W3F1-2021-0039, Application for Technical Specification Change to Revise Pressure/Temperature and Low Temperature Overpressure Protection for 55 Effective Full Power Years

Application for Technical Specification Change to Revise Pressure/Temperature and Low Temperature Overpressure Protection for 55 Effective Full Power Years

ML21237A544
Person / Time
Site:	Waterford
Issue date:	08/25/2021
From:	Gaston R Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2021-0039
Download: ML21237A544 (32)

Text

Entergy Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 W3F1-2021-0039 August 2, 2021 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Application for Technical Specification Change to Revise Pressure/Temperature and Low Temperature Overpressure Protection for 55 Effective Full Power Years Waterford Steam Electric Station, Unit 3 NRC Docket No. 50-382 Renewed Facility Operating License No. NPF-38 In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Entergy Operations, Inc. (Entergy) is submitting a request for an amendment to the Technical Specifications (TS) for Waterford Steam Electric Station Unit 3 (Waterford 3).

The proposed amendment would update the figures contained in TS 3.4.8.1, Pressure/Temperature Limits. The existing pressure/temperature (P/T) limit curves will be extended from 32 to 55 effective full power years (EFPY) for the beltline, extended beltline and nozzle regions. The Low Temperature Overpressure (LTOP) P/T region pressurizer pressure limit is being lowered from the current 554.1 psia to 534 psia to account for 3 reactor coolant pump operation, as described in Attachment 3.

The extension of the current curves follows the Regulatory Guide 1.99, Rev. 2 Position 2.1 ratio procedure and is described in Attachment 4.

The enclosure provides the description and assessment of the proposed changes. Attachments 1 and 2 provide the existing TS pages annotated to show the proposed changes and the clean pages. Attachment 3 provides the Westinghouse analysis, LTR-SDA-20-041, "Extension of Waterford Unit 3 Pressure-Temperature Curves to 55 Effective Full Power Years." Attachment 4 provides figures from the Westinghouse analysis, LTR-SEE-20-48, "Pressure Temperature (P/T) Limit Curve and LTOP Controls Applicability Extension for Waterford 3," from which the TS markups are derived. Note that there are currently no TS Bases changes identified due to this request.

W3F1-2021-0039 Page 2 of 2 It is conservatively estimated that Waterford 3 will reach 32 EFPY by August 4, 2022. Based on this, Entergy requests approval of the proposed license amendment by July 1, 2022 with the amendment being implemented within 30 days.

In accordance with 10 CFR 50.91(a)(1), "Notice for Public Comment," the analysis concerning the issue of no significant hazards consideration using the standards in 10 CFR 50.92 is included in Attachment 1.

In accordance with 10 CFR 50.91(b)(1), "Notice for Public Comment; State Consultation," a copy of this application is being provided to the designated Louisiana Official.

This letter contains no new regulatory commitments.

If you have any questions or require additional information, please contact Paul Wood, Regulatory Assurance Manager, at 504-464-3786.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 2, 2021.

Respectfully, Ron Gaston RWG/ajh

Enclosure:

Description and Assessment to the Proposed Change Attachments: 1. Proposed Technical Specification Changes (Mark-Up)

2. Proposed Technical Specification Changes (Clean pages)

3. Westinghouse Analysis LTR-SDA-20-041, "Extension of Waterford Unit 3 Pressure-Temperature Limit Curves to 55 Effective Full Power Years"

4. Figures from Westinghouse Analysis LTR-SEE-20-48, "Pressure Temperature (P/T) Limit Curve and Low Temperature Overpressure Protection (LTOP) Controls Applicability Extension for Waterford 3" cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector Louisiana Department of Environmental Quality NRC Project Manager

Enclosure W3F1-2021-0039 DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE

W3F1-2021-0039 Enclosure Page 1 of 11 DESCRIPTION AND ASSESSMENT OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION Entergy is requesting approval for proposed amendment to the Technical Specifications (TS),

Appendix A of Renewed Facility Operating License No. NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3).

The proposed amendment would update the figures contained in TS 3.4.8.1, Pressure/Temperature Limits. The existing pressure/temperature (P/T) limit curves will be extended from 32 to 55 effective full power years (EFPY) for the beltline, extended beltline and nozzle regions. The Low Temperature Overpressure Protection (LTOP) P/T region pressurizer pressure limit is being lowered from the current 554.1 psia to 534 psia to account for 3 reactor coolant pump operation, as described in Attachment 3.

Note that the industry uses several synonymous acronyms when referring to pressure/temperature limits. These include "P/T", "P-T", and "PT." This document uses "P/T, however many of the referenced documents use alternative versions which should be considered equivalent.

2.0 ASSESSMENT 2.1 Changes to the Technical Specifications The proposed amendment would modify the two figures in TS 3.4.8.1 to lower the pressure limit at the LTOP region from 554.1 psia to 534 psia. All other heatup and cooldown curves are unchanged.

2.2 Technical Evaluation The heatup and cooldown P/T limit curves for 32 EFPY currently contained in the Waterford Unit 3 TS were implemented in 2004 and revalidated in 2012. Based on the current operating schedule, Waterford 3 is estimated to achieve 32 EFPY around August 4, 2022. The renewed facility operating license has been issued for Waterford 3, with a renewed license term of 60 years (end of license extension, or EOLE). The 60-year EOLE of Waterford 3 corresponds to 55 EFPY. To allow operation to 55 EFPY, the P/T limit curves must be updated.

The proposed amendment would update the figures contained in TS 3.4.8.1, Pressure/Temperature Limits. The existing pressure/temperature (P/T) limit curves will be extended from 32 to 55 EFPY for the beltline, extended beltline and nozzle regions. The LTOP P/T region pressurizer pressure limit is being lowered from the current 554.1 psia to 534 psia to account for starting the third reactor coolant pump at low temperatures.

2.2.1 P/T Limit Curve Applicability Extension The P/T limit curves currently specified by TS 3.4.8.1 Figures 3.4-2 and 3.4-3 for 32 EFPY were developed in Westinghouse Report WCAP-16088-NP, Rev. 2, "Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation," (Reference 1).

These curves were approved by the Nuclear Regulatory Commission (NRC) in

W3F1-2021-0039 Enclosure Page 2 of 11 Waterford 3 License Amendment 196 (Reference 2). The curves were generated in accordance with 10 CFR 50 Appendix G (Reference 3) based on the 1995 ASME Code,Section XI, through the 1996 Addenda (Reference 4). The curves were generated using the methodology documented in WCAP-14040-NP-A, Revision 2 (Reference 5), with the exceptions that the fluence values are calculated fluence values (i.e., comply with RG 1.190 (Reference 6)) and the KIc critical stress intensities are used in place of the KIa critical stress intensities, as taken from ASME Code Case N-641 (Reference 7), and the 1996 version of Appendix G to ASME Section XI was used rather than the 1989 version.

Subsequently, Surveillance Capsule 83° was removed at 24.66 EFPY. The analysis of the materials contained in this capsule was reported in WCAP-17969-NP, Rev. 2, "Analysis of Capsule 83° from the Entergy Operations, Inc. Waterford Unit 3 Reactor Vessel Radiation Surveillance Program" (Reference 8).

The applicability of the 32 EFPY P/T limit curves based on the analysis of Capsule 83° was confirmed as part of license renewal in WCAP-18002-NP, Rev. 0, "Waterford Unit 3 Time-Limited Aging Analysis on Reactor Vessel Integrity" (Reference 9). In addition, adjusted reference temperature (ART) values were calculated for 55 EFPY, which indicate that the current 32 EFPY P/T limit curves could be extended to 55 EFPY. Note that this allows for extension of the P/T limit curves for the beltline and extended beltline materials only. This did not consider nozzle P/T limits, or other reactor coolant pressure boundary ferritic components, which are addressed in LTR-SDA-20-041, which is provided in Attachment 3, to support the LAR for revision to the P/T limit curves.

LTR-SDA-20-041 justifies the extension of the applicability of the current unadjusted P/T limit curves to 55 EFPY. In this letter, the term "unadjusted" is used to indicate P/T limit curves without modification of the pressure-temperature data points for items such as instrument uncertainties, pressure correction factors, or other similar adjustments.

The following was performed by Westinghouse to confirm the continued applicability of the P/T limit curves:

1) Comparison of the design inputs used in the development of the 32 EFPY P/T limit curves with those required for development of 55 EFPY P/T limit curves based on the latest requirements for P/T limit curves.

2) Documentation of the ART values applicable to 55 EFPY and comparison of these values to the ART values used in development of the 32 EFPY values.

3) Documentation of the lowest service temperature (LST) and flange-notch requirements for Waterford 3 and comparison of these values to those used in the Analysis of Record (AOR).

4) Demonstration of the applicability of report PWROG-15109-NP-A (Reference 10) to Waterford 3 to address Regulatory Issue Summary (RIS) 2014-11 (Reference 11).

5) Confirmation of the continued applicability of the key LTOP design basis AOR, including documentation of the continued applicability of existing design analyses through 55 EFPY.

The following is a summary of the approved evaluation:

W3F1-2021-0039 Enclosure Page 3 of 11 2.2.1.1 Fluence Determination As reported in WCAP-18002-NP, estimated Reactor Pressure Vessel (RPV) beltline neutron fluence values applicable through 55 EFPY were calculated for the Waterford 3 RPV beltline and extended beltline materials. The analysis methodologies used to calculate the Waterford Unit 3 (Reactor Vessel) RV fluences satisfy the guidance set forth in RG 1.190. In this analysis, a 3D Westinghouse-developed code, RAPTOR-M3G, is used in neutron transport calculations. The use of the RAPTOR-M3G code was approved for this use for Waterford 3 by the NRC in license amendment 252 (Reference 12). The fluence evaluations include a plant-specific and cycle-specific analysis for fuel Cycles 1 through 19 of Waterford 3. Data is given for the end of Cycle 19 for Waterford 3, at 24.66 EFPY, as well as for projections through 60 EFPY. An evaluation of the most recent dosimetry sensor set from Capsule 83°, withdrawn at the end of Cycle 19, is provided in WCAP-17969-NP (Reference 8). Capsule testing was performed in accordance with 10 CFR 50, Appendix H (Reference 5).

Sister Plant Data: The Vogtle Unit 1 surveillance weld material is weld wire Heat

1. 83653, Flux Type 0091, Lot #3536, which is exactly equivalent to the material used to fabricate the Waterford 3 lower shell longitudinal welds. The chemistry factor used to calculate the ART values for 55 EFPY is calculated in LTR-SDA-20-041 in accordance with RG 1.99 (Reference 13).

2.2.1.2 Adjusted Reference Temperatures The adjusted reference temperature values were calculated for Waterford 3 at 55 EFPY for each applicable reactor vessel material at the 1/4T and 3/4T locations in WCAP-18002-NP; however, this evaluation did not consider the sister plant data discussed above. LTR-SDA-20-041 Tables 2 and Table 3 contain the 55 EFPY ART calculations from WCAP-18002-NP for Waterford 3 at the 1/4T and 3/4T locations, respectively, with the addition of consideration of the Vogtle Unit 1 sister plant data. The calculations were performed per RG 1.99, Rev. 2. These calculations consider all beltline and extended beltline materials (all materials with a projected fluence value of 1 x 1017 n/cm2 [E > 1.0 MeV] at 55 EFPY per RIS 2014-11). LTR-SDA-20-041 Table 4 shows that the ART values at 55 EFPY remain less than or equal to the ART values used to develop the unadjusted P/T limit curves in WCAP-16088-NP. Therefore, the existing 32 EFPY Waterford 3 unadjusted P/T limit curves may be deemed applicable through 55 EFPY for the beltline and extended beltline materials.

2.2.1.3 Nozzle P/T Limits As required by RIS 2014-11, the effect of nozzles on the P/T limit curves was evaluated in LTR-SDA-20-041. As listed in WCAP-18002-NP, Table 2-10, "Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) Projections at the Circumferential Welds and Nozzle to Upper Shell Weld," the calculated maximum fast neutron fluence (E > 1.0 MeV) projection for 55 EFPY for the nozzle to upper shell weld is 3.71 x 1016 n/cm2 (E > 1.0 MeV). This value is used for the bounding fluence value for the inlet and outlet nozzles for Waterford 3. This projected fluence value is below the PWROG-15109-NP-A screening criteria of

W3F1-2021-0039 Enclosure Page 4 of 11 4.28 x 1017 n/cm2 (E > 1.0 MeV); therefore, the nozzles are not limiting compared to the beltline P/T limit curves, and plant-specific nozzle P/T limit curves are not required.

2.2.1.4 Consideration of Other Reactor Coolant Pressure Boundary Ferritic Components Lowest Service Temperature: The LST was analyzed in Section 4.4 of WCAP-16088-NP. As documented in LTR-SDA-20-041, no Class 1 piping, pumps, or valves have been replaced since the approval of WCAP-16088-NP.

As a result, no changes to the LST as determined in WCAP-16088-NP are necessary and no further consideration is necessary for the LST requirement.

The LST is 190°F (without uncertainty).

Flange Notch: The reactor vessel closure head and vessel flange were considered in the P/T limit curves developed in WCAP-16088-NP regarding both the flange notch and the minimum bolt-up temperature of 60°F. The vessel flange has not been replaced; however, per WCAP-16088-NP Table 2-1, the initial Reference Temperature for Nil Ductility Transition, RTNDT, of the replacement closure head (-44°F) is lower than the initial RTNDT of the original closure head (20°F). Since a higher initial RTNDT is conservative, the curves developed in WCAP-16088-NP, which were based on the original reactor vessel closure head, remain applicable with respect to the new closure head, and the boltup temperature remains 60°F (without uncertainty). Furthermore, the replacement closure head and vessel flange have not undergone neutron embrittlement that would affect P/T limits. Therefore, regarding P/T limits, no further consideration is necessary for these components.

Pressurizer: Per UFSAR Table 5.2-1 (Reference 14), the pressurizer was constructed to the 1971 Summer Addenda of Section III of the ASME Code and met all applicable requirements at the time of construction and is original to the plant. Furthermore, the pressurizer has not undergone neutron embrittlement that would affect P/T limits. Therefore, regarding P/T limits, no further consideration is necessary for these components.

Replacement Steam Generators: Per UFSAR Table 5.2-1, the replacement steam generators were constructed to the 1998 Edition through 2000 Addenda of Section III of the ASME Code and met all applicable requirements at the time of construction. Furthermore, the replacement steam generators have not undergone neutron embrittlement that would affect P/T limits. Therefore, regarding P/T limits, no further consideration is necessary for these components.

2.2.1.5 LTOP Limits As stated above, the limiting ART values and unadjusted beltline P/T limits are unchanged. Because of this, the LTOP enable temperature of 200°F (without uncertainty) reported in WCAP-16088-NP remains valid.

W3F1-2021-0039 Enclosure Page 5 of 11 2.2.1.6 Pressurized Thermal Shock (PTS)

As reported in WCAP-18002-NP, the RTPTS values of all of the beltline and extended beltline materials in the Waterford 3 reactor vessel are below the RTPTS screening criteria of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds (per 10 CFR 50.61), through 55 EFPY.

2.2.1.7 Upper Shelf Energy (USE)

As reported in WCAP-18002-NP, the upper-shelf energy (USE) values of all of the beltline and extended beltline materials in the Waterford 3 reactor vessel are projected to remain above the USE screening criterion of 50 ft-lb (per 10 CFR 50 Appendix G), through 55 EFPY.

2.2.1.8 P/T Limit Curve Applicability Conclusion The 1/4T and 3/4T ART values for the 32 EFPY unadjusted P/T limit curves in WCAP-16088-NP were compared to the 1/4T and 3/4T ART values calculated for 55 EFPY in WCAP-18002-NP with additional consideration of Vogtle Unit 1 sister plant data. This comparison demonstrated that the unadjusted P/T limit curves developed for 32 EFPY remain valid for 55 EFPY for the beltline and extended beltline materials. Furthermore, PWROG-15109-NP-A is determined to be applicable to Waterford 3, and thus the nozzle P/T limits will not be limiting compared to the beltline P/T limits. In addition, as documented in LTR-SDA-20-041, it was confirmed that the replacement reactor vessel closure head has been installed validating that the 32 EFPY curves are conservative regarding the flange notch considerations of 10 CFR 50 Appendix G. Finally, as documented in LTR-SDA-20-041, it was confirmed that no Class 1 ferritic components in the Reactor Coolant System (RCS) have been replaced since the development of WCAP-16088-NP; therefore, the LST requirement considered for the 32 EFPY curves remains valid.

2.2.2 LTOP Controls Applicability Extension for Revised 55 EFPY P/T Limits Based on the P/T limit curves Applicability Evaluation described above, which demonstrated that the 32 EFPY P/T limit curves remain applicable to 55 EFPY, Westinghouse performed reviews to confirm the continued applicability of the key LTOP design basis AORs. For this review, Westinghouse identified each of the LTOP AORs on record at Westinghouse that supports the current plant LTOP controls. The pertinent input data and assumptions used in the AORs were identified, and verification that all the input data used at the time of the analyses remains unchanged was performed. Where changed, a logical assessment was made to confirm that the AORs remain valid or bounding. The results of the review performed determined that the design analyses continues to be applicable for the LTOP controls through 55 EFPY.

2.2.3 Reactor Coolant Pump (RCP) Operating Limits in LTOP Enable Region It was noted that the RCP operating procedure identified the operational limitation between two versus three operating pumps is at an indicated temperature limit of 202°F, while the LTOP evaluation of record assumes that no more than 2 RCPs

W3F1-2021-0039 Enclosure Page 6 of 11 operate within the LTOP enable region, which is designated as an indicated value of 230°F. Theoretically, this combination of controls would allow 3 RCPs to operate in the enable region between 202°F and 230°F, which is outside the analysis evaluation and should be corrected. This was documented in the sites corrective action program. The corrective action plan includes the following:

1) Develop a pressure correction factor (PCF) that provides P/T limit adjustment based on the worst case (largest pressure-drop) combination of three operating RCPs.

2) Update the P/T limits from the AOR to replace the existing 2-pump PCF applied to all limits corresponding to the indicated temperature range between the designated minimum bolt-up temperature and the LTOP enable temperature values.

The change to the PCF value is not contrary to what was validated in LTR-SDA-20-41, but instead revises the applicability of the present Waterford 3 P/T limit curves to allow for 3 RCP operation versus 2 (or 4) RCP operation.

Westinghouse has developed a PCF value that is appropriate for 3 RCP operation.

The result of the evaluation is that a PCF of 91 psid (excluding instrument uncertainty) is applicable for 3 RCPs in use within the LTOP region. As applied to TS 3.4.8.1 Figures 3.4-2, "Waterford Unit 3 Heatup Curve" and 3.4-3, "Waterford Unit 3 Cooldown Curve," the actual pressurizer pressure limit is revised from 554.1 psia to 534 psia.

3.0 REGULATORY ANALYSIS

3.1 Applicable Regulatory Requirements The NRC has established requirements in 10 CFR 50 to protect the integrity of the Reactor Coolant Pressure Boundary in nuclear power plants. 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation" (Reference 15), imposes fracture toughness and material surveillance program requirements, which are set forth in 10 CFR 50, Appendices G and H. 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events" (Reference 16),

established the screening criteria for pressurized water reactor (PWR) vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of the license, termed RTPTS.

10 CFR 50 Appendix G, "Fracture Toughness Requirements" (Reference 3), requires that the P/T limits for an operating light water nuclear reactor be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, "Rules for lnservice Inspection of Nuclear Power Plant Components" to generate the P/T limits (Reference 4). 10 CFR 50 Appendix G also provides minimum temperature requirements that must be considered in the development of the P/T limit curves.

10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements" (Reference 17), provides criteria for the design and implementation of reactor pressure vessel (RPV) material surveillance programs for operating light water reactors.

W3F1-2021-0039 Enclosure Page 7 of 11 Adoption of the NRC approved methodology described in Westinghouse Report WCAP-14040-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Reference 5), for the preparation of the P/T limit curves ensures that the requirements of 10 CFR 50, Appendix G will be satisfied.

Generic Letter (GL) 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations" (Reference 18) advised licensees that the NRC staff would use Regulatory Guide (RG) 1.99, "Radiation Embrittlement of Reactor Vessel Materials" (Reference 13), to review P/T limits.

ASME Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1 (Reference 7) permits the postulation of a circumferentially oriented flaw (in lieu of an axially oriented flaw) for the evaluation of the circumferential welds in RPV P/T limit curves. Code Case N-641 also permits the use of alternate reference fracture toughness data (K1c fracture toughness curve instead of K1a fracture toughness curve) for reactor vessel materials in determining the P/T limits. These code cases were approved in RG 1.147, "lnservice Inspection Code Case Acceptability ASME Section XI Division 1," Rev. 13, June 2003 (Reference 19).

RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 6), describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence.

NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," (Reference 11) requires that embrittlement effects must be considered for all materials with a projected neutron fluence of greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV), and the nozzle forging materials must be considered regardless of the projected neutron fluence. Topical Report (TR) PWROG 15109-NP-A, Rev. 0, "PWR Pressure Vessel Nozzle Appendix G Evaluation," (Reference 10) provides a generic disposition of the effects of nozzles on P/T limit curves for the U.S. PWR fleet. Per the NRC safety evaluation for this TR, the use and referencing of this TR is only applicable to U.S. PWR inlet and outlet nozzles with a projected nozzle comer neutron fluence, as calculated by an NRC approved method of fluence evaluation consistent with the plant licensing basis, or another NRC approved method of fluence evaluation, of less than 4.28 x 1017/cm2 (E > 1 MeV).

3.2 Precedence The Waterford 3 P/T Limit Curves were previously revised via TS Amendment 196 (Reference 2),

as requested by Entergy letter W3F1-2003-0075 (Reference 20).

4.0 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Entergy has evaluated the proposed changes to the pressure-temperature limit curves for the Reactor Coolant System (RCS) and using the criteria in 10 CFR 50.92, has determined that the proposed changes do not involve a significant hazards consideration.

As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

W3F1-2021-0039 Enclosure Page 8 of 11

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes update the pressure-temperature (P/T) limits for heatup, cooldown, and inservice leak hydrostatic test limitations for the RCS to a maximum of 55 Effective Full Power Years (EFPY) in accordance with 10 CFR 50, Appendix G. This is the end of the period of extended operation for the renewed Waterford 3 operating license. The provisions of Regulatory Guide 1.99 Positions 1.1 and 2.1 were used to apply sister plant data from Plant Vogtle.

The proposed changes do not involve a significant increase in the probability of an accident previously evaluated because the changes involve no change to the plant or its modes of operation. The proposed changes do not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Completion Time are no different from those during the existing Completion Time.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes implement methods from approved Nuclear Regulatory Commission (NRC) guidance. The updated P/T limits will ensure the continued protection consistent with assuring the integrity of the reactor coolant pressure boundary. The proposed changes do not change the design, configuration, or method of operation of the plant. The proposed changes do not involve a physical alteration of the plant (no new or different kind of equipment will be installed).

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not affect the function of the reactor coolant pressure boundary or its response during plant transients. By calculating the P/T limits using NRC-approve methodology, adequate margins of safety relating to reactor coolant pressure boundary integrity are maintained. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined.

W3F1-2021-0039 Enclosure Page 9 of 11 These changes will ensure that protective actions are initiated and the operability requirements for equipment assumed to operate for accident mitigation are not affected. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.0 CONCLUSION

S In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

7.0 REFERENCES

1. Westinghouse Report, WCAP-16088-NP, Rev. 2, "Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation," June 2012

2. NRC letter to Waterford Steam Electric Station, Unit 3 - "Issuance of Amendment Re:

Pressure Temperature Limit Curves to 32 Effective Full Power Years with Power Uprate," (TAC No. MC1156). [License Amendment 196 (ML041700466), including correction letter of June 23, 2004 (ML041970318)]

3. 10 CFR 50, Appendix G, "Fracture Toughness Requirements"

4. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," Non-mandatory Appendix G

5. Westinghouse Report, WCAP-14040-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," (ML15324A233) May 2004

W3F1-2021-0039 Enclosure Page 10 of 11

6. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Revision 9, March 2001

7. ASME Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1,"

Section XI, Division 1, Approved January 17, 2000

8. Westinghouse Report, WCAP-17969-NP, Rev. 2, "Analysis of Capsule 83° from the Entergy Operations, Inc. Waterford Unit 3 Reactor Vessel Radiation Surveillance Program," (ML19073A302) November 2017

9. Westinghouse Report, WCAP-18002-NP, Rev. 0, "Waterford Unit 3 Time-Limited Aging Analysis on Reactor Vessel Integrity," July 2015

10. Pressurized Water Reactor Owners Group Topical Report PWROG-15109-NP-A, Rev. 0, "PWR Pressure Vessel Nozzle Appendix G Evaluation" (Includes NRC Safety Evaluation) (ML20024E573), January 2020

11. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components (ML14149A165)," October 14, 2014

12. NRC letter to Waterford Steam Electric Station, Unit 3- Issuance of Amendment Re:

Adoption of the RAPTOR-M3G Code for Neutron Fluence Calculations (EPID L-2017-LLA-0399) License Amendment 252 (ML18180A298), including correction letter (ML19035A520), February 6, 2019

13. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988

14. Waterford Steam Electric Station, Unit 3 (Waterford 3), "Updated Final Safety Analysis Report (UFSAR), Revision 312," Table 5.2-1, "Codes and Addenda Applied to the Reactor Coolant Pressure Boundary."

15. 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation"

16. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events"

17. 10 CFR 50, Appendix H, " Reactor Vessel Material Surveillance Program Requirements"

18. Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations"

W3F1-2021-0039 Enclosure Page 11 of 11

19. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability ASME Section XI Division 1," Rev. 13, June 2003

20. Entergy letter W3F1-2003-0075, License Amendment Request, NPF-38-250, Revision to Pressure/Temperature and Low Temperature Overpressure Protection Limits for 32 Effective Full Power Years, Waterford Steam Electric Station, Unit 3, (ML041620063),

October 22, 2003

Attachment 1 W3F1-2021-0039 Proposed Technical Specification Changes (Mark-up)

(4 Pages to Follow)

Replace with INSERT A 55 DELETE

INSERT A Replace with INSERT B 55 DELETE

INSERT B Attachment 2 W3F1-2021-0039 Proposed Technical Specification Changes (Clean Pages)

(2 Pages to Follow)

FIGURE 3.4-2 WATERFORD UNIT 3 HEATUP CURVE - 55 EFPY.

REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS (Curves do not include margins for instrument uncertainties)

WATERFORD - UNIT 3 3/4 4-30 AMENDMENT NO. 106, 160, 196

FIGURE 3.4-3 WATERFORD UNIT 3 COOLDOWN CURVE - 55 EFPY REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS (Curves do not include margins for instrument uncertainties)

WATERFORD - UNIT 3 3/4 4-31 AMENDMENT NO. 106, 160, 177, 196

Attachment 3 W3F1-2021-0039 Westinghouse Analysis LTR-SDA-20-041, "Extension of Waterford Unit 3 Pressure-Temperature Limit Curves to 55 Effective Full Power Years" (1 Pages to Follow)

Westinghouse Non-Proprietary Class 3 To: Dan Flahive Date: May 24, 2021 cc:

From: Donald M. McNutt III Your ref: N/A Ext: (412) 374-4832 Our ref: LTR-SDA-20-041, Rev. 0

Subject:

Extension of Waterford Unit 3 Pressure-Temperature Limit Curves to 55 Effective Full Power Years

References:

1. Westinghouse Report, WCAP-16088-NP, Rev. 2, Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation, June 2012.

2. Westinghouse Report, WCAP-18002-NP, Rev. 0, Waterford Unit 3 Time-Limited Aging Analysis on Reactor Vessel Integrity, July 2015.

3. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Rev. 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

4. Westinghouse Report, WCAP-17009-NP, Rev. 1, Analysis of Capsule W from the Vogtle Unit No. 1 Reactor Vessel Radiation Surveillance Program, April 2009.

5. Westinghouse Report, WCAP-17076-NP, Rev. 0, Vogtle Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, July 2009.

6. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC Presentation, Generic Letter 92-01 and RPV Integrity Assessment, Status, Schedule, and Issues, NRC/Industry Workshop on RPV Integrity Issues, February 1998. [Agencywide Documents Access and Management System (ADAMS) Accession Number ML110070570]

7. Westinghouse Report, WCAP-17969-NP, Rev. 2, Analysis of Capsule 83° from the Entergy Operations, Inc. Waterford Unit 3 Reactor Vessel Radiation Surveillance Program, November 2017.

8. Code of Federal Regulations 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

9. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, October 14, 2014. [ADAMS Accession Number ML14149A165]

10. Pressurized Water Reactor Owners Group Document, PWROG-15109-NP-A, Rev. 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020. [ADAMS Accession Number ML20024E573]

11. Entergy Design Input Record for LTR-SDA-20-023, Rev. 0, Request for Design Inputs Confirmed for PT Curve Development, June 29, 2020.

12. Code of Federal Regulations 10 CFR 50, Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Federal Register, November 29, 2019.

13. Waterford Unit 3 Final Safety Analysis Report, WSES-FSAR-Unit-3, Rev. 311.

© 2021 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 2 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 Background and Purpose The heatup and cooldown pressure-temperature (P-T) limit curves for 32 effective full power years (EFPY) currently contained in the Waterford Unit 3 Technical Specifications were developed in [1]. Since Waterford Unit 3 was approved for license renewal, the P-T limit curves must be updated in order to operate past 32 EFPY.

As a part of the license renewal time-limited aging analysis (TLAA) evaluation provided in [2], adjusted reference temperature (ART) values were calculated for 55 EFPY (considered to be end-of-license extension

[EOLE]), which indicate that the current 32 EFPY unadjusted P-T limit curves could be extended to 55 EFPY.

In this letter the term unadjusted is used to indicate P-T limit curves without modification of the pressure-temperature data points for items such as instrument uncertainties, pressure correction factors, or other similar adjustments. The purpose of this letter is to document the justification for extending the applicability of the current unadjusted P-T limit curves to 55 EFPY. Note that the methods of the analyses performed in this letter are consistent with those in WCAP-16088-NP [1].

Consideration of Sister Plant Data Previous reactor vessel integrity analyses of Waterford Unit 3 did not consider sister plant data from Vogtle Unit 1. The Vogtle Unit 1 surveillance weld material is weld wire Heat # 83653, Flux Type 0091, Lot # 3536, which is exactly equivalent to the material used to fabricate the Waterford Unit 3 lower shell longitudinal welds.

This section provides the Regulatory Guide 1.99 [3] Position 2.1 chemistry factor (CF) in order to calculate the ART values for 55 EFPY.

Sister Plant Regulatory Guide 1.99, Rev. 2 Position 2.1 CF The Position 2.1 CF for the Waterford Unit 3 lower shell longitudinal welds based on Vogtle Unit 1 surveillance weld material was calculated by starting with the data from Table D-1 of [4]. This data was then adjusted using the ratio procedure, as well as a temperature adjustment.

Sister Plant Ratio Procedure 7KHUDWLRSURFHGXUHLVXWLOL]HGWRDGMXVWWKHPHDVXUHG57NDT value when there is a difference in chemistry content between the surveillance material and the corresponding vessel weld. The Vogtle Unit 1 surveillance program weld Position 1.1 CF is 33.7°F from Table 2-3 of [5]. The Waterford Unit 3 Position 1.1 CF value for the Lower Shell Longitudinal Welds 101-142 A, B, and C is 35.0°F from Table 3-5 of [3]. The ratio procedure of Regulatory Guide 1.99, Rev. 2, Position 2.1 is applied to the surveillance data, because the Waterford Unit 3 surveillance program weld CF is not identical to the vessel weld CF.

The ratio calculation is:

CFBeltline Welds = 35.0°F CFSurveillance Weld = 33.7°F Ratio = 35.0°F / 33.7°F = 1.04 The ratio procedure results in a ratio of 1.04; therefore, this DGMXVWPHQWZLOOEHDSSOLHGWRWKH57NDT values.

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 3 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 Sister Plant Temperature Adjustments From NRC Industry Meetings on November 12, 1997 and February 12 and 13, 1998, procedural guidelines ZHUHSUHVHQWHGWRDGMXVWWKH57NDT for temperature differences when using surveillance data from one reactor vessel applied to another reactor vessel. The following is taken from the handout [6] given by the NRC at these industry meetings:

6WXGLHVKDYHVKRZQWKDWIRUWHPSHUDWXUHVQHDU)D)GHFUHDVHLQirradiation temperature ZLOOUHVXOWLQDSSUR[LPDWHO\D)LQFUHDVHLQ571'7.

Thus, for plants that use surveillance data from other reactor vessels that operate at a different temperature, this difference must be considered.

The temperature adjustment is as follows:

7HPS$GMXVWHG57NDT 57NDT, Measured + (Tcapsule - TPlant)

The time-weighted average of the inlet temperature for Waterford Unit 3 was determined to be 544.4°F using the Cycle 19 inlet temperature to represent operation through EOLE. The time-weighted average of the downcomer temperature for Vogtle Unit 1 is different for each capsule. For Capsule U the temperature is 558.8°F, for Capsule Y the temperature is 559.1°F, for Capsule V the temperature is 557.7°F, for Capsule X the temperature is 557.3°F, and for Capsule W the temperature is 557.2°F.

Table 1 Calculation of Chemistry Factor for Waterford Unit 3 Lower Shell Longitudinal Welds using Vogtle Unit 1 Surveillance Capsule Data(1)

Capsule Fluence

'RTNDT(3) FF*'RTNDT Material Capsule (x 1019 n/cm2, FF(2) FF2

(°F) (°F)

E > 1.0 MeV)

U 0.332 0.697 41.1 (25.1) 28.62 0.485 Vogtle Unit 1 Y 1.14 1.037 27.1 (11.4) 28.14 1.075 Surveillance Weld V 1.93 1.180 13.8 (0.0) 16.32 1.392 (Heat #83653) X 3.47 1.325 71.0 (55.4) 94.09 1.755 W 4.36 1.375 49.5 (34.8) 68.05 1.889 SUM: 235.21 6.596 CF Heat # 83653  )) 57NDT * ))2) = (235.21) ÷ (6.596) = 35.7°F Notes:

1. All data taken from Table D-1 of [4], unless otherwise noted.

2. FF = fluence factor = f(0.28-0.10log(f)), where f = fluence x 1019 n/cm2 per [3].

3. The surveillance weld metal RTNDT values have been adjusted by the difference in temperatures between the Waterford Unit 3 RV and the Vogtle surveillance capsules, as previously discussed. These values have then also been adjusted by a ratio factor of 1.04. Note the original RTNDT values are shown in parentheses.

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 4 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 ART Values In order to extend the applicability of the current unadjusted P-T limit curves to 55 EFPY, the ART values at 55 EFPY must be shown to be less than or equal to the ART values used to determine the unadjusted P-T limit curves in [1]. The ART values were originally calculated for Waterford Unit 3 at 55 EFPY for each applicable reactor vessel material at the 1/4T and 3/4T locations in [2]; however, this evaluation did not consider the sister plant data discussed in the previous section. Table 2 and Table 3 contain the 55 EFPY ART calculations from

[2] for Waterford Unit 3 at the 1/4T and 3/4T locations, respectively, with the addition of consideration of the Vogtle Unit 1 sister plant data. These calculations consider all beltline and extended beltline materials (all materials with a projected fluence value of 1 x 1017 n/cm2 [E > 1.0 MeV] at EOLE per Regulatory Issue Summary [RIS] 2014-11 [9]).

As shown in Table 4, the ART values at 55 EFPY remain less than or equal to the ART values used to developed the unadjusted P-T limit curves in [1]. Therefore, the existing 32 EFPY Waterford Unit 3 unadjusted P-T limit curves may be deemed applicable through 55 EFPY for the beltline and extended beltline materials. Since the limiting ART values and unadjusted beltline P-T limits are unchanged, the low temperature overpressure protection (LTOP) enable temperature of 200°F (without uncertainty) reported in [1] also remains valid.

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 5 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 Table 2 Calculation of the Waterford Unit 3 ART Values at the 1/4T Location for 55 EFPY R.G. 1.99, 1/4T Fluence(2)

CF(1) 1/4T RTNDT(U) (4) 'RTNDT I(4) (5) Margin ART Reactor Vessel Material Rev. 2 (x 1019 n/cm2, (qqF) FF(3) (qF) (qF) (°F) (°F) (qF) (qF)

Position E > 1.0 MeV) 5HDFWRU9HVVHOBeltline Materials Intermediate Shell Plate M-1003-1 1.1 20 2.575 1.2535 -25.1 25.1 0 12.5 25.1 25 Intermediate Shell Plate M-1003-2 1.1 20 2.575 1.2535 -20 25.1 0 12.5 25.1 30 Intermediate Shell Plate M-1003-3 1.1 20 2.575 1.2535 -20 25.1 0 12.5 25.1 30 Lower Shell Plate M-1004-1 1.1 20 2.575 1.2535 -37.6 25.1 0 12.5 25.1 13 Lower Shell Plate M-1004-2 1.1 20 2.575 1.2535 0 25.1 0 12.5 25.1 50 Using non-credible Waterford Unit 3 2.1 13.4 2.575 1.2535 0 16.8 0 8.4 16.8 34 surveillance data Lower Shell Plate M-1004-3 1.1 20 2.575 1.2535 -20 25.1 0 12.5 25.1 30 Intermediate Shell Longitudinal Weld 1.1 27 2.575 1.2535 -60 33.8 0 16.9 33.8 8 101-124A (Heat # BOLA & HODA)

Intermediate Shell Longitudinal Welds 1.1 27 2.575 1.2535 -60 33.8 0 16.9 33.8 8 101-124B & C (Heat # HODA)

Lower Shell Longitudinal Welds 1.1 35 2.575 1.2535 -80 43.9 0 21.9 43.9 8 101-142A, B, & C (Heat # 83653)

Using credible sister plant surveillance weld 2.1 35.7 2.575 1.2535 -80 44.8 0 14.0(7) 28.0 -7 data from Vogtle Unit 1 Intermediate to Lower Shell Circumferential 1.1 44.4 2.575 1.2535 -70 55.7 0 27.8 55.7 41 Weld 101-171 (Heat # 88114)

Using credible Waterford Unit 3 2.1 14.8 2.575 1.2535 -70 18.6 0 9.3 18.6 -33 surveillance data 5HDFWRU9HVVHO([WHQGHG%HOWOLQH0DWHULDOV Upper Shell Plate M-1002-1 1.1 92 0.035 0.2389 -15.4 22.0 0 11.0 22.0 29 Upper Shell Plate M-1002-2 1.1 91.3 0.035 0.2389 -1.4 21.8 0 10.9 21.8 42 Upper Shell Plate M-1002-3 1.1 65.5 0.035 0.2389 -20 15.6 0 7.8 15.6 11

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 6 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 R.G. 1.99, 1/4T Fluence(2)

CF(1) 1/4T RTNDT(U) (4) 'RTNDT I(4) (5) Margin ART Reactor Vessel Material Rev. 2 (x 1019 n/cm2, (qqF) FF(3) (qF) (qF) (°F) (°F) (qF) (qF)

Position E > 1.0 MeV)

Heat # 606L40 1.1 125.1 0.035 0.2389 -56(6) 29.9 17(6) 14.9 45.3 19 Upper Shell Longitudinal Heat # FOCA 1.1 27 0.035 0.2389 -30 6.4 0 3.2 6.4 -17 Welds 101-122A, B, & C Heat # JBHA 1.1 41 0.035 0.2389 -40 9.8 0 4.9 9.8 -20 Heat # HODA 1.1 27 0.035 0.2389 -60 6.4 0 3.2 6.4 -47 Upper to Intermediate Shell Heat # 3P4767 1.1 199 0.035 0.2389 -56(6) 47.5 17(6) 23.8 58.4 50 Circumferential Heat # IAGA 1.1 41 0.035 0.2389 -30 9.8 0 4.9 9.8 -10 Weld 106-121 Heat # KOHA 1.1 41 0.035 0.2389 -30 9.8 0 4.9 9.8 -10 Notes:

1. Values taken from Table 3-5 of [2].

2. Values taken from Table 6.1-3 of [2].

3. FF = fluence factor = f(0.28 -  ORJ I .

4. Initial RTNDT values are based on measured data, unless otherwise noted. Values are taken from Table 3-1 and Table 3-2 of [2].

5. Per Appendix D of [7], the Waterford Unit 3 surveillance data for the plate were deemed non-credible, whereas the surveillance data for the weld material were deemed credible. Per the guidance of Regulatory Guide 1.99, Rev. WKHEDVHPHWDO = 17°F for Position 1.1 and for Position 2.1 with non-FUHGLEOHVXUYHLOODQFHGDWDWKHZHOGPHWDO = 28°F for 3RVLWLRQDQGZLWKFUHGLEOHVXUYHLOODQFHGDWD = 14°F for Position 2.1. +RZHYHU QHHGQRWH[FHHG 57NDT.

6. Initial RTNDT value (-56°F) is generic for Linde 0091 flux type welds per [8]7KHUHIRUHI = 17°F.

7. Per Appendix D of [4], the surveillance data for the sister plant weld were deemed credible. Per the guidance of Regulatory Guide 1.99, Rev. 2, with credible weld surveillance data,

= 14°F for Position 2.1. +RZHYHU QHHGQRWH[FHHG 57NDT.

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 7 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 Table 3 Calculation of the Waterford Unit 3 ART Values at the 3/4T Location for 55 EFPY R.G. 1.99, 3/4T Fluence(2)

CF(1) 3/4T RTNDT(U) (4) 'RTNDT I(4) (5) Margin ART Reactor Vessel Material Rev. 2 (x 1019 n/cm2, (qqF) FF(3) (qF) (qF) (°F) (°F) (qF) (qF)

Position E > 1.0 MeV) 5HDFWRU9HVVHO%HOWOLQH0DWHULDOV Intermediate Shell Plate M-1003-1 1.1 20 0.915 0.9750 -25.1 19.5 0 9.7 19.5 14 Intermediate Shell Plate M-1003-2 1.1 20 0.915 0.9750 -20 19.5 0 9.7 19.5 19 Intermediate Shell Plate M-1003-3 1.1 20 0.915 0.9750 -20 19.5 0 9.7 19.5 19 Lower Shell Plate M-1004-1 1.1 20 0.915 0.9750 -37.6 19.5 0 9.7 19.5 1 Lower Shell Plate M-1004-2 1.1 20 0.915 0.9750 0 19.5 0 9.7 19.5 39 Using non-credible Waterford Unit 3 2.1 13.4 0.915 0.9750 0 13.1 0 6.5 13.1 26 surveillance data Lower Shell Plate M-1004-3 1.1 20 0.915 0.9750 -20 19.5 0 9.7 19.5 19 Intermediate Shell Longitudinal Weld 1.1 27 0.915 0.9750 -60 26.3 0 13.2 26.3 -7 101-124A (Heat # BOLA & HODA)

Intermediate Shell Longitudinal Welds 1.1 27 0.915 0.9750 -60 26.3 0 13.2 26.3 -7 101-124B & C (Heat # HODA)

Lower Shell Longitudinal Welds 1.1 35 0.915 0.9750 -80 34.1 0 17.1 34.1 -12 101-142A, B & C (Heat # 83653)

Using credible sister plant surveillance weld 2.1 35.7 0.915 0.9750 -80 34.8 0 14.0(7) 28.0 -17 data from Vogtle Unit 1 Intermediate to Lower Shell Circumferential 1.1 44.4 0.915 0.9750 -70 43.3 0 21.6 43.3 17 Weld 101-171 (Heat # 88114)

Using credible Waterford Unit 3 2.1 14.8 0.915 0.9750 -70 14.4 0 7.2 14.4 -41 surveillance data 5HDFWRU9HVVHO([WHQGHG%HOWOLQH0DWHULDOV Upper Shell Plate M-1002-1 1.1 92 0.012 0.1261 -15.4 11.6 0 5.8 11.6 8 Upper Shell Plate M-1002-2 1.1 91.3 0.012 0.1261 -1.4 11.5 0 5.8 11.5 22

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 8 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 R.G. 1.99, 3/4T Fluence(2)

CF(1) 3/4T RTNDT(U) (4) 'RTNDT I(4) (5) Margin ART Reactor Vessel Material Rev. 2 (x 1019 n/cm2, (qqF) FF(3) (qF) (qF) (°F) (°F) (qF) (qF)

Position E > 1.0 MeV)

Upper Shell Plate M-1002-3 1.1 65.5 0.012 0.1261 -20 8.3 0 4.1 8.3 -3 (6) (6)

Heat # 606L40 1.1 125.1 0.012 0.1261 -56 15.8 17 7.9 37.5 -3 Upper Shell Longitudinal Heat # FOCA 1.1 27 0.012 0.1261 -30 3.4 0 1.7 3.4 -23 Welds 101-122A, B & C Heat # JBHA 1.1 41 0.012 0.1261 -40 5.2 0 2.6 5.2 -30 Heat # HODA 1.1 27 0.012 0.1261 -60 3.4 0 1.7 3.4 -53 Upper to Intermediate Shell Heat # 3P4767 1.1 199 0.012 0.1261 -56(6) 25.1 17(6) 12.6 42.3 11 Circumferential Heat # IAGA 1.1 41 0.012 0.1261 -30 5.2 0 2.6 5.2 -20 Weld 106-121 Heat # KOHA 1.1 41 0.012 0.1261 -30 5.2 0 2.6 5.2 -20 Notes:

1. Values taken from Table 3-5 of [2].

2. Values taken from Table 6.1-4 of [2].

3. FF = fluence factor = f(0.28 -  ORJ I .

4. Initial RTNDT values are based on measured data, unless otherwise noted. Values are taken from Table 3-1 and Table 3-2 of [2].

5. Per Appendix D of [7], the Waterford Unit 3 surveillance data for the plate were deemed non-credible, whereas the surveillance data for the weld material were deemed credible. Per the guidance of Regulatory Guide 1.99, Rev. WKHEDVHPHWDO = 17°F for Position 1.1 and for Position 2.1 with non-FUHGLEOHVXUYHLOODQFHGDWDWKHZHOGPHWDO = 28°F for 3RVLWLRQDQGZLWKFUHGLEOHVXUYHLOODQFHGDWD = 14°F for Position 2.1. +RZHYHU QHHGQRWH[FHHG 57NDT.

6. Initial RTNDT value (-56°F) is generic for Linde 0091 flux type welds per [8]7KHUHIRUHI = 17°F.

7. Per Appendix D of [4], the surveillance data for the sister plant weld were deemed credible. Per the guidance of Regulatory Guide 1.99, Rev. 2, with credible weld surveillance data,

)IRU3RVLWLRQ+RZHYHU QHHGQRWH[FHHG 57NDT.

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Page 9 of 11 Our ref: LTR-SDA-20-041, Rev. 0 May 24, 2021 Table 4 Summary of the Waterford Unit 3 Limiting ART Values Used in the Applicability Evaluation of the Existing Reactor Vessel Heatup and Cooldown Curves at 32 EFPY 1/4T Limiting ART (°F) 3/4T Limiting ART (°F)

Limiting Material Existing 32 EFPY Existing 32 EFPY Evaluation at (Axial Flaw) Curves Documented in Evaluation at Curves Documented 55 EFPY WCAP-16088-NP, 55 EFPY (Table 2) in WCAP-16088-NP, (Table 3)

Rev. 2 Rev. 2 Lower Shell Plate M-1004-2 (Position 2.1 - Using credible 50 --- 42 ---

surveillance data)

Lower Shell Plate M-1004-2

--- 50(1) --- 39 (Position 1.1)

Note:

1. Per Table 2, Upper to Intermediate Shell Circumferential Weld 106-121 (Heat # 3P4767) also has a 1/4T ART value of 50°F. However, since it is a circumferential weld and would, therefore, be analyzed with the less restrictive circumferential flaw method; it is not considered to be as limiting as Lower Shell Plate M-1004-2.

• • • This record was final approved on 5/25/2021 3:31:29 PM. (This statement was added by the PRIME system upon its validation)