Text

10 CFR 50.55a Requests, Relief Requests I6-RR-1, I6-RR-2, and I6-RR-3 Sixth Ten-Year Inservice Inspection Program Interval
	ML23178A142
Person / Time
Site:	Point Beach
Issue date:	06/27/2023
From:	Strand D; Point Beach
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23178A141	List: L-2023-088, 10 CFR 50.55a Requests, Relief Requests I6-RR-1, I6-RR-2, and I6-RR-3 Sixth Ten-Year Inservice Inspection Program Interval;
References
	L-2023-088
	Download: ML23178A142 (1)
	v • d • e

{{#Wiki_filter:NEXTeraM ENERGV4 June 27, 2023 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 10 CFR 50.55a Requests, Relief Requests I6-RR-1, I6-RR-2, and I6-RR-3 Sixth Ten-Year lnservice Inspection Program Interval POINT BEACH 10 CFR 50.55a L-2023-088 In accordance with the provisions of 10 CFR 50.55a(z)(1 ), NextEra Energy Point Beach, LLC (NextEra) requests NRC approval of the attached relief requests for Point Beach Nuclear Plant Units 1 and 2 (Point Beach). Relief is requested from the applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code), Section XI, 2017 Edition for examination and testing requirements identified in this request. Specifically, Enclosure 1 contains Relief Request I6-RR-01 that requests relief from the requirements of ASME Section XI, IWD-5210, Test, IWD-5220, System Pressure Tests, Examination Category D-B, Item No. D2.10 and IWA-5244, Buried Components for the Emergency Diesel Generator (EOG) Class 3 Glycol Cooling (G-03 and G-04 only) and Fuel Oil Subsystems. These Subsystems are routinely inspected and tested by existing Technical Specification Surveillance Requirements that ensure the operability of the EDGs. NextEra requests approval of Relief Request I6-RR-01 prior to July 1, 2025. contains Relief Request I6-RR-02 that requests relief from the requirements of ASME Section XI, Table IWB-2500-1, Category C-B, Item No. C2.21, on the basis that the configuration of the Unit 2 feedwater nozzle extension to nozzle weld does not meet any of the configurations described in Figures IWC-2500-4(a), (b), or (d). NextEra requests approval of Relief Request I6-RR-02 prior to PB2-R44 in Spring 2030. contains Relief Request I6-RR-03 that proposes an extension to Code Case N-770-5, Table 1, Inspection Item A-2, a volumetric examination from every 5 years to an inspection period not to exceed 9 years. NextEra requests approval of Relief Request I6-RR-03 prior to PB2-R41 in Spring 2026. Enclosure 3 contains information proprietary to Westinghouse Electric Company LLC ("Westinghouse"), it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations. Enclosure 3 also provides a copy of LTR-SDA-19-071-NP, Revision 0, dated August 2019, "Point Beach Unit 2 Steam Generator Safe-End Dissimilar Metal Weld Alloy 52 Inspection Extension" (Non-Proprietary). NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

NEXTera*** ENERGY~ Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-19-4934 and should be addressed to Camille T. Zazula, Manager, Infrastructure & Facilities Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Suite 165, Cranberry Township, Pennsylvania 16066. This submittal contains no new Regulatory Commitments or revisions to existing commitments. If you have questions or require additional information, please contact Mr. Kenneth Mack, Licensing Manager at (561) 904-3635. Sincerely, Dianne Strand General Manager Regulatory Affairs Enclosures cc: Regional Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Mr. Mike Verhagan, Department of Commerce, State of Wisconsin NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

ENCLOSURE 1 Point Beach Units 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, $ystem Test Requirements

Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1) --Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected Code Class:

Component Numbers: Examination Category: Item Numbers:

== Description:== 3 NIA D-B D2.10 Emergency Diesel Generator Subsystems Glycol Cooling (G-03 and G-04 only) and Fuel Oil

2. Applicable Code Edition and Addenda

ASME Boiler and Pressure Vessel Code, Section XI, Division 1, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2017 Edition.

3. Applicable Code Requirements

The 2017 Edition of ASME Section XI references the following System Test Requirements in IWD-5200: Table IWD-2500-1 Examination Category D-B Item Number 02.1 O Parts Examined: Pressure-retaining components Test Requirement: System Leakage Test (IWD-5220) Examination Method: Visual, VT-2 Acceptance Standard: IWD-3000 The Extent of Examination: Pressure-retaining boundary Frequency of Inspection: Each inspection period IWD-5210 TEST (a) Pressure-retaining components shall be tested at the frequency stated in and visually examined by the methods specified in Table IWD-2500-1 (D-B). (b)(1) The system pressure tests and visual examinations shall be conducted in accordance with Article IWA-5000 and this Article. The contained fluid in the system shall serve as the pressurizing medium. Page 1 of 8

Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements IWD-5220 SYSTEM LEAKAGE TEST IWD-5221 Pressure (b) For Class 3 [Table IWD-2500-1 (D-B)] components in standby systems (or portions of standby systems) that are not operated routinely except for testing, the leakage test shall be conducted at the system pressure developed during a test conducted to verify system operability (e.g., to demonstrate system safety function or satisfy technical specification surveillance requirements). If portions of a system are associated with more than one safety function, the visual examination need only be performed during the test conducted at the higher test pressures for the respective system safety function. IWD-5222 Boundaries (a) The pressure - retaining boundary for closed systems includes only those portions of the system required to operate or support the safety - related function up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required. (b) The pressure-retaining boundary for nonclosed systems includes only those portions of the system required to operate or support the safety function up to and including the first normally closed valve, including a safety or relief valve or valve capable of automatic closure when the safety function is required. The test boundary shall include open-ended piping periodically pressurized to conditions described in IWD-5221. (c) Portions of systems normally submerged in its process fluid such that the external surfaces of the pressure-retaining boundary are normally wetted during its pressurized conditions are excluded from the test boundary. IWA-5244 Buried Components (a) For buried components surrounded by an annulus, the VT-2 visual examination shall examine evidence of leakage at each end of the annulus and low-point drains. (b) For buried components without an annulus, the following examination requirements shall be met: (1) A VT-2 visual examination shall be performed to identify evidence of leakage on ground surfaces in the vicinity of the buried components and in areas where leakage might be channeled or accumulated. The examination shall be performed after the component has been pressurized to system leakage test pressure for at least 24 hr. Page 2 of 8

Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements Portions of buried components where a VT-2 examination is impractical (e.g., the component is buried beneath impermeable material or encased in concrete) are exempt from VT-2 examination. (2) A test that determines the rate of pressure loss, a test that determines the change in flow between the ends of the buried components, or a test that confirms that flow during operation is not impaired shall be performed. Personnel performing these tests need not be qualified for VT-2 visual examination. (2) The Owner shall specify criteria for the examinations and tests of (1) and (2).

4. Reason for the Request Pursuant to 1 O CFR 50.55a(z)(1 ), relief is requested from the requirements of ASME Section XI, IWD-5210, Test, IWD-5220, System Pressure Tests, Examination Category D-B, Item No.

D2.1 O and IWA-5244, Buried Components for the Emergency Diesel Generator (EOG) Class 3 Glycol Cooling (G-03 and G-04 only) and Fuel Oil Subsystems. These Subsystems are routinely inspected and tested by existing Technical Specification Tests that ensure the operability of the EDGs.

5. Proposed Alternative and Basis for Use

Proposed Alternative: In lieu of performing the system pressure test on the EOG sub-systems each period, in accordance with IWD-5220 and Examination Category D-B, Item No. D2.10, PBNP proposes to use existing Technical Specification Testing of these sub-systems to demonstrate an equivalent level of quality and safety. Basis for Use: The following PBNP Technical Specification Testing that is conducted on a monthly basis will be used to verify appropriate leak tightness of the Class 3 Subsystems during the operation of the EDGs: TS 81, Emergency Diesel Generator G-01 Monthly TS 82, Emergency Diesel Generator G-02 Monthly TS 83, Emergency Diesel Generator G-03 Monthly TS 84, Emergency Diesel Generator G-04 Monthly NextEra believes the proposed alternative to this request provides acceptable quality and safety. Page 3 of 8

Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements The primary intent of Technical Specification surveillance testing differs slightly from Code required examinations. Technical Specifications are intended to demonstrate component operability, whereas the system leakage and hydrostatic tests are intended to demonstrate pressure boundary integrity. No additional examinations are imposed on the EOG Subsystems due to pressure/temperature or size exemptions as allowed by IWO-1220. Therefore, verification of pressure boundary structural integrity on EOG Subsystems is not included in the PBNP ISi Program. Successful EOG operability testing requires the associated subsystems to maintain pressure boundary integrity, providing equivalent quality and safety to ASME Section XI inspections. Those auxiliary support subsystems addressed within the scope of this request for relief include the fuel oil subsystem and the G03 and G04 glycol cooling subsystem. The repeatability of auxiliary subsystem instrumentation (pressure, level, and temperature) recorded during surveillance testing provides supporting data for indirectly verifying component integrity. Operations personnel are specifically trained in testing the standby EOGs and are aware of the necessity to maintain the pressure boundary of the auxiliary subsystems. They also know the necessity to maintain unobstructed flow characteristics for components discharging to a tank vented to the atmosphere, such as the diesel fuel oil transfer pumps. Verification of component pressure boundary integrity is administratively required of personnel performing standby EOG operability testing. If evidence of leakage is identified during the test, a condition report and/or work order is initiated, corrective actions or repairs are implemented, and follow-up confirmatory testing is performed. The following paragraphs provide specific procedural actions which support the use of alternative operability testing and inspection in lieu of ASME Section XI System Leakage Testing, VT-2 visual examination, and alternative operability testing and inspection in lieu of the requirements of IWA-5244, Buried Components. Fuel Oil Transfer Subsystem The Fuel Oil Transfer Subsystem includes components accessible for direct VT-2 examinations and buried components inaccessible for VT-2 examinations. The diesel generator building was constructed approximately 20 years ago to house two additional diesel generators for the site, G-03, and G-04. Included in this building are the two safety-related fuel oil (FO) storage tanks, which are buried beneath the structure. The Class 3 tanks have an outer leak-containment liner and leak-detection capability. The Class 3 FO piping is routed underground from the diesel generator building to the two original diesel generators, G-01 and G-02, which are housed within the control building in the central area of the turbine building. Page 4 of 8

Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements The buried FO piping consists of approximately 500 feet each of two separate 2-inch lines, which are buried in a high-density polyethylene (HOPE) trench which acts as a secondary containment system; leak detection points are provided along the length of this piping as shown in the sketch below. There is no access to these buried components other than excavation. No annulus was provided during original construction that would allow for testing or examination of these buried sections of piping in accordance with IWA-5244(a). Therefore, performing a direct VT-2 examination of this piping is impossible when performing a system leakage test. Alternative operability testing and inspection for the Fuel Oil Transfer Subsystem includes the following: For accessible components between the outlet piping from the day tank to the G-03 and G-04 engines, a VT-2 examination will be performed when the day tank is filled to design capacity, and the transfer pumps demonstrate the ability to provide adequate makeup flow to the day tank during system operation. This test is performed while the day tank is vented to the atmosphere, which is its normal configuration. During the monthly performance of TS-81, 82, 83, and 84, the inventory in the day tank is drained down to the low-level setpoint for pump actuation. The pump is verified to automatically start and allowed to replenish the day tank inventory to the high-level setpoint with verification, the pump automatically stops. During this process, procedure steps require recording the percentage of tank level when the transfer pump automatically starts and the percentage of tank level upon cessation of pump operation. The pump flow rate is recorded during the replenishment of day tank inventory for G03 and G04 with acceptance criteria applied to recorded flow rate values. For inaccessible, buried components supplying G01 and G02, the flow rate to the G01 and G02 day tanks is measured once every eight(8) years, utilizing an ultrasonic flow meter. This data indicates that pressure boundary integrity is being maintained and confirms that flow thru these lines during operation is not impaired, which satisfies IWA-5244(b)(2) requirements. However, PBNP cannot meet the IWA-5244(b)(1) requirement to perform a VT-2 examination after the component has been pressurized to system leakage test pressure for at least 24 hours due to EOG run time limitations. In lieu of this requirement, the installed leak detection systems on the buried tanks and buried piping leak containment trench will be checked weekly by operations personnel, per PC 21, Part 4, Miscellaneous Data. Based on the Technical Specification surveillance testing frequency and the data collected during these alternative tests, PBNP considers the testing performed to satisfy the Technical Specification surveillance requirements to provide an acceptable level of quality and safety as an alternative to ASME Section XI System Leak and Buried Component Testing. Page 5 of 8

Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements SOLID LIO W/SELF SEAL GASKET 4-, *o HOLES FOR BOLTING 2*_11, DIA. Ort UAR.ED 4-1/4* HILT! BDL TS ENLARGED VIEW /I RCC:D VIC:IY TO BE FIELD DETERMINED DETAIL No.1 TYPICAL SECTION OF HDPE TRENCH FOR FUEL OIL LINES Page 6 of 8 FOR TRENCHING AND Fl BACK FILL s~s SPEC 670H-l/02220 A. 02225 HYDi f) HOUSE 0 u 0 z ci 0 c:: LI

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Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements Glycol Cooling Sub-system (G03 and G04 Only) Standby emergency diesel generators G03 and G04 are provided with a glycol cooling subsystem consisting of a coolant-to-air type heat exchanger. During the monthly performance of TS-83 and TS-84, coolant tank level and multiple point temperature indication are recorded prior to starting the engine, after 30 minutes of loaded run time, and before shutting down, or hourly for extended runs. Normal values for all acquired data are provided in the procedure log sheet, and limits for the data are recorded. This data provides a positive indication that pressure boundary integrity is being maintained. Based on the monthly frequency and data collected during these tests, PBNP considers the testing performed to satisfy the Technical Specification surveillance requirements to provide an acceptable level of quality and safety as an alternative to ASME Section XI System Leak testing. Summary The subject subsystems receive these tests every 30 days, which is a much more frequent testing schedule than the 40-month system pressure testing frequency required by the Code. The EDGs are run to test their ability to start when required and to look for any problems that may have occurred while standing idle. During the testing, the EOG systems are examined for leakage. The diesel is walked down three times each day by operations personnel. During the walk down, operations personnel look at the appropriate water level, sump tank fuel level, starting air bank pressure, fuel oil day tank level, service water pressure, glycol expansion tank levels, and storage tanks. The operations personnel also take a general look at the diesel. If the readings are not within specifications, the Shift Manager is informed, and appropriate action is initiated. Additionally, the diesel is thoroughly examined as part of routine maintenance procedures. Any significant discrepancies require initiating an action request and, if appropriate, a work order to correct the identified discrepancies. Any system leakage would be identified by the parameters monitored before a significant reduction in the structural integrity of the components could occur. If evidence of leakage is identified due to surveillance testing, corrective actions or repairs would be implemented, and a follow-up confirmatory test would be performed. These examinations ensure that the components within the fuel oil transfer and glycol cooling subsystems demonstrate pressure boundary integrity and the ability to provide adequate flow for satisfactory standby EOG operation. This demonstrates that the Technical Specification Testing of the EDGs will provide an acceptable level of quality and safety for the Class 3 Subsystems as an alternative to ASME Section XI system pressure testing. Page 7 of 8

Point Beach 1 & 2 Sixth Inspection Interval Relief Request I6-RR-01 Request For Relief from the Requirements of IWD-5200, System Test Requirements

6. Duration of Proposed Alternative

The proposed alternative will be used for the Sixth 10-Year lnservice Inspection Interval of the lnservice Inspection Program for PBNP that commenced on August 1, 2022, and is scheduled to end on July 31, 2032.

7. Precedents

Point Beach Nuclear Plant Units 1 and 2 - Evaluation of Relief Request No. 11 Associated with Emergency Diesel System VT-2 Examinations for the Fourth 10-Year Interval, dated March 21, 2003, (ML030730567). Point Beach Nuclear Plant Units 1 & 2: Fourth 10-Year Interval lnservice Inspection Request for Relief No. 15, Alternate Methods for Pressure Testing of Buried Components, dated September 20, 2007 (ML072480612). Point Beach Nuclear Plant, Units 1 and 2 - Relief Request RR-8, Relief From The Requirements Of The American Society Of Mechanical Engineers Boiler And Pressure Vessel Code For Examination Of Buried Components (TAC Nos. MF4140 and MF4141). Point Beach Nuclear Plant, Units 1 and 2 - Relief Request RR-9, Relief From The Requirements Of The American Society Of Mechanical Engineers Boiler And Pressure Vessel Code System Leakage Test (TAC NOS. MF4142 AND MF4143). McGuire Nuclear Station, Units 1 and 2, Relief Request 09-GO-001, Regarding Alternatives from Pressure Test Requirements for Buried Piping, dated February 18, 2010 (ML100470359). Braidwood Station, Units 1 and 2 -Relief Request 13R-02 for Pressure Testing of Buried Components, dated June 14, 2010 (ML101590696). Page 8 of 8

ENCLOSURE 2 Point Beach Unit 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection

Point Beach 2 Sixth Inspection Interval Relief Request I6-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1) --Alternative Provides Acceptable Level of Quality and Safety-

1.

ASME Code Component(s} Affected Code Class: Component Numbers: Examination Category: Item Number(s):

== Description:== 2 SG-A-8 C-8 C2.21 Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Weld

2.

Applicable ASME Code Edition and Addenda

3.

4.

ASME Boiler and Pressure Vessel Code, Section XI, Division 1, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2017 Edition.

Applicable Code Requirement

Examination Category C-8, Item No. C2.21 requires a surface and volumetric examination in accordance with Appendix I of Section XI during each inspection interval. Code Case N-716-2, Alternative Classification, and Examination Requirements, Section XI, Division 1

Reason for Request

Point Beach is implementing the alternative requirements in Code Case N-716-2 for the sixth inservice inspection interval. This Case requires the Steam Generator to be classified as High safety significant (HSS) and examined in accordance with ASME Section XI. The Class 2 Steam Generator Feedwater Nozzle weld configuration is unique because both replacement steam generators (RSGs) would not fit through the PBNP Unit 2 equipment hatch. As result, the as-designed feedwater nozzles had to be removed and field weld fabricated after the RSGs were in place. The redesigned nozzle configuration includes a welded extension piece which precludes the ability to completely perform a UT examination at 45 and 60-degree scan angles as described in Appendix I, Supplement 9. See Attachment A for a graphic depiction of the nozzle extension weld configuration. Page 1 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection Pursuant to 10 CFR 50.55a(z)(1 ), relief is requested from the requirements of ASME Section XI, Table IWB-2500-1, Category C-B, Item No. C2.21, on the basis that the configuration of the Unit 2 feedwater nozzle extension to nozzle weld does not meet any of the configurations described in Figures IWC-2500-4(a), (b), or (d).

5.

Proposed Alternative and Basis for Use Proposed Alternative: A surface and ultrasonic examination will be performed once at each interval in accordance with Table IWC-2500-1. The ultrasonic examination will utilize 30-degree and 45-degree beam angles to interrogate the lower 1 /3 of the weld and adjacent base material1/4 inch on either side of the weld as described in Figures IWC-2500-4 (a), (b), or (d). The 30-degree angle has been chosen for two reasons: (a) the need to introduce the sound beam from the nozzle taper, and (b) the inside/ outside diameter ratio is such that a shallower beam angle (e.g., 45 degrees) will not strike the inside surface. Due to the short physical length of the nozzle boss flat, scanning with the 45-degree search unit is limited but will be performed to the maximum extent possible, as shown in Attachment A. The proposed technique has been demonstrated to the Authorized Inspection Agency on a calibration block/mock-up with two implanted weld solidification flaws 0.085 inches deep (1.99% through-wall extent). The circumferential flaw was placed in the region of the block, which would equate to the weld centerline to demonstrate the ability of the 30-degree search unit to detect a flaw while scanning on the nozzle taper. The axial flaw was placed at a convenient location on the block away from the calibration reflectors (side and end-drilled holes). The locations of these flaws are shown in Attachment B. The techniques described above do not meet the beam angle requirements of Appendix I, Supplement 9; however, they have been physically demonstrated to detect flaws via the mock-up described above effectively. In addition, the feedwater nozzle is VT-2 examined during each period during the scheduled feedwater pressure tests. Basis for Use: It is PBNP's position based on the demonstrated ability of the alternative examination angles to detect Code-allowable inside-surface connected flaws using a single angle from one direction (in the case of the 30-degree from the nozzle taper) as well as acceptable results of VT-2 visual examinations performed during the feedwater system leakage test; there is a reasonable assurance of continued structural integrity Page 2 of 8

6.

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection of the subject component and, an acceptable level of quality and safety is maintained. Duration of Proposed Alternative The proposed alternative will be used for the Sixth 10-Year lnservice Inspection Interval of the lnservice Inspection Program for the PBNP that commenced on August 1, 2022, and is scheduled to end on July 31, 2032.

7.

Precedents Point Beach Nuclear Plant, Unit No. 2 Safety Evaluation for Relief Request RR-10, "Examination of the Unit 2 Steam Generator Feedwater Nozzle Extension To Nozzle Weld Fifth 10-Year lnservice Inspection Program Interval", (TAC No. MF5012). Page 3 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection 24.59 REF',' Attachment A Unit 2 Feedwater Nozzle Extension to Nozzle Weld Configuration and Ultrasonic Examination Angles STEP 3 PARTITION OF NOZZLE _J ,'4O MAX. PARTING STOCK TYPICAL BOTH SIOES --7PARTING LINE 8.851.02 ~ Depiction of parting line on Replacement Steam Generator Page 4 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection Attachment A (Cont) Unit 2 Feedwater Nozzle Extension to Nozzle Weld Configuration and Ultrasonic Examination Angles .t. 14.06*.0J -¢16,50 REF",---1 n 9,57 REF, STEP_! MACHIH 1110 ITEM OZ AHO 03 Feedwater Nozzle Extension Weld Configuration Page 5 of 8 ~ C [ . 8 R

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection Attachment A (Cont) Unit 2 Feedwater Nozzle Extension to Nozzle Weld Configuration and Ultrasonic Examination Angles 30 Dr:gret' Scan,\\rta Depiction of Examination Angles 4 15116 in. r-3J/8in. 30° Scan Area for Circmnlerential Flaws 4 15116 in. r - 33/Bin. 30' Scan Area for Transverse Flaws Attachment A (Cont) Page 6 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection Unit 2 Feedwater Nozzle Extension to Nozzle Weld Configuration and Ultrasonic Examination Angles 415ll6in. --33/ in.-- ~in.7 '\\ '\\ '\\ I '\\ V. '\\ / I '\\ /,"" / / '\\ / / '\\ / / / '\\ / '\\ 45° Scan Area for Circumferential Flaws Page 7 of 8 / / /

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-02 Steam Generator Feedwater Nozzle Extension to Nozzle Weld Inspection Attachment B Calibration/Demonstration Block Design Showing Weld-Solidification Flaw Locations WS CRAO< .085 DEEP X . 765 LOtiO B.... 6,OQ REF I WS CRACK 1.50 I .085 OffP X.796 LONG 5.00

10.22 ------

---7.97---- ---5.73--- .oo I TYP--f- .085 TYP SECTIOI~ B-8 SCALE 2/1 CROSSHATCHING Ol,411TED FOR ClARITY Page 8 of 8 B _.. +

ENCLOSURE 3 Point Beach Unit 2 Sixth Inspection Interval Relief Request 16-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (DM) Weld Inspection Interval Contains Proprietary Information Contains Non-Proprietary Information Contains Westinghouse Affidavit

Point Beach 2 Sixth Inspection Interval Relief Request I6-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (DM) Weld Inspection Interval Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1) --Alternative Provides Acceptable Level of Quality and Safety--

1. ASME Code Component(s) Affected Class 1 Dissimilar Metal Welds (DMWs) are required to be examined in accordance with Code Case N-770-5.

Drawing Numbers: ISl-2120, ISl-2121 Table 1 Examination Code Case Category Inspection Description Item N-770-5 A-2 RC-34-MRCL-Al-05, Safe-End to "A" Inlet Nozzle N-770-5 A-2 RC-34-MRCL-8I-05, Safe-End to "B" Inlet Nozzle

2. Applicable Code Edition and Addenda

ASME Code Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2017 Edition as modified by 10CFR50.55a.

3. Applicable Code Requirements

Code Case N-770-5, "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." 1 0CFR50.55a(g)(6)(ii)(F) requires that licensees of operating pressurized water reactors as of or after June 3, 2020, implement the requirements of ASME Code Case N-770-5. Inspection Item A-2 of Code Case N-770-5 requires unmitigated butt welds at Hot Leg operating temperatures of :s; 625°F to be volumetrically examined every 5 years. Inspection Item 8-2 requires unmitigated butt welds at Cold Leg operating temperatures of~ 525°F and < 580°F to be volumetrically examined once per interval. Page 1 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (OM) Weld Inspection Interval

4. Reason for the Request The requested extension would allow coordination of the Steam Generator Hot Leg nozzle DMW volumetric examinations, required in March of 2026, with the Steam Generator Cold Leg nozzle DMW volumetric examinations in the Fall of 2030. This coordination would result in draining the reactor coolant system to low levels and opening the Steam Generator manways once instead of twice, thereby supporting nuclear, radiological, and industrial safety.

Technical justification for deferral of the Steam Generator Hot Leg nozzle DMW volumetric examinations is based on a proprietary Point Beach specific crack growth rate analysis which is provided in Attachment 1 of this submittal; the non-proprietary version of the analysis is provided in Attachment 2. The crack growth rate analysis supports the extension of the Steam Generator Hot Leg nozzle DMW inspection interval for the volumetric examination to nominally 9 years while maintaining an acceptable level of quality and safety.

5. Proposed Alternative and Basis for Use

Proposed Alternative Pursuant to 10CFR50.55a(z)(1), NextEra Energy proposes an extension to Code Case N-770-5, Table 1, Inspection Item A-2, a volumetric examination from every 5 years to an inspection period not to exceed 9 years. PBNP Unit 2 has two (2) Model D47F steam generators (SGs), which were installed as replacements in Fall 1996 (U2R22). The SGs are primarily carbon steel with the channel head and nozzles clad with austenitic stainless steel. The SG nozzle to safe-end weld (See Figures 1 A and 1 B) comprises Alloy 82/182 buttering and Alloy 82 weld material. The inside surface of the weld and adjacent base material was clad with Alloy 52 at the factory during fabrication. These welds received ASME Section Ill liquid penetrant and radiography examinations prior to installation. In addition, the ASME Section XI liquid penetrant and ultrasonic examinations pre-service examinations were performed prior to installation at PBNP. The subject welds received ASME Section XI, Appendix VIII demonstrated automated phased array ultrasonic (PA-UT) examinations as well, as ASME Section XI, Appendix IV demonstrated automated eddy current (ECT) examinations in October 2021 delivered with remote tooling from the inside surface during RFO U2R38. Neither the PA-UT nor the ECT recorded indications on either DM weld. Using ECT and PA-UT techniques ensured that neither surface-breaking nor sub-surface flaws were located within the inner 1 /3t of the weld, which could propagate through the Alloy 52 cladding material into the Alloy 82 weld material. Page 2 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (DM) Weld Inspection Interval The welds will continue to have direct bare-metal examinations performed in accordance with Code Case N-722-1 as modified by 10CFR50.55a(g)(6)(ii)(E) and are subject to VT-2 examinations during the RCS pressure test at the end of each refueling outage. NextEra believes the proposed alternative to this request provides acceptable quality and safety. Basis for Use Technical Basis The overall basis used to demonstrate the acceptability of extending the examination interval for Code Case N-770-5, Inspection Item A-2 components is contained in the site-specific weld crack growth analysis performed for PBNP Unit 2, see Attachment 1. The weld crack growth analysis demonstrates that the Point Beach Unit 2 SG primary nozzle OM welds possess adequate thickness to protect against failure due to PWSCC by performing a crack growth evaluation. In the weld crack growth analysis, a 1.5 mm inside surface flaw is postulated in the PWSCC-resistant alloy 52 inlay. The amount of time is determined for the flaw to reach the maximum allowable end-of-evaluation period flaw size. This maximum allowable end-of-evaluation period flaw size would be the largest flaw size that could exist in the OM welds and be acceptable according to the ASME Section XI Code. Crack growth was calculated based on the PWSCC growth mechanism through the Alloy 52 inlay and the Alloy 82 OM welds. The analysis results in Attachment 1 justify a longer examination interval for the SG Hot Leg nozzles than the five years currently allowed for Code Case N-770-5 Inspection Item A-2 welds. Based on the results for the SG Hot Leg nozzle OM welds in Figures 7-1 and 7-2 of the analysis (see Attachment 1 ), an examination interval of up to 8.6 EFPY is acceptable for the Point Beach Unit 2 SG inlet and outlet nozzle OM welds based on the flaw growth evaluation. Therefore, the plant-specific PWSCC growth results in Attachment 1 justify that Point Beach Unit 2 SG inlet and outlet nozzle OM welds can be examined after a duration of at least 8.6 EFPY (9 years) from the previous refueling outage inspections in October 2021, which will allow PBNP Unit 2 to perform the inspection during the Fall 2030 RFO. This technical basis demonstrates that the re-examination interval can be extended while maintaining an acceptable level of quality and safety. Weld Crack Growth Analysis Page 3 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (DM) Weld Inspection Interval An analysis has been performed for the PBNP Unit 2 Steam Generator primary nozzle to safe-end welds using several factors found in "Evaluation of the Inlay Process as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized-Water Reactors" (Reference 4 ), in particular:

1. Weld residual stress calculation

a. Assume a 50% weld repair during fabrication

b. Apply 3 weld layers and then machine to final size

c.

Use minimum thicknesses from construction records in the model

2. Flaw analysis

a. Initial flaw depth is 1.5 mm (half the inlay weld thickness)

b. Assume PWSCC growth of alloy 52 weld material with a factor of improvement of 18 over the Alloy 182 PWSCC rate from MRP-115.

c.

Calculate time to 75% through-wall for axial (c/a of 2) and circumferential flaws (c/a of 10).

d. See the image below for clarification.

Idealized Cmc~. Gl'owth Modd 1 Pip* With Inby (not to scale) a0 = 1.5 mm c0 =5 mm Model 1-OD = 872: mm, I= 68 mm: Inlay= 3 mm I n a =AJ82 c _7~s/ ~ ,ao Based on the crack growth results from Figures 7-1 and 7-2 in Attachment 1, it is demonstrated that it would take more than 8.6 EFPY (9 years) for the postulated 1.5 mm (0.06 inch) deep axial and circumferential flaw in the Point Beach Unit 2 SG inlet nozzle DM weld inlay to grow to the maximum allowable end-of-evaluation period flaw sizes with consideration of Alloy 52 with an FOi of 18 over the Alloy 182 PWSCC rate from MRP-115. It should be noted that for the SG outlet nozzle, the time duration is much longer than those provided in Figure 7-1 and 7-2 for the FOi of 18 for Alloy 52 due to the lower temperature at the outlet nozzle (T = 543°F) as compared to the inlet nozzle (T = 611.1 °F) of the steam generator. Conclusions Page 4 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (DM) Weld Inspection Interval Extending the required PBNP Unit 2 Steam Generator Primary Nozzle Hot Leg DM weld volumetric examination to an inspection period not to exceed 9 years is justified given the: (1) the Alloy 82/182 weld metal has never been exposed to a PWR primary water environment; (2) no recordable indications were identified during the ECT surface examination of the ID surface and volumetric examination of the OM welds in 2021 after more than 24 years of operation, (3) a 50% weld repair was assumed in the weld residual stress calculation, while the actual welds have no repairs (4) an improvement factor of 18 was used in the weld crack growth analysis for Alloy 52 weld metal (no Alloy 152 weld filler metal was used for the inlay), and (5) the PBNP Unit 2 Steam Generator Primary Nozzle to Safe-end Weld Crack Growth Analysis demonstrates that an inside surface flaw with a depth of 1.5 mm would not grow to the allowable flaw size specified by ASME XI rules over the timeframe of the requested inspection interval. The use of this proposed alternative will provide an acceptable level of quality and safety. For these reasons, it is requested that the NRC authorize this proposed alternative in accordance with 1 OCFR50.55a(z)(1 ). Page 5 of 8

Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (DM) Weld Inspection Interval

6. Duration of Proposed Alternative

This request is applicable to the PBNP Unit 2 6th Interval lnservice Inspection until the Fall 2030 RFO.

7. Precedents

NRC Letter to Eric McCartney dated March 22, 2016, "Point Beach Nuclear Plant, Unit 2-Approval of Relief Request 2-RR-11; Steam Generator Nozzle to Safe-End Dissimilar Metal (OM) Weld Inspection Re: (CAC No. MF6615), Adams Accession# ML16063A058. NRC Letter to Don Moul dated December 13, 2019, Point Beach Nuclear Plant, Unit 2-Approval Of Relief Request 2-RR-17 Regarding Steam Generator Primary Nozzle Dissimilar Metal Welds Inspection Interval (EPID L-2019-LLR-0084)

8. References

1. Code Case N-770-5, 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.

2. Westinghouse LTR-SDA-19-071-P, "Point Beach Unit 2 Steam Generator Safe-End Dissimilar Metal Weld Alloy 52 Inspection Extension," August 2019 (Proprietary).

3. Westinghouse L TR-SDA-19-071-NP, "Point Beach Unit 2 Steam Generator Safe-End Dissimilar Metal Weld Alloy 52 Inspection Extension," August 2019 (Non-Proprietary).

4. Evaluation of the Inlay Process as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized-Water Reactors [ML101260554].

Page 6 of 8

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Point Beach 2 Sixth Inspection Interval Relief Request 16-RR-03 Extension of the Steam Generator Primary Nozzle Dissimilar Metal (DM) Weld Inspection Interval Excerpt from Westinghouse Drawing 6147E62 Alloy 52 Cladding Detail ~, *.06 .. ~ *.00: SEE NOTE 31 GTAW, SFA - 5. 14 CL. ERNiCr f"* - ~r.~?~-il-.-.- 7 _______ J~ SMAW, SFA,9f 11 CL. -ENICrFe-7 - RI~. SUFFICIENT WELD STOCK FOR FINAL MACHINING SEE NOTES 5, 7, 8 ANO 31 VT*F SEE NOTE 27 >- RT*F ---t SEE NOTE 28 >- UT *f" ---1 SEE NOTE 30 Page 8 of 8 Westinghouse LTR-SDA-19-071-NP, Revision 0 "Point Beach Unit 2 Steam Generator Safe-End Dissimilar Metal Weld Alloy 52 Inspection Extension" (Non-Proprietary) (39 pages follow)

Westinghouse Non-Proprietaiy Class 3 LTR-SDA-19-071-NP Revision 0 Point Beach Unit 2 Steam Generator Safe-End Dissimilar Metal Weld Alloy 52 Inspection Extension Author: Verifiers: Approved: August 2019 Anees Udyawar*, RV/CV Design & Analysis Alexandria M. Carolan*, RV/CV Design & Analysis Stephen Marlette*, RV/CV Design & Analysis Lynn A. Patterson*, Manager, RV/CV Design & Analysis

Electro11ically approved records are authenticated ill the electronic document ma11ageme11t system.

- - This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 FOREWORD LTR-SDA-19-071-NP Revision 0 This document contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brackets. Coding (a,c,e) associated with the brackets sets forth the basis on which the information is considered proprieta1y. The proprieta1y information and data contained in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. This information is to be withheld from public disclosure in accordance with the Rules of Practice 10CFR2.390 and the info1mation presented herein is to be safeguarded in accordance with 1 0CFR2.390. Withholding of this information does not adversely affect the public interest. This information has been provided for your internal use only and should not be released to persons or organizations outside the Directorate of Regulation and the ACRS without the express written approval of Westinghouse Electric Company LLC. Should it become necessa1y to release this information to such persons as part of the review procedure, please contact Westinghouse Electric Company LLC, which will make the necessaiy an-angements required to protect the Company's proprietaiy interests. The proprieta1y infonnation in the brackets has been deleted in this repott. The deleted info1mation is provided in the proprietary version of this repott (LTR-SDA-19-071-P Revision 0). Page 2 of38

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Westinghouse Non-Proprietary Class 3 1.0 Introduction LTR-SDA-19-071-NP Revision 0 The Point Beach Unit 2 Steam Generators (SG) were fabricated with factory welded stainless steel safe ends attached to the SG primaiy nozzles with Alloy 82 DM (Dissimilar Metal) welds. The inside surface of the welds and adjacent base materials were cladded with Prima1y Water Stress C01rnsion Cracking (PWSCC) resistant Alloy 52 material during fabrication. In November 2012, these welds were examined using ultrasonic volumetric and eddy current surface examination methods with no indication on any of the four DM welds. Per Code Case N-770-2 Inspection Item A-2 [l], unmitigated butt welds at hot leg operation temperature, such as the SG primaiy inlet nozzle DM weld, are inspected eve1y 5 years; and also per Inspection Item B [l], unmitigated butt weld at cold leg temperatures such as the SG outlet nozzle DM weld are inspected eve1y second inspection period not exceeding 7 years. Point Beach Unit 2 received relief [2, 3, 4] from N-770-1 requirements to perform the volumetric examinations after 7.5 EFPY (Effective Full Power Years) from the previous inspection in November 2012 for the SG primary nozzle DM welds examination. Therefore, the next volumetric examination of Point Beach Unit 2 Steam Generator primaty nozzle dissimilar metal welds is planned for the March 2020 refueling outage. In this letter report, Point Beach Unit 2 is providing justification to seek relaxation beyond 7.5 EFPY in order to perform the volumetric examination of the SG DM welds in Fall 2021 refueling outage, which is 9 years (8.6 EFPY) from the previous November 2012 inspection. The technical justification to inspect after 9 years (8.6 EFPY) is based upon two separate analyses:

1. In Section 2 of this repmt, a Factor of Improvement (FOI) comparison evaluation will be perfmmed similar to that used to defer inspections of Alloy 690 reactor vessel head penetration examinations, such as in [ 6, 7]. The methodology consists of detennining the plant specific minimum Alloys 52 FOI for comparison with the laborato1y crack growth rate data presented in MRP-375 [5], and other industiy data in order to support the requested extension period of 9 years. The analysis will calculate the minimum FOI for Point Beach Unit 2 SG nozzle DM welds based on the actual operating temperature and the RIY (Reinspection Years) parameter, per ASME Code Case N-729-4 [9], for the requested examination interval. The calculated RIY for Point Beach Unit 2 is then compared with the Code Case N-729-4 interval for Alloy 600 nozzles of RIY = 2.25 in order to determine the ratio for the factor of improvement for the Point Beach Unit 2 SG DM Alloy 52 weld inlay.

Therefore, the FOI methodology consists of comparing the plant specific calculation of the minimum FOI for Alloy 52 with laboratmy data for Alloy 600 or Alloy 182. MRP-375 [5] stated "Much of the available laboratmy data indicate a factor of improvement of 100 for Alloys 690/52/152 versus Alloys 600/182 (for equivalent temperature and stress conditions) in terms of crack growth rate. Moreover, existing laborat01y and plant data demonstrate a factor of improvement in excess of 20 in terms of the time to PWSCC initiation." The Point Beach Unit 2 plant specific FOI calculation would use the actual temperature in the crack growth rate calculations, and compare it to the FOI of 20 (lesser of PWSCC crack growth rate or the PWSCC initiation) which is provided for Alloy 690/52/152 data in MRP-375 [5]. If the plant specific FOI is bounded by MRP-375, then the analysis would demonsh"ate margin to ensure the potential Page 3 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 of PWSCC is remote in the Alloy 52 weld. This methodology was also used in St. Lucie relief request for Alloy 690/52/152 reactor vessel head penetration [6, 7]. As pa1i of the NRC safety evaluation report in Reference [7], crack growth rate data for Alloy 690/52/152 from Pacific Northwest National Laborato1y (PNNL) and Argonne National Laborato1y (ANL) [8] was also considered. As a result, the summary data report from PNNL and ANL from [8] will also be considered when evaluating the FOI that is calculated for Point Beach Unit 2 SG DM welds.

2.

A separate analysis will also be performed based on a detailed Primaiy Water Stress Corrosion Crack (PWSCC) growth analysis through the Alloy 52 inlay and the Alloy 82 DM weld materials of the Steam Generator nozzle for duration of 9 years (8.6 EFPY) from November 2012. The analysis considers latest plant specific loadings, geometiy, welding residual stresses, and calculates the stress intensity factor and PWSCC growth for postulated axial and circumferential flaws in the nozzle DM weld region. The analysis will determine the minimum calculated FOI needed for the PWSCC growth in Alloy 52 material as compared to the Alloy 182 PWSCC rate based on MRP-115 [14] to demonstrate an inspection period of at least 8.6 EFPY (9 calendar years). The PWSCC growth evaluation is documented in Appendix A as a stand-alone section in this letter repmi. As pa1i of the PWSCC growth analysis, it should be noted that the PWSCC growth rate is highly dependent on the temperature at the location of the flaw, fmihermore, the crack growth rate increases as the temperature increases. Therefore, during periods when the plant is not in operation, such as refueling outages or shutdowns, the temperature at the SG nozzles is low such that crack growth due to PWSCC is insignificant. Therefore, PWSCC growth calculation should be determined for the time interval when the plant is operating at full power. The amount of time when the plant is operating at full power is determined based on previous plant operation data and the anticipated outages scheduled until the next inspections. This operation duration at full power is referred to as Effective Full Power Years (EFPY). For Point Beach Unit 2, based on operational data, the time interval between the previous inspection in November 2012 and the proposed future inspection in Fall 2021, is conservatively determined to be 8.6 EFPY (9 calendar years). For Point Beach Unit 2, this translates to a power availability factor of 95.5% to account for the time the plant is operating at full normal operating temperature. The results from the above two methodologies, along with a discussion on PWSCC initiation of Alloy 52 weld (see Section 3) and a qualitative review of service hist01y for Alloy 52 welds (Section 4), will be used to demonstrate technical justification for Point Beach Unit 2 welds, in order to support the examination extension of9 years (8.6 EFPY) from the previous refueling outage inspections in November 2012 for the SG dissimilar metal welds. References provided in the main body of the letter are located in Section 6, while references for Appendix A are located at the end of the appendix. Page 4 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 2.0 Alloy 52 Factor of Improvement Calculation Based on RIY from Code Case N-729-4 Provided herein is the calculation of the minimum factor of improvement (FOI) on crack growth for the Alloy 52 weld inlay at the Point Beach Unit 2 SG prima1y nozzle DM welds, in order to determine the acceptability of extending the volumetric inspection interval to 9 years (8.6 EFPY). The FOI is calculated by comparing the potential for PWSCC crack propagation between the Alloy 52 and Alloy 600/82/182 materials. The Alloy 52 material potential for PWSCC is calculated based on the Arrhenius equation with the major inputs consisting of the operating temperature at the SG location, and the operating period for inspection deferral (i.e. 9 years or 8.6 EFPY). The methodology for using FOI to demonstrate extension of inspection interval has been implemented numerous times for reactor vessel heads with Alloy 690 nozzles, for example at St. Lucie Unit 1 [6, 7]. The basis for the inspection extension is to address the effect of differences in operation temperature and its impact on the Reinspection Years (RIY) parameter, which is defined in ASME Code Case N-729-4 [9]. The RIY parameter adjusts the time between inspections for the effect of operating temperature using the thermal activation energy appropriate to the PWSCC growth for the material of interest. The RIY parameter used for Alloy 600 material is adjusted to the reference temperature using an activation energy (Q) of 130 kJ/mole (31 kcal/mole) [9]. Based on available laborato1y data for Alloy 690 [ see Attachment 2 of Reference 6], the same activation energy of 130 kJ/mole (31 kcal/mole) is applicable to model the temperature sensitivity of the hypothetical PWSCC flaw growth. Other industty data suggest a higher value of activation energy may be considered for the evaluation of FOI for Alloy 690/52/152. Dming the latest May 2019 lndushy/U.S. Nuclear Regulatmy Commission Materials Programs Technical Information Exchange Public Meeting, a presentation provided jointly by EPRI and NRC [ 1 O] discussed the status of the ongoing cooperative research on PWSCC initiation testing at PNNL. Based on the presentation [ 1 O], an activation energy of 185 kJ/mole ( 44 kcal/mole) was considered for PWSCC initiation times for Alloy 690/52/152 mate1ials. On the other hand, based on MRP-237 Rev. 2 [pg. K-7 of Reference 11], for an Alloy 152 weld test, the activation energy could be conservatively taken as high as 224 kJ/mole (53 kcal/mole). The particular high value of activation energy of 224 kJ/mole is similar to the activation energy assumed for PWSCC initiation in N-729-4, where the value of Q; = 209 kJ/mole (50 kcal/mole). Thus, the activation energy considered in the FOI calculation herein conservatively ranges from 130 kJ/mole (31 kcal/mole) to as high as 224 kJ/mole (53 kcal/mole) for the Alloy 52 material, in order to provide a bounding analysis. Note that based on the discussion in Section 3.1 of MRP-375, laboratory data indicates there is no significant difference in PWSCC susceptibility between Alloys 52 and 152, as was the case for Alloy 82 and 182. Hence the activation energy of Alloy 152 is considered representative of Alloy 52. To begin the analysis, the RIY parameter is calculated first. RIY quantifies the potential for crack propagation between successive volumettic/surface examinations for the Alloy 52 weld based on operating temperature and the activation energy discussed above. The RIY parameter is defined by ASME Code Case N-729-4 [9] as follows: Page 5 of38

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where: RJY R TopJ T,-eJ n nl n2 Westinghouse Non-Proprietary Class 3 n2 RIY = ~ {!J.EFP'9exp [- Qg (- 1 1 )]} ~ R Top) Tref J=n1 LTR-SDA-19-071-NP Revision 0 reinspection years, normalized to a reference temperature of 1059.67°R (588.71 °Kor 600°F) effective full power years accumulated during time period) activation energy for crack growth, for Alloy 52 weld as low as 130 kJ/mole (31 kcal/mole) and as high as 224 kJ/mole (53 kcal/mole) universal gas constant (l.103xl0-3 kcal/mol-0 R) absolute 100% power temperature during time period) ( 0 R =°F+459.67) absolute reference temperature ( 1059.67°R) number of the time periods with distinct 100% power temperature since initial operation number of the first time period with distinct 100% power temperature since time of most recent volumetric/surface inspection ( or replacement) number of the most recent time period with distinct 100% power temperature For conservatism, one interval using the highest temperature was used. The RIY expression simplifies to the following assuming a single representative operating temperature over the period between successive examinations: Conservatively assuming that the EFPY s of operation accumulated at Point Beach Unit 2 since the previous volumetric inspection (November 2012) is equal to the calendar years since inspection, the RIY for the requested extended period of 9 EFPY is considered for calculation. Based on EPU (Extended Power Uprate) program design parameters in WCAP-16983-P [12], the normal operating temperatures for the SG inlet and outlet nozzle are 611.1 °F and 543°F respectively. The calculation herein conse1vatively considers a bounding temperature of 611.1 °F for the RIY, as follows: With activation energy, Qg = 130 kl/mole (31 kcal/mole) RIY = 9 EFPY ex = 9 1.32 = 11.848 [ 31 ( 1 1 )] ( ) P 1.103x10-3 611.1+459.67 600+459.67 ( )( ) With activation energy, Qg = 224 kl/mole (53 kcal/mole) RlY = 9 EFPY ex = 9 1.60 = 14.401 [ 53 ( 1 1 )] ( ) P 1.103x10-3 611.1+459.67 600+459.67 ( ) ( ) Page 6 of38

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Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 The FOI implied by these RIY values for Point Beach SG DM Alloy 52 weld (relative to the limiting RIY for Alloy 600) is calculated as the following ratio: With activation energy, Qg = 130 kl/mole (31 kcal/mole) FOi = ---'--- =---=--= 5.26 [ RIYAlloys 52] (9)(1.32) 11.848 RIYAlloys 600 2.25 2.25 With activation energy, Qg = 224 kl/mole (53 kcal/mole) FOi = ---- = --- = -- = 6.40 [ RIYAlloys 52] (9)(1.60) 14.401 RIYAlloys 600 2.25 2.25 Thus, based on the range of activation energies for Alloy 52 PWSCC propagation, the calculated FOI ranges between 5.26 to 6.40 as compared to the potential for PWSCC for Alloy 600 material. Next, these calculated FOi values can be compared to actual laborat01y PWSCC crack growth rate data for Alloys 690/52/152 versus the crack growth rate for Alloy 600 [13] and Alloy 182 [14]. If the calculated FOI of 6.40 as dete1mined previously are below the FOi determined based on actual laborat01y measmed data then a technical justification can be demonstrated to defer the Point beach SG Alloy 52 weld after 9 years or 8.6 EFPY from the previous volumetric inspection. Alloy 182 material is the appropriate reference for defining the FOI for Alloy 52 weld material. As discussed in Section 3.1 of MRP-375, Alloy 182 weld metal is chosen as the reference for defining the FOI for Alloys 52 and 152 weld metals because Alloy 182 is more susceptible on average to PWSCC initiation and growth than Alloy 82 ( due to the higher Chromium content of Alloy 82). However, the above calculated FOi is compared to both Alloy 600 and Alloy 182 laborato1y data as discussed in MRP-375. MRP-375 investigated laborat01y PWSCC growth rate data for the purpose of assessing FOI values for growth. Data analyzed to develop a conservative factor of improvement include laboratmy specimens with substantial levels of cold work for the Alloy 690 base material. It is important to note that much of the data used to support Alloy 690 CGR (crack growth rates) was produced using materials with significant amounts of cold work, which tends to increase the CGR. Similar processing, fabrication, and welding practices apply to the original (Alloy 600) and replacement (Alloy 690) components. Figure 3-2 ofMRP-375, compares data from Alloy 690 specimens with less than 10% cold work and the statistical distribution from MRP-55 [13] describing the material variability in CGR for Alloy 600. Most of the laborat01y comparisons were bounded by a factor of improvement of 20, and all were bounded by a factor of improvement of 10. Most data support a FOI of much larger than 20. This is similar for testing of the Alloy 690 Heat Affected Zone (RAZ) as shown in Figure 3-4 of MRP-375 (relative to the distribution from MRP-55) and for the Alloy 52/152 weld metal (relative to the distribution from MRP-115 [14]) as shown in Figure 3-6 ofMRP-375. Based on the data, it is conservative to assume a FOI of 20 for PWSCC growth rates for Alloy 52/152 materials. As discussed previously, based on the plant-specific FOI calculated above for Point Beach SG DM Alloy 52 weld, a minimum FOI of 6.40 over Alloy 600 or Alloy 182 materials is necessaiy to achieve an Page 7 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 inspection interval of 9 years or 8.6 EFPY. Based on MRP-375 [5], per the available laborat01y data a factor of improvement of 100 for Alloys 690/52/152 versus Alloys 600/182 (for equivalent temperature and stress conditions) is permissible for PWSCC crack growth. Moreover, existing laborat01y and plant data [5] demonsh*ate a factor of improvement in excess of20 in terms of the time to PWSCC initiation. Another set of laborat01y data compiled by PNNL [18] and presented by Brnemmer, Olszta, and Toloczko were also considered in addition to the laborat01y POI results from MRP-375. The testing in [18] was for Alloy 52M and 152 weldments, which included a mockup inlay repair weld, a mockup overlay weld, a V-groove weld, and two na1rnw gap welds. One of the specimens was for Ringhals mock-up of Alloy 52M inlay applied onto the Alloy 82 material, which can be considered similar to the Point Beach SG DM weld configuration. Most of the laborat01y comparisons from [18] were bounded by a factor of improvement of 20, and all were bounded by a factor of improvement of 10. Per Figures 28 and 29 of [18], the data support a POI of much larger than 20 for Alloy 52M inlay material as compared to the Alloy 600 MRP-55 and Alloy 182 MRP-115 crack growth rates. The NRC in its review of St. Lucie Alloy 690 head extension [6, 7] also considered Alloy 690/152/52 crack growth date from PNNL and ANL rep01ts [8], some of these data were previously published in [ 18]. Based on a review of the PNNL/ ANL data presented in [8], the application of a POI of 6.40 for the Point Beach SG DM Alloy 52 weld to the 75th percentile curves in MRP-55 and MRP-115 bounded essentially all of the data included in the PNNL and ANL data summaiy report. This provides another basis to support justification of extending the inspection of the Alloy 52 DM welds at Point Beach Unit 2 SG to 9 years or 8.6 EFPY. The plant specific minimum POI of 6.4 for the Point Beach Unit 2 SG Alloy 52 DM weld, as calculated in this section, is bounded by various laborat01y data that demonstrates a much larger POI of approximately 20 to 100 over the Alloy 600/182 rates. Therefore, given the lack of PWSCC initiation or cracking detected (discussed later in this report) to date in any Pressurized Water Reactor (PWR) plant applications of Alloys 690/52/152, the simple POI assessment can be used as a supporting basis to extend the inspection inte1val to 9 years or 8.6 EFPY for Point Beach Unit 2 SG DM weld. A plant specific PWSCC evaluation is also perf01med in Appendix A to demonstrate that crack growth for postulated axial and circumferential flaws in the Alloy 52 weld (with a POI of 18 over the Alloy 182 PWSCC rate from MRP-115 [14]) will take longer than 8.6 EFPY (9 years) to reach the maximum allowable end-of-evaluation flaw size calculation per ASME Section XI. The calculated POI of 18 for Alloy 52 is also bounded by various laboratory data that demonstrates a POI in excess of 20 to 100 over the Alloy 600/182 rates. Therefore, the detailed fracture mechanics basis provided in Appendix A supplements the POI calculations pe1formed in this section of the repo1t. Page 8 of38

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Westinghouse Non-Proprietary Class 3 3.0 PWSCC Crack Initiation of Alloy 52 Materials LTR-SDA-19-071-NP Revision 0 Due to the excellent service histoty for Alloy 690 base material and Alloy 152/52 weld material this section provides a comparison of PWSCC initiation of the Alloy 690 base and weld materials as compared to the Alloy 600 base and weld metals based on actual laborato1y measured data. Based on the discussions provided in Section 3.2.1.3 of MRP-375 [5], laboratory tests by EDF (Electric de France) showed that Alloy 182 cracked after 95 hours, and Alloy 82 cracked after 570 hours. In comparison, Alloy 52/152 still had not cracked after >21,000 hours. This resulted in a FOi for Alloy 52/152 of 37 compared to Alloy 82 and over 150 compared to Alloy 182. Also per MRP-375, KAPL (Knolls Atomic Power Laborato1y) tested Alloy 52/152 welds for 2300 hours at 640°F (338°C) and 5300 hours at 680°F (360°C). The f01mer tests showed no indications of PWSCC, while the latter only had a few, isolated "pockets." KAPL estimated the FOi of Alloy 52/152 over Alloy 82 to be approximately 100. Tests by MHI (Mitsubishi Heavy Industries) have demonstrated specimens of both Alloys 52 and 152 that have not cracked after 107,000 hours [Section 3.2.1.3 of [5)). Futthermore, based on more recent data by MHI for the Alloy 52 and 152 welds specimen, there was no evidence of cracking after > 122,535 hours at temperature of 360°C; this translates to 71 years of operation at 325°C ( 6 l 7°F) [Table 6 of Reference [ 15)). During the latest May 2019 lndustry/U.S. Nuclear Regulatory Commission Materials Programs Technical Iriformation Exchange Public Meeting, the presentation provided jointly by EPRI and NRC [IO] discussed the status of the ongoing cooperative research on PWSCC initiation testing at PNNL. Based on the presentation [Slide 9 of Reference 10], temperature-adjusted PWSCC initiation times provided based on the latest test data show that Alloy 690/52/152 have not experience any PWSCC initiation at > 181,752 hours (20. 7 years) for an adjusted temperature of 325°C ( 6 l 7°F). When compared to the slowest PWSCC initiation time for Alloy 182 weld at 17,514 hours for an adjusted temperature of 325°C [Slide 9 of Reference 10], this represents a FOi of> lOx for PWSCC initiation of Alloy 52 welds. Therefore, the discussion provided here for PWSCC initiation supports that the Alloy 52 DM weld at Point Beach will take a significantly long time (more than 20 years up to even 71 years) before any evidence of crack initiation is detected at the SG DM locations. Provided in the next section is a qualitative review of the operation experience for cracked Alloy 600/82/182 materials as a comparison to the operation time for Point Beach Alloy 690/52/152 materials to demonstrate the highly PWSCC resistance characteristics of the later materials as compared to the fonner when operating in the same environment and temperature. Page 9 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 4.0 Operation Experience of Alloy 82/182/600 versus Alloy 52/152/690 Welds This section provides a comparison of the operating experience of components installed with highly PWSCC resistance materials of Alloy 52/152/690 in relationship to the operating experience of Alloy 82/182/600 in PWR reactor coolant system service. To date, there have been no occurrences of PWSCC cracking in Alloy 52/152/690 components in PWR environment. Information related to Alloys 690, 52, and 152 in PWRs is located in MRP-110 [16] and in MRP-111 [17]. Steam Generators Replaced steam generators with Alloy 690 tubing have been in operation since 1989 at D. C. Cook 2, Indian Point 3, and Ringhals 2. No corrosion induced flaws have been detected at these plants or any subsequent plants that have statted up since that time with Alloy 690 tubes in either replacement or original steam generators. In contrast, PWSCC was detected after one cycle of operation at several units with Alloy 600MA tubes, e.g., Doe! 3, Tihange 2 and V. C. Summer, and after the second cycle at a number of other plants with Alloy 600MA tubes. This expetience indicates that there is a service demonstrated factor of improvement of about 20 or more, with the value increasing as the PWSCC-free service life of Alloy 690 tubes continues to accumulate. Steam generator tubes are joined to the tube sheet using autogenous welds between the tube and Alloy 52/152 cladding on the primaty face of the tubesheet. Thus, each steam generator has thousands of welds and heat affected zones. There have been no reports of PWSCC initiation or cracking detected at these weld joints between Alloy 690 tubes and the cladding on the tubesheet (the cladding on early Alloy 690 steam generators was Alloy 82/182, while for later units has been Alloy 52/152). The earliest U. S. steam generator with Alloy 52 welds installed was in 1994 for V. C. Summer, which has over 25 years of experience with no cracking. While the Alloy 690 tube to tubesheet weld joints are not routinely inspected with sensitive methods, significant cracking would likely have been detected as result of leakage or visible cracks, as has occuned occasionally with Alloy 600 tube to tubesheet welds. Many steam generator tubes have also been plugged using Alloy 690TT tube plugs since the late 1980s. There have been no reports of PWSCC being detected in these plugs. The plugs have been of two main kinds: mechanical plugs with an internal expanding mandrel that expands and seals the plug envelope and tube to the tube sheet, and rolled in thimble tubes. In both cases, the plugs are made from thick wall rod material rather than from thin tubes. In contrast to the over 30 years of trouble free service with Alloy 690TT plugs, plugs made of Alloy 600MA and even Alloy 600TT experienced PWSCC within one to two years of service. Therefore, over thirty (30) years of experience for steam generators indicates that there is a service demonstrated factor of improvement of about 30 or more for Alloy 52/152/690, with the value increasing as the PWSCC-free service life of Alloy 690 tube plugs continues to accumulate. Pressurizers Alloy 690 and its weld metals were also used to repair pressurizer components in more than 17 plants since 1994. The high se1vice temperature in the pressurizer, (typically 653°F) is particularly aggressive regarding PWSCC initiation and growth. Many of these pressurizer components that have been repaired with Alloy 690 and its weld metals, have the equivalent of more than 100 EDY (Effective Degradation Years) of time-temperature exposure since installation without cracking. Inspections performed have shown no evidence of PWSCC indications. Page 10 of38

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Westinghouse Non-Proprietaiy Class 3 Reactor Vessels Heads with Alloy 690 Nozzles L TR-SDA-19-071-NP Revision 0 New and repaired reactor pressure vessel heads with Alloy 690 nozzles started to be used in the industty from 1991. From 1992 onwards, a significant number of reactor pressure vessel heads have been replaced with wrought Alloy 690 tubes and Alloy 52/152 weld metal (early replacements in France). In the USA, over 45 heads have been replaced with Alloy 52/152/690 materials and the service perfmmance has been excellent. Most of these heads have over 40 to 100 penetration nozzles made of Alloy 690 and welded with Alloy 52/152 welds. These replacement heads have operated over 25 years with no inspection findings, and most all have been inspected at least once, with no findings. In both France and the USA, detailed inspections are required at 10 year intervals for the reactor vessel heads. More recently, based on the high resistant to PWSCC of Alloy 690/52/152 in replaced reactor vessel head, the NRC will allow plants to volumetrically inspect every 20 years as part of 10 CFR Part 50 proposed rulemarking on Code Case N-729-6 [22]. Repairs with Alloy 52/152 Material Alloy 52 and 152 weld metals started to be used in repairs and in replacement components starting in the mid-1990s. There have been no reports of service induced PWSCC in these welds. Furthe1more, Alloy 52/152 welds pe1form well in prima1y water mostly because of its Chromium chemistty. Weld inlays were installed at Ringhals 3 and 4 in Sweden. For the hot leg nozzle at Ringhals 4, the inlay was applied in 2002, and then inspected with UT (Ultrasonic testing) and ECT (Eddy CmTent Testing) in 2005 with no indications. Fmthe1more, inspection based on UT and ECT in 2010 also demonstrated no indication after 11. 7 EDY. This particular location is cmTently on a IO year re-inspection frequency. Similarly, Ringhals 3 hot leg DM weld had inlay applied in 2003, and inspected with UT and ECT in 2006 with no indications. Subsequently, this location was re-inspected with UT and ECT in 2010 with no indication after 10.4 EDY, now this location is also on a 10 year re-inspection frequency. The experience of Ringhal inlays on hot leg nozzle DM welds demonstrates a good precedence for the beneficial use of Alloy 52 inlay material applied to the Alloy 82 weld. Next, a comparative discussion will be provided for cracking of hot leg nozzles Alloy 82/182 welds based on operation conditions and duration as compared to the time Point Beach Unit 2 SG DM welds have been in operation composed of Alloy 52 material. Based on cmTent operation data, the Alloy 52 inlay weld at Point Beach Unit 2 SG DM weld have been operating for 19 EFPY [19], with no cracking at hot leg temperatures of 605-611 °F since the time the replaced steam generators were put in service in 1997. In contrast, V. C. Summer discovered through-wall axial cracking at hot leg DM weld in October 2000 after being in-service since 1984 when the plant was commercially started. One of the leading factor of cracking at V. C. Summer was extensive weld repairs on the inside surface and cold work of the weld due grinding [20]. Compared to V. C. Summer, Point Beach Unit 2 SG DM welds does not have repairs and have been in operation with the Alloy 52 PWSCC resistant weld for a longer time period. Another example of cracking in Alloy 182/82 welds at hot leg temperature is that of Seabrook. In October 2009, the IO-year in-service inspection at Seabrook identified an axial indication in the Alloy 182/82 DM at one of the reactor outlet nozzles. The plant has been in operation since 1990 and was at 16.53 EFPY with hot leg operating temperatures of 621 °F when the flaw was discovered in 2009. In contt*ast, Point Beach Unit 2 SG Alloy 52 welds have been operating at hot leg temperature for over 19 EFPY with no evidence of PWSCC initiation or cracking. Page 11 of38

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Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 In 2012, No1th Anna Unit 1 discovered two through-wall axial flaws at the SG inlet nozzle Alloy 182 welds when preparing to apply full strnctural weld overlays at the location during a planned outage. Note that no leakage was observed during operation prior to the outage. Fabrication records indicated extensive ID (Inside Diameter) weld repairs were perfmmed for the SG in question which had the through-wall flaws [21]. Unlike North Anna Unit 1, Point Beach Unit 2 SG DM Alloy 52 does not have any ID repairs that could result in PWSCC initiation due to high tensile residual stresses from repair processes; thus operating experience demonstrates a low likelihood of PWSCC initiation or growth. The discussion in this section demonstrates the excellent service history of Alloy 52/152/690 materials, whether installed or used as repair, as compared to Alloy 82/ 182/600 materials. Based on service histmy, many of the Alloy 52/ l 52 welds have been in operation longer than Alloy 82/ 182 welds at hot leg conditions, with no evidence of PWSCC initiation or cracking. Fmthe1more, in contrast with other plants that have PWSCC cracking at hot leg nozzles with histo1y of ID weld repairs, the inside surface of Point Beach Unit 2 SG DM does not have any evidence of repair processes due to welding, which demonstrates a very favorable condition to negate PWSCC initiation or cracking in the Alloy 52 weld layer. Page 12 of38

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Westinghouse Non-Proprietary Class 3 5.0 Conclusion LTR-SDA-19-071-NP Revision 0 This letter report provides technical basis to justify that the examination interval of the Point Beach Unit 2 SG DM welds can be longer than 8.6 EFPY (9 years) from the previous refueling outage inspections ih November 2012. The technical justification is based on the factor of improvement calculation for plant specific operating temperatures at hot leg conditions for the Point Beach Unit 2 SG Alloy 52 weld. The calculated FOI for Point Beach Alloy 52 weld is 6.4 (per Section 2) over Alloy 600 or Alloy 182 materials. Based on a review of the PNNL/ANL laboratory data presented in [8] which has been considered in previous NRC reviews [6, 7], the calculated FOI value of 6.4 for the Point Beach SG DM Alloy 52 weld is bounded by the FOI needed for PWSCC growth for Alloy 52 welds based on a comparison of the of Alloy 600 (MRP-55) and Alloy 182 (MRP-115) curves against measured laborat01y data [8] for Alloy 52/152 material. This simplistic FOI approach provides technical support for extending the inspection of the Alloy 52 DM welds at Point Beach Unit 2 SG to 8.6 EFPY (9 years). A plant specific PWSCC growth analysis is also performed in Appendix A of this letter report for Point Beach Unit 2 SG nozzle DM weld. The crack growth analysis statted with an initial flaw depth of 1.5 mm (half of the Alloy 52 weld inlay thickness), for a postulated axial flaw with aspect ratio (flaw length/flaw depth) of 2 and a postulated circumferential flaw with aspect ratio of 10. The evaluation includes PWSCC growth of the postulated initial flaws through the Alloy 52 inlay with a factor of improvement of 18 above the MRP-115 [14] crack growth rate for Alloy 182 material. The PWSCC growth then continues for the postulated flaw through the Alloy 82 material of the Point Beach SG DM weld, till it reaches the maximum allowable end-of-evaluation flaw size. Based on the crack growth results from Figures 7-1 and 7-2 (in Appendix A of this letter rep01t), it is demonstrated that it would take more than 8.6 EFPY (9 years) for the postulated 1.5 mm (0.06 inch) deep axial and circumferential flaw in the Point Beach SG inlet nozzle DM weld inlay to grow to the maximum allowable end-of-evaluation period flaw sizes. It should be noted that for the SG outlet nozzle, the time duration is much longer than those provided in Figure 7-1 and 7-2 (Appendix A) for a FOI of 18 for Alloy 52. Operating experience and current data indicate that Alloy 690 and associated weld metal Alloy 52/152, are highly PWSCC resistant materials. These materials are extremely PWSCC initiation resistant, with very slow PWSCC growth from starter indications. PWSCC initiation data from laboratory results suppo1ts that the Alloy 52 DM weld at Point Beach will take a significantly long time (more than 20 years up to even 71 years) before any evidence of crack propagation is detected at the SG DM locations. Lastly, it should be noted that Point Beach Unit 2 SG DM welds have been in operation with the Alloy 52 weld for more than 19 EFPY, with no cracking at hot leg temperatures of 605-611 °F since the time the replaced steam generators were put in service in 1997. On the other hand, several instances of cracking were already discovered for Alloy 600/82/ l 82 materials at the beginning of life for components operating at hot leg temperatures. Hence, there is strong basis from service histo1y that components fabricated and repaired with Alloy 690/52/152 materials have not experienced any PWSCC initiation or cracking to date. Therefore, based on the quantitative FOI calculations in Section 2, along with the plant specific PWSCC growth analysis in Appendix A, and the qualitative review of service hist01y for Alloy 52 welds, there is technical justification for Point Beach Unit 2 SG DM welds to operate safely for 8.6 EFPY (9 years) from the previous refueling outage inspections in November 2012. Page 13 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6.0 References LTR-SDA-19-071-NP Revision 0 I. ASME Section XI Division l Code Case N-770-2, "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," Approval Date June 9, 2011.

2. NextEra Energy Point Beach Unit 2 Letter to NRC, "10 CFR 50.55a Request, Relief Request 2-RR-11, Unit 2 Steam Generator Nozzle to Safe-End Dissimilar Metal (DM) Weld Inspection," August 13, 2015, NRC ADAMS Accession No. ML15225A104.

3. NextEra Energy Point Beach Unit 2 Letter to NRC, "Point Beach Nuclear Plant Unit 2 - Request for Additional Information for Relief Request 2-RR MF6615," November 19, 2015, NRC ADAMS Accession No. ML15324A152.

4.

NRC letter to NextEra Energy Point Beach Unit 2, "Point Beach Nuclear Plant, Unit 2 - Approval of Relief Request 2-RR-11; Steam Generator Nozzle to Safe-End Dissimilar Metal (DM) Weld Inspection RE: (CAC No. MF6615)," March 22, 2016, NRC ADAMS Accession No. ML16063A058.

5. Materials Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375). EPRI, Palo Alto, CA: 2014.

3002002441.

6.

Florida Power and Light Relief Request L-2017-026, "St. Lucie Unit 1 Docket No. 50-335 In-Service Inspection Plans Fourth Ten-Year Interval Unit 1 Relief Request 12," Febrna1y 14, 2017, NRC ADAMS No. MLI 7045A357.

7. U.S. NRC Safety Evaluation Report, "St. Lucie Unit No. I -Relief from Requirements of the ASME Code Regarding Relief Request 12 for the Fourth 10-Year Inspection Interval (CAC NO. MF9273),"

August 31, 2018, NRC ADAMS Accession No. ML17219Al 74.

8. U. S. NRC Memorandum TO: David Alley, Office of Nuclear Reactor Regulation, FROM:

Makuteswara Srinivasna, Office of Nuclear Regulat01y Research, "Transmittal of Preliminruy Primary Water Stress c01rnsion Cracking Data for Alloys 690, 52, and 152," October 30, 2014. NRC ADAMS No. ML14322A587.

9.

ASME Section XI Division 1 Code Case N-729-4, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Section XI, Division l" (Approval Date: June 22, 2012)

10. Focht, E., Crooker, P., Toloczko, M., Zhai, Z., Jenks, A., "EPRI-NRC Cooperative Research Project:

PWSCC Crack Initiation Characterization of Alloys 600/182 and Alloys 690/52/152 - Status Update," Presented at the Indust,y I U.S. Nuclear Regulato,y Commission Materials Programs Technical Information Exchange Public Meeting, May 21-22, 2019, Rockville, MD, NRC ADAMS No. ML19134A252.

11. Materials Reliability Program: Resistance of Alloys 690, 152, and 52 to Prima1y Water Stress Conosion Cracking (MRP-237, Rev. 2): Summa1y of Findings Between 2008 and 2012 from Completed and Ongoing Test Programs. EPRI, Palo Alto, CA: April 2013. 3002000190.

12. Westinghouse Document, WCAP-16983-P, Revision 1, "Point Beach Units I and 2 Extended Power Uprate (EPU) Engineering Rep01t," October 2014 Page 14 of38

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Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0

13. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primaiy Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55) Revision l, EPRI, Palo Alto, CA: 2002. 1006695.

14. Materials Reliability Program Crack Growth Rates for Evaluating Primaiy Water Stress C01TOsion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004.

1006696.

15. Sakima, K., Maeguchi, T., et. al., "An Update on Alloys 690/52/152 PWSCC Initiation Testing," l ih International Conference on Environmental Degradation of Materials in Nuclear Power Systems -

Water Reactors, August 9-13, 2015, Ontario, Canada.

16. Materials Reliability Program, Reactor Vessel Closure Head Penetration Safety Assessment for U.S.

Pressurized Water Reactor (PWR) Plants (MRP-110): Evaluations Supporting the MRP Inspection Plan, EPRI, Palo Alto, CA: 2004. 1009807.

17. Materials Reliability Program: Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111), EPRI, Palo Alto, CA: 2004. 1009801.

18. Toloczko, M., Olszta, M., Brnemmer, S., "Stress C01TOsion Crack Growth of Alloy 52M in Simulated PWR Primary Water," 151h International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 7-11, 2011, Colorado, USA, pg 225-243.

19. Nextera Report, Docket 50-301, "Point Beach Nuclear Plant, Unit 2, Fall 2012 Unit 2 (U2R32) Steam Generator Tube Inspection Report." May 16, 2013, NRC ADAMS Accession No. ML13140A015.

20. OBI 1505 - Hairline Crack Found in Weld Connecting RCS Hot Leg Pipe to Reactor Vessel Nozzle.

Event No. 395-001007-1, Event Date: 10/07/2000.

21. License Event Report No. 50-338/2012-001-00, "North Anna Power Station, Unit 1, Degraded Reactor Coolant System Piping Due To Primary Water Stress Corrosion Cracking," NRC ADAMS Accession No. ML12151A441.

22. NRC Proposed rulemaking for ASME 2015-2017 Code Editions, 10 CFR Part 50, Federal Register, Vol. 83, No. 218, November 9, 2018.

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Westinghouse Non-Proprietmy Class 3 Appendix A LTR-SDA-19-071-NP Revision 0 Point Beach Unit 2 Steam Generator Primary Nozzle to Safe-End Weld PWSCC Growth Analysis Page 16 of38

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Westinghouse Non-Proprietaiy Class 3 LTR-SDA-19-071-NP Revision 0 Point Beach Unit 2 Steam Generator Primary Nozzle to Safe-End Weld PWSCC Growth Analysis 1.0 Introduction The potential for Ptimmy Water Stress Cmrnsion Cracking (PWSCC) of the Point Beach Unit 2 Steam Generator (SG) inlet and outlet nozzle Dissimilar Metal (DM) weld requires an appropriate assessment of the examination frequency as well as the overall examination strategy for nickel-base alloy components and weldments. The Point Beach Unit 2 SGs were fabricated with facto1y welded stainless steel safe ends attached to the SG prima1y nozzles with Alloy 82 DM welds. The inside swface of the welds and adjacent base materials were cladded with PWSCC resistant Alloy 52 material during fabrication. In December 2013, NextEra Energy submitted Relief Request 2-RR-7 to the Nuclear Regulatmy Commission (NRC) per ML13365A310 [l], as supplemented with Request for Additional Info1mation (RAI) responses per ML14206A929 [2], to request re-categorization of the primary steam generator (SG) nozzle to safe-end welds to Inspection Item G, "Uncracked Butt Weld Mitigated With an Inlay" per Code Case N-770-1 [3]. Re-categorization of the welds to Inspection Item G would allow inspection of the welds once eve1y 10-year inspection interval in lieu of once eve1y 5 years for Inspection Item A-2 (SG inlet nozzle) and eve1y second inspection period not exceeding 7 years for Inspection Item B (SG outlet nozzle). Note that ASME Section XI Code Case N-770-2 [4] is the latest NRC approved version per l OCFR50.55a. The inspection frequency guidelines in Code Case N-770-2 for Inspection Item A-2, B, and Gare the same as those provided in N-770-1. In response to the NextEra Energy Relief Request [1], the NRC requested a flaw evaluation be performed to demonstrate that the SG DM welds possess adequate thickness to protect against failure due to PWSCC. This request was completed in Westinghouse letter report LTR-PAFM-15-11-P Revision O [5] in June 2015. The crack growth evaluation in the letter repmt provided a technical basis for extending the examination interval for the steam generator dissimilar metal welds from 5 and 7 years for the steam generator inlet and outlet nozzle, respectively, to a duration of7.5 EFPY. In August 2015, NextEra Energy submitted Relief Request 2-RR-11 m ML15225Al04 [6] and ML15324Al52 [7] to the NRC which included the LTR-PAFM-15-11-P Revision O [5] PWSCC growth analysis of the SG inlet and outlet nozzle dissimilar metal welds. The Relief Request 2-RR-11 was approved by the NRC per MLl 6063A058 [8] in March 20 I 6, which allowed the SG DM welds to be inspected 7.5 EFPY from the last examination for these welds. The volumetric and eddy current surface examinations were previously performed for the Point Beach Unit 2 SG primary nozzle DM welds during the November 2012 refueling outage with no indication on any of the four DM welds. The next scheduled volumetric examination for the SG primaiy nozzle DM welds is planned for the March 2020 refueling outage which is 7.5 EFPYs past November 2012 refueling outage. Point Beach Unit 2 is seeking relaxation beyond 7.5 EFPY previously dete1mined in LTR-PAFM-15-11-P [5] to perform the volumetric inspection in Fall 2021, which is at least 9 years (8.6 EFPY) past November 2012, by performing an updated crack growth evaluation in this Appendix A of the letter repmt. The crack growth evaluation will be used to dete1mine the maximum time allowed between Page 17 of38 0

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Westinghouse Non-Proprietary Class 3 inspections for the SG DM welds. LTR-SDA-19-071-NP Revision 0 The crack growth analysis herein will start with an initial flaw depth of 1.5mm (half of the Alloy 52 weld inlay thickness), for a postulated axial flaw with aspect ratio (flaw length/flaw depth) of 2 and a postulated circumferential flaw with aspect ratio of 10. The evaluation includes PWSCC growth of the postulated initial flaws through the Alloy 52 inlay with a factor of improvement of 18 above the MRP-115 [ 11] crack growth for Alloy 182 material. The justification for the use of the factor of improvement of 18 for the Alloy 52 weld is provided in the main body of this letter report, as supported by numerous expert panel reviews and laboratmy data. Also note that once the postulated flaw is through the Alloy 52 inlay, additional PWSCC growth will be performed through the Alloy 82 DM weld thickness based on the crack growth rate in MRP-115. The flaw evaluation will be consistent with the methodology m ML101260554 [9] and the previously accepted flaw evaluation in LTR-PAFM-15-11-P [5]. Page 18 of38

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Westinghouse Non-Proprietary Class 3 2.0 Methodology LTR-SDA-19-071-NP Revision 0 In order to support the technical justification for a proposed extension to the examination intervals for the Point Beach Unit 2 SG primaty nozzle DM welds, it is necessaty to demonstrate the strnctural integrity of the SG prima1y nozzle DM welds subjected to the PWSCC growth mechanism. To demonstrate the strnctural integrity of the DM welds, it is essential to determine the operation duration for which it would take the postulated initial axial and circumferential flaws to grow to the maximum allowable end-of-evaluation period flaw size. A postulated initial flaw depth of 1.5 mm (0.06 inch) (half the inlay weld thickness) is used for the flaw evaluation, consistent with ML101260554 [9] and LTR-PAFM-15-11-P [5]. Per Relief Request 2-RR-11 in ML15225Al04 [6], the SG prima1y nozzle welds had performed an ASME Section XI, Appendix VIII qualified automated phased anay ultrasonic (PA-UT) examination as well as ASME Section XI Appendix IV automated eddy current (ECT) examination in November 2012 with remote tooling from the inside surface. Neither the PA-UT nor the ECT recorded indications on any of the four DM welds. The use of both ECT and PA-UT ensured that neither surface breaking nor subsurface flaws were located within the lower 1/3 thickness of the weld which could propagate through the Alloy 52 cladding material into the Alloy 82 weld material. Therefore, based on the nondestrnctive examination (NDE) results, the initial postulated flaw depth of 1.5 mm (0.06 inch), which is half the inlay weld thickness, is appropriate for the SG DM welds. The maximum allowable end-of-evaluation period flaw size is determined in accordance with the 2007 Edition with 2008 Addenda of the ASME Section XI Code [10]. To detennine the maximum allowable end-of-evaluation period flaw sizes and the crack tip stress intensity factors used for the PWSCC analysis, it is necessary to establish the stresses, crack geomehy, and the material prope1iies at the locations of interest. The applicable loadings which must be considered consist of piping loads acting at the DM weld regions and the welding residual stresses which exist in the region of interest. The loadings considered in the analysis included the latest piping loads, taking into consideration the replacement steam generator and the Extended Power Uprate (EPU) programs. In addition to the piping loads, the effects of welding residual stresses are also considered.. The nozzle geomehy and piping loads used in the fracture mechanics analysis are shown in Section 3.0. A discussion of the plant specific welding residual sh*ess distt*ibutions used for the DM welds is provided in Section 4.0. The determination of the maximum allowable end-of-evaluation period flaw sizes is discussed in Section 5.0. The flaw growth is dete1mined due to the PWSCC growth mechanism in the SG primaiy nozzle Alloy 82 DM weld and Alloy 52 inlay material. The PWSCC growth model for the Alloy 82 DM weld material is per MRP-115 [ 11]. For the Alloy 52 inlay, the PWSCC growth considers a factor of improvement of 18 over the MRP-115 [ 11] crack growth rate for the Alloy 182 material. This factor of improvement is used to justify perf01ming the volumetric examination of the SG DM welds in Fall 2021 refueling outage, which is 9 years (8.6 EFPY) from the previous November 2012 inspection. The PWSCC growth is calculated based on the normal operating temperature and the crack tip stress intensity factors resulting from the normal operating steady state piping loads and welding residual stresses as discussed in Section 6.0. Section 7.0 provides the flaw growth curves used in dete1mining the allowable inspection interval for the Point Beach Unit 2 SG prima1y nozzle DM welds. Page 19 of38

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Westinghouse Non-Proprietary Class 3 3.0 Nozzle Geometry and Loads Geometty, Material Properties, and Normal Operating Parameters L TR-SDA-19-071-NP Revision 0 The Point Beach Unit 2 SG inlet and outlet nozzle dissimilar metal weld geometries were based on SG drawing [12]. The aimensions are shown in Table 3-1. The limiting material properties used were based on those for the weaker stainless steel safe end material in lieu of the DM weld material. The nozzle normal operating temperatures were based on EPU program design parameters in WCAP-16983-P [13]. A n01mal operating pressure of2250 psia was used in the analysis. Table 3-1 Point Beach Unit 2 Steam Generator Primary Nozzle Geometry, Normal Operating Temperatures, and Material Properties a,c,e Piping Loads The piping loads due to pressure, deadweight, I 00% power n01mal operating thermal expansion, seismic events, and Loss of Coolant Accident (LOCA) events were considered for the analysis of the SG inlet and outlet nozzles. The Operation Basis Eatthquake (OBE) loads are assumed to be the same as the Safe Shutdown Ea1thquake (SSE) loads for conservatism. The axial force and moment components for various loadings are summarized in Table 3-2. The loadings considered in this analysis included the effects of the replacement steam generator program and the EPU program in WCAP-16983-P [13]. The loads in Table 3-2 bound both loops of the Point Beach Unit 2 reactor coolant system. Page 20 of38

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Westinghouse Non-Proprietary Class 3 Table 3-2 Point Beach Unit 2 Piping Loads Page 21 of38 L TR-SDA-19-071-NP Revision 0

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Westinghouse Non-Proprietary Class 3 4.0 Dissimilar Metal Weld Residual Stress Distribution LTR-SDA-19-071-NP Revision 0 The residual stresses used in the generation of the crack growth evaluation charts are obtained from the finite element residual stress analysis for the Point Beach Unit 2 Steam Generator dissimilar metal weld geomet1y in C-8850-00-01 [15]. The finite element analyses in C-8850-00-01 [15) were pe1formed to simulate the weld fabrication process for the nozzle safe end weld region assuming an initial 50% inside surface weld repair in accordance with the guidelines in MRP-287 [16]. A two-dimensional axisymmetric finite element model of the nozzle was used in the finite element analysis. The finite element model geometly includes a portion of the low alloy steel nozzle, the stainless steel safe end, a pmtion of the stainless steel piping, the DM weld attaching the nozzle to the safe end (along with an inlay on the inside surface), and the stainless steel weld attaching the safe end to the piping. Figure 4-1 shows a sketch of the final nozzle DM weld configuration. The following fabrication sequence was simulated in the finite element residual stress analysis: The SG nozzle is buttered with weld-deposited Alloy 82 material. The inside surface of the buttering and the nozzle is cladded with weld deposited Alloy 52 material. The nozzle and buttering are post weld heat treated at 1100°F. The nozzle is welded to the safe end ring forging with an Alloy 82 weld, with a layer of Alloy 52 on the inside surface. A repair cavity of 50% of the wall thickness is machined out of the weld region as per the guidance in MRP-287 [16). The repair cavity is filled with Alloy 82 weld metal, with a layer of Alloy 52 on the inside surface. The outside and inside diameters of the weld region are machined to final size. A shop hydrotest is performed at a pressure of 3110 psig and temperature of 300°F. The safe end is machined with the piping side weld prep. The machined safe end is welded to a long segment of stainless steel piping using a stainless steel weld. A plant hydrotest is performed at a pressure of 2485 psig and a temperature of 300°F. After the plant hydrotest, normal operating temperature and pressure was uniformly applied three times to consider any shakedown effects, after which the model was set to normal operating conditions. The resulting hoop and axial welding residual stt*esses under normal operating conditions in the DM weld region are shown in Figures 4-2 and 4-3 respectively. The nonnal operating hoop welding residual stresses have been modified slightly from C-8850-00-01 [15) as shown in Figure 4-2 and used in the crack growth analysis. This is done so that the 3rd order polynomial fit of the modified hoop stress will adequately represent the original hoop stresses at the inside surface of the inlay from C-8850-00-01 [15), as shown in Figure 4-2. The 3rd order polynomial stress fit provides a better fit for the hoop stresses than a 4th order polynomial. The axial residual stresses from C-8850-00-01 [15) have also been modified as shown in Figure 4-3, and the modified stresses are fitted to a 4th order curve fit in order to adequately represent the original axial Page 22 of38

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Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 stresses from C-8850-00-01 [ 15]. The modified axial stresses shown in Figure 4-3 are used in the crack growth analysis. The axial stresses used in the PWSCC growth analysis are based on the combination of the axial welding residual stresses and the stresses due to pressure, normal operating the1mal expansion loads, and the deadweight loads. As stated in ML101260554 [9], the weld residual stress calculation should model the inlay in three weld layers before the weld is machined to final size; however, in the C-8850-00-01 [15] residual stress evaluation, the inlay is made of two weld layers and is then machined to the final size. It is important to note that the Point Beach Unit 2 SG inlay was applied during fabrication, and not as an inservice repair. Since the inlay was applied during fabrication, the inlay is built out even with the inside surface of the cladding and safe-end and the final machining is applied to the entire inside surface of the component, including the inlay, nozzle cladding, and safe-end. For a traditional inservice repair, the inlay is built out past the inside surface of the surrounding material and only the inlay is machined, which would result in higher stresses in the inlay. Since, for the Point Beach Unit 2 SG, the inlay is machined to final size along with the surrounding material, the differences that would result in the residual sh*ess dish*ibution due to applying two weld layers as opposed to three will be insignificant based on engineering judgment. The consideration of the 50% inside surface weld repair which was conservatively modeled in the finite element analysis even though the Point Beach Unit 2 SG primary nozzle DM welds are free of repairs, provides additional basis that the residual stress evaluation is conservative. The welding residual stress used herein is consistent with that originally considered in LTR-PAFM 11-P [5] and approved by the NRC in ML16063A058 [8]. Page 23 of38

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Weld Inlay-Alloy 52 Filler Metal Channel Head Cladding - Stainless Steel Filler Metal Westinghouse Non-Proprietary Class 3 Safe End - Stainless Steel Base Metal Alloy 52 Filler Metal LTR-SDA-19-071 -NP Revision 0 Safe End to Nozzle Weld - Alloy 82 Filler Metal Nozzle Buttering - Alloy 82 Filler Metal Nozzle (Channel Head Forging) - Carbon Steel Base Metal Figure 4-1: Point Beach Unit 2 Steam Generator Dissimilar Metal Weld Configuration Page 24 of38

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-I u;* cil 8 a. ~ (/) ~ a, "O "O ~ Q. 0 ::s CXl ~ a, ~ ~ co .I:,. 6 w c.i, 0 -0 ~ -=i u;* (/) ru ct 3 (I) =- ~ (/) a, Q. Q. (I) Q. O" '< s: (I) -0

0

~ m (/) 'Zi ct 3 C: "O 0 ::s ~ 5: 9?. 5* 2-Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 a,c,e Figure 4-2: Through-Wall Hoop Residual Stress at the SG Inlet and Outlet Nozzle DM Welds Under Normal Operating Conditions Page 25 of38

-{ <ii" m () 0 0.. ~ (/) ~ e!. a, -0 -0 a ct> a. 0 a, a, ~ co -1>-o w ij, 0 lJ ~ -=i

r cn*

(/) iii (b 3 ct> ;:;.

E a,

(/) a,

a.

ct> a. CT =r ct> lJ

0

~ m ~ (b 3 C: -0 0 ~ e!. a: e!. 5* 2, Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 a,c,e Figure 4-3: Through-Wall Axial Residual Stress at the SG Inlet and Outlet Nozzle DM Welds Under Normal Operating Conditions Page 26 of38

Westinghouse Non-Proprietary Class 3 5.0 Maximum Allowable End-of-Evaluation Period Flaw Size Determination LTR-SDA-19-071-NP Revision 0 In order to develop the technical justification for a longer interval between examination of the SG primaiy nozzle DM welds, the first step is the determination of the maximum allowable end-of-evaluation period flaw sizes. The maximum allowable end-of-evaluation period flaw size is the size to which an indication can grow to until the next inspection or evaluation period. This particular flaw size is determined based on the piping loads, geometry, and the material properties of the component. The evaluation guidelines and procedures for calculating the maximum allowable end-of-evaluation period flaw sizes are described in paragraph IWB-3640 and Appendix C of the ASME Section XI Code [10]. Rapid, nonductile failure is possible for feITitic materials at low temperatures, but is not applicable to the nickel-base alloy material. In nickel-base alloy material, the higher ductility leads to two possible modes of failure, plastic collapse or unstable ductile tearing. The second mechanism can occur when the applied J integral exceeds the J1c fracture toughness, and some stable tearing occurs p1ior to failure. If this mode of failure is dominant, then the load-caITying capacity is less than that predicted by the plastic collapse mechanism. The maximum allowable end-of-evaluation period flaw sizes of paragraph IWB-3640 for the high toughness materials are determined based on the assumption that plastic collapse would be achieved and would be the dominant mode of failure. However, due to the reduced toughness of the DM welds, it is possible that crack extension and unstable ductile tearing could occur and be the dominant mode of failure. To account for this effect, penalty factors called "Z factors" were developed in ASME Code Section Xl, which are to be multiplied by the loadings at these welds. In the current analysis for Point Beach Unit 2, Z factors based on MRP-216 [17] are used in the analysis to provide a more representative approximation of the effects of the DM welds. The Z-factors for Alloy 82/182 from MRP-216 [17] have been incorporated into the latest NRC approved ASME Section Xl 2013 Edition per 10CFR50.55a. The use of Z factors in effect reduces the maximum allowable end-of-evaluation period flaw sizes for flux welds and thus has been incorporated directly into the evaluation pe1formed in accordance with the procedure and acceptance criteria given in IWB-3640 and Appendix C of ASME Code Section XI. It should be noted that the maximum allowable end-of-evaluation period flaw sizes is 75% of the wall thickness in accordance with the requirements of ASME Section XI paragraph IWB-3640 [10]. The maximum allowable end-of-evaluation period flaw sizes detennined for both axial and circumferential flaws have incorporated the relevant material properties, pipe loadings, and geometry. Loadings under normal, upset, emergency, and faulted conditions ai*e considered in conjunction with the applicable safety factors for the conesponding service conditions required in the ASME Section Xl Code. For circumferential flaws, axial stress due to the pressure, deadweight, thermal expansion, seismic, and pipe break loads are considered in the evaluation. As for the axial flaws, hoop stress resulting from pressure loading is used. The maximum allowable end-of-evaluation pe1iod flaw sizes for the axial and circumferential flaws at the SG prima1y nozzle DM welds are provided in Table 5-1. The maximum allowable end-of-evaluation period axial flaw size was calculated with an aspect ratio (flaw length/flaw depth) of 2. The aspect ratio of 2 is reasonable because the axial flaw growth due to PWSCC is limited to the width of the DM weld configuration. For the circumferential flaw, an aspect ratio of 10 is used. Page 27 of 38

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Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 It should be noted that the resulting maximum allowable end-of-evaluation period flaw sizes were limited by the ASME Code limit of 75% of the weld thickness for both flaw configurations. Table 5-1 Maximum End-of-Evaluation Period Allowable Flaw Sizes (Flaw Depth/Wall Thickness Ratio - a/t) Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10) 0.75 0.75 Page 28 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietaty Class 3 6.0 PWSCC Growth Analysis LTR-SDA-19-071-NP Revision 0 The PWSCC growth analysis involves postulating an inside flaw in the dissimilar metal weld inlay for the nozzles of interest. The objective of this analysis is to detetmine the service life required for a postulated inside surface flaw to propagate to a size that exceeds the maximum allowable end-of-evaluation period flaw depth as described in Section 5.0. An initial flaw depth of 1.5 mm (0.06 inch) into the inlay will be used in the crack growth evaluation. Note that for all postulated inside surface flaws, the governing crack growth mechanism for the SG primary nozzle DM welds is PWSCC. Crack growth due to PWSCC was calculated for both axial and circumferential flaws based on the norn1al operating condition steady-state stresses combined with the welding residual stresses. For axial flaws, the hoop stresses are due to pressure and residual stresses. For circumferential flaws, the axial stresses considered are due to pressure, thetmal expansion, deadweight, and residual stresses. The input required for the crack growth analysis is basically the information necessaty to calculate the crack tip stress intensity factor (K1), which depends on the geometty of the crack, its smrnunding strncture, and the applied stresses. The geometty and loadings for the nozzles of interest are discussed in Section 3.0 and the applicable residual stresses used are discussed in Section 4.0. Once K1 is calculated, PWSCC growth due to the applied stresses can be calculated using the crack growth rate for the Alloy 82 nickel-base alloy from MRP-115 (11]. For PWSCC through the Alloy 52 inlay thickness, a factor of improvement of 18 over the MRP-115 (11] crack growth for the Alloy 182 material is conservatively used, in order to justify perfotming the volumett*ic examination of the SG DM welds in Fall 2021 refueling outage, which is 9 years (8.6 EFPY) from the previous November 2012 inspection. Using the applicable stresses at the DM welds, the crack tip stress intensity factors can be detetmined based on the stt*ess intensity factor expressions from NASA database (18] and API-579 2007 Edition (19]. Since the hoop welding residual stt*esses are best fitted with a 3rd order polynomial, the stress intensity factor expression for the axial flaws will be based on the NASA database from (18]. The axial residual stresses were fitted with a 4th order polynomial; therefore, the circumferential stress intensity factor expressions are from API-579 (19]. A 4th order polynomial stress distribution profile is defined as: Where: a0, a 1, a2, a3, and a4 are the stress profile curve fitting coefficients; x is the distance from the wall surface where the crack initiates; tis the wall thickness; and a is the stress perpendicular to the plane of the crack. Page 29 of 38

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Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 The stress intensity factor calculations for semi-elliptical inside surface axial and circumferential flaws are expressed in the general form as follows: Where: a C t R <t:> n Gj CTj Q Crack depth Half crack length along smface Thickness of cylinder Inside radius Angular position of a point on the crack front Order of polynomial fit Gj is the influence coefficient for j1 stress distribution on crack surface (i.e., G0, Gr, G2, G3, G4) CTj is stress profile curve fitting coefficient for l stress distribution (i.e., a 0, ar, <J2, <J3, <J4) The shape factor of an elliptical crack is approximated by: Q = 1 + l.464(a/c)1-65 for a/c::: 1 or Q = 1 + l.464(c/a)1-65 for a/c > 1 Once the crack tip stress intensity factors are determined, PWSCC growth calculations can be performed using the crack growth rate discussed below with the applicable normal operating temperature. The Point Beach SG inlet and outlet nozzle to safe end dissimilar metal weld regions are primarily made of nickel based alloys (Alloy 82) with an Alloy 52 inlay on the inside surface. The Alloy 52 inlays were installed as a protective banier for the Alloy 82 weld against PWSCC. Alloy 52 weld metal is known to be more resistant to PWSCC than the lower chromium content Alloy 82. Current industty data, as discussed in the main body of this letter, suggests that the PWSCC crack growth for Alloy 52 has a factor of improvement of 100 over the PWSCC crack growth of Alloy 182 or better. However, for the evaluation contained herein, a conservative improvement factor of 18 over the Alloy 182 crack growth rate will be used to represent the crack growth rate of Alloy 52. The PWSCC growth rate for the Alloy 82/182 material based on MRP-115 [11] is: da [ Qg ( 1 1 )] a =exp -- -(K)P dt R T Tref FOi Page 30 of38

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Where: da dt Qg R T Tref a ~ K FOI Westinghouse Non-Proprietary Class 3 = Crack growth rate in m/sec (in/hr) LTR-SDA-19-071-NP Revision 0 = Thermal activation energy for crack growth = 130 kJ/mole (31.0 kcal/mole) = Universal gas constant = 8.314 x 10-3 kJ/mole-K ( 1.103 x 10-3 kcal/mole-0R) = Absolute operating temperature at the location of crack, K ( 0R) = Absolute reference temperature used to nonnalize data= 598.15 K (1076.67°R) Crack growth amplitude 1.50 X 10-12 at 325°C (2.47 X 10-7 at 6l 7°F) Exponent = 1.6 Crack tip stress intensity factor MPam (ksiin) = Improvement Factor = 2.6 for Alloy 82 (MRP-115), = 18 considered for Alloy 52 (see main body of letter repoli); The PWSCC growth rate is highly dependent on the temperature at the location of the flaw, fmihermore, the crack growth rate increases as the temperatme increases. Therefore, during periods when the plant is not in operation, such as refueling outages or shutdowns, the temperatme at the SG nozzles is low such that crack growth due to PWSCC is insignificant. Therefore, PWSCC growth calculation should be dete1mined for the time interval when the plant is operating at full power. The amount of time when the plant is operating at full power is determined based on previous plant operation data and the anticipated outages scheduled until the next inspections. This operation duration at full power is refened to as Effective Full Power Years (EFPY). For Point Beach Unit 2, based on operational data, the time interval between the previous inspection in November 2012 and the proposed future inspection in Fall 2021, is conservatively set to 8.6 EFPY (9 calendar years). For Point Beach Unit 2, this translates to a power availability factor of 95.5% to account for the time the plant is operating at 100% power normal operating temperature of 611.1 °F. Page 31 of38

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Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 7.0 Technical Justification for Deferring the Volumetric Examination Point Beach Unit 2 previously performed qualified automated phased array ultrasonic (PA-UT) and ECT inspections of these welds in November 2012 and no indications were identified during that inspection. Point Beach Unit 2 is seeking relaxation from the ASME Code Case N-770-2 [4] requirement in order to defer the volumetric examination of the SG prima1y nozzle DM welds planned during the March 2020 refueling outage to Fall 2021. Therefore, the technical basis herein is to justify acceptable PWSCC growth for an undetected postulated flaw for an operation duration of 8.6 EFPY (9 years) for the SG p1imary nozzle DM welds. A 1.5 mm (0.06") initial flaw depth is postulated for the flaw growth evaluation and an aspect ratio of 2 is used for the axial flaw and an aspect ratio of 10 for the circumferential flaw. The PWSCC growth was calculated in two stages. The first stage is growth through the Alloy 52 inlay material. As discussed in Section 6.0 for Alloy 52 an improvement factor of 18 over the crack growth rate for Alloy 182 in MRP-115 is used for growth through the inlay. The second stage is growth through the Alloy 82 DM weld based on MRP-115. The crack growth through each material is then combined to determine the length of time it would take for the postulated initial flaw to grow to the maximum allowable end-of-evaluation period flaw size. Deadweight, normal thermal expansion, and pressure (2.25 ksi) loadings along with through wall axial residual stresses were used to generate the through wall axial stresses used in the PWSCC analysis for the circumferential flaw. Since the axial welding residual stresses are compressive at locations through the wall, PWSCC analyses were perf01med with and without residual stress in order to detennine the most limiting PWSCC growth results. Only through wall n01mal operating hoop residual stresses were used in the PWSCC analysis for the axial flaw. It should be noted that no fatigue crack growth calculation was performed since crack growth due to PWSCC is the controlling crack growth mechanism. The PWSCC growth curves for an axial flaw and a circumferential flaw are shown in Figures 7-1 and 7-2 respectively for SG inlet nozzle. The horizontal axis displays service life in Effective Full Power Years (EFPY), and the vertical axis shows the flaw depth to wall thickness ratio (a/t). The maximum allowable end-of-evaluation period flaw sizes are also shown in these figures for the respective flaw configurations. The SG inlet nozzle crack growth results in te1ms of EFPY are based on a temperature of 611.1 °F. Note that the PWSCC growth results for the SG outlet nozzle (temperature= 543°F) are not reported because they are bounded by the SG inlet nozzle results as the hot leg temperature results in faster PWSCC growth through the DM weld. Moreover, the consistency between the SG outlet and inlet nozzle geometty (Table 3-1) results in the same welding residual stresses between the two nozzle DM weld regions; fmihe1more, the piping loads between the two nozzles are similar in magnitude and do not have any significant difference on the PWSCC growth. Furthermore, the maximum end-of-evaluation flaw sizes (see Table 5-

1) are the same for the SG inlet and outlet nozzle regions. As a result, the SG inlet nozzle PWSCC growth results in Figures 7-1 and 7-2 are bounding for the SG outlet nozzle DM weld region as well.

Based on the crack growth results from Figures 7-1 and 7-2, it is demonstrated that it would take more than 8.6 EFPY (9 years) for the postulated 1.5 mm (0.06 inch) deep axial and circumferential flaw in the Point Beach Unit 2 SG inlet nozzle DM weld inlay to grow to the maximum allowable end-of-evaluation period flaw sizes with consideration of Alloy 52 with a FOi of 18 over the Alloy 182 PWSCC rate from Page 32 of 38

- - This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 LTR-SDA-19-071-NP Revision 0 MRP-115. It should be noted that for the SG outlet nozzle, the time duration is much longer than those provided in Figure 7-1 and 7-2 for the FOI of 18 for Alloy 52 due to the lower temperature at the outlet nozzle (T = 543°F) as compared to the inlet nozzle (T = 611.1 °F) of the steam generator. Therefore, the plant specific PWSCC growth results, in this appendix, justifies that Point Beach Unit 2 SG inlet and outlet nozzle DM welds can be examined after a duration of at least 8.6 EFPY (9 years) from the previous refueling outage inspections in November 2012. Page 33 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

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'.7 7 0 0 2 4 6 s 10 Effective Full Power Years (EFPY) Figure 7-2 Flaw Growth Evaluation Chart for SG Inlet Nozzle (T = 611.1 °F) Circumferential Flaw (AR=lO) For Alloy 52 PWSCC growth rate, a conservative FOI of 18 over the Alloy 182 crack growth rate is used Page 35 of38 12

Westinghouse Non-Proprietary Class 3 8.0 Summary and Conclusions LTR-SDA-19-071-NP Revision 0 In August 2015, NextEra Energy submitted Relief Request 2-RR-11 in ML15225A104 [6] and ML15324Al52 [7] to the NRC which included the Westinghouse PWSCC growth analysis in LTR-PAFM-15-11-P [5] of the SG inlet and outlet nozzle dissimilar metal welds. The PWSCC growth analysis extended the examination interval from 5 and 7 years for the steam generator inlet and outlet nozzle, respectively, to 7.5 EFPY. The Relief Request 2-RR-11 was approved by the NRC per ML16063A058 [8] in March 2016. A volumetric examination and eddy current examination of the SG primaiy nozzle DM butt welds were performed in November 2012 at Point Beach Unit 2 with no indication on any of the four DM welds. The next required volumetric examination is planned during the March 2020 refueling outage, which is 7.5 EFPY past November 2012. Point Beach Unit 2 is seeking further relaxation from ASME Code Case N-770-2 [4], beyond the 7.5 EFPY to at least 8.6 EFPY (9 years) by perfotming a flaw evaluation to demonstrate that the SG DM welds possess adequate thickness to protect against failure due to PWSCC. The PWSCC growth analysis herein is consistent with the methodology in ML101260554 [9] and the previous flaw evaluation LTR-P AFM-15-11-P [5]. A 1.5 mm (0.06 inch) inside surface flaw is postulated in the inlay and the amount of time is determined for the flaw to reach the maximum allowable end-of-evaluation period flaw size. This maximum allowable end-of-evaluation period flaw size would be the largest flaw size that could exist in the DM welds and be acceptable according to the ASME Section XI Code [10]. Crack growth was calculated based on the PWSCC growth mechanism through both the Alloy 52 inlay and the Alloy 82 DM weld. The evaluation herein considered the PWSCC crack growth through the Alloy 82 DM weld based on MRP-115 [l l], while for the PWSCC growth though the Alloy 52 weld inlay, a factor of improvement of 18 over the crack growth rate for Alloy 182 based on MRP-115 [ 11] was used. The justification for the factor of improvement of 18 for the Alloy 52 weld mate1ial is provided in numerous laborato1y data as discussed in the main body of this letter report. The results in Figures 7-l and 7-2 demonstrate that it would take more than 8.6 EFPY (9 years) for the postulated 1.5 mm (0.06 inch) deep axial and circumferential flaw in the Point Beach Unit 2 SG inlet nozzle DM weld inlay to grow to the maximum allowable end-of-evaluation period flaw size. The PWSCC growth results for the SG outlet nozzle DM weld are bounded by the results in Figures 7-1 and 7-2 for the SG inlet nozzle, because the outlet nozzle is operating at a lower temperature of 543°F as compared to the inlet nozzle which is at 611.1 °F. Furthe1more, all other geometric inputs, welding residual stresses, loading, and maximum end-of-evaluation period flaw sizes are consistent between the SG inlet and outlet nozzles; thus, the results for the SG inlet nozzle DM weld conservatively bound the SG outlet nozzle. Therefore, in conclusion, the Point Beach Unit 2 plant specific analysis performed herein justifies an examination interval of 8.6 EFPY (9 years). Page 36 of 38

- - This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietaiy Class 3 9.0 Appendix A References LTR-SDA-19-071-NP Revision 0

1. NextEra Energy Point Beach, LLC Letter to NRC, dated December 27, 2013, "10 CFR 50.55a Request, Relief Request 2-RR-7 Re-categorization of Unit 2 Steam Generator Nozzle to Safe-End Welds." [ML13365A310]

2.

NextEra Energy Point Beach Unit 2 Letter to NRC, "10 CFR 50.55a Request, Relief Request 2-RR-7, Re-Categorization of Unit 2 Steam Generator Nozzle to Safe-End Welds, Response to Request for Additional Information," July 24, 2014. [ML14206A929]

3. 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 l," Approval Date December 25, 2009.

4.

ASME Code Case N-770-2, Section XI Division 1, "Alternative Examination Requirements and Acceptance Standards for Class l PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86 l 82 Weld Filler Material With or Without Application of Listed Mitigation Activities Section XI, Division l," Approval Date June 9, 2011.

5.

Westinghouse Letter LTR-PAFM-15-11-P, Revision 0, "Point Beach Unit 2 Steam Generator Primary Nozzle to Safe-end Weld Crack Growth Analysis," June 2015.

6. NextEra Energy Point Beach Unit 2 Letter to NRC, "10 CFR 50.55a Request, Relief Request 2-RR-11, Unit 2 Steam Generator Nozzle to Safe-End Dissimilar Metal (DM) Weld Inspection," August 13, 2015. [ML15225Al04]

7. NextEra Energy Point Beach Unit 2 Letter to NRC, "Point Beach Nuclear Plant Unit 2 - Request for Additional Information for Relief Request 2-RR MF6615," November 19, 2015.

[ML15324Al52]

8. NRC letter to NextEra Energy Point Beach Unit 2, "Point Beach Nuclear Plant, Unit 2 - Approval of Relief Request 2-RR-11; Steam Generator Nozzle to Safe-End Dissimilar Metal (DM) Weld Inspection RE: (CAC No. MF6615)," March 22, 2016. [ML16063A058]

9.

Rudland, David L. et al, "Evaluation of the Inlay Process as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized Water Reactors," April 2010. [ML101260554]

10. ASME Boiler & Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition with 2008 Addenda.

11. Materials Reliability Program: Crack Growth Rates for Evaluating Prima,y Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004.

1006696.

12. Westinghouse Drawing 6147E62, Revision 3, Sheets 1 through 6. "Point Beach Unit 2 Model 47F Replacement Steam Generator Channel Head Welding Machining and Assembly."

13. Westinghouse Document, WCAP-16983-P, Revision 1. "Point Beach Units 1 and 2 Extended Power Uprate (EPU) Engineering Report," October 2014.

Page 37 of38 ... This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 L TR-SDA-19-071-NP Revision 0

14. Technical Manual No. TM 1440-C370, Revision 1. "Vertical Steam Generator Instrnctions for Wisconsin Electric Power Company Point Beach Nuclear Plant - Unit 2." Westinghouse General Order No. MK-77054. June 2012.

15. Dominion Engineering, Inc., Document C-8850-00-01, Revision 0, "Welding Residual Stress Calculation for Steam Generator Nozzle DMW."

16. Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287). EPRI, Palo Alto, CA: 2010, 1021023.

17. Materials Reliability Program: Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in Pressurizer Nozzle Dissimilar Metal Welds (MRP-216, Rev. 1): Evaluations Specific to Nine Subject Plants. EPRI, Palo Alto, CA: 2007. 1015400.

18. 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.

19. American Petroleum Institute, API 579-1/ASME FFS-1 (API 579 Second Edition), "Fitness-For-Service," June 2007.

Page 38 of 38

- - This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

LTR-SDA-19-071-NP Revision 0

I his page was added lo the quality record by the PRIMb syslem upon 1ls validal1on and shall noi be considered in the page numbering of lh1s document."

Approval Information Author Approval Udyawar Anees Aug-26-2019 13:36:35 Verifier Approval Marlette Stephen Aug-26-2019 13:48:20 Verifier Approval Carolan Alexandria M Aug-26-2019 14:10:48 Manager Approval Demetri George J for Patterson Lynn Aug-26-2019 16:03:50 Files approved on Aug-26-2019

- - This record was final approved on 8/26/2019 4:03:50 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse CAW-19-4934 Application for Withholding Proprietary Information from Public Disclosure (4 pages follow)

Westinghouse Non-Proprietary Class 3 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: COUNTY OF BUTLER: CA W-19-4934 Page 1 of3 (1) I, Korey L. Hosack, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse). (2) I am requesting the proprietary portions of LTR-SDA-19-071-P be withheld from public disclosure under 10 CFR 2.390. (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 10 CFR 2.390, 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 and is not customarily disclosed to the public. (ii) 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.

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CAW-19-4934 Page 2 of3 (5) Westinghouse has policies in place to identify proprietary information. 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 Westinghouse's 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. (t) It contains patentable ideas, for which patent protection may be desirable. (6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means oflower case letters (a) through (t) 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

Westinghouse Non-Proprietary Class 3 AFFIDAVIT CA W-19-4934 Page 3 of3 refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5)(a) through (f) of this Affidavit. I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief. I declare under penalty of perjury that the foregoing is true and corre t. Executed on: ~r:fs2k Korey L. Hosack, Manager Product Line Regulatory Support

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 Commission's 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)(l). 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.}}

v t e Letter Sequence Request
EPID:L-2024-LLR-0005, 10 CFR 50.55a Requests, Relief Requests I6-RR-1, I6-RR-2, and I6-RR-3 Sixth Ten-Year Inservice Inspection Program Interval (Open)
Initiation Request, Request Acceptance Supplement Administration Questions Answers Results Approval... Other: ML24200A161
MONTHYEAR2024-02-21 [Table View]

L-2023-088, 10 CFR 50.55a Requests, Relief Requests I6-RR-1, I6-RR-2, and I6-RR-3 Sixth Ten-Year Inservice Inspection Program Interval

Text

2. Applicable Code Edition and Addenda

3. Applicable Code Requirements

5. Proposed Alternative and Basis for Use

6. Duration of Proposed Alternative

7. Precedents

Applicable Code Requirement

Reason for Request

2. Applicable Code Edition and Addenda

3. Applicable Code Requirements

5. Proposed Alternative and Basis for Use

6. Duration of Proposed Alternative

7. Precedents

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