ML12171A517

Pre-Filed Hearing Exhibit NYS000386, H. Cothron, EPRI Report 1019040 (Redacted) Steam Generator Management Program Steam Generator Divider Plate Cracking Engineering Study, Non-Proprietary Version, (Dec. 31, 2009)

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NYS000386 Submitted: June 19, 2012 t='~~~~ HECTR!C POWER

~~-~~ R£S£.ARCH !NSHTUTE Steam Generator Management Program Steam Generator Divider Plate Cracking Engineering Study Non-Proprietary Version 1019040

Steam Generator Management Program Steam Generator Divider Plate Cracking Engineering Study 1019040 Technical Update, Decem~er 2009 EPRI Project Manager H. Cothron ELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California 94304-1338

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THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI).

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ORGANIZATION(S) THAT PREPARED THIS DOCUMENT Westinghouse Electric Company, LLC This is an EPRI Technical Update report A Technical Update report is intended as an informal report of continuing research, a meeting, or a topical study. It is not a final EPA I technical report.

NOTE For further information about EPRI, call the EPRI Customer Assistance Center at 800.313.3774' or e-mail askepri@epri.com. /

• Electric Power Research Institute, EPRI, and TOGETHER*SHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc.

Copyright© 2009 Electric Power Research Institute, Inc. All rights reserved.

CITATIONS This document was prepared by Westinghouse Electric Company, LLC P.O. Box 158 Madison, PA 15663 Principal Investigator C. Cassino This document describes research sponsored.by the Electric Power Research Institute (EPRI).

This publication is a corporate document that should be cited in the literature in the following manner:

Steam Generator Management Program: Steam Generator Divider Plate Cracking Engineering Study. EPRI, Palo Alto, CA: 2009, 1019040.

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PRODUCT DESCRIPTION Cracking in steam generator (SG) divider plate to stub runner welds has been reported by Electricite de France (EdF) plants. EPRI Report 1014982 describes results of a conservative Phase I analysis of a crack in the divider plate to stub runner weld of a domestic Westinghouse-designed SG. The results of the Phase I analysis show that such a crack could grow in the heat-affected zone and the divider plate welds between the divider plate and the stub runner during normal operating conditions. Therefore, Phase II work began in 2008 to better understand crack initiation and growth in a divider plate to stub runner weld and its impact on plant accident analyses, ASME stress report fatigue limits, alternate repair criteria, and installed plugs and sleeves. EPRI technical update 1016552, published in 2008, provides a detailed analysis of the effect of a cracked divider plate on the behavior of a steam generator during normal and accident operating conditions. This technical update addresses the concern that a cracked divider plate may result in exceeding a design criterion in the SG design documentation for Model F, Model 44F, Model D5, Model51F, and Model 51 SG. This update also addresses the potential of a degraded divider plate to affect alternate repair criteria and installed plugs and sleeves.

Results and Findings All of the structures and related components affected by the divider plate were reviewed for the potential effect of a degraded divider plate condition during operations. In this context, a degraded divider plate is considered to be a divider plate with a flaw or crack that extends the full length of the divider plate to tubesheet connection (100% of the length) and fully through the width of the weld (100% through-wall).

• The design and technical justification of the tubesheet, channelhead, divider plate, and lower shell are not affected by the presence of a degraded divider plate. *

• The.design and technical justification of the significanttubesheet junctions and welds are not affected by a degraded divider plate. *

• The repair devices (plugs and sleeves) and alternate repair criteria for the tube portion within the tubesheet are sensitive to a degraded divider plate assumptions, but no changes to the conclusions from those analyses or the level of conservatism in the results are expected in the case of a degraded divider plate condition.

Challenges and Objectives This report is intended for steam generator analysts and engineers in nuclear power. This report is mainly applicable to nuclear power plants that have Westinghouse-designed steam generators in the existing domestic fleet with Alloy 600 divider plates. This report does not apply to steam generators with center stays or floating divider plates. The purpose of this report is to establish that divider plate cracking is not a safety significant issue for domestic plants. Specifically, the purpose of the analysis is to determine the impact of SG divider plate cracking on:

• ASME Stress Report fatigue limits

• Alternate Repair Criteria

• Installed plugs and sleeves v

Applications, Values, and Use This research was done in response to operating experience atEdF. It is intended, if possible, to provide a technical justification for not inspecting SG divider plates in the domestic fleet. This determination would save utilities on inspections costs and radiation exposure.

EPRI Perspective This study focuses on Westinghouse-designed steam generators with Alloy 600 divider plates in the domestic fleet. This is the only study on the impact of a degraded divider plate on steam generator operations in the United States. '

Approach The project team reviewed more than 600 reports, calculation notes, drawings, and change notices for this study with a focus on information that relates most closely to concerns about maintaining the safety margin in the lower SG complex in the event that a degraded divider plate condition occurs during operations.

Keywords Divider plate cracking Lower steam generator complex Channel head Vl

CONTENTS 11NTRODUCTION .................................................................................................................... 1-1 2 ASME STRESS REPORT UPDATE

SUMMARY

................................................................... 2-1 2.1 Tubesheet, Channelhead, Lower Shell and Divider Plate ..................... :......................... 2-1 2.2 Tube-to-Tubesheet weld .: ............................................................................................... 2-2 2.3 Lower SG Thermal Hydraulic Boundary Conditions ........................................................ 2-2 31NSTALLED PLUGS AND SLEEVES .................................................................................... 3-1 3.1 TIG and Laser Welded Sleeves ...................................................................................... 3-1 3.2 Leak Limiting Alloy 800 Tube Sleeves ............................................................................ 3-1 3~3 Mechanical Tube Plugs ................................................................................................... 3-2 4 SG TUBE ALTERNATE REPAIR CRITERIA ......................................................................... 4-1 4.1 C*- Alternate Repair Criteria forCE SGs ....................................................................... 4-1 4.2 F*/L *-Alternate Repair Criteria for Westinghouse SGs ................................................. A-1 4.2.1 F*, L* and W*- Hard Roll and WEXTEX Expansions .............................................. .4-1 4.2.2 H*- Hydraulic Expansions ....................................................................................... 4-2 4.2.3 Interim ARC ...... .'.... *.................................................................................................. 4-2 5 CONCLUSIONS ..................................................................................................................... S-1 6'REFERENCES .......................................................................................................................6-1

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1 INTRODUCTION The divider plate is a component in the lower SG complex. It is connected to the channel head and tubesheet, but it is not part of the primary pressure boundary for the reactor coolant system

[ 1]. The divider plate influences the deformation and stress distributions in the lower SG complex. Multiple indications of cracking in the divider plate welds have been reported in foreign steam generators [2]. Two EPRI reports [2] and [4]discuss the initial investigation ofthe impact that such degradation, if found, could have on domestic SGs. The cracks in the divider plate welds in the foreign fleet occur on the welds that connect the stub runner to the divider plate and the stub runner to the tubesheet. Cracks were observed on both the cold leg and hot leg side of the divider plate. The most likely cause of the degradation is a combination of primary waterstress corrosion cracking (PWSCC) and fatigue. However, it is also possible that factors such as loose parts impingement on the primary face of the hot leg material and manufacturing artifacts have contributed to the observed divider plate degradation.

Pictures of the as installed divider plate and the interior hot leg chamber of the lower steam generator complex are shown in Figures 1 and 2. Figure 3 provides a schematic of the divider plate and channel head geometry and the site of observed cracking at EdF.

j Figure 1 Divider Plate Welds in an _91der Model of Steam Generator, With a Two Piece Stub Runner 1-1

Figure 2 Divider Plate Welds in a Recent Replacement Model of Steam Generator, With a Single Piece Stub Runner Figure 3 Sketch of Divider Plate Geometry 1-2

The United States Nuclear Regulatory Commission (US NRC) has also expressed concerns about the foreign divider plate cracking phenomena. Most of the US NRC concerns on record in topically related requests for additional information [5] focus on the application oftube-in-tubesheet alternate repair criteria. However, this project focuses on all the concerns below from both industry and the NRC.

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• Is it inore limiting to have a cracked or an uncracked divider plate in an operating SG?

• Does a degraded divider plate affect the operation of the SG under any design conditions?

• Does a degraded divider plate invalidate the qualifications of installed plugs or sleeves?

• Does a SG with a cracked divider plate violate the ASME Code structural criteria?

• Is it necessary to inspect the divider plate and the connecting welds between the tubesheet, stub runner and divider plate?

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The. purpos~ of this technical update is to address the concern of a cracked divider plate resulting in a violation of an ASME Code design condition or concern in the SG design documentation for Model F, Model44F, Model D5, Model51F, and Model 51 SG. It also addresses alternate repair criteria and installed plugs and sleeves.

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The conclusions in this report do not apply to any SG design or model that directly takes credit for the divider plate in order to meet ASME Code design requirements, which includes some of the recent replacement steam generator designs in the domestic fleets. It is assumed that recent replacements include improved material that would not be susceptible to PWSCC and other degradation mechanisms in the existing domestic fleet. The conclusions in this report also do not apply to Combustion Engineering (CE) design steam generators with a central stay coh.imn in the lower steam generator complex. .

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2 ASME STRESS REPORT UPDATE

SUMMARY

The design and analysis of the major components, welds and repair devices in the lower SG complex must satisfy the requirement of the ASME Code [1]. There have been different lower SG complex designs over the last 40 years. However, the requirements and criteria that apply to the design of those structures have not changed with respect to the ASME Code. The following summarizes the role of the divider plate structure in the lower SG complex with respect to the design and operation of a SG. More than 600 reports, calculation notes, drawings, and change notices were reviewed for this study. The information in References [6] through [48] represents the most significant information that directly relates to the concern of the lower SG complex maintaining its safety margin during operations in the event that a degraded divider plate condition occurs.

~ 2.1 Tubesheet, Channelhead, Lower Shell and Divider Plate

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2.2 Tube-to-Tubesheet weld 2.3 Lower SG Thermal Hydraulic Bou.ndary Conditions 2-2

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INSTALLED PLUGS AND SLEEVES There are several types of Westinghouse* SG tube repair sleeve designs. These sleeve designs include: laser welded sleeves, TIG sleeves and leak limiting Alloy 800 sleeves. A tube sleeve design that is intended for the repair of the tube portion within the tube sheet may be affected by the change in tubesheet displacement due to a degraded divider plate because a change in the tubesheet displacement can effect the contact pressure distribution between the tube, sleeve and tubesheet. Some tube sleeve designs do not take credit for tubesheet displacement wheri con~idering the forces holding the tube repair sleeve in the tube. The summary contained in this report will address only the Westinghouse tube sleeve designs. It is important to note that there are no tubes currently in operation with Westinghouse tube sleeve repairs.

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3.1 TIG and Laser Welded Sleeves 3.2 Leak Limiting Alloy 800 Tube Sleeves 3-1

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SG TUBE ALTERNATE REPAIR CRITERIA Many different alternate repair criteria (ARC) have been employed in SGs over the last 40 years. The discussion below pertains only to the alternate repair criteria that involve the tube portion within the tubesheet. This is because ARC for the tube portion within the tubesheet could be directly affected by tube sheet displacements. Several of the ARC discussed below are no longer used in the domestic fleet (e.g., C*, L *). In those cases, the details of how past ARC may have been affected b'y a degraded divider plate condition are included for the sake of completeness.

4.1 C* - Alternate Repair Criteria for CE SGs

-4.2 F*/L *-Alternate Repair Criteria for Westinghouse SGs 4.2.1 F*, L*and W*- Hard Roll and WEXTEX Expansions

4.2.2 H* - Hydraulic Expansions .

4.2.3 Interim ARC

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5 CONCLUSIONS

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REFERENCES

1. ASME Boiler and Pressure Vessel Code, Section 111, "Rules for Construction of Nuclear Power Plant Components," The American Society of Mechanical Engineers, New York, NY, 1989, No Addenda.

2. MRP-EDF-SGPP02, "DAMPIERRE 112 NPP - Unit 1 - Steam Generator #27 Retired from Loop 3 when the SGs Were Replaced in 1990- Chemical, Metallurgical and Mechanical Characterizations of the Weld Joining the Partition Stub, the Divider Plate and the Channel Head Bowl."

3. Divider Plate Cracking in Steam Generators: Results ofPhase II: Evaluation of the Impact of a Cracked Divider Plate on LOCA and Non-LOCA Analyses. EPRI, Palo Alto, CA:

2008.1016552.

4. CN-TAS-08-11, "Quantitative Evaluation of SG Divider Plate Crack Open Conditions for Non-LOCA Safety Analyses," May 8, 2008.

5. NRC Letter, "Wolf Creek Generating Station- Issuance of Amendment Re: Revision to Technical Specification 5.5.9 on the Steam Generator Program (TAC No.MD8054)," United States Nuclear Regulatory Commission, Washington, D.C, April4, 2008.

6. WNET-153 VOL. 4, "Model D5 Steam Generator Stress Report- Tubesheet and Shell Junction Analysis."

7. CN-NCE-DCPPRSG-5, "Delta 54 Replacement Steam Generator Tube to Tubesheet Weld Analysis Diablo Canyon Power Plant Units 1 and 2," January 6, 2008.

8. AEP-85-791, "Plugging Margin for Tubesheet Tube Indications, August 22, 1985.

9. CN-SGDA-00-37, "Evaluation of the Plugged/Burst R1C6 HL Tube at Indian Point Unit 2,"

April 4, 2000.

10. DDP2-5413, "51 Series Channel Head-Tubesheet-Stub Barrel Transient Analysis," August 1970. '

11. DDP2-5543, "51 Series Channel Head-Tubesheet-Stub Barrel Steady State Analysis,"

January 1970.

12. PGE-04-62, "Final Westinghouse Responses to NRC Request for Additional Information on Alloy 800 Sleeve," May 2004.

13. SG-81-07-029-SECTION 3.3, 51, "Series Steam Generator Divider Plate Analysis," April 1972.

14. SM-94-85, "Prairie Island Units 1 and 2, 51 Series SGs F* and L* Plugging Criteria for Tubeffubesheet Roll Expansions," November 17, 1994.

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15. WCAP-14797-P, Rev. 2, "Generic W* Tube Plugging Criteria for 51 Series Steam Generator Tubesheet Region WEXTEX Expansions," March 2003.

16. WTD~ED(SA)-69-028, "51 Series Tubesheet Equivalent Plate Interaction Stress Equations,"

July 25, 1969.

17. WTD-ED(SA)-70-037, "Influence of the Divider Lane on Tubesheet Stresses," September 1970.

18. WTD-SM-75-081, "Analysis of Tube Implant for ALA (51 Series) Steam Generators,"

August 1975.

19. WNEP-8733-V4, SG-89-03-017, "NYPA Model44F Replacement Steam Generator Stress Report," December 1989. *

20. WNEP-8399, "Model44F Steam Generator Replacement Unit Primary Chamber Divider Plate Analysis," November 1983.

21. P. DeRosa, "51 Series Steam Generator Channel Head-Tubesheet-Stub Barrel Transient Analysis," 1970. '

22. T.C. Watson, "Model D4-2 Steam Generator Stress Report: Tubesheet and Shell Junction Analysis," December 1977.

23. WCAP-16290-P Volume 5, "Model54F Replacement Steam Generator Divider Plate Analysis FirstEnergy Nuclear Operating Company Beaver Valley Power Station Unit 1,"

August 2005 ..

24. WCAP-15919-P, Rev. 2, "Steam Generator Tube Repair for Westinghouse Designed Plants with 7/8 Inch Inconel600 Tubes Using Leak Limiting Alloy 800 Sleeves," January 2006.

25. SM-98-43, Rev. 1, "Effect of Weld Width on the Peak to Average Stress Ratio for Laser Welded Sleeves in 7/8 Inch SG Tubes," June 1998.

26. SM 12, "Verification of Computer Program 2SAXIBC for Obtaining Displacement Boundary Conditions from a Global 2-D Axisymmetric Model for a Local 2-D Axisymmetric Model," November 1998.

27. SM-97-02, Rev. 0, "Farley Model 51 Steam Generator Laser Welded Elevated Tubesheet Sleeeving- Calculation of Tube Length Need to effectively Model Steam Generator Tube Axial Stiffness," January 1997. .

28. CEN-294-P, Revision 3-P, "Prairie Island Steam Generator Tube Repair Using Leak Tight Sleeves," December 1991.

29. STD-7.3.1-8069, "Plugging of Zion 1, SG 'C', Row 5, Column 54 Sleeved Tube," March 1988.

30. PGE-04-62, "Final Westinghouse Responses to NRC request for additional information on Alloy 800 sleeves," May 10, 2004. *

31. NSD-JNE-3066, "Alloy 690 versus Alloy 800 Sleeve Mechanics," April 12, 1994.

32. DCL-04-124, "Withdrawal of License Amendment request 03-15, 'Steam Generator Tube Repair Using Leak Limiting Alloy 800 Sleeves and Revision to Technical Specification Table 5.5.9-2, 'Steam Generator (SG) Tube Inspection."'

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33. AEP-85-791, "Ameridm Electric Power Service Corporation D.C. Cook Units 1 and 2 Plugging Margin for Tubesheet Tube Indications," August 22, 1985.

34. Updates to section 8.0 of the Alloy 800 Tube Sleeve Licensing Report for Application in Westinghouse 7/8 inch Steam Generator Tubes.

35. NSD-JLH-7103, "Structurally Limiting Distance from Weld to Nearest Tube Indication for Farley Unit 1 Laser Welded Sleeving," March 24, 1997.

36. LTR-SGMP-09-109 P-Attachment, "Response to NRC Request for Additional Information on H*; RAI #4; Model F and Model D5 Steam Generators," August 25, 2009.

37. WCAP-17071-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," April2009.

38. WCAP-17072-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model D5)," May 2009.

39. WCAP-17091 ~P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model44F)," June 2009.

40. WCAP-17092-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model51F)," June 2009.

41. SG-84-02-067, "Model44F Replacement Steam Generator Stress Report for Wisconsin ElectriC Power Company, Point Beach Unit 1 Field Modifications Analysis."

42. WNET-150, Volume 3, "Model E2, Steam Generator Stress Report, Primary Chamber Components Interactions Analysis," December 1978.

43. SG-96-09-014, "Specific Application of Laser Welded Sleeves for Farley Units 1& 2Steam Generators," September 1996.

44. WCAP-15919, Rev. 2, "Steam Generator Tube Repair for Westinghouse Designed Plants with 7/8 Inch Inconel600 Tubes Using Leak Limiting Alloy 800 Sleeves," January 2006.

45. WCAP-12244, Rev. 3, "Steam Generator Tube Plug Integrity Summary Report." November 1989.

46. CN-SGDA-04-79, "Verification and Validation of Tube/Tubesheet Contact Pressures and C*

Spreadsheet forCE System 80 Steam Generators," August 2004.

47. CN-CDME-08-4, Rev. 1, "Structural Evaluation of the Minimum Circumferential Ligament Required as Part of the WCNOC IARC," March 2008.

48. R. Flury, "51 Series Steam Generator Tube-to-Tubesheet Junction Analysis," January 1971.

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Electric Power Research Institute, EPRI, and TOGETHER. .. SHAPING THE FUTURE OF ELECTRICITY are registered service marks *of the Electric Power Research Institute, Inc.

1019040 Electric Power Research Institute 3420 Hillview Avenue, Palo Alto, California 94304-1338

• PO Box 10412, Palo Alto, California 94303-0813

• USA 800.313.3774

• 650.855.2121

• askepri@epri.com

• www.epri.com