ML071920158

Comment (23) of Charles A. Tomes on Proposed Rules Pr 50 Regarding Industry Codes and Standards; Amended Requirements

ML071920158
Person / Time
Site:	Millstone, Kewaunee, Point Beach, Prairie Island, Surry, Farley, Robinson, San Onofre, Fort Calhoun
Issue date:	06/27/2007
From:	Tomes C Dominion Energy Kewaunee
To:	NRC/SECY/RAS
SECY RAS
References
72FR16731 00023, PR-50, RIN 3150-AH76
Download: ML071920158 (85)

Text

DOCKETED PR 50 USNRC (72FR16731)

July 5, 2007 (3:30pm)

June 27, 2007 OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF Secretary ATTN: Rulemakings and Adjudications Staff Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

RIN 3150 - AH76, Response to NRC Request for Public Comment to Incorporate ASME Code Case N-729 Revision 1 With Supplemental Requirements into the Code of Federal Regulation, Revision 1

References:

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

4. Letter from Charles A Tomes to Secretary (NRC) dated June 15, 2007 entitled "RIN 3150 -

AH76, Response to NRC Request for Public Comment to Incorporate ASME Code Case N-729 Revision 1 With Supplemental Requirements into the Code of Federal Regulation" This letter is provided in response to the Nuclear Regulatory Commission (NRC) solicitation of public comments through 10 CFR 50, RIN 3150-AH76, Industry Codes and Standards; amended requirements (Federal Register, Volume 72, Number 65, Page 16731, Dated April 5, 2007). The original version of my letter was transmitted to NRC in an attempt to meet the stated comment date of June 19, 2007. Dissemination of recently made available primary water stress corrosion cracking (PWSCC) initiation and growth test data for Alloy 690/52/152 materials hampered my ability to complete review of the NRC proposal to incorporated supplemental requirements into CFR. Since submittal of my original letter to NRC on June 15, 2007 changes have been brought to my attention.

This letter reflects changes that I am aware of and supercedes the original letter transmitted to NRC on June 15, 2007. As such I respectfully request that NRC replace the body of the original letter located in the public document room with this letter. Attachments 1 through 3 of the original letter remain unchanged.

The author of this letter wishes to thank the NRC for an opportunity to provide comments on NRC's plans to incorporate ASME Code Case N-729 Revision 1 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.

--- y l1 of 6 r6 C -o*

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 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 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 I 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 inspections during future plant operation. The following enhancements were included 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 j-groove welds,

2. The base metal grain size from 4 to 7 (with a target of 5) was selected to optimize PWSCC resistance and also ensure ultrasonic examination,

3. Narrow groove j-groove welds were used to reduce the residual stress,

4. During j-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 Marmon clamps on the thermocouple 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 permitted on the Alloy 690 tubing,

10. Penetrant testing was performed at pre-defined increments during welding of the j-groove weld,

11. A "PT White" criteria was used on the final surface of the j-groove weld and alloy 690 tubing,

12. The surface of the j-groove welds were polished smooth to permit eddy current testing and penetrant testing,

13. The distance between the thermal sleeve and top of the funnel was increased to better accommodate inspection probes during future inservice inspections, 2 of 6

14. Preservice Inspection (PSI) included bare metal visual (BMV) inspections of the reactor vessel head, eddy current testing of the j-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 1 - 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 1 and 2) from Mitsubishi Heavy Industries and Kansai Electric Power Company, 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 and Kansai Electric Power Company at the EPRI 2007 International PWSCC of Alloy 600 Conference and Exhibit Show, Atlanta, GA, June 11 - 14, 2007 is included in Attachment 1. 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 hrs, and 85,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, respectively. All testing was performed at 360 C (680 F). Testing to date confirms no crack initiation has occurred in Alloy 690, 52, and 152 materials. Alloy 600 materials tested in the same conditions all 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. For example, when this adjustment is made using the activation energy data for Alloy 600 in Code Case N729 Revision 1, the equivalent time period for base metal is approximately 80 years. 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 mm/s, which are of no engineering consequence. A copy of 3 of 6

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 byNRC 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 verify that the materials are highly resistant to PWSCC and expected to perform inservice for excess of 80 years without experiencing PWSCC initiation, based on activation energy data provided in ASME Code Case N 729 Revision 1. 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, 2007 verify that the materials are highly resistant to PWSCC and expected to perform inservice for excess of 80 years without experiencing PWSCC initiation, based on activation energy data provided in ASME Code Case N-729 Revision 1. Alloy 690 steam generator tubing has been inservice in the PWR industry for over 18 years without incident of PWSCC.

5. Utilities performed 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.materials. 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 4 of 6

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 Revision 1. It may be possible that a USA industry group of Owners that recently purchased replacement reactor vessel heads from Mitsubishi Heavy Industries will be formed to formally propose alternative inspection requirements based upon research data discussed herein (and provided by Mitsubishi Heavy Industries and Kansai Electric Power Company) should the NRC endorse ASME Code Case N729 Revision 1 (along with the cited NRC supplemental requirements).

7. It is recommended that if the NRC endorses requirements of ASME Code Case N729 Revision 1 through adoption into the Code of Federal Regulation it be limited to reactor vessel heads with CRDM tubing and j-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 1 and 2

" Nuclear Management Company, Prairie Island Units 1 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.

Charles A. Tomes 2045 Fawn Lane Green Bay, WI 54304 Attachments 5 of 6

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 6 of 6

Attachment'l 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 PWR Materials I

EPRI PWSCC of Alloy 600 2007 International Conference & Exhibition June 11-14, 2007 Renaissance Waverly Hotel Atlanta, GA SeijiAsada, Akira Konish, Koji Fujimoto Mitsubishi Heavy Industries Ltd.

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

LI I

INTRODUCTION PWSCCs 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 3600C in simulated PWR primary water as a Joint Research Program between the Japanese PWR utilities and Mitsubishi Heavy Industries, Ltd. (MHI)

~CONTENTS,

• 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

I

~ZZ Experience of PWSCC on SG ECT indications were Lessons Learned found in a large, numbers

• Choice of Material of Steam Generator (SG)

Tubes and the root cause >- 690 > 600 was PWSCC >-TT 4.

> MA I

Joint Development Programs on SG TT690

• Establishment of an Tube Material Data experimental

[Japanese PWR utilities and MHI]

method on PWSCC for Alloy 6001690

w Experience of PWSCC on RPV Head Penetration l Sept. 1991 : Bugey-3 in France First through wall crack Le ssons Learned

.-- in Alloy 600 RPV head penetration

~ii~%I Root cause was PWSCC of Inlet Outlet CRDIVI Alloy 600 head penetration Nozzle ii-(Tcold)

I II Joint Development Fuel-/

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

III Experience of PWSCC on RPV Head Penetration L

oay 2004 :Ohi-3 in Japan .

2700 I CRDM Head Penetration I Layout of Head Penetration I 180/

1800 I*,

This seems to be a remaining leakage from the TIC nozzle seal operation (1991).

Experience of PWSCC on RPV Head Penetration 0-weld

\ \ *-i* 'Aqo!600) 270.'\ \ ""

• Location: RV 2600 to 280o in the J-weld m I

Experience of PWSCC on RPV Head Penetration

• Counter Measures : Repair & Replacement Step 1: Weld repair for #47 J-weld Defect 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 using Alloy 690. weld of RV Head Penetration (Alloy 690 weld)

Step 2: RV Head Replacement Chanqe of material of nozzles v

Alloy 600 -" Alloy 690 (Improvement of resistance for PWSCC)

Chanae of material of J-welds Alloy 600 --* Alloy 690 (Improvement of resistance for PWSCC)

L I

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 initiation data for Alloy 690 materials to verify high reliability of Alloy 690.

• The Japanese PWR Utilities and MHI are conducting constant load PWSCC tests for Alloy 690 materials.

1o~

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 Weld Metals_______ Chemical Composition (mass%)

C Si Mn P S Ni Cr Cu Ti Nb Al Weld Joint Alloyl52 (SMAW) 0.030 0.46 3.37 0.007 0.007 55.9 28.93 <0.01 0.12 1.62 0.16 Alloy 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 1(GTAW)II II IIII

~11~

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

Heat Mechanical Properties (R.T.)

Treatment Yield Tensile Grain Alloys Strength Strength Elongation Size MA (MPa) (MPa) (%)

SG Tube MA600 (Rfeene (Reference) 9750C 346 680 42 RVH Pene. 10750C+TT 286 650 50 4.0 TT690 BMI Nozzle 1075*C+TT 284 661 51 5.9

Water Chemistry of Simulated PWR 12 Primary Water (MOC)

Test Items Conditions pH (at 250 C) 6~-8 Conductivity (pS/cm at 25 0C) 5~30 H3 B0 3 (ppm as B) 400600 LiOH (ppm as Li at 25 0C) 0.2~ 2.2 Dissolved Hydrogen (cc STPIkg H20) 25-- 35 Dissolved Oxygen (ppb) <5 Cl1(ppm) <0.05 Temperature (*C) 360

Test Loop for Uni-axiai Constant Load

• . Stress Corrosion Cracking Test Test Chamber (I) Test Chamber (11) Test Chamber (111)

RCS

1<

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

15~

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

Icj o 2-ý66 (1) For SG Tube (2) For Alloys (114 Tubular Type) (Plate Type)

16~

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

Weld Metal

_ .I /

1 ,

1/ý4 Detail of B Weld Metal B I1 Iq I -

1I\2I I * - I 60 60 120 Thickness of Specimens :1mm (3) For Weld Metal

L7iE Use of Material Data Base on Alloy 600.

Estimation of PWSCC based on Material Data Base on Alloy 600 SG tubes 1000 ----------- ----- ------------- -------I MA600 SG Tube 7--------------- ------------- ------- I ---- ...... --------------- ------

A:

CA 400 -------------

C/)

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

---

----------------- --------------- 4 ------------ : ----------------

A  : Ruptured Not ruptured 360"C (680°F

) Simulated PWR RCS Wate 100 100 1,000 10,000 100,000 During Time (hr)

18~

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.  !

.. MA6 0 0 S G T u b e A MA600 SG Tube A A__

CD co U) 400 3001

).

OL Max. 73,0004r

. --I---!-.----.

.A  : Ruptured 0l-4.A-4.  : Not ruptured 360°C (680F ) Simulated PWR RCS Water 100 *- - - I III . . . . .

100 1,000 10,000 100,000 During Time (hr).

19~

Status of RPV Heads in Japan

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

2O~

Status of RPV Head Replacement in Japan 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 OHI 2 4 loops Kansai 1999 OHI 1 4 loops Kansai 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 1 Status of RPV Head Replacement in Japan Unit Loop Utility RVH R Year IKATA 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~

Latest Test. Results of Alloy TT690 for BMI Nozzle Comparison of PWSCC Initiation Data on TT 690 BMI Nozzle Material with those on Alloy 600 SG tubes 1000 TT690 BMI Nozzle El MA600 SG Tube A MA600 SG TubeA - .....-----D-.

-A 4001 ~ -.....-

ci, 3001 Max 58,000Hr 6I.

CL CL A  : Ruptured 0'J. Lýr  : Not ruptured 360°C (680°F) Simulated PWR RC.S Water 100 100 1,000. 10,000 100,000 During Time (hr)

23~

b Preventive Maintenance for BMI Nozzle Water Jet Peening (WJP) is applied to relieve tensile stress.

BMI nozzle RN BMI noz WJP nozzle RN 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

2<

Preventive Maintenance for BMI Nozzle Water Jet Peening (WJP) is also applied to J-welds of BMI Nozzles.

BMI nozzle WJP for J-weld BMI nozzle near the center BMI nozzle near the circumference

(*) Taniguchi,M, Hori, 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 on Alloy 600 SG tubes 1000

---

Alloy 152 (SMAW) E MA600 SG Tube A J MA600 SG Tube Ul) 400 -'. -.. .-i. . .. .; . . . . . . .. . . . .. . .- - -- .--. ---

c/) 300 V ---i-- -

QL Max.8.5000Hr  :

(I, A  : Ruptured U-0. LY*  : Not ruptured 360°C (6800F) 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

---

.--------...... t~il Alloy 52 (G TAW )

---

. .M

- A 600 S G T ube MA600 SG Tube A - "

U) 4001 U) ----

2 3001 ----------

C) Max. 84,OOOHi A  : Ruptured L'  : Not ruptured 360°C (680°F ) Simulated PWR RCS Water 100 .--.1--..--1... -- I 100 1,000 10,000 100,000 During Time (hr)

127 ]

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

Manipulator crane -

Ner Jet Feed mechanism

-Guide pole nozleZI/ mechanism IReactor Vessel Test block °*,Lim Image of WJP device with guide pole for MCP safe-end

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

1281 Preventive Maintenance for Alloy 600 Welds Alloy 690 Cladding, spool piece replacement, etc. are also preventive maintenance methods.

Shelter b~efre a dinqAfter cladding

-owAII.ov ~wdI /I~m AM loy .aedI V Sw~ink-~316 Safe-end Swik' 31 I-I x

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

1291

• • • CONCLUSIN

• 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.

1301 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

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Presentation Outline Pacific Northwest National Laboratory Introduction

> PNNL Crack Growth Systems

>Test Setup

>Alloy 152 CGR Results

>Alloy 152 Fracture Surface Summary & 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 requiring very long tests to achieve measurable crack growth even in the best systems.

>Variations in material (CW, rolling orientation, heat treatment) and environmental (temperature, impurities) are being evaluated.

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

PNNL Crack Growth Systems Porthwest National Laboratory SOutlet conductivity < 0.065 pS/cm under BWR test conditions

>Reversing DCPD, automated K control, autoclave flow rate of 220 J2 cc/min.

>Continuous measurement of load, inlet conductivity, outlet conductivity, DCPD voltage, DCPD current, autoclave water temperature, and Other parameters

>Water conductivity in conjunction with manual pH measurement for B/Li determination.

Specimen Pacific Northwest National Laboratory notch should be 1.0 mm

>Alloy 152 weld supplied by above.,intersection. of crosshairs EPRI NDE Center.

>Weld material is a mockup made by MHI for Kewaunee ,vertical reactor. ~~ crossha ir*

>Sample 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 SetupPacific National Northwest Laboratory

>Conditions: 30 MPav/m, 325 0 C, 1000 ppm B, 2 ppm Li, 29 cc/kg H2.

> Pre-cracked in-situ using sequence to transition from fatigue to SCC.

> Crack length measurement calculated from DCPD data using reference DCPD potential correction

>reference DCPD potential taken from probes on back-face of sample

> reference potential correction algorithm designed for reference probes in this location

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

SCC response evlated by approaching constant K through a series of steps with gentle cycling and increasingly longer hold time.

Step Kmax Load Frequency CGR length of crack ext (MPa4m) Ratio (mrris) 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 5.4x10 8 580 100 8 30 0.7 0.001 Hz + 6.8x10 9 540 10 9000 s 9 30 0.7 0.001 Hz+ 1.0xOx0- 9 800 3.5 86,400 s 10 30 constant K *1.0xl0 1 ° 2250 1

Crack Transitioning: Steps 1-7

• National Laboratory CT013 CGR O.5-CT MHI Kewaunee 152 mockup, sample NRC 152-C 12.5 3250C, 1000 ppm B, 2.0 ppm U, 29 cc/kg H2 60 12.3 I. I

. 55 51 1-4 12.1 -50 ---.... ............ + 5-1 1 .9 ------------------... . . . . . .. . . . . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . . ... . .. . . . . 45 1 1 .7 ------ - 40 E 1 E 1 1 .5 -. . . . . . . . . . . . . . . .-- .. .. ..-. ..-

. - - - -- . . .. . - . - -- L..

.... 35 S1 1 .3 . . . - .... ....................... ................................................................. 3 .

C U1o.11 .1 -- . .--------

. . . . . . . ------ . . . . . . . . . . . . . . ............................................... 5 =0 02 U

0 10.7 .. ............................................................................................... is 1 0 .3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - - - - - - - - - - - - - - - - - -5 10.1 0 0 100 200 300 400 500 600 700 800 time (hrs)

Crack Transitioning: Steps 8-9 Pacific Northwest National Laboratory Cr013 CGR 0.STCT MHI Kewaunee 152 mockup, sample NRC 152-C 325°C, 30 MPa'Vm, 1000 ppm B, 2.0 ppm Li, 29 cc/kg H2 11.905 35 WOo P 6 C, 5 pm 11.900 30 U

.E 4j 11.895 25 U 0 1- U 0

11.890 20 11.850 0 1 1 . -15 700 B0O 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 1900 2000 2100 time (hrs)

Step 10: Constant K Pacific Northwest National Laboratory C17013 CGR 0.5TCT MHI Kewaunee 152 mockup, sample NRC 152-C 11.906 30 11M 11.905 25 E

11.904 20 u E

2  :.a 11.903 U

U 0 U

11.902 10 0

11.901 5 11.900I I . 0 1800 2000 2200 2400 2600 2800 3000 3200 3400 3600 3800 4000 4200 4400 4600 time (hrs)

Results - CGR Summary Pacific National Laboratory Step Kmax Load Frequency CGR length of crack ext (MPalm) Ratio (mrnis) time (hrs) (gtm) 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 8 - 580 100 8 30 0.7 0.001 Hz 6.8x10-9 540 10

+2.5h 9 30 0.7 0.001 Hz 1.OxlO"9 800 3.5

+24h 10 30 constant K *l.OxlO"'O 2250 1

Observations Pacific Northwest National Laboratory SLow CGRs measured in alloy 152 weld metal decreasing to "'7x10- 9 and '1x10-9 mm/s during 0.001 Hz cyclic loading with hold times of 2.5 or 24 h, respectively.

DCPD 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.

> CGR under constant K is clearly less than that during the cycle + hold conditions and approaches ,v,10-O mm/s.

Post-Test Crack Profile Pacific Northwest National Laboratory

> Total crack length in this cross-section is about 2.6 mm.

> Fatigue or corrosion fatigue growth does not show significant tendency to flow dendrite boundaries.

> SCC growth limited to final conditions and last few [tm.

CT013 Crack Profile 1.0 mm

Post-Test Fracture Surface Pacific: North-vvest National I aoratory 4

Interdendritic SCC seen at crack front I m ~ n I

0

'4 0

-'I I I-'

SCC CGR from Crack Surface ,atv,.

Nation I L a ta.a AL Rough Estimates of CGR in "Local" Regions along Crack Front:

Assume SCC crack growth limited to regions extending beyond straight line drawn across crack front.

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

-Assume SCC growth begins at onset of first step with hold time (at 800 h).

> Calculated SCC CGR: ".3x10 9 mm/s in these "local" regions; consistent with DCPD-measured rates for final steps.

V

Summary & Conclusions Pacific Northwest National Laboratory

> Alloy 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.

> Stable CGR measured at 'v5x10 -8 mm/s during cycling at 0.001 Hz and decreases to "10-9 mm/s during SCC transitioning at 0.001 Hz + hold time of 24 h.

> DCPD 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-O mm/s.

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

> Additional 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 Ahluwalia 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 SCC-resistant materials, stable, sustained SCC growth - albeit at very low growth rates (2 - 7 x 10-9 mm/s) - 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 - 400X 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 de 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 SCC 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 OF) 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 -50X compared to prior studies on T-L orientation. The relevance of such deformation and orientation is not cleatr, but such

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

No effect of pH/B/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 thermo-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 fugacity and test temperature; etc.

SCC of Alloy 690 PWR SCC Growth Rates of Cold Worked Alloy 690

&Alloys 52/15:2 Weld Metal PeterAndresen, Al Ahluwalia2 &John Hickling3 GE Global Research Center 2EPRI, 3C/C Alloy 600 Conference Atlanta June 2007 I

SCC of Alloy 690 Testing Approach Crack growth rates conditions for alloy 690:

cold worked by forging at 25 OC by 20 - 40% (thickness) 9 cold work simulates weld residualstrain in HAZ

• recent work on 1-dimensionalcold rolled (no cross-roll)

• used resistivity coupon for dcpd correction 0.5T CT specimens in 340 &360 °C PWR primary water testing at 25 - 35 ksivtn, including "Varying-K" (GE) 18 - 20 cc/kg H2 to be near Ni/NiO good water chemistry: -2 volume exchanges per hour, full-flow demineralization,and active H2 sparging measured potentials of 690 & Pt vs. Cu/Cu2 O/ZrO2 2

SCC of Alloy 690 1800FAnneal 20% CW Alloy 690 SC(#2 -c248 -6S0, 2R NWS244HK111,18SO0F Aimedl

.1 02 0

2000FAnneal SC&=2. c249 - 69O, ORA, tWWBSI4-11,Z OOF Anmre E

0-

-04 =L 0

C

-0.8 E

-O.8 E 500 1000 1!0 20M 25M 30M 35D 4M0 45W Test Tirm, hours 60D 1000 Ism8 200 250 m 3500 408) 450 WOO Test Tam. hours Well-behaved, low crack growth rate response during earlierproof-of-concept testing 3

SCC of Alloy 690 Alloy 690 CRDM Material CRDM housing of Alloy 690 (heatWN415) provided by Duke Power LOcatiori C iin uw C check 0.018 0.31 10.14 0.0007 .29.07 59.67 29.1 _ 0.016 ladle 0.02 0.31 10.1 0.0007 0.28 0.007 59.75 29.04 0.015 Reported average yield strength = 37.7 ksi Reported average tensile strength 89.1 ksi Annealed at -721C for 0-11 hours 4

C OMIF6 CRDMh of Alloy 690 Alloy 52 & 152 weld metal

(,heot WN415, Duke) (from B&W) 5

SCC of Alloy 690 41% Cold Work Alloy 690 CRD04 S=C#2a - c280 - 690, 41/6RA WN415 CRDM 11.005 . . . . . 0.4 11 + -0 niTIS 0.2 10.995-0 E 10.99-E 0 mmni eq -0.2 0 co-10.985- E

-0.4 "

10.98- $

+

6.2 x 10. c280 - 0.5TCT of 690 + 41%RA, 340C -0.6 vo mnms -o.

10.975- 25 kshin, 550 B /1.1 U, 18 cc/kg H 2 0 Al 340C, pH 7.60. At 300C, pH= 6.93 10.97- and potential would be -165 mV higher -0.8 Pt potential CT potential 10.9654- 0 1 3 -- 1 500 1000 1500 2000 2500 3O000 3500 Test Time, hours GE tests at Constant & Varying K (dK/da) 6

SCC of Alloy 690 41% Cold Work Alloy 690 CR0M SCC#7 - c280 - 690, 41%RA, WN415 CRDM 11.28 - . . . . . . . . I . . . . . I . . . . . , . . .. .

. . 0.2 11.26 Outlet conductivity xO.01 mm/s - 0 11.243 x 10" mms

-0.2.

E 00 E 0 0 01 0 0 1. It x 1 -0.4 E c280 - 0.5TCT of 690 + 41%RA, 340C .-. >

11.18- 35 ksis/in, 550 B I 1.1 Li, 18 cclkg H2 0

At 340C, pH = 7.60. At 300C, pH = 6.93 11.16 and potential would be -155 mV higher -0.8 Pt potential CT potential 11.14- -1 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 Work Alloy 690 CRDM SCC#8 - c280 - 690, 41%RA, WN415 CRDM 0.2 11.34 0

11.32

-0.2

• EE: 11.3 0

0 E 11.23 -0.4 E 1 .

11.28 -0.6 "

0 11.24 -0.8 11.22 4 -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% Cold Work Alloy 690 CR0M SCC#9 - c280 - 690, 41%RA, WN415 CRDM 11.332 11.327 E 0 E I-I..

0

= 11.322 E aI 0 Zo C. 3 11.317 11.3124-15300 15500 15700 15900 16100 16300 16500 Test Time, hours GE tests at Constant & Varying K (dK/da) 9

C, SCC of Alloy 690 41% Cold -Work AIoy 690 CRDM jv 10

SCC of Alloy 690 20%1~f cold Wor-k Allo-y.690 CROM 2M5- O.5rC of SM + MPRA 360rm00o 5 sim n560 BI1.1 Li, 1 cc/kgH 2 G2 11.0 256 W 0B 1U, "'

9 180PR;;7[]2 0.

2xielOnys 10 .. 11.020 E0 0 co rfn's t:~ 0)

H0 Q~~4~ A -10.98' Q

-165v 1096 0 t 0*

pHit360Cn~9Id vell&fted. pHat360C notY define.

-. ACH=M At.~d=pH=S A[3oCp=7vso At3D00.pH=6J93 ur~dpateWivuidbe-156nVNW" -as 10.94 andpctertidvoddD456nltJIV01H Pt~xe,1t crpcgeuvsc 1097 100 7W 7 70D 7 100 1300 15M 194 300 SD 7 1100 17W 15M TestTI~inS tEunD Test Terne houis EPRI Program - Constant Km,,

11

SCC of Alloy 690 20% Cold Work Alloy 690 CRD0 S&- c35-MOYIo9,/ ViM41509M Saw -cm-AIoy 6O0, W^"/A V"I15OMi 1117 4 I I.. I i.OA

-02 1115 11.13 E

E .0

£ -02 ?

0)

• 11.11, 10

-0.4 =L 11.09 46B 11.07, 11UD -1 Boo 140D 190D 240D 2WD 3400 390 4400 Test Trh hours Test "fir, houms EPRI Program- Constant Kmax 12

CCof Alloy 690 20% Cold Work Alloy 690 CRDM c285 c286 13

SCC of Alloy 690 c285/c286 SEM Fractography 14

SCC of Alloy 690 Testing on 1D Cold Rolled 690 Evaluation of two O.5T CT specimens of Alloy 690:

" cold worked alloy 690 by 1D rolling by 20 -26%

" use worst S-L orientation: crack plane = rolling plane

" tested near peak in CGR (nearNi/NiO transition)

" tested at 360C to acceleratetesting

" used periodic "gentle"cyclic loading to activate SCC Observed increasedgrowth rates at constant K 15

SCC of Alloy Er 690%

oo iDp, 26% Cold Wotrked'Alloy 69,0 13 0.4 c372 - 0.5TCT of 690 + 26%RA ID, 360C 4 10-12.9 25 ksi/in, 600 B 1 1 Li, 26 cclkg H2 mm/s /S 0.2 I

12.81 Outlet conductivity x 0.01 0..:

oU 7

4-.

E 12.7 4 3 x10-x~ 0 E 0%

-0.7 mm of growth mnOs -0.2 a, 0

• at constant K c 12.6' i-, E

~Cd -0.4 =-

'U 12.5

-0.6 M 12.4 o 0 0

Est pH at 360C = 8.2 used fc

-0.8 12.3 At 340C, pH = 7.53. At 300C, pF = 6.86 Pt potential CT potential 12.2 -1 650 750 850 950 1050 1150 1250 1350 1450 Test Time, hours Increasedgrowth rates in S-L orientation 16

SCC of Alloy 690 1D, 20% Cold Worked Alloy 690 11.69- 0.4 c373 - 0.5TCT of .690 + 20%RA I D, 360C 11.68-

- 25 ksi/in, 600 B / I Li, 26 cc/kg H2 0.2 11.67-11.661-Outlet.conductivity x 0.01 4.2 x 10- r r-0 L

mm/S 0 E

E 11.65-

-0.09 mm of growth. -0.2 1 at constant K

• 11.64- E

'"0

-0.4 u-

' 11.63-0 11.62- -0.6 "

o 0 0

11.61 Est. pH at 360C = 8.2 used for IC -0.8 11.6- At 340C, pH = 7.53. At 300C, pH = 6.86 Pt potential CT potential 11.59 . I I. - . . .  ; .. . . . . . . . . . . . ... . .- , . - e , -1 650 750 850 950 1050 1150 1250 1350 1450 Test Time,, hours Increasedgrowth rates in S-L orientation 17

SCC of Alloy 690 Comparison of GE & Bettis Data 24% Cold Rolled Lots 10 10-6 mm/s Bettis Labs

. *A3 1

A 2 New GE Data on S  % _I D CW, S-L Orient.

(.9 0.1

• 5-7 S-T

• 5-7 S-L

~I),

• 5-9 S-T 0.01 A5-9 S-L A A Prior GE Data - forged HTA1 (1925,F) + TT (Lot 5-7): Blue HTA2 (2000°F) +'1TT (Lot 5-9): Red 0.001 0 10 20 30 40 50 Average Applied K (ksi\mn)

Increasedgrowth rates in S-L orientation 18

SCC of Alloy 690 10, 26% Cold Worked Alloy 690 SCC#7 - c372 - Alloy 690, 26%RA ID, NX3297HK12, ANL I

14.4 0.4 c372 - 0.5TCT of 690 + 26%RA ID, 360C 25 ksi/in, 600 B I 1 Li, 26 cclkg H2 1.3 x 10 mm/s 0.2 Outlet conductivity x 0.01 0 14.35

+,

E 0 E

-0.2 a 0

14.3' E I.0 -0.4 {n "Sb yli odn After -1.2 mmn of growl th -0.6 "

14.25+ by cyclic loading U0 1.3 x 10'-Est. pH at 360C = 8.2 used for ýc

-0.8 MIS0 At 340C, pH = 7.63. At 300C, pH = 6.86 Pt potential CT potenti.al 14.2 -1 1550 1650 1750 -1850 1950 2050 2150 Test Time, hours Somewhat lower CGR after fatigue crack advance 19

SCC of Alloy 690 iD, 20% Cold Worked Alloy 690 SCC#7 - c373 - Alloy 690, 20%RA ID, Heat B25K 12.16 0.4 c373 - 0.5TCT of 690 + 20%RA I D,' 360Co 25 ksi*in, 600 B 1 1 Li, 26 cclkg H2 -0.2 12.15 Outlet conductivity x 0.01 2.1 x 10'

______.____________,.,,____,..______-0.2 mm/s 0 .-

+ j 12.14 E

• 00 o E M Co 0 4j A-0.2 &.

0)

(U c 12.13 0 10-000

-o -

-0.4 V 12.12 " msEt Ha 60 . sdfr -0.4 After -0.4 mm of growth -0.

by cyclic loading 12.11 MMSEst. pH at 360C =8.2 used for +c 0.

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 Somewhat lower CGR ofter fatigue crock advonce 20

SCC of Alloy 690 Summary of EPRI Alloy 690 Results.

> Crack growth is broadly consistent with otherAlloy 690 specimens - some, but slow, SCC growth.

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

Difficulty in sustaininggrowth at longer hold times.

SEN exam showed strong evidence of IGcracking.

Summary. Typical heats & microstructuresof Alloy 690 are shown to be susceptible to IG SCC growth in primary water, although growth rates are very low.

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

21

SCC of Alloy 690 Alloy 152 & 52 Weld Metal SCC#2 0300 - Alloy 152 As-welded - heat WC1OET

~~0.4

• SCC#3i- c301 - Alloy 52 As-welded - heat NX2579JK

.* . . 0.4 onductivity x0.01 ON0 21 - D.STCT Win of 152 B I5As-welded, IS 360C Sksn H2 1113-01 -s0.5TCT lut.et IuI . 0.2 -5 Outlelt conductivity x 0.01 -, 25 ksBin, 550ofB52 tn 55 II IAs-welded, 360CI clkgkg LI, 18 cclkg H2 0.2

+ 1 11.155 - #4 m*

C; ID~ 6 10m , E 0- 11.135

-0 Minis -00.2 t. T

• MMI - .2

-0at.6CCnotreltdefind. pH at S6OCnot well defined. S

.3x1*At 3400, pH = 7.60. At 300C, pH = 6.93 J*At $40C, pH1 = 7.60. At DO0C, pH = 6.93 ree nd potential would be -165 mV higher -. 1.15"and potential would be -165 mV higher --.

Pt potential CT potential IPt potential CT potential

- 11.135 . .: -4 100 1800 2800 3800 4800 5800 800 1800 2800 3800 4800 6800 Test Time, hours dime, Test hours c300 (alloy 152) c301 (alloy52)

EPRI Program- Constant Kmax + Cycling 22

SCC of Alloy 690 Alloy 1.52 & 52 Weld Metal SCC#4 - c3090 - Alloy 152 As-welded - heat WCIOE7 SCC#4 - c301 - Alloy 52 As-welded - heat NX2575JK 11.23 c301 - O.5TCT of 52 As-welded, 360C 11.22- Outetcodutiit x0.1 25 ksi~in, 550 B 1 1.1 Li, IS cclkg H2 "

- 0.2 I

11.21 - ~

Oullel ~ ~ 5.xodctvt 1001i

-0 -:

(U 11.2 E

E 0 E 0

a. E 11.19. -- 0.2 *"

0* 00 E c 11.16. -- 0.4 m.

"A 11.17.

11.16-08 0 --046 0

• 0 -448~

0 U

11.15 -

Est. pH at 360C = 8.2 used for *4 - -0.8 At 340C, pH = 7.60. At 300C, pH - 6.93 11.14 -

Pt potential CT potential 11.13 . .  :;: ' 4 -1 4400 5400 6400 7400 8400 9400 1040D 4400 540D 6400 7400 8400 940D 10400 Test Time, hours Test Time, hours c300 (alloy 152) c301 (alloy 52)

EPRI Program- Constant Kmax + Cycling 23

(t, of~l 56 c300 (alloy 152) a0130 (alloy 5?)

SCC of Alloy 690 Alloy 152 & 52 Weld Metal 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 Alloy 152 & 52 Weld Metal SCC#4 - c336 - Alloy 152 As-welded - heat 307380 - EPRUIMh 11.2 .I ., I' . . I . . . . * . . ..

I ... - . ,", I .

+f02 Outlet conductivity x 0.01 1x10'mmns 11.19 11.18 t1117 SCC#4 - c337 - Alloy S2 As-welded - heat NXoB05TS - GENE E

--0.4 11.115 T 'z

"*o 11.15, C a.)J 11.11 Outlet conductivity x 0.01 0

.-0.8 .

11.15. c336 - 0.5TCT of 162 As-welded, 360C -0 11.105 26 ksiin, 600 B I I Li, 26 cclkg H, .-0.6 U Est. pH at 360CC- 8.1 used for+, 11.1 11.14. --0.2 E

E At 340C, pH = 7.53. At 300C, pH -6.86 Pt potential CT potential .: 11.095 Eat p" at 3C=61 u*~msedfr4 I, 1,27 - .TTof2A-ele,3 k

3300 3500 3700 3900 4100 4300 4500 4700 S11.09 r At30, ;5.A 30,sH=68 *-0.64~

Test Time, hours 11.085

1. IC - At37 , p1,TC7a52As-welded,pH 8.6 0 11.08

-0.8 11.075 Pt potential CT potential 11.07 33( O0 3500 3700 3900 4100 4300 4500 4700 Test Time, hours c336 - 2nd Alloy 152 c337 = 2nd Alloy 52 Growth rates are very low 26

SCC of Alloy 690 Summary for Alloy 152 52 Weld Metal Evaluation of Alloy 152 and 52 weld metal indicates similar susceptibility to that observed in Alloy 690:

• prototypical heats and welding processes

" sustainedgrowth is difficult at long hold times

" no major difference between Alloys 152 and 52 Must await post-test fractographyto confirm response, cracking morphology, and growth rates.

Vulnerabilitiesmust be probed and understood,including weld heat affected zones, off-microstructures& cold work.

27

SCC of Alloy 690 Conclusions Results obtained to date under acceleratedconditions show:

• slow crack growth at constant K appears to occur in some (but not all) 2D CW Alloy 690, &Alloys 152/52 welds

• increasedgrowth rates at constant K in 1-D cold rolled Alloy 690 with crock plane = rolling plane (S-L orientation)

• rising dK/da loading shows somewhat higher CGRs and may be relevant in certain field situations

• truly interqranulorcrack propagationhas been demonstrated for Alloy 690 base materials Future work should examine:

• possibility of increasedPWSCC susceptibility in HAZ

" PWSCC in alternate cold work orientations

• effect of "off-microstructures"from materialprocessing 28