ML071770248

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Comment (16) Submitted by Charles A. Tomes on Proposed Rule Pr 50 Regarding Industry Codes and Standards; Amended Requirements
ML071770248
Person / Time
Site: Millstone, Kewaunee, Point Beach, Prairie Island, Surry, Farley, Robinson, South Texas, San Onofre, Fort Calhoun  Duke Energy icon.png
Issue date: 06/15/2007
From: Tomes C
- No Known Affiliation
To:
NRC/SECY/RAS
SECY RAS
References
72FR16731 00016, PR-50, RIN 3150-AH76
Download: ML071770248 (84)
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Text

PR 50 (72FR16731)

June 15, 2007 DOCKETED USNRC Secretary June 15, 2007 (2:45pm)

ATTN: Rulemakings and Adjudications Staff Nuclear Regulatory Commission OFFICE OF SECRETARY Washngtn, C 2055-001RULEMAKINGS AND Wash 55-001ADJUDICATIONS ngt n, C 20 STAFF Charles A Tomes 2045 Fawn Lane Green Bay, WI 54304

Subject:

RIN 3150 - AH76, Response to NRC Request for Public Comment to Incorporate ASME Code Case N-729 Revision I With Supplemental Requirements into the Code of Federal Regulation

Dear Sir:

References:

I1. PWSCC Lifetime Evaluation on Alloy 690, 52, and 152 for PWR Materials, MHI, EPRI PWSCC of Alloy 600 2007 International Conference & Exhibition, June 11 - 14, 2007, Atlanta, GA

2. Crack Growth Response in Simulated PWR Water of Alloy 152 Weld Metal, PNNL, ICG-EAC, Hualien, Taiwan, April 2007
3. PW.SCC Growth Rates in Alloy 690 and Its Weld Metal, GE Global Research Center, EPRI PWSCC of Alloy 600 2007 International Conference. & Exhibition, June 1.1 - 14, 2007, Atlanta, GA The author of this letter wishes to thank the Nuclear Regulatory Commission (NRC) for an opportunity to provide comments on NRC's plans to incorporate ASME Code Case N-729 Revision I as amended by NRC supplemental requirements into the Code of Federal Regulation for, nondestructive testing of replacement reactor vessel head control rod drive mechanism (CRDM) tubing and j-groove weld metal.

The comments provided herein are based on planning, research, and replacement reactor vessel head activities spanning back to the early 1990's. Following the initial reports of cracking at Bugey Unit 3, the commercial nuclear power industry initiated research projects to develop alternate materials that are highly resistant to PWSCC. Coincident with incidents of CRDM j-groove weld cracking in the USA, utilities initiated plans to replace reactor vessel heads with materials that are highly resistant to primary water stress corrosion cracking (PWSCC).

While employed at the Nuclear Management Company I was involved with development of contracts and oversight activities to fabricate and install replacement reactor vessel heads at five (5) nuclear plants: Kewaunee Power Station, Prairie Island Nuclear Generating Station Units 1 and 2, and Point Beach Nuclear Plant Units 1 and 2. As part of this project, Mitsubishi Heavy Industries fabricated the five (5) replacement reactor vessel heads under contract to Westinghouse Electric Company. To date, all five (5) reactor vessel heads have been replaced with CRDM tubing and j-groove weld metal Ye mrpI104- 36Z 0c o&7 See 1-0 a

fabricated from Alloy 690, 52, and 152 materials. Since completion of this project my employment status has changed from Nuclear Management Company to Dominion Energy Kewaunee, Inc as Dominion purchased the Kewaunee Power Station in Summer 2005. Information provided herein is applicable to Kewaunee Power Station, Prairie Island Nuclear Generating Station 1 and 2, and Point Beach Nuclear Plant Unit 1 and 2.

A primary goal of the project was to build quality into the replacement reactor vessel heads to prevent leakage and reduce the need for detailed inspectidns during future plant operation. The following enhancements were included in to reduce the likelihood of PSWCC, leakage, and problems encountered during future inspections:

1. Alloy 690, 52, and 152 was used for fabrication of the CRDM tubing and i-groove welds,
2. The grain size from 5 to 7 was selected to optimize PWSCC resistance and also ensure ultrasonic examination,
3. Narrow groove i-groove welds were used to reduce the residual stress,
4. During ji-groove welding the ID surface of the Alloy 690 tubing was cooled with water to minimize stresses,
5. The threaded joint on the CRDM latch mechanism was replaced with a butt weld to eliminate the possibility of leakage,
6. Vents on top of the CRDM rod housings were eliminated to reduce the possibility of leakage,
7. The Man-non clamps on the thermnocouple ports were replaced with leak free CETNA to reduce the possibility of leakage,
8. Removable insulation with inspection ports was installed,
9. No repairs were pen~nitted on the Alloy 690 tubing,
10. Penetrant testing was performned at pre-defined increments during welding of the i-groove weld,
11. A "PT White" criteria was used on the final surface of the i-groove weld and alloy 690 tubing,
12. The surface of the i-groove welds were polished smooth to permnit eddy current testing and penetrant testing,
13. The distance between the thermnal sleeve and top of the funnel was increased to better accommodate inspection probes during future inservice inspections,
14. Preservice Inspection (PSI) included bare metal visual (BMV) inspec tions of the reactor vessel head, eddy current testing of the i-groove weld and alloy 690 tubing above and below the j-groove weld, and ultrasonic inspection of the alloy 690 tubing.

Upon completion of the replacement reactor vessel head projects for the Nuclear Management Company, I am pleased to communicate that the goals and objectives to improve PWSCC resistance, reduce the possibility of leakage above the reactor vessel head, and reduce potential problems with future inspections are successful.

The decision to award a contract to Mitsubishi Heavy Industries was heavily influenced by the knowledge that they had conducted extensive PWSCC crack initiation testing under accelerated PWR water conditions. Up to presently, most of this information was considered proprietary and had not been released to the public.

As part of the replacement reactor vessel head project Nuclear Management Company contracted' Mitsubishi Heavy Industries to fabricate eight (8) linear feet of alloy 52 and alloy 152 weld metal to be used for future PWSCC testing. Nuclear Management Company donated this material to the Electric Power Research Institute in order for it to be included in various industry PWSCC testing programs. To date, some of this weld metal has been tested under NRC contract by Pacific Northwest National Laboratory and also by GE Global Research.

References I - 3 document PWSCC laboratory test results (applicable to replacement reactor vessel heads at Kewaunee Power Station, Prairie Island Nuclear Generating Station Units 1 and 2, and Point Beach Nuclear Plant Units I and 2) from Mitsubishi Heavy Industries, Pacific Northwest National Laboratory, and GE Global Research that have been recently released to the public.

A copy of the presentation made by Mitsubishi Heavy Industries at the EPRI 2007 International PWSCC of Alloy 600 Conference and Exhibit Show, Atlanta, GA, June 11 - 14, 2007 is included in . The Mitsubishi Heavy Industries PWSCC test results show that cracking has not initiated for Alloy 690, 52, and 152 materials in a simulated PWR environment for approximately 73,000 hrs, 84,000 l-lrs, and 85,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, respectively. All testing was performned at 360 C (680 F).

Testing to date confinrms no crack initiation has occurred in Alloy 690, 52, and 152 materials. Other materials including Alloy 600, 82, or 182 were included in the test matrix and showed evidence of crack initiation early on during testing consistent with industry experience. One factor for understanding the quality of Alloy 690, 52, and 152 is to make adjustments for testing performed at 680 F to operating temperature. When this adjustment is made, a factor of 6.4 applies for base metal and a factor of 14.9 applies for weld metal. These factors can be multiplied directly to the test duration to adjust for differences in temperature. From this data, the equivalent time without initiation for Alloy 690 base metal is approximately 467,200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. Similarly, the equivalent time without crack initiation for weld metal is approximately from 1,251,600 to 1,266,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> for Alloy 52 and Alloy 152, respectively. This equivalent time period for base 'metal and weld metal are 53 years and 142 to 144 years, respectively. It is my understanding that these quality improvements apply to the entire replacement reactor vessel head population supplied by Mitsubishi Heavy Industries to the USA market from 2003 through 2012.

The PWSCC crack growth rates for the Alloy 690, 52, and 152 materials fabricated by Mitsubishi Heavy Industries, donated to the Electrical Power Research Institute by the Nuclear Management Company, and independently tested by Pacific Northwest National Laboratory and GE Global Research are on the order of 10-9 minis, which are of no engineering consequence. A copy of presentations recently made by Pacific Northwest National Laboratory and GE Global Research is included in Attachment 2 and 3, respectively.

It is with this understanding, after several enhancements and extensive verification that the Alloy 690, 52, and 152 materials are highly resistant to PWSCC that the following comments -are made:

1. The NRC requirement to perform both eddy current testing and ultrasonic testing on the wetted surface of the alloy 690 tubing and j-groove weld is too stringent. This requirement being imposed by NRC (and not endorsed by ASME Code) will nearly double the time duration required to conduct the examinations, from 7 days to as much as 14 days, thus

increasing cost and radiation exposure to employees and vendors. Radiation levels are typically on the order of 5 R/hr under the reactor vessel head.

2. The leak path method has -provided accurate supplemental information as confirmation of leakage for when assessment is needed of other indications such as eddy current signals and evidence of boric acid crystals observed during visual examinations. The leak path method is considered to be reliable for reactor vessel heads with interference fits such as those fabricated by Combustion Engineering and Mitsubishi Heavy Industries. The supplemental requirements imposed by NRC to perform both eddy current testing and ultrasonic testing is too conservative and may be misdirected as some crack patterns may not be detectable by ultrasonic techniques and must be confirmed by combination of eddy current testing and leak path assessment.
3. The requirement to perform the first NDE examination for replacement reactor vessel heads with Alloy 690, 52, and 152 materials after 10 years is too stringent. The crack initiation and crack growth data discussed herein and attached to this letter verify that the materials are highly resistant to PWSCC and expected to perform inservice for excess of 53 to 142 years without experiencing PWSCC initiation. Alloy 690 steam generator tubing has been inservice in the PWR industry for over 18 years without incident of PWSCC.
4. The requirement to perform successive NDE examination for replacement reactor vessel heads with Alloy 690, 52, and 152 materials after 7 years is too stringent. The crack initiation and crack growth data discussed herein and presented at the 2007 International PWSCC of Alloy 600 Conference in Atlantic, Georgia on June 11 - 14, 2Q07 verify that the materials are highly resistant to PWSCC and expected to perform inservice for excess of 50 to 140 years without experiencing PWSCC initiation. Alloy 690 steam generator tubing has been inservice in the PWR industry for over 18 years without incident of PWSCC.
5. Utilities performned economic analysis to justify replacement of reactor vessel heads based upon using materials that are highly resistant to PWSCC to reduce or eliminate the need to perform unnecessary NDE underhead examinations on the Alloy 690, 52, 152 tubing and j-groove weld m-aterials. The economic analysis typically includes consideration of radiation exposure to employees and vendors. Adoption of these aggressive NDE testing requirements by NRC will result in unnecessary radiation exposure to nuclear employees and vendors.
6. USA utilities who purchased replacement reactor vessel heads from Mitsubishi Heavy Industries understand that some level of field verification may be needed or desired, by NRC, in the future to confirm the laboratory test results observed by Mitsubishi Heavy Industries, Pacific National Laboratory, and GE Global Research (discussed herein). To this end it may be desirable for USA utilities who purchased replacement reactor vessel heads from Mitsubishi Heavy Industries to propose an integrated replacement underhead NDE inspection program at a frequency of 5 years starting 10 years after installation of the first replacement reactor vessel head as opposed to the continued inservice inspections at predefined durations specified in Code Case 729 Rev 1. Mitsubishi Heavy Industries is in the process of investigating the formation of a USA industry group of Owners that recently purchased replacement reactor vessel heads from Mitsubishi Heavy Industries to formally propose alternative inspection requirements based upon research data discussed herein should the NRC endorse ASME Code Case N729 Rev 1 (along with the cited NRC supplemental requirements).
7. It is recommended that if the NRC endorses requirements of ASME Code C 'ase N729 Revision 1 through adoption into the Code of Federal Regulation it be limited to reactor vessel heads with CRDM tubing and i-groove welds fabricated from Alloy 600, 82, and 182 materials. This

approach will give industry adequate time to formulate and to agree to appropriate NDE requirements for the replaced reactor vessel heads fabricated from Alloy 690, 52, and 152 tubing and. j-groove welds with NRC and ASME. The USA commercial nuclear power industry has replaced all of the reactor vessel heads classified as highly susceptibility to date so adequate time exists to reach an agreement with industry and ASME Code.

USA utilities that recently purchased replacement reactor vessel heads from Mitsubishi Heavy Industries include:

" Dominion Generation Kewaunee Power Station

  • Dominion Generation Surry Unit 2
  • Dominion Generation Millstone Unit 2
  • Southern Nuclear Company, Farley Units 1 and 2

" Progress Energy, HB Robinson

" Omaha Public Power District, Fort Calhoun

" Southern California Edison, SONGS Units 2 and 3

" South Texas Project Units. I and 2

  • Nuclear Management Company, Prairie Island Units I and 2

" Nuclear Management Company, Point Beach Units 1 and 2 Thank you for considering these comments. Questions regarding the nature of this information may be directed to Mr. Charles Tomes of Dominion Energy Kewaunee, Inc at 920-388-8192 and Mr.

Joseph Hutter, Vice Mitsubishi Nuclear Energy Systems, Inc. at 412-374-7395.

Sincerely, Charles A oe Attachments Cc Leslie Hartz, Vice President - Site Vice -President Kewaunee Power Station, Dominion Jerry Bischof, Vice President Nuclear Engineering, Dominion Dennis Koehl, Site Vice President Point Beach Nuclear Plant, Nuclear Management Company Mike Wadely, Site Vice President Prairie Island Nuclear Generating Station, Nuclear Management Company Joseph E Hutter, Vice President Mitsubishi Nuclear Energy Systems, Inc

Attachment 1 PWSCC Lifetime Evaluation on Alloy 690, 52, and 152 for PWR Materials Presented by Mitsubishi Heavy Industries EPRI PWSCC of Alloy 600 2007 International Conference & Exhibition June 11 - 14, 2007 Atlanta, GA

PWSCC Life Time Evaluation on Alloy 690, 52 and 152 for PWVR Materials EPRI PWSCC of Alloy 600 2007 International Conference & Exhibition June 11-14, 2007 Renaissance Waverly Hotel Atlanta,. GA Seii~sdaAkira KonishJ Koji Fujim"oto Mitsubishi Heavy Industries Ltd.

Shinro Hirano, Hajime Ito The Kansal Electric Power Co.,, INC.

2 INTROD'UCTION

>PWSC.Cs in Alloys 600, 82 and 182 for Reactor Vessel Head Penetration material and its weld metals were reported in Bugey-3 and other PWR plants.

In order to evaluate the PWSCC integrity of Alloys 690, 52, and 152 for RV base material and its weld metals, the authors have started uni-axial constant load stress corrosion cracking tests at 360 0C in simulated PWR primary water as a Joint Research Program between the Japanese PWR utilities and Mitsubishi Heavy Industries, Ltd. (MHI)

  • 1. Experience of PWSCC on SG & RV Head Penetration
  • 2. Latest PWSCC Test Results of Alloy TT690 RV Head Penetration Material and Maintenance
  • 3. Latest PWSCC Test Results of Alloy TT690 BMI Nozzles and Maintenance
  • 4. Latest PWSCC Test Results of Alloy 690 Weld Metals
  • 5. Conclusions

Experience of PWSCC on SG ECT indications were Lessons Learned found in a large numbers *Choice of Material of Steam Generator (SG) 6O 60 Tubes and the root cause was PWSCC >-TT > MA

-7 TT690_

Joint Development Programs on SG *-Establishment of an Tube Material Data eprmna

[Japanese PWR utilities and MHIJ eprmetoontalSC for Alloy 600/690

5 Experience of PWSCC on RPV Head Penetration Sept. 1991:m Bugey-3 in France First through wall crack L ssons Learned

-~in Alloy 600 RPV head penetration.

/

I Root cause was PWSCC of Inlet IOutlet pCRIDMV Alloy 600 head penetration Nozzle Nozzle 1*

(Tcold) (Thot)

Joint Development Fuel Programs on RV CR Materials Experimental Method Developed by the Joint Development Programs on SG Tube

Experience of PWSCC on RPV Head Penetration lay 2004 : Ohi-3 in Japan _______

Layout of Head 2700 I CRDM Head Penetration I Penetration 1800 This seems to be a remaining leak print of the Investigation leakage from. the TIC nozzle seal portion during test operation (1991 ).

7 I

Experience of PWSCC on RPV Head, Penetration J-weld h1oy 600) 1800 NDE Result

" Location: RV 2600 to 2800 in the TApprox J-weld 9O00

  • Length : Max. 5mm

( (1) 'Is' Grinding (0.5mm) : Surface (2) 3rd Grinding (total 3mm): Inside

- Linear-like cracks a long the dendrite - Branched along the dendrite

- Length becomes longer further inside.

Anorox Apro 14mm 14-Approx 15mm

-q Appr N N 4 Approx 0.7mm

,Approx Approx 0.9mm* 0.6mm t ~APprox 74- Approx 7-

\1" \1

Experience of PWSCC on RPV Head Penetration

'NCounter Measures Repair & Replacement U

U Step 1: Weld repair for #47 J-weld Defect i In order to maintain the (estimated) integrity of RCS pressure boundary and avoid PWSCC propagation, weld repair was Weld repair for the leaked J-performed by us-ing Alloy 690. weld of RV Head Penetration (Alloy 690 weld)

Step 2: RV Head Replacement Change of material of nozzles Alloy 600 --- Alloy 690 (Improvement of resistance for PWSCC)

Change of material of J-welds Alloy 600 --+ Alloy 690 (improvement of resistance for PWSCC)

Experience of PWSCC on RPV Head Penetration

  • It is well-known that Alloy 690 has high resistance against PWSCC and we use Alloy 690 as the counter -measure to PWSCC,
  • We should obtain PWSCC i~nitiation data for Alloy 690 materials to verify high reliability of Alloy 690.
  • The Japanese PWR Utilities .and MHJ are conducting constant load PWSCC' tests for Alloy 690 materials,

10 a

Chemical Compositions of Test Materials (Alloys MA600, TT690, 152 & 52)

Chemical Composition (mass%)

Alloys C Si Mn P S Ni Cr Fe Cu MA600 SG Tube 0.027 0.35 0.30 0.008 0.001 74.50 15.90 8.51 0.02 (Reference)

RVH Pene. 0.020 0.35 0.32 0.010 0.001 60.10 30.10 8.65 0.01 TT690 BMI Nozzle 0.021 0.32 0.28 0.008 0.001 60.15 29.70 9.00 0.03 Chemical Composition (mass%)

C Si Mn P S Ni Cr Cu Ti Nb Al Aly52 (SMAW) 0.030 0.46 3.37 0.007 0.007 55.9 28.93 <0.01 0.12 1.62 0.16 Allo 52 Weld Joint 0.030 0.17 0.24 0.005 <0.001 60.41 28.95 <0.01 0.56 0.01 0.63 (GTAW)

Heat Treatment Condition and Mechanical Properties of Test Materials (Alloys MA600,, TT690)

Mechanical Properties (R.T.)

Heat I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

AlosTreatment Yield Tensile Elnain Grain Alloys Strength Strength Elnain Size MA (MPa) (MPa)(%

MA600 SG Tube 95C36604 MA600 (Reference) 95C36604 RVH Pene. I 075 0C+TT 286 650 50 4.0 TT690 BMI Nozzle' 1075 0C+TT 284 661 .51 5.9

Water Chemistry of Simulated PWR L1 Primary Water (MOC)

Items Test Conditions pH (at 25 0C) 6~'8 Conductivity (j[iSlcm at 251C) 5~30____

1-3130 3 (ppm as B) 400~'600 LIOH (ppm as Li at 250C) 0.2~'2.2

.Dissolved Hydrogen (cc STP/kg H 2 0) 25~35 Dissolved Oxygen (ppb) <5

-CI (ppm) <0.05

-Temperature (OC) 360

Test Loop for Uni-axial Constant Load Stress Corrosion Cracking Test Test Chamber (1) Test Chamber (11) Test Chamber (111)

RCS

Loading Mechanism of Uni-axial Constant Load Stress Corrosion Cracking Test Instrument Test Specimen Air Cylinder Test ChamberInt Inlet

15 Test Specimens for Uni-axial Constant Load SCC Test (1/2)

C~~J 1~

6 93 (1) For SG Tube (2). For Alloys (1/4 Tubular Type) (Plate Type)

16 Test Specimens for Uni-axialI Constant Load SCC ;Test (2/2)

Weld Metal Detail of B 32*

.9-Thickness of Specimens: I1mm (3) For Weld Metal

17 I

Use of Material Data Base on Alloy 600 Estimation of PWSCC based on Material Data Base on Alloy 600 SG tubes 1000 -

U- U- U U U U

MlIA600 SG TubeA

---------- t t 0u - - - -- - - - - - - ----------------

A 400 ----------- 4I4 300 - -- - - - - - - - -

CL 0-J A :Ruptured A-.*  : Not ruptured 3600C (680 0F ) Simulated PWR RCS Water 100 - I-100 1,000 10,000 100,000 During Time (hr)

18 a

Latest Test Results of Alloy TT690 for RPV Head Penetration Comparison of PWSCC Initiation Data on TT 690 RPVH Penetration Material with those on Alloy 600 SG tubes 1000 - . . - . . - .

TT690 RPVH Pene. ~Th1

L  : : M 600 SG Tube cu MA600 SG Tube A:

400 1 co~

w

+j-U0 3001[

,,,Max. 73,000 r 0l A  : Ruptured

Not ruptured 360*C (680*F ) Simulated PWR RCS Water 100 I I 100 1,000 10,000 100,000 During Time (hr)

19 a

Status of RPV Heads inJapan

()Taniguchi,M, Honi, N., "Maintenance Technology Development for Alloy 600 PWSCC Issue", 12th International Conference on Nuclear Engineering, 2004,

20 0

Status of RPV Head Replacement in Japan

--.1 Unit Loop Utility RVH R Year TAKAHAMA 1 3 loops Kansai 1996 MIHAMA 3 3 loops Kansai 1997 TAKAHAMA 2 3 loops Kansai 1997 MIHAMA 2 2 loops Kansai 1999 01H112 4 loops Kan sai 1999 OHI111 4 loops Kansal 2000 GENKAI 1 2 loops Kyushu 2001 GENKAI 2 2 loops Kyushu 2001 IKATA 1 2 loops Shikoku 2001 MIHAMA 1 2.loops Kansai 2001

[continued]

21 Status of RPV Hea-d Replacement in Japan

-. 11 Unit Loop Utility RVH R Year I KATA 2 2 loops Shikoku 2002 OHI 3 4 loops Kansai 2006 TAKAHAMA 4 3 loops Kansai 2007 (Plan)

OHI 4 4 loops Kansai 2007 (Plan)

TSURUGA 2 4* loops JAPC 2007 (Plan)

TAKAHAMA 3 3 loops Kansai 2008 (Plan)

TOMARI 2 2 loops Hokkaido 2009 (Plan)

TOMARI 1 2 loops Hokkaido 2008 (Plan)

SENDAI 1 3 loops Kyushu 2008 (Plan)

SENDAI 2 3.loops Kyushu_ 2008 (Plan)

22 I

Latest Test Results of Alloy T690 for BMI Nozzle Comparison of PWSCC Initiation Data on TT 690 BMI Nozzle Material with those on Alloy 600 SG tubes 1000 -- - - - - - - ----- - - - - -

51

,TT690 -BMIl Nozzle Lv:!MA60 SG Tube cu MA600 SG Tube

------ AAr 4001----------- ------ E),~

a,)

L.

3001----------- -------

a,) Max. 58,OOOHr A :Ruptured

Not ruptured 360 0C (6800F ) Simulated PWR RCS~ Alater 100 100 1,000 10,000 100,000 During Time (hr)

23 Preventive Maintenance for BMI Nozzle Water Jet Peening (WJP) is applied to relieve tensile stress, BMVI nozzle RN BMIV WJP nozzle R/V High pressure jet water

()Koji Okimura et al., "Residual Stress. Improved By Water Jet Peening Using Cavitation For Small-Diameter Pipe Inner Surfaces", 9th international Conference On Nuclear Engineering, 2001-

24 Preventive Maintenance for BMI Nozzle Water Jet Peening (WJP) is also applied to 3-welds of BMI Nozzles, BMI nozzle

, alloy 600 Vessel bottom WJP for J-weld BMI nozzle near the center BMVI nozzle near the circumference

()Taniguchi,M, Honr, N.,"Maintenance Technology Development for Alloy 600 PWSCC Issue", 12th International Conference on Nuclear Engineering, 2004

25 Latest Test Results of Alloy 152 (SMAW)

Comparison of PWSCC Initiation Data on Alloy 152 (SMAW)

Material with those o-n Alloy 600 SG tubes 1000 -----------

Alloy 152 (SMAW ) E

I------ ------------ MA600 SG TubeA MA600 SG Tube Ak

------ -~ I CD 4001------------ I------

I----------:--- .---

9 CD) 3001-----------

a/

a)

A :Ruptured LJ-*~ LZY*  : Not ruptured, 3600C (680 0F ) Simulated PWR RCS Wate r 100 100 1,000 10,000 100,000 During Time (hr)

26 Latest Test Results of Alloy 52- (GTAW)

Comparison of PWSCC Initiation Data on Alloy 52 (GTAW)

Material with those on Alloy 600 SG tubes 1000 Alloy 52 (GTAW)

i :MA600 SG Tube MA600 SG Tube AA El~

0~

U, 400 L

C,,

0) )

-I-I U) 300 Max.,84,OOOH-r U) 0.

a A :Ruptured LJ LY-10  : Not ruptured 360*C (680'F ) Simulated PWR RCS Water 100 100 1,000 10,000 100,000 During Time (hr)

27 Preventive Maintenance for Alloy 600 Welds Water Jet Peening (WJP) is applied to relieve tensile stress.

Manipulator crane I

Feed mechanism Reactor Vessel Image of WJTP device with guide pole for MCP safe-end

()Taniguchi,M, Hori, N.,"Maintenance Technology Development for Alloy 600 PWSCC Issue", 12th International Conference on Nuclear Engineering, 2004,

28

,Preventive Maintenance for Alloy 600 Welds

-1.1-Alloy 690 Cladding, spool piece replacement , etc. are also preventive maintenance methods.

Shelter Before cladding After cladding 6003Almu Weld nietal Ic Safe-end

.AIklt' Ste Iýowx iainkess 316 ) omv Alloy stel siiies, 16 iih~tIl 600allow Cladding withi Inconowl 6903alloy filler iewal RIV

(*) Taniguchi,M, Honi, N.,"Maintenance Technology Development for Alloy 600 PWSCC Issue", 12th International Conference on Nuclear Engineering, 2004,

CONCLUSI29

-The Japanese PWR utilities and MHI have been accumulating material data for Ni-based alloys and maintenance for Alloy 600 material in the plants , since ECT. indications were found in the SG tubes.

  • As the leak in Bugey-3 in 1991 was. a turning point, the maintenance strategies for RV Head have been also established based on the estimation for PWSCC initiation by use of the Alloy 600 database, and the Japanese PWR utilities started RVH replacement where new RV heads have Alloy 690 head penetrations.

CONCLUSIONS (continue)

  • The constant load PWSCC tests for Alloy 690 materials are being performed and it is ascertained that Alloy 690.

materials including weld metals have excellent reliability against PWSCC.

  • Preventive maintenance measures, such as WJP, etc. for Alloy 600 portions are also being performed for the Japanese PWR plants.

Attachment 2 Crack Growth Response in Simulated PWR Water of Alloy 152 Weld Metal Presented by Pacific Northwest National Laboratory ICG-EAC April 2007 Huallien, Taiwan

I,

  • .::*:*',,+2,7, *:.*.-*% *'*;:'*'Z".:.,' * :,+,*

CrckGrwtRspne -in-Simu; at d PWR' Water o0f AIoylS2,WeId etal>.- .,'.-

.. B.-Toloczko S.fM.Brtemmel Pacific Northw~est. Na~tional: .Laboratory Richand, WA

  • .... N --'

ICG-EAC Hualien, Taiwan Aprill-ý, 2.007

Presentation Outline Pacific Northwest National Laboratory Introduction SPNNL Crack Growth Systems

~Test Setup "ioAIloy 152 CGR Results

)ýAlloy 152 Fracture Surface

,ýSumma-ry & Conclusions

Introduction Pacific Northwest National Laboratory

~Stress corrosion crack growth testing of alloy 690 CRDM tube heats and prototypic alloy 152 weldments are underway at PNNL in simulated PWR primary water.

)ýAvailable information on these materials suggest very low crack growth rates requiringý very long tests to achieve measurable crack growth even in the best systems.

ý>Variations in material (CW, rolling orientation, heat treatment) and environmental (temperature,, imp~urities) are being evaluated.

)ýInitial data presented on an as-received alloy 152 mockup weld under simulated PWR primary water conditions.

PNNL Crack Growth Systems Paii otws NainlLaboratory 11 "11

)oOutlet conductivity *5 0.065 pS/cm under BWR test conditions

)oReversing DCPD, automated K nH control, autoclave flow rate of 220 cc/min.

)o-Continuous measurement of load, inlet conductivity, outlet conductivity,, DCPD voltage,, DCPD current,, autoclave water temperature, and other parameters

)ýWater conductivity in conjunction with man ual pH measurement for B/Li determination.

Specimen Pacific Northwest National Laboratory

.notch shouldbe 10mm.

-Alloy 152 weld supplied by bove intersect ion of crossh ir EPRI NDE Center.

);ýWeld material is a mockup made by MHI for Kewaunee reactor. V crosshiýý SSample is composed entirely -

of weld material.

~Crack root oriented to allow SCC testing in the middle of a weld pass with crack oriented roughly along dendrite direction.

Test Setup Pacific Northwest National Laboratory

)ýConditions: 30 M.Pa-Vm, 325 0C, 1000 Pp B, 2 ppm Li, 29 cc/kgP H2.

SPre-cracked -in-situ usinggseq uence to transition from fatigue to SCC.

SCrack length measurement calculated from DCPD data using reference DCPD potential correction

>reference DCPD potential taken from probes on back-face of sample Sreference potential correction algorithm designed for reference probes in this location

Results - CGR Summary Pacific Northwest National Laboratory Expect extremely low or zero SCC CGRs in PWR primary water.

SCC response eviated by approaching constant Kthrough a series of steps with gentle cycling and increasingly longer hold time.

Step Km..ax Load Frequency CGR length of crack ext (MN4'ln) Ratio (MminS) time (hrs) (gim) 1 25 0.3 1 Hz 2 28 0.5 1 Hz 3 30 0.6 1 Hz 4 30 0.7 1 Hz 5 30 0.7 0. 1 Hz 6 30 0.7 0.01 Hz 7 30 0.7 0.001 Hz 5.4x 10-' 58010 8 30 0.7 0.001lHz + 6.8x109' 540 10 9000 S 9 30 0.7 0.001lHz+/-+ I.Ox 10- 800 3.5 86,400 s 10 30 constant K *1Ox10-10 2250 1

Crack Transitioning: Steps 1-7 PaicNotws NainlLaboratory CT013 CGR 0.5TCT MHI Kewaunee 152 mockup, sample NRC 152-C 12.5 32511C, 1000,ppm B, 2.0 ppmn U, 29 cc/kg H2 to60 0

12.3~! --.

jgb cc

-------. A ---------------------------------- 55 12.1 -- ---- ------ ------- ------------------------ 50 Eu E 0 10.5

-- ----- 20 300

400----- -- 500--- 600----8---- ----- 700--

time (hrs

Crack Transitioning: Steps 8-9 Pacific Northwest National Laboratory CTO013 CGR O.STCT MHI- Kewaunee 152 mockup, sample NRC 152-C 11.905 35 5Spun 11.900 30 E

U' E

Eu 25~

C 0

U 20 11.890 11.8851 ', II I I 11is 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900 2000 2100 time (hrs)

Step 10: Constant K NaionlLbrtr CT013 CGR O.5TCT MHI Kewaunee 152 mockup, sample NRC 152-C 11.906 - 325 0 C, 30 MWav'mi, 1000 ppm ,B, 2.0 ppm Li, 29 cc/kg H2 30 JimN 11.905- ----------- -------------------------- ------------------ ---------------------------- 25 E F 1A E

O~ 11 90 C

U 0 11.901--- ----------------------- ------------------ DCPD -indicates -CGR------ +

11.900 MIT I -0 1800 2000 2200 2400 2600 2800 3000 3200 3400 3600 3800 4000 4200 4400' 4600 time (hrs)

Results - CGR Summary Pacific Northwest National Laboratory Step Kmax Load Frequency CGR length of crack ext (MWa"m) Ratio (mini/s) time'(lirs) (gim) 1 25 0.3 1 Hz.

2 28 0.5 1 Hz 3 30 0.6 1 Hz 4 30 0.7 1 Hz 5 30 0.7 0. 1 Hz 6 30 0.7 0.01 Hz 7 30 0.7 0.001 Hz 5.4x10- 580 100 8' 30 0.7 0.001 Hz 6.8x 10-9 540 10

+ 2.5 h 9 30 0.7 0.001 Hz 1.0x1091 800 3.5

+ 24 h 110 30 constant K *.lOxlOlO 2250 1

Observations Pacific Northwest National Laboratory SLow CGRs measured in alloy 152 weld metal decreasing to ^-,7x10-9 and fvlx10-9 mm/s during 0.001 Hz cyclic loading with hold times of 2.5 or 24 h, respectively.

SDCPD measurements suggest consistent crack advance under constant K. but at an extremely low propagation rate where years would be required to obtain sufficient crack extension for a confident assessment.

SCGR under constant K is clearly less than that during the cycle + hold conditions and approaches 'r'10-10 mm/s.

Post-Test Crack Profile Pacific Northwest National Laboratory STotal crack length in this cross-section is about 2.6 mm.

SFatigue or corrosion fatigue growth*

does not show significant tendency to flow dendrite boundaries.

SSCC growth limited to final conditions and -last few [tm.

CT013 Crack Profile 1.0 MM

Post-Test Fracture Surface Pacific Northwest National Laboratory Interdendritic SCC seen at crack front A

0 C

I.,.

0

-'I t~J

SCC CGR from Crack Surface PaicNotws NainlLaboratory Rough Estimates of CGR in ""Local"

~iRegions along Crack Front:

-Assume scc crack growth limited to at regions extending beyond straight line

ýh. drawn across crack front.

0.

Calculate area of extensions and divide by width of section analyzed.

3

.3 ~~~Assu me SCC growth beg ins at onset of first step with hold time (at 800 h).

SCalculated SCC CGR: "'3x10-9 mm/s in these "local" regions; consistent with DCPD-measu~red rates for final steps.

Summary & Conclusions Pacific Northwest NainlLaboratory SAlloy 152 weld metal found to be SCC resistant in simulated primary PWR water at 3250C even when the pre-crack is oriented along dendrite boundaries in a single pass.

SStable CGR measured at iv5x10-8 mm/s during cycling at 0.001 Hz and decreases to ,v10-9 mm/s during SCC transitioning at 0.001 Hz + hold time of 24 h.

SDCPD suggests consistent crack advance under constant K., but at an extremely low propagation rate where years would be required to obtain sufficient crack extension. CGR under constant K is clearly less than that during the cycle + hold conditions in previous steps and approach "~10-10 mm/s.

SFractography indicates a reasonably straight crack front during cyclic loading with interdendritic SCC during final steps.

SAdditional long-term, higher-temperature tests are underway on as-welded and stress-relieved alloy 152 samples in series.

Attachment 3 PWSCC Growth Rates in Alloy 690 and Its Weld Metal, GE Global Research Center Presented by GE Global Research Center EPRI PWSCC of Alloy 600 2007 International Conference & Exhibition June 11 -14, 2007 Atlanta, GA

PWSCC Growth Rates in Alloy 690 and Its Weld Metals Peter Andresen, John Hickling, Al Ahiuwalia and John Wilson GE Global Research The goal of this on-going program is to perform initial evaluations of the environmental crack growth rates on Alloy 690 and Alloys 152 / 52 weld metals. As has been consistently shown for many other SCC-resistant materials, some inherent susceptibility to SCC exists, and the concept of SCC immunity should be replaced with concepts such as adequately low crack growth rates. Thus, while Alloy 690 and 152/52 weld metals have lived up to their good reputation as 5CC-resistant materials, stable, sustained SCC growth - albeit at very low growth rates (2 - 7 x 10-9 minis) - was observed at constant K in simulated primary water at 340 and 360 'C.

When compared with industry standard estimates for the crack growth rates of Alloy 600 and Alloys 182 weld metal, Alloy 690 and its weld metals exhibited rates~z 70 - 4.00X lower, a very sizeable difference. Note that these approximate factors of improvement must be considered preliminary until more specimens, more conditions, more heats, heat affected zones, etc. are evaluated to provide sufficient confidence in the comparisons being made.

The agreement between dc potential drop and the actual crack length determined from post-test fractography was reasonable (in the range of 4 - 40% error), giving confidence in the reliability of the technique to monitor these very low crack growth rates. Other factors, including statistical measures of linearity of behavior, the magnitude of the resistivity correction, etc. provide a strong basis for confidence in the reported crack growth rate observations.

The crack morphology at (or near) constant K was primarily intergranular in many cases for the base metal, and there was further evidence of intergranular secondary cracking.

Some transgranular cracking was also observed, especially in the weld materials, leading to the encouraging conclusion that the grain boundaries, which are usually the weak point in the microstructure from an SCC perspective, possess inherently high resistance to 5CC in Alloy 690 and its weld metals.

The CRDM form of Alloy 690 used in these studies is much more homogeneous that the Alloy 690 plate used in prior studies. The plate material, particularly after the 982 'C (1800 'F) final anneal, exhibited compositional and carbide banding, less uniformity in grain size, and a lower density of carbides in the grain boundary. But all forms of Alloy 690 tested to date have exhibited similar, very low crack growth rates.

Recent observations on 1-dimensional cold rolled Alloy 690 in the S-L orientation revealed growth rates elevated by as much as -50OX compared to prior studies on T-L orientation. The relevance of such deformation and orientation is not clear, but such

observations must be understood and the nature of deformation during fabrication and weld shrinkage must be characterized.

No effect of pHIB/Li water chemistry parameters was observed on Alloy 690, although only very limited data were. obtained. This agrees with a large body of data on Alloy 600 and stainless steel.

While the results of the tests to date are very promising, only a limited range of

  • conditions and microstructures have been evaluated to date. Additional testing, some of which is now in progress, is needed to confirm and better quantify the factor of improvement in PWSCC resistance for Alloy 690 and its weld metals as a function of such key variables as: other heats; different types of cold work and orientation vs. the plane and direction of cracking; the thermno-mechanical and residual strain conditions associated with weld heat affected zones; off-microstructure conditions that might be developed during non-optimal processing; weld dilution effects; variation in H2 fuagacity and test temperature; etc.

SCC ofAlloy 690-PWIRi SCC Grow tht Rattes o' ColId W re lo 9 Peter Anidresen,,ý AlAhluwalia2 &JOhn Hickling3 GE Globabl Research Ce~n-ter 2EPRI -CC Alloy 600 Conference Atlanta Junhe 2007 1

SCC of Alloy 690

~~~ e'rmtin' Approach______________

Crack growth rates conditionsfor alloy 690:

Scold worked by forging at 25 0Cby ?0 - 40% (thickness)

  • cold work simulates weld residual strain in HAZ
  • recent work on 1-dimensional cold rolled (no cross-roll)
  • used resistivitycoupon for dcpd correction O.5T CT specimens in 340 & 360 OC PWR primary water Stesting at 25 - 35 ksi yIn, including "Varying-K"'(GE)

S18 - 20 cc/kg H2 to be near Ni/NiC Sgood water chemistry: -2 volume exchanges per hour, full-flowdemineralization, and active H2 sparging Smeasured potentials of 690 &Pt vs. CulCu 2O/ZrO 2 2

SCC of Alloy 690 1800F Anneal 0

20%' CW Alloy -6) 90 W=04 - c=4 - 690, W5 hA, NXB2A ill,1 ISOOF Anneal 0.2 A. ,

  • 0 2090F7 Anneal1 E

I0 SCC#2 - c249 - 690, 20P/oA, NX8244HKIl12,20F Anneal In 0

I.. 0

-0.6 E

E CF 500 10DO 1tco 200 2500 3000 Test Tune, hours 5m) 1C0) 1500 200) 2500 3000 3500 400 4500 5000 Test Time, hours Well-behavedl, low cra,,ck,growth rate response during earlie r proof-of-concept testilng 3

SCC of Alloy 690 Altloy 6,90 -CRDM Material CRDN housing of Alloy 690 (heat WN415) provided by Duke Power check 0.018 0.31 10.14 0.0007 0.29 0.007 59.67 29.1 0.016 ladle 0.02 10.31 10.1 0.0007 10.28 0.007 597 29.0 001 Reported average yield strength = 37.7 ksi Reported average tensile strength = 89.1 ksi Annealed at -721C for -11 hours 4.-

cr of AlJoy 690 A if A I'loy,69u C KO.M Oy""S, 52ýq 52 CRDfr of Alloy 690 Alloy 52 & 152 weld metal (heat WN415, Duke) (from B&W) 5

SCC of Alloy 690 4/6Col Wo.rkl Allo~y 690 CR0 H' SCC#2a - c280 - 690, 41%0/RA, WN415 CRDM 1~Il f III -0.4 11+ 0.2 10.995-

-0 co 3.x.31009 E jo.99 E CD D co 0 0~ tz4d .- 0.2 aL

"- 01A~ 4.001U 0 E

U 7nW S10.985

+0 0 1. '

-04 =

+

r310.975. 6.2 x10'~ c280 - 0.5TCT of 690 + 410/^RA 340C

-0.6 25 ksilin, 550 B /11.1 U, 18 cc/kg H2 10.975- At 340C, pH = 7.60. At 300C, pH = 6.93 and potential would be -155 mV higher -0.8 Pt potential CT potential 10.965 1- -- 1

) 1000 1500 2000 2500 3000 3500 4000 Test limne, hours GE tests at Constant & Varying K (dK/da) 6

SCC of Alloy 690 41% Cold Work Alloy 6190 CRDH' SCC#7 - c280 - 690, 41%RA, WN415 CRDM 11.26 .

11.24-3x09m /

1- -0.2

~~0

.6~ 0

-040 CC 0

At 340C, pH = 7.60. At 300C, pH = 6.93 1116and potential would be -155 mV higher -- 0.8 Pt potential CT potential 8600 9100 9600 10100 10600 11100 11600 12100 12600 13100 Test Time, hours GE tests at Constant & Varying K.(dK/da) 7

SCC of Alloy 690 41% Cold W1.ork Alloy 690 CR08M SCC#8 - c280 - 690, 41%RA, WN415 CRDM 0.2 11.34 0

11.32

-0.2 E 11.3 E

-0.

11.28

-0.4 11.26 11.24 -0.8 11.224I- -1 10000 11000 12000 13000 14000 15000 16000 Test Time, hours GE tests at Constant & Varying K (dK/da) 8

SCC of Alloy 690 _ _ _ _ _ _ _ _ _ _ _ _ _ _

41% Co/ldli W1olrk Alfloyr 690 CRDHj, SCC#9 - c280 - 690, 41%RA, WN415 CRDM 11.332 0.2 0

11.327 E 42-E 0 a.

0 S11.322 -0.4 E U)

-0.6 11.317 0

-0.8 11.312 4- -1. -1 15300 15500 15700 .15900 16100 16300 16500 Test Time, hours GE tests ait Constant & Varying K (dK/da) 9

SCC of Alloy 590 41./ -iWb. Alý.,690CD 10

SCC of Alloy 690 .--

-20% ColO-di Work-Alloy 6.90 CRDH I IL . . . . .

0<=~ - CME- NI~CWM69, VRA~ V*J15 QIFJ

.,fA IwIk" .. . . I..

Sccm - CM6 - AIOY 690, 20JAK VMS41 O~FM

. . . 0ý4 C285 - 0.6=c of 690 + 200/.RA, 360C] c286 - 0.5=-C of 690 + 200/cRA 360CI 25ksWin, 550 B/I1.1 U, 18 cckg H2 -0.2 11.04- 0.2 11.07- Qiist condueity x OLO 0

2. 16n~Y E -o k~r 5 B/11L,1 c~gi E Ouetco~k~vtyx .0

~ý80eyd 6-dlvt x 0A S11.0 S2 -0.

O+

0 E 6xa 0 0 0 6 040 11.01 tz, +

p~iat 36GC rit vMI defined .

Al 340C,pH -7.6D. At 300C,pji - ~6.93 -Q6 1Q94 pH at 36OCnot I defined.

lo099 A 340C, IM= 7.60. ft 300C, ri-= 6.93 W-and potendal vmWd be -155 MV and pdtnial v~tid be -156 MV Ugher

'K Pt poterffa CT pctendal Ptpatental Cr pte~a

. 1.-I Iuw1 I 100 30D s 700 9w 1100 130D 15M 1WD 30 5w 700 90D 1100 1300 150)

Test 1Tirm houm Test 11nM hauws EPRI Program - Constant Kmax I1I

SCC of Alloy 690 2-0-.% Cold-'W Work Allfoyi 690 CRDM 0

S=~~-c28-Alloy 690, 20%R^ VM*J45 FW S=~2- c286- Alloy690 W3 cA,VYt4415am 0

-02 0

0 900 140 i9M 2400 20 3400 390W40 900 140W 1900 2400 20 3400W0 4400 Test Tirrp, hours Test Tinp, hours EPRI Program - Constant Km,,

12

SCC of Alloy 1,390 .

20%Cot W or Aloy690 CR0 c285 c286 13

SCC of Alloy 690 .c285/c286 SEH Fractography 14

SCC of Alloy 690 Tresting on 10 Coldr'Rvoiled 6,90 SEvaluation of two O.5T CT specimens of Alloy 690:

" cold worked alloy 690 by i D rolling by 20 -26%

  • use worst S-L orientation: crack plane = rolling plane
  • tested near peak in CGR (near Ni/NiO transition)
  • tested at,360C to accelerate -testing
  • used periodic "gentle" cyclic loading to activate SCC SObserved increased growth rates at constant K 15

SCC of Alloy 690 10t 26% Cold W,oirked Alloy 690 13- 0.4 c372 - 0.5TCT of 690 + 26%RA 1ID, 360C 4 x 10' 12.9- - 25 ksi~in, 600 B /1 Li, 26 cclkg H2 mii -0.2 128 - - Outlet conductivity x 0.01 40 E 12.7 3 x10-1 E 0

-0.7 mm of growth mii --0.2 0, 0

at constant K S12.6- E

-Cd C) -~ 0

-0.4 =L 0

0 0 --0.6 'a 12.4 -

0 0 0

Est. pH at 360C =8.2 used fciroc -0.8 12.3 -

At 340C, pH =7.53. At 300C, pH-I= 6.86 Pt potential CT potential 12.2 -1 0 750 850 950 1050 1150 1250 1350 1450 Test Time, hours

.Increasedgrowth rates in S-L orientation 16

SCC of Alloy 690 iD~t 201% Cold Wjo rked Alloyl 6-90 11.69- I I -. I I I I I I *0.4 c373 - 0.5TCT of 690 + 20%RA 1ID, 360C 11.68-25 ksi~in, 600 B / 1 Li, 26 cclkg H2 - 0.2 11.67- -j Outlet conductivity x 0.01 11.664 4.2 x 10-' -0

.5 MminS E .-S 0

E 11.65* t -0.09 mm of growth -4 0-'

0 at constant K c 11.64- E 0

~~0 0 -0.4 ca 11.63- W CU 00 11.62-

- -0.6 v 00 0 11.61 Est. pH at 360C =8.2 used for 'C - -0.8 11.6- At 340C, pH = 7.53. At 300C, pH = 6.86 Pt potential CT potential 11.59+ IR!N= - "I'll" . . . . I . . . . . . . . . . . . . . . . . . .. . . I -1 65 0 750 850 950 1050 1150 1250 1350 1450 Test Time, hours Increased growth rates in.S-L orientation 17

SCC of Alloy 690 Comparison of GE & -S.etftisDatar 24% Cold Rolled Lots 10 10-6 mm/s Bettis Labs A,

(~3 I (I)

A' 2 New GE Data on E A k ~I DDCW, S-L Orient.

(9 0.1 *.5-7 S-T 0

U) o 5-7 S-L

~J)

A,5-9 S-T (TS I-U) 0.01 5-*-

AAPrior GE Data - forged HTAI (19253~F) + TT (Lot 5-7): Blue HTA 200) TT(o5-)Re 0.001 0 10 20 30 40 50 Average Applied K (ksi'Vin).

Increased growth rates in S-L orientation

-18

SCC of Alloy 690 ID1 26)% Goldj Worked Alloy 619_,0 SCC#7 - c372 - Alloy 690, 26%RA 11), NX3297HK12, ANL 14.4 0.4 c372 -0.5TCT of 690 + 26%RA 1ID, 360C 25 ksi~in, 600 B 11 LU, 26 cclkg H2 1.3 x 10O7 0.2

___________________________________________mm/s Outlet conductivity x 0.01 14.35 0

E 0 E -U4 '

0 0 N 0 14.3 410 -0.24 J9 0

M Afe -. rno got

'U 4by cyliladn 0 s.P t 6C=82ue o

1. 0

-0.48 14.25 MM/ 30,PH at 7.53. At A~Et. use fo 8.2C 6.8 Pt potential CT potential 14.2 -1 1550 1650 1750 18501 950 2050 2150O Test Time, hours Somnewhat lower CGR after fatigue crack advance 19

SCC of Alloy 690_____________

10,20,,% Cold Worked Alloyl 690)

SCC#7 -c373 - Alloy 690, 20%RA 1ID, Heat B25K 12.16 0.4 0.2 12.15 0

12.14 E t4-E CU) 4J'. 0

-0.2 a, r-oC 0 S12.13 E J) U)

-0.4 ~

12.12 After -0.4 mm of growth -0.6 zo C

by cyclic loading 0 12.11 mm/s Est. pH it 360C = 8.2 used for 4c -0.8 At 340C, pH = 7.53. At 300C, pH = 6.86 Pt potential CT potential 12.1 1550 1650 1750 1850 1950 2050 2150 Test Time, hours Somnewhat lower CGR after fatigue crack advance 20

SCC ofAlloy 690 Summary of ErPRI Alloyl 690 Results.

SCrack growth is broadly consistent with other Alloy 690 specimens - somne,. but slow, 5CC growth.

SMuch higher growth rates in 1-dimensional cold rolled material with crack plane = rolling plane (S-L orientation)

SDifficulty insstaining growth at longer hold times.

SSEN exam showed strong evidence of IGcracking.

Summay Typical heats &microstructures of Alloy 690 are shown to be susceptible to IG5CC growth in primary water, although growth rates are very low.

Vulnerabilities must be probed and understood, including weld heat affected zones, off-rnicrostructures & cold work.

21

SCC of Alloy 690 AIlloy, 152)&52 Weld Metal SCC#2 - 0300 - Alloy 152 As-welded - heat WCIOE7 SCC#3 - 0301 - Alloy 52 As-welded - heat NX2579J.K 1 . . . U.4 11.165 -

0301 - 0.5TCT of 52 As-welded, 360C

0. Outlet conductivity x0.01 25 ks-i,50BII i 8cgH .0.2 11.155-00s f12 swldd30 -0.4T 0 .0 E

E oE 11.145 -

145 A+ 10E 0I

-0. E N

11.135 -

U -.4J4~

  • 0 00Cntwlldfnd pH atl

-40C10= At pH 7.60. At3,pH080 11.125 - an+oeta ol e-5 Vhge .-0.6 135 .- -1U C 0 0

0 0 11.115 - *-0.8

(

11. Pt potential CT potential 11.105 1.

800 1800 2800 3800 4800 5800 800 1800 2800 3800 4500 5800 Test Time, hours Test Time, hours c300 (alloy 1,52) c301 (alloy 52)

EPRI Program - Constant Km,, + Cycling 22

SCC of Alloy 690 Alloy 152 & 52 Weld Metal WXC - &=00 -ia 152YIAem ed - MMM WIOE7 accW" - eSG - AfOW 52 Asm*Wde beat NXXIGMK 0

-0.2 -

44

-0.

11.12 4ý 440) 6=W 7400 9400 swO 10w) 7400 Tait Tiff". how, Test lknt, hours c300 (alloy 152) c3C 1 (alloy 52)

EPRI Program - Constant Km,, + Cycling 23

5CC of Alloy 690 Alloy 152 &.52 Weld Metal c300 (alloy 152) c301 (alloy 52)

/LA

SCC of Alloy 690 Allo1y 15m2%&52Werld,, Mretal c300 (alloy 152) c301 (alloy 52)

EPRI Program - Constant Kmax + Cycling Plan to shift to 85)400- s hold & then constant K 25

SCC of Alloy 690 Alloby, 152,&52, Wel d Metall SCC#4 - c336 - Alloy 152 As-welded - heat 307380 -EPRIIMHI 11.2-T -'

E E

-0.2 SCC#4 - c337 - Alloy 52 As-welded - heat NXOB06TS - GENE

-11.16-2.56 X 10-TC of12%swldd 30 t 0.2 0

Outlet conductivity x 0.01 11.105-0-.

Pt poeta CTpoenia 1lx 10" mm/ds -0 11.15. 11.11

  • -0.6 1.2 x10' +.

mm/s

-4.S -0.2 11.14 E

.e.11.05 -

11.134-CD E 3300 3500 3700 3900 4100 43D0 4500 4700 x 11.09- :9 ; 09 Teat Time, hours 04U 11.085 - 06 V c337 - O.5TCT of 52 As-welded, 360C 11.08- 25 ksi'in, 600 B / I Ui, 26 cclkg H2 0 C.

Eat. pH at 360C = 8.1 used for

  • 11.075- At 340C, pH = 7.53. At 300C, pH = 6.86 Pt potential CT potential

'8 11.07......................................... _0

-1 3300 3500 3700 3900 4100 4300 4500 4700 Test Time, hours c336 -2nd Alloy 152 C33 7 = 2nd Alloy 52 Growth rates are very low 26

SCC of Alloy 690_

Summary for AIlloyl 152/152- Weld Metal/,'

SEvaluation of Alloy 152 and 52 weld metal indicates similar susceptibilityto that observed inAlloy 690:

  • prototypicalheats and welding processes
  • sustained growth isdifficultat long hold times e no major difference between Alloys 152 and 52 SMust await post-test fractography to confirm response, cracking morphology, and growth rates.

Vulnerabilities must be probed and understood, including.

weld heat affected zones, off-microstructures&cold work.

27

SCC2 of Alloy 690 Conclusions Results obtained to date under accelerated conditions show:

  • slow crack growth at constant 1< appears to occur in some (but not all) 2D CWAIloy 690, &Alloys 152/5? welds
  • increased growth rates at constant 1* in 1-D cold rolled Alloy 690 with crack plane = rolling plane (S-L orientation)
  • rising dK/da loading shows somewhat higher CGRs and may be relevant in certain field situations
  • truly inteqrgrnular crack propagation has been demonstrated for Alloy 690- base materials Future work should examine:
  • possibility of increased PWSCC susceptibility in HAZ
  • PWSCC inalternate cold work orientations
  • effect of "off-microstructures" from material processing 28
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