed mitigative measures i

ti t

t ti l f

di ti b ittl t i l di i ti 3

investigate potential of radiation embrittlement, including synergistic effects of thermal & neutron embrittlement

evaluate effects of void swelling, including the associated decrease in ductility

Void Swelling

Void swelling of most reactors internals is Void Swelling

Void swelling of most reactors internals is not expected to be limiting over the current licensing period licensing period

Continued research work will explore the extent of void swelling over the the extent of void swelling over the extended operating life of present reactors (i.e 54 EFPY) 4 reactors (i.e 54 EFPY)

Helium bubbles in Mono crystalline Nickel (110) Irradiated with 100 keV ions 5

y

(

)

6

=

Background===

Neutron irradiation changes the material microstructure (radiation hardening) & micro-chem. (radiation induced segregation) that leads to:

increase in yield strength & decrease in ductility

degradation of fracture toughness

increased susceptibility to IASCC

void formation, coalescence and swelling

creep relaxation

Neutron irradiation also changes the water chemistry (radiolysis);

BWR normal water chemistry results in an increase in corrosion y

potential,

Ecorr 200 mV (SHE);

PWR water contains H2 toscavenge radiolysis products,

E

800 mV (SHE);

Ecorr - 800 mV (SHE);

Extent of micro-structural & micro-chemical changes vary with irradiation temperature, neutron fluence, flux, & energy spectrum 7

Materials series with low dose exposure Help to understand the thresholds for irradiation effects:

Fracture and tearing toughness, Irradiation-assisted stress corrosion cracking (IASCC)

Irradiation assisted stress corrosion cracking (IASCC)

Materials with high dose exposure Help to understand if saturation of mechanical Help to understand if saturation of mechanical properties occurs:

Radiation-induced segregation (RIS)

Fracture toughness (FT) and tensile properties Fracture toughness (FT) and tensile properties Irradiation-assisted stress corrosion cracking (IASCC)

Help to understand the dose dependence for the :

Onset for the void swelling and the extent of the void swelling (if any at a given temperature and exposure) 8

Threshold and Saturation Slow Irradiation : No shock Threshold : Lower Boundary diti t

b ifi ff t

Slow Irradiation : No shock Time for stress induced damage recovery condition to observe a specific effect LOW DOSE / dpa Fast Irradiation : Not enough time for Saturation : Upper Boundary condition to observe a specific effect any stress recovery before damage 9

co d t o to obse ve a spec c e ect High Dose / dpa

Chemical composition of austenitic stainless steels austenitic stainless steels 10 Ref: Y, Chen et al., NUREG/CR - 7018(2010) and 6965 (2008)

Neutron irradiation of SSs leads to the formation of:

Point defect clusters (black dots)

Dislocation loops Dislocations network s Cavities

(

id f

i d/

b bbl

)

Based on experimental observations it has been (voids of vacancies and/or gas bubbles)

Precipitates Ref: O. Chopra and A. S. Rao, NUREG/CR-7027 (2010)

Based on experimental observations it has been suggested that:

Microstructural changes vary with irradiation condition, i.e., temperature, fluence, g

y p

dose rate, & spectrum, and material condition & composition Below 300oC: black spot defect clusters & faulted dislocation loops Above 300oC: large faulted loops, network dislocations, cavities/voids, & precipitates Depending on material & irradiation condition, loop density & size reach saturation 11 p

g p

y Microstructural changes correlate well with change in yield strength

Continuing increase in IASCC susceptibility at higher dose can not be explained readily

Irradiation Microstructure -

Density/Size of Loops with Dose Density/Size of Loops with Dose

N t

i di ti f SS 11 12

Neutron irradiation of SSs leads to the formation of:

Point defect clusters (black dots)

Dislocation loops 1022 1023 8

9 10 nsity (m-3)

Diameter (nm)

Loop Size Loop Density Dislocation loops Network dislocations Cavities (voids of vacancies and/or gas bubbles) 1021 5

6 7

8 Type 304 SS Loop Den Average Loop LWR Irradiation at 275-290C and/or gas bubbles)

Precipitates 1020 4

5 0

1 2

3 4

5 6

7 8

Type 316 SS Irradiation Dose (dpa)

Ref: O. Chopra and A. S. Rao, NUREG/CR-7027 (2010)

Microstructural changes vary with irradiation condition, i.e., temperature, fluence, dose rate, & spectrum, and material condition & composition Below 300oC: black spot defect clusters & faulted dislocation loops Above 300oC: large faulted loops, network dislocations, cavities/voids, & precipitates 12 Above 300 C: large faulted loops, network dislocations, cavities/voids, & precipitates Depending on material & irradiation condition, loop density & size reach saturation Microstructural changes correlate well with change in yield strength,

Continuing increase in IASCC susceptibility at higher dose can not be explained readily

Density & Size of Loops in LWR BOR 60 I di ti vs BOR-60 Irradiation Small symbols LWR LWR 275°C BOR-60 320°C y

Big symbols BOR-60 Edwards et al., also observed (Ref: 12th Intl. Conf. Env. Deg., 2005)

Loop size is similar for both LWR and BOR -60 reactor irradiations 13 Loop size is similar for both LWR and BOR -60 reactor irradiations.

Loop density is higher for BOR-60 than LWR Both temperature and thermal spectrum has effect.

Irradiation temperature 275oC versus 320oC has some effect in loop evolution initially (dose < 5 dpa).

However for dose levels > 5 dpa irradiation temperature has very little effect.

Recent NRC Research on Irradiation Microstructure - Density/Size of Loops with Dose (Y Chen et al NUREG/CR 7128 2012) with Dose (Y. Chen et al., NUREG/CR-7128, 2012) 14

15 Typical microstructure of un-irradiated solution annealed 304 stainless steel

Irradiated SA 304 SS Irradiation BOR 60 reactor Irradiation BOR -60 reactor Bright Field (BF) images Dark Field (DF) images

(

)

g

loop size distributions 16 Microstructure of irradiated solution annealed 304 SS

Irradiated SA 304 SS Irradiation BOR 60 reactor Irradiation BOR -60 reactor 17 Microstructure of irradiated solution annealed 304 SS with low sulfur

Irradiated 316 LN SS Irradiation BOR 60 reactor Irradiation BOR -60 reactor 18 Microstructure of 316 LN SS irradiated to doses of (a) 24.5 and (c) 45.0 dpa. (b) and (d) are magnified view of (a) and (c) respectively.

Irradiation BOR -60 reactor 19 Quantitative data on the average loop size and loop density and qualitative conclusion on the presence of voids and precipitates in the irradiated stainless steel samples.

Mi h

i l

h t th i

b d

f Micro-chemical changes at the grain boundary of Steels due to Neutron Irradiation (Ref: O. Chopra and A. S. Rao, NUREG/CR-7027 (2010))

20

Micro-chemical changes at the grain boundary of Steels versus N

t I

di ti Neutron Irradiation RIS results in GB depletion of Cr, Mn, Mo & enrichment of Ni, Si, P, C, B Segregation depends strongly on irradiation temperature, dose, & dose rate In LWRs RIS increases with neutron dose peaks at intermediate temp In LWRs, RIS increases with neutron dose, peaks at intermediate temp,

& increases at lower dose rates At 300°C, saturates at 5 dpa 21

Dose Dependence of Grain Boundary Cr, Ni & Si Contents for Stainless Steels Irradiated in LWRs & Fast Reactors Irradiated in LWRs & Fast Reactors PNNL fast 320°C Fujimota fast 330°C Fujimoto PWR 295 320°C Cr Ni Si Fujimoto PWR 295-320°C RIS results in GB depletion of Cr and the enrichment of Ni, Si. (also P, C, B - not shown here)

Stronger RIS in LWRs than BOR-60 (except Fujimoto HP), particularly above 5 dpa Irradiation temperature comparable differences most likely due to dose rate Data from Edwards et al., 13th Intl. Conf. Env. Degrad., P0139, 2007 Irradiation temperature comparable, differences most likely due to dose rate Edwards et al. conclude, BOR60-produced GB compositions provide reasonable representation of LWR-irradiated SSs; however, PWR-irradiated SSs contain cavities that may effect fracture properties Fujimoto et al. conclude from SSRT data, IASCC susceptibility of FBR-irradiated SSs is j

p y

significantly lower than PWR irradiated SSs - <5%IGSCC vs 50-100%IGSCC Data suggest FBR irradiations may not be prototypical of LWRs, particularly >5 dpa 22

Irradiation Hardening 600 800 MPa)

Type 304 SS Irradiated & Tested at 289C High Purity Water with 8 ppm DO 600 800 MPa)

Type 316 SS Irradiated & Tested at 289C High Purity Water with 8 ppm DO 1.35 dpa 1.35 dpa 3.0 dpa 3.0 dpa 400 ngineering Stress (M 400 600 ngineering Stress (M Unirradiated Un irradiated 0.45 dpa 0.45 dpa 0

200 0

10 20 30 40 50 60 Unirradiated 0.45 dpa 1.35 dpa 3.00 dpa En 0

200 0

10 20 30 40 50 Unirradiated 0.45 dpa 1.35 dpa 3.00 dpa En Defect structure & precipitates act as obstacles to dislocation motion that lead to Engineering Strain (%)

Engineering Strain (%)

Ref: Y, Chen et al., NUREG/CR - 7128(2012) matrix strengthening - increase in yield strength & decrease in ductility In general, cavities (voids) are strong barriers, large faulted dislocation loops are intermediate barriers, & small loops & bubbles are weak barriers 23

Effect of Irradiation Dose on Tensile Yield Strength Irradiation temperature 100-427oC Test temperature 125-427oC Yield strength can increase up to five times by 3 to 5 dpa g

p y

p Increase in yield strength follows a square root dependence on dose Yield strength of solution annealed SSs saturates between 3 & 5 dpa At higher dose(>5 dpa), drastic change in deformation mode - dislocation channeling 24

Increase in Yield Stress -

T 304 SS Type 304 SS 1000 600 800 (MPa)

Proposed Curve for BWR Core Shrouds 400 304L/LWR 304/LWR 304L/LWR 304/LWR 304/LWR 304/LWR Yield Stress Irradiation temperature 100-427oC test temperature 125-427oC 0

200 0

2 4

6 8

10 12 304L/LWR 304/LWR 304L/LWR 304/LWR 304/LWR 304/LWR 304/Fast 304/Fast 304/Fast 304L/Fast YS of solution annealed SS increases from 150-200 to 800 MPa at 3-5 dpa Yield stress of SA SSs saturates between 3 & 5 dpa; nearly all SSs show strain Neutron Dose (dpa) 25 softening at higher dose & little or no uniform elongation Proposed curve for BWR core shrouds represent lower bound of the data for Types 304 & 304L SS & their welds 25

Increase in Yield Stress Type 316 SS

- Type 316 SS Irradiation temperature 90-427oC, test temperature 100-427oC 800 1000 800 1000 400 600 316CW 316 316L 316LN ield Stress (MPa)

Fast Reactors 400 600 316CW 316L eld Stress (MPa)

LWRs 0

200 0

2 4

6 8

10 12 316L 316-20%CW 316-20%CW 316-12%CW 316LN 316L Y

0 200 0

2 4

6 8

10 12 316L 316 316L 316 316 316CW Yie YS of SA SS increases from 180-250 to 800 MPa at 3-5 dpa YS of cold worked SS increases from 500-700 to 1000 MPa at 3-5 dpa 0

2 4

6 8

10 12 Neutron Dose (dpa) 0 2

4 6

8 10 12 Neutron Dose (dpa) 26 26 p

Both Fast reactor and LWR irradiation results on the YS of materials is the same SSs irradiated >3 dpa show strain softening & little or no uniform elongation 26

SLOW STRAIN RATE SLOW STRAIN RATE TENSILE TESTING (SSRT)

Ref: Y, Chen et al., NUREG/CR - 7018(2010) and 6965 (2008)

Ref: Y, Chen et al., NUREG/CR 7018(2010) and 6965 (2008) 27

Tests performed in PWR water p

48 10 5

Dose (dpa)

Material Material Type 48 10 5

Dose (dpa)

Material Material Type

304L CW 304 304L

304L SA

304 CW 48 10 5

Type

304L CW 304 304L

304L SA

304 CW 48 10 5

Type

316LN 2 SA

HP 304L 1 SA, Low O

HP 304L 1 SA, High O

304L CW 304, 304L

316LN 2 SA

HP 304L 1 SA, Low O

HP 304L 1 SA, High O

304L CW 304, 304L

316 CW

316 SA

316LN-Ti 2 SA 316, 316L

316 CW

316 SA

316LN-Ti 2 SA 316, 316L 1

High purity 304L stainless steel with high (0.047 wt.%) and low oxygen (0.008 wt.%) content 2 Low carbon, nitrogen 316 stainless steels with and w/o Ti addition (N0 06 0 1 wt % Ti0 027 wt %)

(N0.06-0.1 wt.%, Ti0.027 wt.%)

SA - Solution Annealed, CW - Cold Worked

Irradiation and SSRT Tests Irradiated in BOR-60 (a fast flux reactor)

- Irradiation temperature ~ 320°C

- Damage rate ~ 10-6 dpa/s ( <10-7 dpa/s in Halden)

- Three doses: 5, 10, 48 dpa

SSRT strain rate: 7.4 x 10-7 s-1

Test Conditions:

Sample grip

- In PWR water DO < 10 ppb Temperature: 315°C Pressure: 1800 psig Sample A fractured sample Pressure: 1800 psig Conductivity: 20 mS/cm pH: 6.6 ECP: - 650 mV (ss)

- 700 mV (Pt)

Flow rate: 25 ml /min Flow rate: 25 ml /min 29

Type 304 and 304L SSs Irradiation BOR -60 reactor Irradiation BOR -60 reactor

• CW samples exhibit much higher yield stress and less elongation.

Microstructure of SA and CW Type 304L SS CW Type 304L SS Small dimples with some brittle areas Large dimples 304L SA, 10 dpa 304L CW, 10 dpa p

brittle area brittle area brittle area brittle area area

• SA samples possess fully ductile features while brittle features can be seen in CW samples.

Microstructure of irradiated Type 304L CW SS More brittle areas in higher dose sample and cleavage on sample surface 10-dpa 48-dpa S

ll di l

ith b ittl and cleavage on sample surface.

Small dimples with some brittle areas brittle area brittle area brittle area brittle area 10

Type 316 SS - Effect of Dose Irradiation BOR 60 reactor Irradiation BOR -60 reactor

• Yield stress is higher and elongation is lower for CW SS than that of SA SS at 48 dpa.

• For CW SS, yield stress (0.2) increases with dose from 5 to 10 dpa, and saturates above 10 dpa.

Type 316 SS Irradiated to 47 dpa b

a Irradiated to 47 dpa SA IG cracking IG cracking c

d CW c

d TG cracking TG cracking TG cracking

• Some IG features in SA samples, only TG cracking in CW samples.

• Despite a much lower elongation, the failure surface of the CW sample does not appear more susceptible to cracking than the SA sample at 48 dpa.

15 34

Type 316 SS - CW, 5, 10 d 48 d and 48 dpa 5 dpa 10 dpa 48 dpa Mixed mode cracking Mixed mode mode cracking Mixed mode Mixed mode cracking

Dimple fracture with small mixed-mode cracking areas at all dose levels.

mode cracking 35

Yield Stress - Dose effects SA versus CW SA versus CW

• The increase of yield stress by CW is not affected by irradiation beyond 10 dpa.

• The yield stress differences between SA and CW materials are consistent between 10 to 48 dpa.

p

• The yield stress seems saturate at 5-10 dpa.

36

Loss of ductility - Dose effects SA versus CW Halden SA versus CW Test in Environment Total elongation Reduction of area BOR60

Total elongation and reduction of area decreases with increasing dose up to 48 dpa.

37

SSRT Tests - Effect of S Content 1000 Type 304 SSs, SA Irr. Temp ~320oC Dose 4 8 dpa a

Irr. Temp 320oC D

4 8 d 600 800 Stress (MPa)

Dose = 4.8 dpa C9 C1 C12 Low-S Dose ~ 4.8 dpa 0

200 400 0

2 4

6 8

10 12 14 S

Strain rate = 7.4x10-7 s-1 C1 Low-S (0.003%)

Low-S High-S Strain (%)

38 IG cracking is severe in the high-S Type 304 SS, but no IG fracture in the low-S Type 304 SS High-S (0.016%)

HP Type 304L SS SA with high - O (0.008%) and low-O (0.0047%)

Low-O, RA ~ 80%

High-O, RA ~ 60%

700 HP 304L SS, SA 500 600 700 a)

High-O, 9.6 dpa HP 304L SS, SA Low-O, 9.6 dpa 9.6 dpa 200 300 400 Stress (MPa High-O, 47.5 dpa p

Low-O, 47.5 dpa 9.6 dpa 0

100 0

1 2

3 4

5 6

7 8

Strain (%)

Test temp. = 315oC Strain rate = 7.4 x 10 -7 s-1 47.5 dpa

• A load drop beyond yield is observed for all HP 304L samples, regardless of their O content.

The low O specimens are more ductile than the high O specimens 47.5 dpa

• The low-O specimens are more ductile than the high-O specimens.

• No IG cracking was observed in low-O specimens.

RA - reduction of area

500 600 700 High-O, 9.6 dpa HP 304L SS, SA 300 400 500 Stress (MPa)

High-O, 47.5 dpa Low-O, 9.6 dpa Low-O, 47.5 dpa 0

100 200 0

1 2

3 4

5 6

7 8

St i

(%)

Test temp. = 315oC Strain rate = 7.4 x 10 -7 s-1 p

Strain (%)

40

HP 304L SS - 10 dpa, low O vs high O low-O vs. high-O Low-O High-O 10 dpa, 10 dpa, RA=60%, dimples RA=82%, dimples g

p,

RA - reduction of area

HP 304L SS - 48 dpa, low O vs high O RA76%, dimples RA58%, dimples Low-O High-O low-O vs. high-O 48 dpa, 48 dpa,

Fracture morphology was unchanged with increasing dose from 10 to 48 dpa.

Dimples remain the dominant features on failure surface.

Reduction of area (RA) was similar to that of 10-dpa, ~60% for high-O, and

~80% for low-O specimens.

Crack Growth Rate (CGR) Testing 43

Crack Growth Rate (CGR) Tests

• Fatigue cyclic loading with triangle waveform at 1-2 Hz and load ratio 0.2-0.3 are used to obtained a sharp crack.

p

• Cyclic loading with saw tooth waveform of increasing load ratio and rise time.

- The obtained CGR is compared with CGR in air to evaluate environmental enhancement for each step.

- If the observed CGR is higher than CGR in air, then we continue to increase the rise time, load ratio (R) up to 1000 s rise time and R=0.5-0.7. Load ratio = (Minimum load/Maximum load)

(

)

- If the observed CGR fall back to CGR in air, we repeat the cyclic loading steps until we observe the environmental enhancement again.

S t th t

t i t

t l d

ith ith t

i di

• Set the test in a constant load with or without periodic partial unloading (PPU)

- Eg. CASS is known to be difficult to crack, especially in low-corrosion-potential environments. Therefore PPU is applied for CASS sample 44 potential environments. Therefore PPU is applied for CASS sample testing.

- PPU is done every 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> depending upon the situation (whether crack is initiated or not).

Effect of Environment on Crack Growth Rate on Crack Growth Rate Under more rapid cycling loading condition pre-cracking occurs &

crack growth will be dominated by mechanical fatigue Note:

For stress intensity Kmax 15-18 MPa m1/2 45 y

max environmental enhancement typically occurs at R 0.5 & rise time 30 s; fracture morphology changes from transgranular (TG) to intergranular(IG)

Note:

To transition TG fatigue crack to IG SCC fracture Change rise time from 30 1000 seconds & loading conditions (R) to R = 0.5 0.7.

Cyclic CGR on austenitic stainless steels in 300-335 ppb DO environment 10-7 10-6 Austenitic SS 0.45 dpa 289°C, 300-350 ppb DO 10-7 10-6 Austenitic SS 1.35 dpa 289°C, 300-350 ppb DO 10-10 10-9 10-8 GRenv (m/s) pp 10-10 10-9 10-8 GRenv (m/s) pp 10-13 10-12 10-11 10 304L SS (Spec. C3-A) 316 SS (Spec. C21-A)

CG 10-13 10-12 10-11 10 304L SS (Spec. C3-B) 316 SS (Spec. C21-B)

CG 10 3 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 CGRair (m/s) 10 13 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 CGRair (m/s)

Cyclic CGRs for austenitic SSs irradiated to (a) 0.45 dpa, (b) 1.35 dpa, and (c) 3 dpa.

At 0 45 dpa no environmental enhancement was detected for Type 304L SS

At 0.45 dpa no environmental enhancement was detected for Type 304L SS.

Moderate enhancement was observed for Type 316 SS (specimen C21-A).

With increasing dose, environmentally enhanced cracking also increases.

The difference in cyclic CGRs for different austenitic SSs tend to decrease.

At 1.35 and 3 dpa the cyclic CGRs for Type 304L and 316 SS are nearly identical.

46 y

y y

The low-carbon Type 316 SS (specimen C16-B) showed slightly lower cyclic CGRs than the normal-carbon-content Type 316 SS (specimen C21-C) at 3 dpa

Dose dependence of constant load CGR for austenitic stainless steels in BWR (NWC) environment steels in BWR (NWC) environment Constant-load CGRs versus stress intensity for austenitic SSs irradiated to (a) 0 45 dpa (b) 1 35 dpa and (c) 3 dpa 47 (a) 0.45 dpa, (b) 1.35 dpa, and (c) 3 dpa.

Dose dependence of cyclic CGRs for Type 304L and 316 SSs 10 7 10-6 Austenitic SSs (304L, 316) 289°C, 300-350 ppb DO for Type 304L and 316 SSs.

10-9 10-8 10-7 m/s) 1.35 dpa 3 dpa 10-11 10-10 10 CGRenv (m 0 45 dpa 10-13 10-12 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 0.45 dpa 1.35 dpa 3 dpa 0.45 dpa 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 CGRair (m/s)

The IASCC susceptibility is observed to increase with an increase in neutron dose 48

44 CGRI-JR48, Spec. BR-01 H

t CIR BR i 5 d c

10-8 CGR_JR-48, Spec. BR-01 7.700 7.750 28 32 36 40 ength (mm)

Pa m0.5)

Heat: CIR BR, irr. ~ 5 dpa Low-DO water, 320oC 18a Partial unloading, R=0.7, every 2 hrs K

a 10-9 m/s)

NUREG-0313 Curve Heat: CIR BR, irr. ~ 5 dpa Low-DO water, 320oC 7 550 7.600 7.650 12 16 20 24 Crack Le K (MP 18b Partial unloading, R=0.7, every 4 hrs a

10-10 mental CGR (m Curve 7.550 2390 2395 2400 2405 2410 2415 2420 2425 2430 Time (h) 10-11 Experim 287°C Constant K

~K1.3 CGRs for a 2-hour hold time are about a factor of two greater than the CGRs of a 4-hour hold time, suggesting the stepped crack growth also occurred at 10-12 5

10 15 20 25 30 35 40 unloading every 2 hrs unloading every 4 hrs constant K suggesting the stepped crack growth also occurred at lower stress intensity levels.

The ligaments are being broken to support additional crack growth and or The passive layer is broken with unloading to result in 49 Stress Intensity K (MPa m1/2) p y

g higher CGR.

Summary Summary

- Microstructural changes vary with irradiation condition, i.e., temperature, fluence dose rate & spectrum and material condition & composition fluence, dose rate, & spectrum, and material condition & composition

- Below 300oC: black spot defect clusters & faulted dislocation loops

- Above 300oC: large faulted loops, network dislocations, cavities/voids, &

precipitates p

p

- RIS results in GB depletion of Cr, Mn, Mo & enrichment of Ni, Si, P, C, B

- Segregation depends strongly on irradiation temperature, dose, & dose rate In LWRs, RIS increases with neutron dose, peaks at intermediate temp,

, p p,

& increases at lower dose rates

- Defect structure & precipitates act as obstacles to dislocation motion that lead to matrix strengthening (work hardening) - increase in yield strength &

decrease in ductility

- Yield stress is higher and elongation is lower for CW SS than that of SA SS 50

Summary Cont.

The increase of yield stress of either SA or CW is not affected by neutron irradiation beyond 10 dpa, however the total elongation and reduction of Summary Cont.

y p

g area tends to continuously decrease with increasing dose up to 48 dpa.

SA samples possess fully ductile features while brittle features are seen in CW samples.

While some inter granular (IG) cracking is observed in SA samples, predominant trans granular (TG) cracking is noticed in CW samples.

With an increase in the neutron dose, the environmentally enhanced cracking increases in both 304 and 316 steels cracking increases in both 304 and 316 steels.

IG cracking is severe in the high-S Type 304 SS, but not in the low-S 304 SS The low-O specimens are more ductile than the high-O steels.

At 0 45 dpa no environmental enhancement was detected for Type 304L SS At 0.45 dpa no environmental enhancement was detected for Type 304L SS.

Moderate enhancement was observed for Type 316 SS (specimen C21-A).

With an increase in the neutron dose, the environmentally enhanced cracking increases in both 304 and 316 steels 51 cracking increases in both 304 and 316 steels.

The IASCC susceptibility of both 304 and 316 steels increase with an increase in neutron dose up to 5 dpa.