ND-18-1185, Request for Alternative: Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds (VEGP 3&4-PSI/ISI-ALT-06)

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Request for Alternative: Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds (VEGP 3&4-PSI/ISI-ALT-06)
ML18292A789
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 10/19/2018
From: Whitley B
Southern Nuclear Operating Co
To:
Document Control Desk, Office of New Reactors
Shared Package
ML18292A787 List:
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ND-18-1185, VEGP 3&4-PSI/ISI-ALT-06
Download: ML18292A789 (82)
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{{#Wiki_filter:Brian H. Whitley Southern Nuclear Director, Regulatory Affairs Operating Company, Inc. 3535 Colonnade Parkway Birmingham, AL 35243 Tel 205.992.7079 October 19, 2018 Docket Nos.: 52-025 ND-18-1185 52-026 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3 and 4 Request for Alternative: Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds (VEGP 3&4-PSI/ISI-ALT-06) Ladies and Gentlemen: Pursuant to 10 CFR 50.55a(z)(1), Southern Nuclear Operating Company (SNC) hereby requests NRC authorization to use an alternative to the requirements of Section XI, IWB-2500, of the ASME Boiler and Pressure Vessel (B&PV) Code, 2007 Edition through 2008 Addenda (code of record) for Vogtle Electric Generating Plant (VEGP) Units 3 and 4. The proposed request for alternative is applicable to preservice and inservice inspection (PSI/ISI) of specific valve-to-pipe welds listed in Section XI Table IWB-2500-1 Examination Category B-J and Table IWC-2500-1 Examination Category C-F-1, and ASME Class 3 valve-to-pipe welds treated as Class 2 due to leak-before-break (LBB) criterion. The details of the 10 CFR 50.55a(z)(1) request are contained in enclosures 1 and 2 to this letter. Approval is requested by April 18, 2019, to support preservice inspections. contains information that is considered proprietary; therefore, Enclosure 1 is requested to be withheld from disclosure to the public under 10 CFR 2.390. contains the non-proprietary version of Enclosure 1. provides the SNC affidavit for withholding proprietary information contained in . provides the Westinghouse affidavit for withholding proprietary information contained in Enclosure 1. This letter contains no regulatory commitments. Should you have any questions, please contact Mr. Corey Thomas at (205) 992-5221.

U.S. Nuclear Regulatory Commission ND-18-1185 Page 2 of 4 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 19th day of October 2018. Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY Brian H. Whitley Director, Regulatory Affairs Southern Nuclear Operating Company : Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds (Withheld Information) : Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Proposed Alternative VEGP 3&4-PSlllSl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds : Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Affidavit from Southern Nuclear Operating Company for Withholding Under 10 CFR 2.390 (VEGP 3&4-PSl/ISl-ALT-06)  : Vogtle Electric Generating Plant (VEGP) Units 3 and 4 - Westinghouse Authorization Letter CAW-18-4797, Affidavit, Proprietary Information Notice and Copyright Notice (VEGP 3&4 PSI/ISi-ALT-06)

U.S. Nuclear Regulatory Commission ND-18-1185 Page 3 of 4 cc: Southern Nuclear Operating Company / Georgia Power Company Mr. S. E. Kuczynski (w/o enclosures) Mr. D. G. Bost (w/o enclosures) Mr. M. D. Meier (w/o enclosures) Mr. D. H. Jones (w/o enclosures) Mr. J. B. Klecha Mr. G. Chick Mr. D. L. McKinney (w/o enclosures) Mr. T. W. Yelverton (w/o enclosures) Mr. B. H. Whitley Ms. C. A. Gayheart (w/o enclosure 1) Mr. C. R. Pierce Ms. A. G. Aughtman Mr. D. L. Fulton Mr. M. J. Yox Mr. E. W. Rasmussen Mr. J. Tupik Mr. W. A. Sparkman Ms. A. C. Chamberlain Ms. A. L. Pugh Ms. P. Reister Ms. K. Roberts Document Services RTYPE: VND.LI.L00 File AR.01.02.06 Nuclear Regulatory Commission Mr. W. Jones (w/o enclosures) Ms. J. Dixon-Herrity Mr. C. Patel Ms. J. M. Heisserer Mr. B. Kemker Mr. G. Khouri Ms. S. Temple Mr. F. Brown Mr. C. J. Even Mr. A. Lerch Mr. S. Walker State of Georgia Mr. R. Dunn (w/o enclosure 1)

U.S. Nuclear Regulatory Commission ND-18-1185 Page 4 of 4 Oglethorpe Power Corporation Mr. M. W. Price (w/o enclosure 1) Ms. A. Whaley (w/o enclosure 1) Municipal Electric Authority of Georgia Mr. J. E. Fuller (w/o enclosure 1) Mr. S. M. Jackson (w/o enclosure 1) Dalton Utilities Mr. T. Bundros (w/o enclosure 1) Westinghouse Electric Company, LLC Mr. L. Oriani (w/o enclosures) Mr. C. Churchman (w/o enclosures) Mr. M. Corletti Mr. M. L. Clyde Ms. L. Iller Mr. D. Hawkins Mr. J. Coward Other Mr. S. W. Kline, Bechtel Power Corporation Ms. L. A. Matis, Tetra Tech NUS, Inc. (w/o enclosure 1) Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc. (w/o enclosure 1) Mr. S. Roetger, Georgia Public Service Commission (w/o enclosure 1) Ms. S. W. Kernizan, Georgia Public Service Commission (w/o enclosure 1) Mr. K. C. Greene, Troutman Sanders (w/o enclosure 1) Mr. S. Blanton, Balch Bingham Mr. R. Grumbir, APOG (w/o enclosure 1) NDDocumentinBox@duke-energy.com, Duke Energy (w/o enclosure 1) Mr. S. Franzone, Florida Power & Light (w/o enclosure 1)

Southern Nuclear Operating Company ND-18-1185 Enclosure 2 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds (Enclosure consists of 64 pages, including this cover page.)

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Plant Site-Unit: Vogtle Electric Generating Station (VEGP) - Units 3 and 4 Interval-Interval Applies to Preservice Inspection and the First Inservice Dates: Inspection Interval Approval is requested by April 18, 2019 to support Requested Date performance of the ASME Section XI Preservice and Inservice for Approval: Inspection weld examinations Components affected consist of ASME Section III, Subsection NB Class 1 and ASME Section III, Subsection NC Class 2 Valve-to-Pipe welds (ASME Section XI, Table IWB-2500-1 Examination Category B-J and Table IWC-2500-1 Examination Category C-F-1, respectively) as identified in Table 1 and ASME Code Table 2 of this Alternative Request. Components Affected: Consistent with the AP1000 UFSAR and the ASME Section III Piping Design Specifications, ASME Class 3 welds identified as leak-before-break (LBB) shall be inspected to the same preservice/inservice requirements as ASME Class 2 piping welds. Therefore, welds identified as Class 3 LBB Valve-to-Pipe welds are also included in this Alternative Request. Applicable Code ASME Section XI Code, 2007 Edition through the 2008 Edition and Addenda (code of record). Addenda: In accordance with ASME Section XI, IWA-2200 (c), all nondestructive examinations of the required examination surface or volume shall be conducted to the maximum extent practical. When performing VT-1, surface, radiographic, or Applicable Code ultrasonic examination on a component with defined surface or Requirements: volume, essentially 100% of the required surface or volume shall be examined. Essentially 100% coverage is achieved when the applicable examination coverage is greater than 90%; however, in no case shall the examination be terminated when Page 2 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds greater than 90% coverage is achieved, if additional coverage of the required examination surface or volume is practical. ASME Section XI, Figure IWB-2500-8 and Figure IWC-2500-8 require that for Category B-J and Category C-F-1 piping welds, respectively, the inner 1/3-t (where t is the wall thickness) of the weld be examined for a distance of 1/4-inch into the base metal on each side of the weld. In this Alternative Request, the examination volume consists of the valve and pipe base metal and the weld. 10 CFR 50.55a(g)(3)(i) and (ii) states that components classified as ASME Code Class 1, 2, and 3 must be designed and be provided with access to satisfy the preservice examination requirements set forth in ASME Section III and Section XI. The Vogtle Units 3 and 4 UFSAR (Section 5.2.4.2) states that components and welds requiring inservice inspection (and by default preservice inspection) are designed to allow for the application of the required inservice inspection methods (and preservice inspection methods). This application requires sufficient clearance for personnel and equipment, maximized examination surface distances, two-sided access, favorable materials, weld-joint simplicity, elimination of geometrical interferences, and proper weld surface preparations. Per Section III of the ASME Code, valve design must satisfy the requirements of ASME B16.34 and ASME Section III NB/NC-4000 for minimum wall thickness (tm). The valve-wall thickness must be greater than the pipe-wall thickness to withstand the nozzle loading. These design requirements define the configuration of the outer surface of the valve that impacts the Reason for ultrasonic examination processes. Request: The ASME Section III design requirements have been met for the pipe-to-valve welds included in this alternative. However, the reason for this request is that the Section XI examination requirements for dual-sided access cannot be met due to the valve outer surface geometry. Recent ultrasonic examination technology has been limited in its capability to qualify equipment, procedures, and personnel for single-sided exams of austenitic stainless-steel pipe-to-valve welds as defined in 10 CFR 50.55a(b)(2)(xvi)(B). Several AP1000 Class 1 and 2 austenitic stainless-steel valve-to-pipe welds do not allow for essentially 100% of the required Page 3 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds examination volume coverage as defined in ASME Section XI, IWA-2200 (c). Consistent with the AP1000 UFSAR and the ASME Section III Piping Design Specifications, ASME Class 3 welds identified as LBB shall be examined to the same preservice/inservice requirements as ASME Class 2 piping welds. Therefore, welds identified as Class 3 LBB that also have inspectability limitations are included in this alternative. There are two categories of austenitic valve-to-pipe welds addressed in this request. Category 1 are those valve-to-pipe welds with a cast austenitic valve body welded to a wrought austenitic pipe. The cast austenitic material combined with the outer surface configuration of the valve prevent the successful application of ultrasonic techniques needed to detect and size inner diameter surface initiated planar flaws. ASME Section XI, Mandatory Appendix VIII, VIII-3100 establishes the requirements for qualification tests. Table VIII-3110-1 specifically provides a list of qualification supplements for various components. The entry for cast austenitic refers to note 1 which indicates the supplement is In the course of preparation. Therefore, Appendix VIII, Table VIII-3110-1 does not specifically address the performance demonstration requirements for cast austenitic materials. However, Appendix VIII-3110(c) does mandate that the requirements of Section XI, Appendix III shall be met for piping welds whose requirements are in the course of preparation. Reason for Request Utilizing ultrasonic test techniques applied from the cast (Continued): austenitic valve body and consistent with Section XI, Appendix III are not practical given the restricted outer surface configuration of the valve. In addition to the outer surface geometry challenges, the Category 1 weld limitations in examination coverage are also driven by the cast austenitic stainless steel (CASS) valve bodies. For the current operating fleet, qualified ultrasonic examination coverage per the latest revision of PDI-UT-2 is approximately 50% of the required examination volume. The examination volume not qualified for coverage is that associated with the Page 4 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds austenitic valve material and one-half of the weld material adjacent to the valve. [There weld.] (a,c) Category 2 valve-to-pipe welds are those with a wrought austenitic stainless-steel valve body welded to a wrought austenitic stainless-steel pipe. Per Table VIII-3110-1 of ASME Section XI, Mandatory Appendix VIII, Supplement 2 shall be used to qualify procedures, equipment, and personnel to be used for detecting flaws. Similar to the Category 1 welds, the outer surface configuration of the valves limit the application of ultrasonic techniques from the valve side of the weld consistent with the requirements of Section XI, Appendix VIII, Supplement 2. The qualified ultrasonic examination coverage will be approximately 50% of the required Section XI examination volume. The examination volume not qualified for coverage is that associated with the wrought austenitic valve material and one-half of the weld material adjacent to the valve. Figure 1 also identifies the required ASME Section XI examination volume and the non-credited 50% of ASME Section XI required examination volume on the valve side of the weld for Category 2 welds. The examination volume not qualified for coverage is that associated with the wrought austenitic valve material and one-half of the weld material Reason for adjacent to the valve. Request (Continued): [There 

                                                           ".] (a,c)

In summary, the achievable exam coverage for both Category 1 and 2 welds are limited due to the outer surface configuration of the valve and the required dual-sided ultrasonic examination requirements. In addition to the outer-surface geometry Page 5 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds challenges, the Category 1 weld limitations in exam coverage are also driven by the CASS valve bodies. The proposed alternative is to eliminate the volumetric examination requirement for the austenitic stainless-steel valve material. The current valve-to-pipe examination volume is shown in Figure 2 for Category 1 and Category 2 welds, including Class 3 LBB valve-to-piping welds. Figure 2 also illustrates the AP1000 design features implemented on the Vogtle Units 3 and 4 valve-to-pipe welds intended to improve inspectability and provide better access to the valve-side examination volume. These design features include a flush weld on the outer surface with a 1/32-inch per inch flatness with a smooth transition from the weld to pipe. Additionally, Figure 2 shows a reduced maximum outside diameter (OD) taper of 18° on the valve for a distance of 1.5 times the minimum pipe wall thickness from the valve-to-pipe weld in comparison to the 30° maximum OD taper as specified in ASME Section III, Subsection NB, Figure NB-4250-1 and Subsection NC, Figure NC-4250-1. These design features are included in the Vogtle Units 3 and 4 piping specifications. These AP1000 design features have made it possible to qualify a new ultrasonic examination procedure, specifically for Vogtle Units 3 and 4. This new ultrasonic exam procedure has been developed based on the requirements of the latest revision of PDI-UT-2. In accordance with Appendix VIII-3130(a), any two procedures with the same essential variables are considered equivalent. This new procedure has been demonstrated to extend the qualified ultrasonic examination coverage beyond Proposed the current 50% to include the far side of the weld material, up Alternative: to the fusion line, resulting in a larger qualified valve-to-pipe examination volume, illustrated in Figure 2. All of the flaws contained in the mock-ups were detected in the sample set during the procedure qualification. To support development of the new ultrasonic examination procedure, 25 mock-ups were developed using the valve-to-pipe configurations at Vogtle Units 3 and 4. These mock-ups contained axial and circumferential flaws throughout the weld, including pipe and valve side, and contained flush weld crowns as contained in the Vogtle Units 3 and 4 design specifications. These mockups range from 3nominal OD to 14OD with nominal wall thicknesses ranging from 0.322to 1.406. These dimensions are currently within the qualified range of PDI-UT-2. Page 6 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds The sizes utilized in the mockups bound the Vogtle 3 and 4 configurations. In addition, these mockups have been used in conjunction with existing mock-ups to develop the Vogtle Units 3 and 4 specific procedure. The proposed alternative implements a similar approach for the elimination of the valve examination volume that has been adopted in the ASME Code Section XI for Examination Category B-M-2: Valve Bodies. In Section XI editions prior to the 2008 Addenda, welds in valve bodies were required to undergo volumetric examinations. However, in the 2008 Addenda of Section XI, the volumetric examination requirement was removed and substituted with a VT-3 visual examination of the internal surfaces when a valve is disassembled for maintenance or repair. This change was justified based on the high flaw tolerance of cast and wrought austenitic materials. However, this alternative request is not recommending any additional visual examinations. While the proposed alternative is to eliminate the requirement for a volumetric examination of the austenitic valve material in the valve-to-pipe welds, site specific qualified outer diameter (OD) surface ultrasonic examinations of the valve side weld and fusion line examination volume using ultrasonic test techniques from the pipe side of the weld, on the conditioned weld surface, and where practical, from the valve side of the weld will be performed. These site-specific qualified OD surface applied ultrasonic examinations will include:

1. Longitudinal wave ultrasonic techniques are currently defined in PDI-UT-2 for the detection of circumferentially-oriented flaws from the pipe side of the weld; until recently, these techniques were not qualified in accordance with the requirements in 10 CFR 50.55a(b)(2)(xvi)(B) for austenitic material single side coverage. Vogtles procedure is qualified to detect flaws over the entire weld volume, Proposed including the fusion line. A best effort examination will be Alternative performed on the ASME Section XI volume for the valve.

(Continued): Longitudinal wave probes will be utilized for the detection of circumferential flaws utilizing the Vogtle 3 and 4 specific procedure. These longitudinal wave probes will be contoured to the outer diameter surface in accordance with the procedure requirements and will be used for axial beam scanning from the pipe side of the weld toward the valve for the detection and length sizing of circumferential flaws. Page 7 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Figure 5 and Figure 6 show the expected examination coverage for circumferentially-oriented flaws (for both tapered and flat welds, respectively) utilizing the qualified examination techniques described above for both Category 1 and Category 2 welds.

2. Longitudinal wave ultrasonic techniques are not currently defined in PDI-UT-2 for the detection of axially-oriented flaws from the conditioned weld surface and where practical from the valve surface; until recently, these techniques were not qualified in accordance with ASME Section XI Appendix VIII.

Vogtles procedure is qualified to detect flaws over the entire weld volume, including the fusion line. A best effort examination will be performed on the ASME Section XI volume for the valve. Longitudinal wave probes will be utilized for the detection of axial flaws utilizing the Vogtle 3 and 4 specific procedure. These longitudinal wave probes will also include a skew or beam correction angle to compensate for the valve taper angle and provide for more direct impingement on axial planar flaws. The probes will be contoured to the outer diameter surface in accordance with the site-specific procedure requirements and will be used for circumferential beam scanning from the conditioned weld surface and where practical from the valve surface for the detection and length sizing of axial flaws. Circumferential scans of the weld will be performed primarily from the top of the weld. Some minor Proposed loss of coverage is expected at the transition of the base Alternative metal-to-weld toe, due to the taper transitions. Shear wave (Continued): probes are not practical in this application due to attenuation within the weld material. Figure 3 and Figure 4 show the expected examination coverage for axial oriented flaws (for both tapered and flat welds, respectively) utilizing the qualified examination techniques described above for both Category 1 and Category 2 welds. Individuals qualified to the Vogtle 3 and 4 specific procedure shall be qualified to PDI-UT-2. These examination techniques will result in essentially 100% ASME Section XI coverage for axial and circumferential flaws included in the piping base metal and the weld. The Vogtle 3 and 4 specific procedure requires a prerequisite to have the weld crown removed to allow scanning on top of weld and a flatness of 1/32 over the size of the transducer. The scope of the Vogtle 3&4 procedure includes Page 8 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds the pipe base material and the entire weld up to fusion side of the valve weld and weld figures have been added to show coverage for the pipe base material and entire weld. A table was added to the Vogtle 3 and 4 specific procedure to add specific transducers for circumferential scanning of axial flaws on top of weld. No changes were necessary to the Vogtle 3 and 4 specific procedure for the identification of circumferential flaws from the pipe side of the weld. These outer diameter surface applied ultrasonic test techniques will be applied for the preservice examination in order to obtain a baseline volumetric examination of the ASME Code Section XI defined examination volume. It is noted that volumetric examination of the weld using the radiographic examination method will have already been performed in accordance with ASME Code Section III. Additionally, per the examination requirements of ASME Section III NB-2541 and NB-2571, a liquid penetrant examination of all external and accessible interior portions of the valve bodies and machined surfaces (including the weld prep) has been completed prior to N-stamping the valves. For cast austenitic valves, visual examinations will be performed in accordance with design specifications. Application of other volumetric methods such as ID applied ultrasonic techniques is not practical due to access and the as-welded surface conditions of the weld. The technical justification for this proposed alternative is included in Attachment 1 to this alternative. Attachment 2 provides historical information relative to compliance with Code Case N-481 and NRC acceptance of the code case. Attachment 3 provides a summary of a representative flaw tolerance evaluation for a typical AP1000 CASS valve-to-pipe weld to demonstrate that these locations are flaw tolerant for postulated flaw depths that are up to 40% of the wall thickness. Basis for Use: This proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1) based on research, prior approval and implementation of ASME Section XI Code Case N-481, and the results of the representative flaw tolerance evaluation as detailed in the attachments. Page 9 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Duration of The duration of the proposed Alternative is the Section XI Proposed Preservice Examinations and for the First Inservice Inspection Alternative: Interval. SV3-GW-GEI-100, AP1000 Preservice Inspection Program Plan for Vogtle Unit 3 SV4-SW-GEI-100, AP1000 Preservice Inspection Program Plan for Vogtle Unit 4 ASME Boiler & Pressure Vessel Code, Section III, Division 1, Subsection NB, Class 1 Components and Subsection NC, Class 2 Components, Rules for Construction of Nuclear Power Plant Components, 1998 Edition and 2000 Addenda. ASME Boiler & Pressure Vessel Code, Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition through the 2008 Addenda.

References:

Vogtle Electric Generating Station (VEGP) - Units 3 and 4 UFSAR, Revision 7.1 ASME Code Case N-481: Alternative Examination Requirements for Section XI, Division 1, Examination of Pump Casings. ASME Code Case N-770-1: Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1. PDI-UT-2, Latest Revision: PDI Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds NDE 20180702-002 Vogtle 3-4 Rev. 5: EPRI NDE Letter to Vogtle 3&4 Construction Status: Awaiting NRC Authorization Page 10 of 64

ND-18-1 185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and lnservice Inspection Requ irements for Specific Valve-to-Pipe Welds Figure 1: Valve-to-Pipe Welds - Current Qualified PDl-UT-2 Examination Volume Aust enitic Valve Body Non-credit ed Exam Volume (50%) Wrought Aust enitic Pipe c 

                                                                +

D F 

                                      +

Required ASME Section XI Examination Volume (d) ASME Code Examination Volume, A-8-C-D. (e) Valve-to-Pipe weld with current qualified PDl-UT-2 examination volume shown as blue parallel lined area, E-8-C- F. (f) Representation of coverage limitations; the non-credited ASME Section XI examination volume is shown as red cross-hatched area, A-E-F-D (50% of examination volume cross-section). Page 11of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and lnservice Inspection Requ irements for Specific Valve-to-Pipe Welds Figure 2: Category 1 and 2 Alternative Request Examination Volume and AP1000 lnspectability Design Features Based on New Qualified UT Exam Procedure for Vogtle Units 3 & 4 Blend flush or smooth across entire weld surface Transition to be blended Austenitic Valve smooth Body 

           '~est_ Effort'        'i              F         '  "!-C-. .                --------

Alternative Request Examination Volume * 'fl Examination Volume Required ASME Section XI Examination Volume (c) Valve-to-Pipe weld with Alternative Request examination volume based on new site specified qualified UT procedure, shown as blue parallel lined area, E-8-C-F. (d) lnspectability features include flush weld surface, smooth transition, and a reduced maximum outer surface taper of 18°. Figure 3: Valve-to-Pipe Weld Representing Austenitic Valve-to-Pipe Welds (Tapered) - Alternative Request Qualified Examination Volume for Axially-Oriented Flaws a*oo 0.911 T 316 SS ASME Code Examination Volume Qualified Examination Volume A-B-C-0 E-B-C-F Page 12 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Figure 4: Weld Representing Austenitic Valve-to-Pipe Welds (Flat) - Alternative Request Qualified Examination Volume for Axially-Oriented Flaws ASME Code Examination Volume Qualified Examination Volume A-B-C-D E-B-C-F Figure 5: Weld Representing Austenitic Valve-to-Pipe Welds (Tapered) - Alternative Request Qualified Examination Volume for Circumferentially-Oriented Flaws ASME Code Examination Volume Qualified Examination Volume A-B-C-D E-B-C-F Figure 6: Weld Representing Austenitic Valve-to-Pipe Welds (Flat) - Alternative Request Qualified Examination Volume for Circumferentially-Oriented Flaws ASME Code Examination Volume Qualified Examination Volume A-B-C-D E-B-C-F Page 13 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds Category 1 Cast Austenitic Valve to Wrought Austenitic Pipe Welds Table 1 (a,c) Page 14 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z )(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds Category 1 Cast Austenitic Valve to Wrought Austenitic Pipe Welds Table 1 (a,c) Page 15 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z )(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds Category 1 Cast Austenitic Valve to Wrought Austenitic Pipe Welds Table 1 (a,c) Page 16 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z )(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds Category 1 Cast Austenitic Valve to Wrought Austenitic Pipe Welds Table 1 (a,c) Page 17 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z )(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds Category 1 Cast Austenitic Valve to Wrought Austenitic Pipe Welds Table 1 (a,c) Page 18 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z )(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds Category 1 Cast Austenitic Valve to Wrought Austenitic Pipe Welds Table 1 (a,c) Page 19 of 64

ND-18-1185 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z )(1) Regarding Preservice and lnservice Inspection Requirements for Specific Valve-to-Pipe Welds Category 2 Wrought Austenitic Valve to Wrought Austenitic Pipe Welds Table 2 (a,c) Page 20 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds ATTACHMENT 1: Technical Basis for Proposed Alternative: Pre-service Examination Limitations for Austenitic Class 1 and Class 2 Valve-to-Pipe Welds in AP1000 Plants Introduction For Class 1 and Class 2 valve-to-pipe welds, ASME Code Section XI Figure IWB-2500-8 requires that the inner third of the wall thickness of the weld be volumetrically examined over an extent that includes the weld and 0.25 inch of base metal on each side of the weld. As such, a portion of the valve base material is part of the required examination volume. ASME Code Section XI, IWA-2200 requires that essentially 100% of the total examination volume be examined. Essentially 100% coverage is achieved when the defined examination volume coverage is greater than 90%. For austenitic valve-to-pipe welds the required 100% examination volume coverage cannot be obtained due to a combination of the geometry on the valve end near the weld and the austenitic base and weld materials associated with the weld joint. Current ASME Code Section XI, Appendix VIII qualifications for austenitic piping welds require access to both sides of the weld, which cannot be met. These examination limitations are generally consistent with those which exist in the current operating fleet, and are primarily associated with Class 1 and Class 2 austenitic valve (and pump) to pipe welds. The purpose of this work is to provide the technical basis for a proposed alternative that includes an alternative of the examination volume defined in ASME Code Section XI, Figure IWB-2500-8 for austenitic valve-to-pipe welds. The Proposed Alternative for the Pre-Service Examination is summarized as follows: Eliminate the austenitic valve base material from the required examination volume. Perform a qualified examination of the ASME Section XI volume of the wrought piping material and the weld. There are several complementary bases for this proposed alternative, as will be described in the text to follow. It should be noted that the actual situations described here are not identical to the valve-to-pipe welds of interest here, but the situation is sufficiently similar that the same arguments are applicable. For example, the materials and water chemistry are the same, so the operating history discussed herein would be expected to be directly applicable, and the conclusion that any flaw indication acceptable by examination would remain acceptably small throughout service life remains valid. It will be seen that the issue of inspectability of stainless steel has existed for many years. The purpose for showing the examples is to demonstrate that regulatory treatment of the issue has been consistent throughout this period of time, even up to the present day. The details of the technical basis are outlined below: Page 21 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds

  • The first example of a situation where volumetric examinations of cast stainless steel were addressed is in Code Case N-481. This code case replaced the requirement of a volumetric examination of pump casings with a visual examination. The technical basis for this Code Case will be reviewed in detail, but the key points were that cast stainless steel is highly resistant to corrosion, has low stress due to the cast design, and has not had any issues with degradation during service. A requirement for a flaw tolerance evaluation was added to the code case, and many submittals were made, to cover the range of different cast austenitic stainless-steel (CASS) pump casing designs.
  • Code Case N-481 was incorporated into the ASME Code at some time in the 1980s, at which time valve bodies were added to the exemption. The NRC insisted that the flaw tolerance requirement be dropped, because the staff had seen enough such evaluations to be convinced of the flaw tolerance of these components. The existing inservice examination rules for pump and valve welds in ASME Section XI are now allowed to be done visually.
  • Recently, Code Case N-770 was developed to provide enhanced examination requirements for Alloy 600 and its associated welds because of observed degradation during service. It is important to note that the technical basis for N-770-1 [7], mentions the intention that only a best effort volumetric examination should be required of the cast stainless steel portions of these susceptible welds. The key arguments of the technical basis are summarized below.

The NRC staff has approved both of these Code documents, and has not identified any concerns in this area, even with the recently issued Generic Aging Lessons Learned document for second license renewal (GALL-SLR). Technical Basis for Code Case N-481 The technical basis for Code Case N-481 was based on application of the key aspects of LLB to the reactor coolant pump casings of Westinghouse plants. Since the results were similar for all the different pump models, it was assumed that applications to similar cast structures, such as other pump designs would yield the same results. A summary of the technical basis paper [1] is provided in the paragraphs to follow. In the years preceding the acceptance of Code Case N-481, Section XI of the ASME Boiler and Pressure Vessel Code required examination of reactor coolant pump casing welds as well as pump internal surfaces once during each ten (10) years of plant operation. Category B12.10 (B-L-1) specified a volumetric examination of the pump casing weld(s). Category B12.20 (B-L-2) specified a visual examination of the internal pressure boundary surfaces of the pump casing. One pump in each group of pumps performing similar functions in the system was required to be examined during each 10-year interval. Page 22 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Both the volumetric and visual examinations required by ASME Section XI at that time would require disassembly of the pump. The only effective volumetric examination method available for welds in the heavy sections cast stainless-steel pump casings was radiography using the miniature linear accelerator (MINAC). In order to perform the radiographic examination, the pump must be disassembled completely, including removal of the diffuser adapter and casing adapter. This amount of disassembly is significantly greater than that performed during normal pump maintenance. Because the pump bowl must be dry for installation of the MINAC, a complete core unload is typically required since the reactor coolant system water level must be drained below the level necessary to maintain shutdown cooling for the core. Disassembly, preparation, examination, and reassembly of the pump also resulted in high levels of radiation exposure to maintenance and examination personnel. Utilities performing such examinations reported overall radiation exposures ranging from 35 to 100 Man-Rem. The remainder of this summary reviewed the safety and serviceability for the class of reactor pump casings. Three-dimensional finite element stress analysis results of the dominant pump casing designs have been summarized. Fatigue and fatigue crack growth (FCG) results have been reviewed. Extensive elastic-plastic fracture mechanics results were performed for large postulated through-wall flaws sized by leak-rate calculations to criteria an order of magnitude greater than the detection capacity of the plants. Significant margins in flaw size and load were demonstrated. In short, the LBB approach applied to reactor coolant loop piping systems was applied to pump casings. The methodology used in the Westinghouse Owners Group study for evaluating the LBB for the pump casings is consistent with that recommended in NUREG 1061, Volume 3 [2]. Note that this methodology was later implemented into the US NRC Standard Review Plan [3]. The methodologies involved:

9. Establishing material properties including fracture toughness values
10. Performing stress analyses of the structure
11. Review of operating history of the structure
12. Selection of locations for postulating flaws
13. Determining a flaw size giving a detectable leak rate
14. Establishing stability of the selected flaw
15. Establishing adequate margins in terms of leak rate detection, flaw size and load
16. Showing that a flaw indication acceptable by examination remains small throughout service life Page 23 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds The pump casings are all fabricated from cast stainless steel. Such materials are highly resistant to corrosion issues and exhibit high-fracture toughness values (i.e., JIc and Tmat) in the pre-service product form; however, such materials are subject to thermal aging embrittlement at nominal operating temperatures of nuclear plants. The materials for each pump casing were evaluated for thermal aging embrittlement consistent with the methodology approved by the NRC and were all found to meet the acceptance criteria. The pump casings were designed to the ASME code and other requirements in force at the time of fabrication. Design, normal, operating, and faulted applied loads on the nozzles are very conservative when compared to actual service conditions. The stresses are such that code conditions are easily met. Elastic-plastic fracture mechanics J-integral analyses were made at critical locations. The selection of critical locations was based on fracture toughness, normal stress level and faulted stress level. For example, the six critical locations for the model 93A pump casing are shown in Figure A1-1. For all the pump casings examined, a total of fourteen critical locations were evaluated. Margin on flaw size was demonstrated by showing stability (i.e., the fracture criteria are met) exists under faulted loading for a flaw twice the size giving the selected leak rate (10 gpm). Margin on load was established by showing stability for faulted loads increased by up to a factor of 1.4 using the flaw size giving the selected leak rate. To determine the sensitivity of the reactor coolant pump casing to cyclic fatigue loading, fatigue analyses were performed. Two types of fatigue analyses were included, one a conventional ASME Section III evaluation to determine a maximum cumulative usage factor, and the other an ASME Section XI FCG analysis assuming various initial flaw sizes. The three-dimensional finite element stress analysis results at the worst stressed location were used. Deadweight, pressure and all normal and upset transient loadings were considered. A cumulative usage factor of approximately 0.15 was found, which is much less than the allowable usage factor of 1.0. Fatigue crack growth was small. A flaw having the maximum acceptable depth allowed by the ASME Code was seen to grow only 0.16 in. for 40 years of service. In summary, service conditions are such that flaws are not expected to occur throughout the life of the plant. If small flaws are present, they will grow only a small amount throughout the service life. If a flaw should penetrate the wall it will leak at a very detectable rate and will not be unstable thus allowing a shutdown of the plant to address the condition. In conclusion, successful demonstrations of LBB were provided which are applicable to the primary pump casings of all Westinghouse design PWRs. Since the results were similar for all the different pump models, it was assumed that applications to similar cast structures, such as other pump designs would yield the same results. As the Code Case was applied, it became clear that similar conclusions could be reached for cast valve bodies, and the applicability of the Code Case was expanded to cover valve bodies in the year 2000. Page 24 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Adoption of Code Case N-481 into the ASME Code The applicability to Westinghouse units was demonstrated formally in [4], which is summarized in Attachment 2, Section A2-1. The NRC issued a formal approval of this report in 1993, and the text of the approval appears in Section A2-2 of Attachment 2. Several examples of plant-specific applications are shown in Section A2-3 of Attachment 2. The general applicability of the Code Case N-481 became clear as it was applied at more and more operating plants. In the year 2000, the Code Case was incorporated into the body of the ASME Code (as reproduced in Table A1-1). However, the flaw tolerance calculation was removed, because multiple flaw tolerance calculations had been completed and reviewed by NRC, which demonstrated that these components were both flaw tolerant and resistant to degradation by corrosion mechanisms (the NRC representative on the appropriate Code Committee insisted on this action, since he considered the continued review of such calculations an unwarranted expenditure of funds.). In 2008, valve bodies were treated in the same manner because they were of similar design and had likewise been trouble-free for their entire operating life. Similar examination requirements for valve bodies were inserted in Table IWB-2500, which is reproduced as Table A1-2. In 2008, this table of examination requirements was also simplified to have one category for pumps and a second category for valves. Again, this change was driven by the excellent operating experience in both components. There was no formal written basis for the 2008 revision, other than the explanation which accompanied the action, which is reproduced below, courtesy of Reedy Engineering: Page 25 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds In addition to the explanation above, a search was made of INPO and other industry databases, and failed to identify any examples of service degradation of pump casings or valve bodies resulting from corrosion or any other cracking mechanism. The NRC formally accepted the 2000 Addenda of the ASME Code, through endorsement in 10 CFR 50.55a, in 2002 [5]. Similarly, The NRC endorsed the 2008 Addenda to the 2007 Edition of the Code, publishing the endorsement in 10 CFR 50.55a, in 2011 [6]. In both cases, there were no exceptions taken in these portions of ASME Section XI. Since the code case had been incorporated in an edition of the Code endorsed by NRC, the code case was annulled in 2004. Code Case N-770: The Latest Treatment of This Issue in Section XI To demonstrate the consistency with which the NRC has treated the issue of examination requirements for Cast Stainless Steel, a summary is provided of the background and basis of this code case. The basis was published as a technical paper at the ASME Pressure Vessels and Piping Conference in 2010 [7]. The goal of Code Case N-770, and its first revision, N-770-1, was to increase the examination requirements for Alloy 600, Alloy 82, and Alloy 182 materials, to account for their susceptibility to Primary Water Stress Corrosion Crack (PWSCC), and to identify appropriate examination requirements after mitigations were performed. Since these examinations often involved welds, where the base metal on one side of the weld was stainless steel, the issue of inspectability of the stainless steel had to be addressed. In section 3.2 of the technical basis for N-770-1 Ultrasonic Testing, the following statement is made: For cast stainless-steel items for which no supplement is available in Appendix VIII, the examination volume shall be examined by Appendix VIII procedures to the maximum extent practicable If 100% coverage of the required volume for axial and circumferential flaws cannot be met, 100% coverage for circumferential flaws (of the susceptible material) is to be met. This is the practical solution for the examination of the weld and buttering of the susceptible material when the base metal is cast stainless steel, or otherwise not completely inspectable. The examination coverage requirements will then be considered to be met. Code Case N-770-1 was endorsed and made mandatory by the NRC with the endorsement of the 2008 Addenda to the 2007 Code [6], further confirming the treatment of cast stainless steel materials which started with the original acceptance of Code Case N-481. Treatment of Cast Austenitic Stainless-Steel (CASS) Structural Integrity by Grimes [8] A structural integrity evaluation based on a fracture mechanics analysis was also performed for CASS valves as part of the NRC Grimes letter [8] for license renewal applications. The conclusion of the Grimes letter was that CASS valve bodies are flaw tolerant and have adequate fracture toughness even for very large through-wall flaws. The fracture mechanics evaluation in the Grimes letter was based on an elastic-plastic assessment using the Failure Assessment Diagram (FAD) methodology for a representative NPS 4 SA-351 CF8 valve. The evaluation used conservative input assumptions and demonstrated that the CASS valve could sustain a 1.35-in. long through-wall crack. Thus, the Grimes letter performed a conservative Page 26 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds bounding integrity analysis to estimate a large crack size that would be required in the valve bodies before challenging the integrity of the valve and the adjacent weld. Below is the excerpt from the Grimes letter attachment, on Integrity of <4-inch NPS Valve Bodies Made from CAST Stainless Steel. With these conservative input assumptions, the FAD analysis shows that the small diameter CASS valve bodies could withstand a through-wall defect approximately 1.35 inches long. While the specific value would be specific to the application, the analysis demonstrates that the CASS valve bodies are flaw-tolerant, even for severely aged materials. Based on the fact that we did not identify any service failure history associated with these small diameter CASS valve bodies, and the fact that they can withstand very long throughwall cracks, even under high stresses, suggests that additional inspections during a license renewal period are not warranted. We therefore conclude that the present requirements for in-service inspection are adequate. Summary and Conclusions The issue of examination capabilities for stainless steel, and their implementation into ASME Code requirements, has been recognized for many years. Stainless steel in general and cast stainless steel in particular is highly resistant to corrosion and has very high fracture toughness, which counteracts the concern about inspectability. Also, service experience in PWR plants over the past 40 years has been excellent. The inspectability issue was first addressed in Code Case N-481, which eliminated the need for volumetric examination of pump casings, and experience was good with this case, so it has been implemented into the ASME code for more than 14 years. The NRC approved the Code Case initially and also was part of the driving force behind its implementation into the Code, which was endorsed with no exceptions in the Code of Federal Regulations. Recently, the approach was continued by the NRCs endorsement of Code Case N-770 as mandatory, again with no exceptions in the area of cast stainless steel examinations. Therefore, the basis which has been employed for many years is offered to support the proposed alternative. Page 27 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Table A1-1: Initial Implementation of Code Case N-481 into Section XI (2000 -2007) Page 28 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Table A1-2: Revision of Examination Requirements for Pumps and Valves in the 2008 Edition Page 29 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Figure A2-1: Location of the Six Flaws Postulated for Analyses in the Model 93A Pump Casing Page 30 of 64

ND-18-1185 Attachment 1 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Non-Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds References

1. Kammerdeiner, Greg, Code Case on Alternative Examination Requirements for Section XI, Division 1, Examination of Pump Casings, ASME Committee Correspondence, Aug. 18, 1988.
2. Report of the US Nuclear Regulatory Commission Piping Review Committee:

Evaluation for Pipe Breaks NUREG-1061, Vol.3, November 1984.

3. Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol.52, No.167/Friday, August 28, 1987/Notices, pp. 32626-32633.
4. Witt, F.J. and Petsche, J.F., Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Nuclear Steam Supply Systems, Westinghouse Electric Report WCAP-13045, Sept. 1991.
5. NRC endorsement of ASME Code, Section XI, 2000 Addenda: 67FR60520, Sept. 26, 2002
6. NRC Endorsement of ASME Code, Section XI, 2008 Addenda to 2007 Edition:

76FR36232, July 21, 2011.

7. Donavin, P., Elder, G.G., and Bamford, W.H., Technical Basis Document for Alloy 82/182 Weld Inspection Code Case N-770 and N-770-1, in Proceedings, ASME Pressure Vessels and Piping Conference, July, 2010.
8. Letter from Christopher I. Grimes, U.S. Nuclear Regulatory Commission, License Renewal and Standardization Branch, to Douglas J. Walters, Nuclear Energy Institute, License Renewal Issue No.98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components. ML003717179. May 19, 2000.

Page 31 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds ATTACHMENT 2: Compliance with Code Case N-481, and NRC Acceptance Thereof Code Case N-481 was accepted by the NRC, which led to the submittal of a topical report, WCAP-13045. The highlights of this report are provided in Attachment A2-1, which follows. The NRC accepted this report in a letter dated April 13, 1993, which is reproduced in Attachment A2-2 below. This letter stated that individual submittals by individual plants were not necessary, but the documentation of compliance with the Code Case should be maintained at each plant site, available for NRC audit. Summaries of several such reports are provided in Attachment A2-3 of this appendix. The use of these reports has also been extended for applicability to the license renewal period from 40 to 60 years, and an example of NRC acceptance in this situation is found in Attachment A2-4. Page 32 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Attachment A2-1: Summary of WCAP-13045, Demonstrating Compliance with N-481 for Westinghouse Pump Models A2-1: Summary of WCAP-13045, Compliance to ASME Code CASE N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems, September 1991. WCAP-13045 is a generic integrity evaluation, applicable to all Westinghouse design pump casings, which demonstrates compliance to ASME Code Case N-481. WCAP-13045 does not provide plant-specific pump casing evaluations, because it was not found feasible to qualify each plant specifically. Instead, WCAP-13045 provides enveloping or bounding criteria whereby a specific utility need only show that the pump casings of interest fall under the umbrella established in WCAP-13045. There are eight different models of pumps in the Westinghouse-type pressurizer water reactors, Models 63, 70, 93, 93A, 93A-1, 93D, 100A, and 100D. Pump Models 63, 70, 93, and 93D have a tangent outlet nozzle while pump Models 93A and 93A-1 have outlet nozzles that are radially oriented. Pump Models 100A and 100D have a general shape somewhat like Model 93, but a radially-oriented outlet nozzle like Model 93A. Models 93A and 93A-1 are single-casting pump casings with no welds. About 90% of the domestic plants use the Models 93, 93A, and 93A-1. The materials for the pump casings are SA351 CF8 and SA351 CF8M for all but three plants which are made of SA351 CF8. Instead of modeling each different model of pump, a model representative of each of the outlet nozzle configurations were chosen for the 3D finite element stress analyses and fracture evaluation. Model 93A was chosen for the type of pumps with a radial outlet nozzle, and Model 93 was chosen for the type of pumps with a tangential outlet nozzle. Models 93 and 93A were chosen, because they are the most typically used pump in the domestic plants. Axisymmetric models for the Model 63 and Model 100A pumps were developed which did not contain the outlet nozzles. The loads used in the analysis were selected to be conservative for a majority of the plants. The primary loads of interest are internal pressure and the force and moment loadings on the nozzles. The highest stresses in the 3D models were found to be in the outlet nozzle crotch region. The surface stress exceeded 50 ksi for the Model 93A pump casing and 40 ksi for Model 93 pump casing. The selections of the locations for the postulated quarter thickness cracks were selected by looking at the stress results for the 3D and axisymmetric models. The selected flaw locations were then used in the plant-specific WCAP evaluations, depending on the plants model pump casing. In total, 11 locations on the 3D models and 4 locations on the axisymmetric models were chosen for postulated flaws, due the stress and locations of the welds. A stability evaluation was performed in order to determine if enough margin is available to meet the stability criteria. The stability analysis completed in WCAP-13045 required re-evaluation for each plant thats plant-specific loads are not bounded by WCAP-13045. The stability analysis was conducted for Models 93, 93A, 63, and 100A at the previously selected critical locations for postulating quarter-thickness flaws. Flaws had a 6-to-1 aspect ratio with exceptions in certain pump casing models, where the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. The critical locations for the four different pump casings are shown in the Figures below. The normal and Page 33 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds faulted loadings for the pump models are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the pump casings per the ASME Code Case N-481, when subject to the normal and faulted loadings outlined in WCAP-13045, are determined to be stable. In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience. Therefore, the sensitivity of the postulated cracks to cyclic loadings was evaluated as a generic FCG analysis. The highest stress location for each pump casing model was chosen for FCG. Through the FCG analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses. The Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion. Strict pipe-cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Contaminant concentrations are kept below the thresholds known to be conducive to stress-corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. Thus, during plant operation, the likelihood of stress-corrosion cracking is minimized. Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled lines. Additionally, the operating transients of the RCS primary piping are such that no significant water hammer can occur. An evaluation of the low- and high-cycle fatigue loadings was carried out as part of this study. An assessment of the low-cycle fatigue was performed in the FCG assessment. High-cycle fatigue loads in the system would result primarily from pump vibrations. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for FCG. It is concluded that the primary loop pump casings of all models are in compliance with Item (d) of ASME Code Case N-481 as long as the plant-specific loads are bounded by the loads used in WCAP-13045. If the loads are not bounded the information provided in WCAP-13045 requires to be updated based on the plant-specific loads. Page 34 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Critical Locations for Postulating Quarter-Thickness Flaws Page 35 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Critical Locations for Postulating Quarter-Thickness Flaws Page 36 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Critical Locations for Postulating Quarter-Thickness Flaws Page 37 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and lnservice Inspection Requ irements for Specific Valve-to-Pipe Welds Critical Locations for Postulating Quarter-Thickness Flaws 

                                  .) .

l*100.A Page 38 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Attachment A2-2: NRC Acceptance of WCAP-13045 Page 39 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Attachment A2-3: Example Summaries of plant Specific Reports Demonstrating Compliance with Code Case N-481 A2-3.1: Summary of WCAP-16957-P, A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casing of Plant A for the License Renewal Program, March 2009. The primary loop pump casings for the Plant A Units are the Westinghouse Model 93A design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless-steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant A pump casings, the loadings used in WCAP-13045 needed to be compared to the unit-specific loads for Plant A. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant A Units are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant A normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant A loads compare to the generic WCAP-13045 loads as follows:

  • The Plant A normal forces at the inlet and outlet nozzle are not bounded by WCAP-13045. The Plant A normal forces at the inlet and outlet nozzle are larger than the loads in WCAP-13045.
  • The Plant A normal moments at the inlet and outlet nozzle are bounded by WCAP-13045. The Plant A normal moments at the inlet and outlet nozzle are smaller than the loads in WCAP-13045.
  • The Plant A faulted forces, moments and total stresses at the inlet and outlet nozzle are bounded by WCAP-13045. The Plant A faulted force and moment at the inlet and outlet nozzle are smaller than the loads in WCAP-13045.

Since not all the normal loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant A Units to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant A loads. The stability analysis for Plant A was conducted at four critical locations for postulating quarter-thickness flaws in 93A pump casings as described in WCAP-13045. Flaws have a 6-to-1 aspect ratio with exception of the two flaws in the outer outlet nozzle knuckle region 3-93A and 4-93A, where the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. One postulated flaw was a circumferential crack in the inlet nozzle, the location selected being in the high-stress region near the nozzle to pipe weld. The second flaw is in a nominal weld location in the casing proper. The last two flaws were selected in the high-stress regions of the outlet nozzle associated with the nozzle crotch. The critical locations for the Plant A pump casings are shown in the below figure. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant A pump casings per the ASME Code Case N-481, when subject to the normal and faulted loadings are determined to be stable. In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience. Therefore, the sensitivity of the postulated cracks-to-cyclic loadings was evaluated as a generic FCG analysis. The highest stress location, Flaw 4-93A, was chosen for FCG. The postulated Page 40 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds flaw at the outlet nozzle knuckle is in the axial direction of the pump casing. Through the FCG analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses. Any reasonably sized flaws in the pump casings will exhibit only minimal crack extension during 40 years, such flaws remaining well below the flaw sizes shown to be stable. Since the transients and cycles for the 60-year plant are the same as the 40-year plant, the FCG assessment is applicable to 60 years. The Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion. Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Contaminant concentrations are kept below the thresholds known to be conducive to stress-corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. Thus, during plant operation, the likelihood of stress-corrosion cracking is minimized. Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled lines. Additionally, the operating transients of the RCS primary piping are such that no significant water hammer can occur. An evaluation of the low- and high-cycle fatigue loadings was carried out as part of this study. The assessment of the low-cycle fatigue was performed in the FCG assessment. High-cycle fatigue loads in the system would result primarily from pump vibrations. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for FCG (See the figure below). It is concluded that the primary loop pump casings of Plant A are in compliance with Item (d) of ASME Code Case N-481 and applicable for the license renewal period. Page 41 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 42 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 43 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 44 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds A2-3.2: Summary of WCAP-15355, A Demonstration of Applicability of ASME Code Case N-481 to the Primary Pump Casings of Plant B, January 2000. The primary loop pump casings for the Plant B Units are the Westinghouse Model 93 design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless-steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant B pump casings, the loadings used in WCAP-13045 needed to be compared to the unit-specific loads for Plant B. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant B Units are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant B normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant B loads compare to the generic WCAP-13045 loads as follows:

  • The Plant B normal moments at the inlet and outlet nozzle is bounded by WCAP-13045. The Plant B normal moments at the inlet and outlet nozzle is smaller than the load in WCAP-13045.
  • The Plant B normal forces at the inlet and outlet nozzle is not bounded by WCAP-13045. The Plant B normal forces at the inlet and outlet nozzle is larger than the load in WCAP-13045.
  • The Plant B faulted force and moment at the inlet and outlet nozzle is bounded by WCAP-13045. The Plant B faulted force and moment at the inlet and outlet nozzle is smaller than the load in WCAP-13045.

Since not all the normal loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant B Units to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant B loads. The stability analysis for Plant B was conducted at seven critical locations for postulating quarter-thickness flaws in 93 pump casings as described in WCAP-13045. Flaws have a 6-to-1 aspect ratio with one exception. The exception is the flaw selected at the outlet nozzle knuckle. An aspect ratio is not defined but the crack front curvature is representative of the crack front curvature for a crack having a 6-to-1 aspect ratio. For the outlet nozzle knuckle, the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. The critical locations for the Plant B pump casings are shown in the figure below. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant B pump casing per the ASME Code Case N-481, when subject to the normal and faulted loadings, are determined to be stable. In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience. Therefore, the sensitivity of the postulated cracks-to-cyclic loadings was evaluated as a generic FCG analysis. The highest stress location, Flaw 5-93, was chosen for FCG. The postulated flaw at the outlet nozzle knuckle is in the plane of the weld. Through the FCG analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses. Any reasonably sized flaws in the pump casings will exhibit only minimal crack extension during 40 years, such flaws remaining well below the flaw sizes shown to be stable. Since the transients and cycles for the 60-year plant are the same as the 40-year plant, the FCG assessment is applicable to 60 years. Page 45 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds The Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to cracking failure from the effects of corrosion. Strict pipe cleaning standards prior to operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Contaminant concentrations are kept below the thresholds known to be conducive to stress-corrosion cracking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. Thus, during plant operation, the likelihood of stress-corrosion cracking is minimized. Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled lines. Additionally, the operating transients of the RCS primary piping are such that no significant water hammer can occur. An evaluation of the low- and high-cycle fatigue loadings was carried out as part of this study. The assessment of the Low-cycle fatigue was performed in the FCG assessment. High-cycle fatigue loads in the system would result primarily from pump vibrations. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for FCG (See the figure below). It is concluded that the primary loop pump casings of Plant B are in compliance with Item (d) of ASME Code Case N-481. Page 46 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 47 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds A2-3.3: Summary of WCAP-16377, A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casing of Plant C for the License Renewal program , January 2005. The primary loop pump casings for the Plant C Unit are the Westinghouse Model 93A design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant C pump casings, the loadings used in WCAP-13045 needed to be compared to the unit specific loads for Plant C. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant C Unit are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant C normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant C loads compare to the generic WCAP-13045 loads as follows:

  • The Plant C normal force and moment at the inlet nozzle are bounded by WCAP-13045. The Plant C normal force and moment at the inlet nozzle is smaller than the load in WCAP-13045.
  • The Plant C normal force at the outlet nozzle is bounded by WCAP-13045. The Plant C normal force at the outlet nozzle is smaller than the load in WCAP-13045.
  • The Plant C normal moment at the outlet nozzle is not bounded by WCAP-13045.

The Plant C normal moment at the outlet nozzle is larger than the load in WCAP-13045.

  • The Plant C faulted force and moment at the inlet nozzle is bounded by WCAP-13045. The Plant C faulted force and moment at the inlet nozzle is smaller than the load in WCAP-13045.
  • The Plant C faulted force at the outlet nozzle is bounded by WCAP-13045. The Plant C faulted force at the outlet nozzle is smaller than the load in WCAP-13045
  • The Plant C faulted moment at the outlet nozzle is not bounded by WCAP-13045.

The Plant C faulted moment at the outlet nozzle is larger than the load in WCAP-13045. Since not all the normal and faulted loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant C Unit to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant C loads. The stability analysis for Plant C was conducted at two critical locations for postulating quarter thickness flaws in 93A pump casings as described in WCAP-13045. Flaws have a 6-to-1 aspect ratio and quarter thickness depths where the depth of the crack is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch. The two flaws were selected in the high stress regions of the outlet nozzle associated with the nozzle crotch, axially oriented with respect to the casing. The critical location for the Plant C pump casings are shown in the below figure. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant C pump casing per the ASME Code Case N-481, when subject to the normal and faulted loadings are determined to be stable. In the stability analyses, cracks are postulated at various locations in the pump casings. These postulated cracks are subject to various cyclic conditions the pump casings experience. Therefore, the sensitivity of the postulated cracks to cyclic loadings was evaluated as a generic FCG analysis. The highest stress location, Flaw 4-93A, was chosen for FCG. The postulated Page 48 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds flaw is at the outlet nozzle knuckle in the axial direction of the pump casing. Through the FCG analysis, the end-of-period flaw values are well less than the stable flaw sizes established in the stability analyses. Any reasonably sized flaws in the pump casings will exhibit only minimal crack extension during 40 years, such flaws remaining well below the flaw sizes shown to be stable. Since the transients and cycles for the 60-year plant same as the 40-year plant, the FCG assessment is applicable to 60 years. It is concluded that the primary loop pump casings of Plant C are in compliance with Item (d) of ASME Code Case N-481 and applicable for the license renewal period. Page 49 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 50 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds A2-3.4: Summary of WCAP- 15169, A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casing of Plant D, March 1999. The primary loop pump casings for the Plant D Units are the Westinghouse Model 100A design. These pump casings are fabricated from SA351 CF8 (this material has Type 304 stainless steel chemistry) cast stainless steel. In order to ensure the generic WCAP-13045 report was bounding for the Plant D pump casings, the loadings used in WCAP-13045 needed to be compared to the unit-specific loads for Plant D. The faulted nozzle loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Plant D Units are compared with the Level A screening (i.e., enveloping) faulted loads from WCAP-13045. The Plant D normal loads are compared with the Level C screening normal loads for evaluating the loss-of-load condition in WCAP-13045. The Plant D loads compare to the generic WCAP-13045 loads as follows:

  • The Plant D normal force at the inlet nozzle is not bounded by WCAP-13045. The Plant D normal force at the inlet nozzle is larger than the load in WCAP-13045.
  • The Plant D normal moment at the inlet nozzle is bounded by WCAP-13045. The Plant D normal moment at the inlet nozzle is smaller than the load in WCAP-13045.
  • The Plant D normal force and moment at the outlet nozzle are not bounded by WCAP-13045. The Plant D normal force and moment at the outlet nozzle is larger than the load in WCAP-13045.
  • The Plant D Faulted force and moment at the inlet nozzle are bounded by WCAP-13045. The Plant D faulted force and moment at the inlet nozzle is smaller than the load in WCAP-13045.
  • The Plant D faulted force and moment at the outlet nozzle are not bounded by WCAP-13045. The Plant D faulted force and moment at the outlet nozzle is larger than the load in WCAP-13045.

Since not all the normal and faulted loading conditions are bounded by WCAP-13045, a stability evaluation had to be performed in order to determine if enough margin is available to allow the Plant D units to meet the stability criteria. The stability analysis completed in WCAP-13045 was re-evaluated using the Plant D loads. The stability analysis for Plant D was conducted at one critical location as described in WCAP-13045. The critical location for the Plant D pump casings was 1-100A, as shown in the below figure. The crack was chosen at the stress concentration which is representative of the weld location. The normal and faulted loadings are determined to be stable because the stability results are lower than the bounding criteria. It can be concluded that flaws postulated in the Plant D pump casing per the ASME Code Case N-481, when subject to the normal and faulted loadings are determined to be stable. It is concluded that the primary loop pump casings of Plant D are in compliance with Item (d) of ASME Code Case N-481. Page 51 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 52 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Attachment A2-4: NRC Acceptance of Compliance with N-481 for License Renewal Page 53 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Key Excerpts From the Report: Page 54 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 55 of 64

ND-18-1185 Attachment 2 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Page 56 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds : Flaw Tolerance Evaluation in Support of Technical Basis for Alternative Request on Preservice and Inservice Inspection of Stainless Steel Valve-to-Pipe Welds Introduction The Class 1 and 2 cast austenitic stainless steel (CASS) valve-to-pipe welds do not allow for essentially 100% of the required examination volume coverage as defined in ASME Section XI, IWA-2200 (c) due to restrictions imposed on the ultrasonic examination of the CASS material. Qualified ultrasonic examination coverage will be approximately 50% of the examination volume, even if the exam from the pipe side of the weld was perfect because there is no ultrasonic qualification in place for exams from the cast valve body side of the weld. No credit can be given to the examination volume, which consist of the austenitic valve material and one-half of the weld material adjacent to the valve material. Therefore, to support the alternative request for the examination of these particular locations, a sample flaw tolerance evaluation is provided in this attachment to demonstrate that the valve-to-pipe weld location is flaw tolerant (i.e., a postulated flaw in the required examination region does not grow to the maximum allowable end-of-evaluation period flaw size within the life of the plant). [The 

                                                                                      ](a, c)

Page 57 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Flaw Tolerance Evaluation Methodology The CASS valve-to-pipe weld location is evaluated based on the guidelines in paragraph IWB-3640 and Appendix C of the 2007 Edition with 2008 Addenda ASME Section XI code [1]. The maximum allowable end-of-evaluation period flaw size is calculated at the representative valve location discussed above for a postulated inside surface axial and circumferential flaw with an Aspect Ratio (AR), flaw length/flaw depth equal to 6:1. This particular aspect ratio will conservatively bound any flaws discovered during pre-service and in-service inspections; furthermore, the postulated flaw shape is typical of ASME Section XI flaw evaluations. The evaluations of inside surface flaws are considered herein, because the required examination for this valve to weld location is of the inner one-third thickness of the wall. Furthermore, the evaluation of postulated inside surface flaws conservatively covers evaluations for embedded flaws and outside surface flaws as the stress intensity factors and crack growth rates for inside surface flaws are more limiting than embedded and outside surface flaws. The primary crack growth mechanism for postulated flaws in the stainless-steel valve-to-pipe weld location is Fatigue Crack Growth (FCG). Crack growth due to PWSCC growth does not need to be investigated since the base metal (CASS valve) and stainless steel weld material have a very low susceptibility to stress corrosion cracking due to the lack of oxygen in a PWR environment. The maximum allowable postulated flaw size is determined by subtracting the 60-year (design life) FCG from the maximum allowable end-of-evaluation flaw sizes. The FCG rate as well as the stress intensity factor equations required to complete an FCG analysis is further discussed below. The purpose of this evaluation is to demonstrate that the maximum allowable postulated flaw size for 60 years which would be acceptable is large compared to the area which requires examination. Maximum Allowable End-of-Evaluation Period Flaw Sizes The calculation of the maximum allowable end-of-evaluation period flaw sizes for stainless-steel base metals and stainless-steel welds are based on the guidelines from paragraph IWB-3640 and Appendix C of ASME Section XI [1]. These guidelines were used to determine the maximum allowable end-of-evaluation period flaw size for axial and circumferential flaw configurations. An aspect ratio of 6 is conservative for both axial and circumferential flaws. The maximum end-of-evaluation period flaw sizes determined for both axial and circumferential flaws have incorporated the limiting material properties, loadings and geometry. Loadings under normal, upset, test, emergency and faulted conditions were considered with the applicable safety factors for the corresponding service conditions required in the ASME Code Section XI. For circumferential flaws, axial stress due to the pressure, deadweight, thermal, seismic, and pipe break loads were considered in the evaluation. As for the axial flaws, hoop stress resulting from pressure loading was used since it is the only stress which affects an axial flaw. The dimensions and operating parameters for a representative CASS valve to piping weld location are shown in Table A3-1. A normal operating temperature of 660°F was conservatively used in determining the maximum allowable end-of-evaluation period flaw size. The CASS Page 58 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds valve-to-pipe weld loads are given in Table A3-2 in the local coordinate system, where the x-axis is along the component centerline, and the y-axis and z-axis by right-hand-rule. (a,c) Table A3-1: CASS Valve to Piping Weld Geometry and Operating Parameters Table A3-2: CASS Valve-to-Piping Loads (a,c) [The Page 59 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds order to 

                                          .] (a,c) Therefore, the maximum end-of evaluation period allowable flaw size can be determined with the consideration of the CASS material properties without the need to account for reduced toughness due to thermal aging. Furthermore, the valve-to-pipe welds are based on GTAW (gas tungsten arc weld), as this required weld process for the welds included in this alternative. Thus, the maximum end-of-evaluation flaw sizes will be based on limit load with Z = 1, per the guidance of ASME Section XI Appendix C.

Therefore, the calculated maximum allowable end-of-evaluation period flaw sizes for the CASS valve-to-pipe weld location are shown in Table A3-3. Table A3-3: Maximum Allowable End-of-Evaluation Period Flaw Size (Aspect ratio = 6) Maximum End-of-Evaluation Flaw Configuration Period Flaw Size (af/t) Axial Surface Flaw 0.53 Circumferential Surface Flaw 0.42 From this calculation, it is concluded that the flaw tolerance of the valve-to-pipe weld is very large. The second step in this process involves a series of FCG analyses which are performed to determine the largest postulated allowable initial flaw size for 60 years of plant operation such that the final crack growth flaw sizes will not reach the maximum-end-of-evaluation period flaw sizes shown in Table A3-3. Fatigue Crack Growth Analysis Fatigue crack growth is the only credible mechanism for crack growth in the material between the CASS valve-to-pipe weld since both the weld and the base metals have very low susceptibility to PWSCC. The FCG analysis procedure involves postulating an initial flaw at the weld region and predicting the growth of that flaw due to an imposed series of loading transients. The input required for a FCG analysis is essentially the information necessary to calculate the range of crack tip stress intensity factor, which depends on the crack size and shape, geometry of the structural component where the crack is postulated, and the applied cyclic stresses. The FCG analysis involves calculating growth for a flaw on the inside surface of the CASS valve-to-pipe weld, for both an axial and circumferential (AR = 6) flaw. The postulated flaws are subjected to cyclic loads due to the transients. Welding residual stresses were added to the transients operating stresses. The inputs required for the FCG analysis is the range in stress intensity factor, K, and the R ratio (Kmin/Kmax), all of which are functions of the stresses and flaw size. Page 60 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds The methodology for calculating the crack tip stress intensity factors is documented in Mettu and Raju [5] for axial flaws and Chapuliot, Lacire, and Le Delliou [6] for circumferential flaws. When evaluating axial and circumferential flaws, semi-elliptical surface flaws with aspect ratios (flaw length/flaw depth) of 6 for axial flaws and circumferential flaws are considered. Once R (load ratio = Kmin/Kmax) and K are calculated, the crack growth due to any given stress cycle can be calculated for each transient. This increment of crack growth is then added to the original crack size, and the analysis proceeds to the next transient. The procedure is continued in this manner until all the transients known to occur in the period of evaluation have been analyzed. The design transient load cycles used in the analysis for the CASS valve-to-pipe weld are listed in Table A3-4. The reference crack growth law used for the CASS valve-to-pipe weld location was taken from Appendix C of Section XI [1] for air environments. A factor of 2 for PWR environments per Reference [4] is included in the reference crack growth law of Appendix C of ASME Section XI code for air environments. Welding Residual Stress For the FCG analysis, the stainless-steel welding residual stresses in the CASS valve-to-pipe weld are also considered along with the transient stresses. The welding residual stress values contained in Reference [4] were used. Page 61 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSl/ISl-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and lnservice Inspection Requ irements for Specific Valve-to-Pipe Welds Transients and Cycles The thermal and pressure transient hoop and axial stress data used in the FCG analysis for the CASS valve-to-pipe location are listed in Table A3-4. The transient stresses are combined with the welding residual stresses and then used in the FCG analysis. It should be noted that the emergency and faulted transients are considered for conservatism in the FCG evaluations. Table A4-4: Design Transients and Cycles (a,c) Page 62 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds Fatigue Crack Growth (FCG) Results The FCG evaluation is performed based on the methodology discussed above for the CASS valve-to-pipe weld locations. The FCG results are subtracted from the maximum allowable end-of-evaluation flaw size to determine the maximum allowable initial flaw size. The maximum initial flaw size is a sufficiently large postulated flaw which would not reach the maximum allowable end-of-evaluation flaw size in 60 years. The growth of an initial postulated flaw size is provided in increments of 10 years over a period of 60 years for an axial and circumferential flaw with an aspect ratio of 6:1 in Table A3-5, along with the maximum allowable end-of-evaluation flaw sizes previously determined. The maximum allowable initial axial and circumferential flaw sizes (ai/t) are calculated to be 0.52 and 0.41, respectively. Any initial axial and circumferential flaw sizes less than 52% and 41% of the wall thickness, respectively, is encompassed by this analysis and will not grow to the maximum allowable end-of-evaluation flaw size in 60 years. The results in Table A3-5 encompass any postulated flaw (including embedded and outside surface flaws) in the CASS valve, stainless steel weld and pipe material. Table A3-5: Maximum Allowable Initial Flaw Size, End-of-Evaluation Period Flaw Size, and FCG Results for Postulated Axial and Circumferential Flaws (AR = 6) Maximum Maximum a/t at End of Year End-of-Flaw Initial Evaluation Configuration Flaw Size 10 20 30 40 50 60 Flaw Size (ai/t) (af/t) Axial Surface 0.52 0.5209 0.5222 0.5231 0.5244 0.5253 0.5266 0.53 Flaw Circumferential 0.41 0.4106 0.4114 0.4120 0.4128 0.4135 0.4143 0.42 Surface Flaw Summary and Conclusions A representative flaw tolerance evaluation was completed for a typical cast austenitic stainless steel (CASS) valve-to-pipe weld location since the weld cannot be fully volumetrically examined because of the ultrasonic examination requirements of Section XI for the CASS material. The intent of this analysis is to demonstrate that a large postulated flaw at the CASS valve-to-pipe weld region in the examination region will not grow to the maximum end-of-evaluation flaw size for the design life of the plant (60 years). The CASS valve-to-pipe weld location were evaluated per the guidelines in paragraph IWB-3640 of ASME Section XI and Appendix C [1]. Postulated inside surface axial and circumferential flaws with aspect ratio of 6:1 were evaluated at the CASS valve-to-pipe weld locations. AP1000 specific geometry, loadings, and material properties were considered in the maximum end-of-evaluation period flaws and the FCG analysis. The flaw tolerance evaluation incorporated the limiting ASME code material properties based on the base metals. The welding process is Page 63 of 64

ND-18-1185 Attachment 3 Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds required to be GTAW (gas tungsten arc welding) for the welds included in this alternative, and thus the maximum end-of-evaluation flaw sizes are based on limit load with Z = 1. As shown in Table A4-5, the maximum end-of-evaluation flaw size (af/t) is 0.53 for an axial flaw and 0.42 for a circumferential flaw with aspect ratio of 6. The FCG analysis demonstrates that it would require a very large axial flaw (greater than 52% of the wall thickness) or circumferential flaw (greater than 41% of the wall thickness) to reach the maximum end-of-evaluation flaw size in 60 years. Thus, the results from Table A3-5 can be used to demonstrate the structural integrity of the valve-to-pipe location, i.e., any flaw in the required inner one-third wall thickness examination region would not grow to the maximum end-of-evaluation flaw size per ASME Section XI in 60 years (design life). References

1. ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition with 2008 Addenda.
2. ASME Code Section II, Part D - Properties, 1998 Edition with 2000 Addenda.
3. Letter from Christopher I. Grimes, U.S. Nuclear Regulatory Commission, License Renewal and Standardization Branch, to Douglas J. Walters, Nuclear Energy Institute, License Renewal Issue No.98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components, ML003717179, May 19, 2000.
4. Evaluation of Flaws in Austenitic Steel Piping, Trans ASME, Journal of Pressure Vessel Technology, Vol. 108, Aug. 1986, pp. 352-366.
5. S. R. Mettu, I. S. Raju, "Stress Intensity Factors for Part-through Surface Cracks in Hollow Cylinders," Jointly developed under Grants NASA-JSC 25685 and Lockheed ESC 30124, Job Order number 85-130, Call number 96N72214 (NASA-TM-111707), July 1992.
6. S. Chapuliot, M. H. Lacire, and P. Le Delliou, Stress Intensity Factors for Internal Circumferential Cracks in Tubes Over a Wide Range of Radius Over Thickness Ratios, ASME PVP Vol. 365, 1998.

Page 64 of 64

Southern Nuclear Operating Company ND-18-1185 Enclosure 3 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Affidavit from Southern Nuclear Operating Company for Withholding Under 10 CFR 2.390 (VEGP 3&4-PSI-ALT-06) (This Enclosure consists of 2 pages, plus this cover page.)

ND-18-1185 Affidavit from Southern Nuclear Operating Company for Withholding Under 10 CFR 2.390 (VEGP 3&4-PSI-ALT-06) Affidavit of Brian H. Whitley

1. My name is Brian H. Whitley. I am the Regulatory Affairs Director of Southern Nuclear Operating Company (SNC). I have been delegated the function of reviewing proprietary information sought to be withheld from public disclosure and am authorized to apply for its withholding on behalf of SNC.
2. I am making this affidavit on personal knowledge, in conformance with the provisions of 10 CFR Section 2.390 of the Commissions regulations, and in conjunction with SNCs filing on dockets 52-025 and 52-026, Vogtle Electric Generating Plant Units 3 and 4, Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1)

Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds. I have personal knowledge of the criteria and procedures used by SNC to designate information as a trade secret, privileged or as confidential commercial or financial information.

3. Based on the reason(s) at 10 CFR 2.390(a)(4), this affidavit seeks to withhold from public disclosure Enclosure 1 of SNC letter ND-18-1185 for Vogtle Electric Generating Plant Units 3 and 4, Proposed Alternative VEGP 3&4-PSI/ISI-ALT-06 (Proprietary) in Accordance with 10 CFR 50.55a(z)(1) Regarding Preservice and Inservice Inspection Requirements for Specific Valve-to-Pipe Welds.
4. The following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
a. The information sought to be withheld from public disclosure has been held in confidence by SNC and Westinghouse Electric Company.
b. The information is of a type customarily held in confidence by SNC and Westinghouse Electric Company and not customarily disclosed to the public.
c. The release of the information might result in the loss of an existing or potential competitive advantage to SNC and/or Westinghouse Electric Company.
d. Other reasons identified in Enclosure 4 of SNC letter ND-18-1185 for Vogtle Electric Generating Plant Units 3 and 4 , Westinghouse Authorization Letter CAW-18-4797, Affidavit, Proprietary Information Notice and Copyright Notice (VEGP 3&4-PSl/ISl-ALT-06) .
5. Additionally, release of the information may harm SNC because SNC has a contractual relationship with the Westinghouse Electric Company regarding proprietary information.

SNC is contractually obligated to seek confidential and proprietary treatment of the information.

6. The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.
7. To the best of my knowledge and belief, the information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method.

I declare under penalty of perjury that the foregoing is true and correct. ___a_.-J_._U_._~_g _.~---------- Executed on *0/'°'/'6 Brian H. Whitley Date

Southern Nuclear Operating Company ND-18-1185 Enclosure 4 Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Westinghouse Authorization Letter CAW-18-4797, Affidavit, Proprietary Information Notice and Copyright Notice (VEGP 3&4-PSI/ISI-ALT-06) (This Enclosure consists of 11 pages, including this cover page.) September 6, 2018 SVP_SV0_005287 Westinghouse Non-Proprietary Class 3 Page 2 of 11 @Westinghouse Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4372 Document Control Desk Direct fax: (724) 940-8505 11555 Rockville Pike e-mail: monohajs@westinghouse.com Rockville, MD 20852 CAW-18-4797 September 6, 2018 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Westinghouse input for SNC Vogtle Electric Generating Plant Units 3 and 4, Request for Alternative: Preservice Inspection Requirements for Specific Valve to Pipe Welds (VEGP 3&4-PSI-ALT-06) The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC ("Westinghouse"), pursuant to the provisions of paragraph (b )( 1) of Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence. The proprietary information for which withholding is being requested in the above-referenced response is further identified in Affidavit CAW-18-4 797 signed by the owner of the proprietary information, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations. Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Southern Nuclear Operating Company. Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-18-4797, and should be addressed to Edmond J. Mercier, Manager, Fuels Licensing and Regulatory Support, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2 Suite 256, Cranberry Township, Pennsylvania 16066. 9~i~~~ Licensing Inspections and Special Programs 

                       © 2018 Westinghouse Electric Company LLC. All Rights Reserved.

September 6, 2018 SVP_SV0_005287 Page 3 of 11 Westinghouse Non-Proprietary Class 3 Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Enclosures to CAW-18-4797

1. AFFIDAVIT
2. PROPRIETARY INFORMATION NOTICE and COPYRIGHT NOTICE
                 © 2018 Westinghouse Electric Company LLC. All Rights Reserved.

September 6, 2018 SVP_SV0_005287 Page 4 of 11 ENCLOSURE 1 to CAW-18-4797 AFFIDAVIT September 6, 2018 SVP_SV0_005287 Page 5 of 11 CAW-18-4797 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: SS COUNTY OF BUTLER: I, Jill S. Monahan, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse") and declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. Executed on: 9 -{y-d () / 0 QLQQS(Ylo~ Jill S. Monahan, Manager Licensing Inspections and Special Programs September 6, 2018 SVP_SV0_005287 Page 6 of 11 3 CAW-18-4797 (1) I am Manager, Licensing Inspections and Special Programs, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse. (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commissions (Commissions) regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit. (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commissions regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse. (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of September 6, 2018 SVP_SV0_005287 Page 7 of 11 4 CAW-18-4797 Westinghouses competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability). (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product. (d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. (iii) There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position. (b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense. September 6, 2018 SVP_SV0_005287 Page 8 of 11 5 CAW-18-4797 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. (iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission. (v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in ND-18-1185, Vogtle Electric Generating Plant Units 3 and 4, Request for Alternative: Preservice Inspection Requirements for Specific Valve to Pipe Welds (VEGP 3&4-PSI-ALT-06), for submittal to the Commission, being transmitted by Southern Nuclear Operating Company letter. The proprietary information as submitted by Westinghouse is that associated with Southern Nuclear Operating Company Alternative Request Number 6, and may be used only for that purpose. (a) This information is part of that which will enable Westinghouse to (i) Manufacture and deliver products to utilities based on proprietary designs. September 6, 2018 SVP_SV0_005287 Page 9 of 11 6 CAW-18-4797 (b) Further, this information has substantial commercial value as follows: (i) Westinghouse plans to sell the use of similar information to its customers for the purpose of licensing of new nuclear power stations. (ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications. (iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money. In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended. Further the deponent sayeth not. September 6, 2018 SVP_SV0_005287 Page 10 of 11 ENCLOSURE 2 to CAW-18-4797 PROPRIETARY INFORMATION NOTICE and COPYRIGHT NOTICE September 6, 2018 SVP_SV0_005287 Page 11 of 11 PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC in connection with requests for generic and/or plant-specific review and approval. In order to conform to the requirements of 10 CFR 2.390 of the Commissions regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1). COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.}}

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