ML19269B775

Proposed License Amendment Request: Update of Reactor Coolant System Heatup and Cooldown Pressure-Temperature Limitations Figures for Subsequent License Renewal

ML19269B775
Person / Time
Site:	Surry
Issue date:	09/19/2019
From:	Mark D. Sartain Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
19-374
Download: ML19269B775 (101)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 September 19, 2019 U. S. Nuclear Reg,ulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST Serial No.:

NRA/GDM:

Docket Nos.:

License Nos.:

10 CFR 50.90 19-374 R1 50-280/281 DPR-32/37 UPDATE OF REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMITATIONS FIGURES FOR SUBSEQUENT LICENSE RENEWAL Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) requests amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating License Numbers DPR-32 and DPR-37 for Surry Power Station (Surry) Units 1 and 2. The proposed change revises TS Figures 3.1-1 and 3.1-2, Surry Units 1 and 2 Reactor Coolant System Heatup Limitations and Surry Units 1 and 2 Reactor Coolant System Coo/down Limitations, respectively, to: 1) update the cumulative core burnup applicability limit (Effective Full Power Years; EFPY) from 48 to 68 EFPY, and 2) revise and relocate the limiting material property basis from the TS figures to the TS Basis. The cumulative core burnup applicability limit is also updated for the Low Temperature Overpressure Protection System (L TOPS) Setpoint and the L TOPS Enabling Temperature (T-enable) at Surry Units 1 and 2; however, no additional TS changes are required since these two values remain conservative with respect to the existing TS limits. The proposed change is being requested as a result of evaluations performed for the Surry Subsequent License Renewal (SLR) effort. Associated TS Basis changes are included for information. provides a discussion of the proposed change, and Attachment 2 includes an excerpt from the Surry SLR application dated October 15, 2018 (Serial No.18-359)

[ADAMS Accession No. ML18291A842] that provides the supporting technical evaluation for the proposed change. Marked-up and typed TS pages reflecting the proposed change are provided in Attachments 3 and 4, respectively.

We have evaluated the proposed amendment request and have determined that it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is included in Attachment 1. We have also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released off-site or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement

Serial No.19-374 Docket Nos. 50-280/281 Page 2 of 3 or environmental assessment is needed in connection with the approval of the proposed change. The proposed TS change included in this license amendment request has been reviewed and approved by the Facility Safety Review Committee.

Dominion Energy Virginia requests approval of the proposed license amendment request by June 30,. 2020, with a 60-day implementation period, to coincide with the issuance of the renewed Surry Units 1 and 2 SLR operating licenses.

Should you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.

Respectfully, Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support Commitments contained in this letter: None Attachments:

1. Discussion of Change

2. Supporting Technical Information from SLR Application (SLRA), Section 4.2, "Reactor Vessel Neutron Embrittlement Analysis"

3. Marked-up Technical Specifications and Bases Pages

4. Proposed Technical Specifications and Bases Pages COMMONWEAL TH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. Mark D. Sartain, who is Vice President - Nuclear Engineering and Fleet Support, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this f 9f\\... day of S~'cu-", 2019.

My Commission Expires: fw..2,'v-,,b.\\- 3\\, "2..cl-3 GARY DON MILLER Notary Public Commonwealth of Virginia Reg.# 7629412 My Commission Expires August 31, 20~

cc:

U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, GA 30303-1257 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street Suite 730, Richmond, VA 23219 Mr. Bill Rogers, Senior Reactor Engineer License Renewal Projects Branch Division of Materials and License Renewal Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-11 F1 Washington, DC 20555 Ms. Angela Wu, Project Manager License Renewal Projects Branch Division of Materials and License Renewal Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-11 F1 Washington, DC 20555 Mr. Vaughn Thomas NRC Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 04 F-12 11555 Rockville Pike Rockville, MD 20852-2738 Mr. G. Edward Miller NRC Senior Project Manager - North Anna

. U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Surry Power Station Serial No. 19-37 4 Docket Nos. 50-280/281 Page 3 of 3

Serial No.19-374 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures DISCUSSION OF CHANGE Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures DISCUSSION OF CHANGE TABLE OF CONTENTS 1.0 Summary Description 2.0 Detailed Description 2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change 2.4 Description of the Proposed Change 3.0 Technical Evaluation 4.0 Regulatory Evaluation 4.1 Applicable Regulatory Requirements 4.2 No Significant Hazards Consideration 4.3 Conclusions 5.0 Environmental Consideration 6.0 References Page1of15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures DISCUSSION OF CHANGE 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) is submitting a license amendment request (LAR) to revise the Surry Power Station (Surry) Units 1 and 2 Technical Specifications (TS). Specifically, TS Figures 3.1-1 and 3.1-2, Surry Units 1 and 2 Reactor Coolant System Heatup Limitations and Surry Units 1 and 2 Reactor Coolant System Coo/down Limitations, respectively, are being revised to: 1) update the cumulative core burnup applicability limit (Effective Full Power Years; EFPY), and 2) revise and relocate the limiting material property basis from the TS figures to the TS Basis.

The cumulative core burnup applicability limit is also updated for the Low Temperature Overpressure Protection System (L TOPS) Setpoint and the L TOPS Enabling Temperature (T-enable) at Surry Units 1 and 2. The proposed changes are being implemented as a result of evaluations performed for the Surry Subsequent License Renewal (SLR) application (SLRA)

[Reference 6.1]. Associated TS Basis changes are included for information.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation RCS components are designed to withstand the effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.

The pressure/temperature (P-T) limitations curves limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation.

The P-T limit curves are for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature. Each P-T limit curve defines an acceptable region for normal operation. The typical use of the curves is operational guidance during heatup or cooldown maneuvering when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The P-T limit curves establish operating limits that provide a margin to brittle failure of the reactor vessel (RV) and piping of the reactor coolant pressure boundary (RCPB). The RV is the component most subject to brittle failure, and the P-T limit curves' limits apply mainly to the RV. The limits do not apply to the pressurizer, which has different design characteristics and operating functions.

Page 2 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures 10 CFR 50, Appendix G, requires the establishment of P-T limits for specific material fracture toughness requirements of the RCPB materials and requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code, Section Ill, Appendix G. The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RT NOT) as exposure to neutron fluence increases.

The actual shift in the RT NOT of the RV material is established periodically by removing and evaluating the irradiated RV material specimens, in accordance with ASTM E 185 and Appendix H of 10 CFR 50. The operating P-T limit curves are adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials.

The P-T limit curves are calculated using the most limiting value of RTNoT corresponding to the limiting beltline region material for the RV. The heatup curve represents a different set of restrictions than the coo Id own curve because the directions of the thermal gradients through the RV wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

The L TOPS controls RCS pressure at low temperatures so the integrity of the RCPB is not compromised. The L TOPS actuation logic monitors RCS pressure and determines when a condition is not acceptable. Each time the P-T limit curves are revised, the L TOPS setpoint and T-enable value must also be re-evaluated to ensure functional requirements can still be met to ensure the integrity of the RCPB.

2.2 Current Technical Specifications Requirements The current Surry TS Figures 3.1-1 and 3.1-2 RCS Heatup and Cooldown P-T Limitations curves reflect the cumulative core burnup applicability limit of 48 EFPY and the material property bases associated with the period of extended operation for the renewed Surry Units 1 and 2 operating licenses.

Surry TS 3.1.G.1 specifies an L TOPS arming temperature of 350° F and an L TOPS pressurizer PORV setpoint of ~390 psig for operation.

2.3 Reason for the Proposed Change By letter dated October 15, 2018, [Reference 6.1], Dominion Energy Virginia submitted an application to the US NRC for the subsequent license renewal of Renewed Facility Operating License Nos. DPR-32 and DPR-37 for Surry Units 1 and 2. The operating licenses will be extended from 60 years to 80 years thereby requiring a TS change for the Surry TS RCS heatup and cooldown P-T Limits figures, L TOPS Setpoint, and T-enable value from 48 EFPY to 68 EFPY cumulativ~ core burnup applicability limit.

Page 3 of 15 J

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures 2.4 Decription of Proposed Change The proposed change updates TS Figures 3.1-1 and 3.1-2, Surry Units 1 and 2 Reactor Coolant System Heatup Limitations and Surry Units 1 and 2 Reactor Coolant System Coo/down Limitations, respectively, to reflect the increased cumulative core burnup applicability limit for RCS P-T Limits, L TOPS Setpoints, and L TOPS T-Enable values from 48 to 68 EFPY cumulative core burnup applicability limit. The proposed change also revises and relocates the limiting material property bases information from the TS figures to the TS 3.1.B Basis for Surry Units 1 and 2.

The proposed revisions to TS Figures 3.1-1 and 3.1-2 and the TS 3.1.B Basis are summarized as follows:

1. TS Figures 3.1-1 and 3.1-2 The proposed change revises the existing TS Figure 3.1-1 and 3.1-2 titles, respectively, from:

Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°Flhr) Applicable to 48 EFPY and Surry Units 1 and 2 Reactor Coolant System Coo/down Limitations (Coo/down Rates up to 100°Flhr) Applicable to 48 EFPY to Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°F/hr) Applicable to 68 EFPY [BOLD emphasis added]

and Surry Units 1 and 2 Reactor Coolant System Coo/down Limitations (Coo/down Rates up to 100°Flhr) Applicable to 68 EFPY [BOLD emphasis added]

The proposed change revises and relocates to the TS 3.1 Basis the existing Material Property Basis information contained in the informational block located above each curve. (See Item 3 below.) The revised information includes the Limiting Material and Limiting Adjusted Reference Temperature (ART) values associated with plant operation up to the SLR cumulative core burnup applicability limit of 68 EFPY, as opposed to the existing limit of 48 EFPY associated with the first license renewal for Surry Units 1 and 2.

Page 4 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures

2. L TOPS Setpoint and L TOPS T-enable value The L TOPS Setpoint and L TOPS T-enable for Surry Units 1 and 2 were also evaluated for cumulative core burnups up to 68 EFPY.

It was determined that the existing L TOPS setpoint and the T-enable value remain valid. The evaluation determined an L TOP enabling temperature of 283°F and L TOP PORV setpoint of 399.6 psig. TS 3.1.G.1 requires the system to be operable (i.e., arming temperature) when the RCS temperature is 350°F and specifies a PORV lift setting of.:::. 390 psig. Thus, the existing TS values, which are also reflected in UFSAR Section 4.3.4, are bounding and require no revision.

3. TS Basis The proposed TS 3.1.B Basis Insert for SLR is as follows:

The technical basis for the data points and the associated Adjusted Reference Temperature (ART) values used to generate the heatup and coo/down curves is provided in WCAP-14177 (Reference 2) and were determined to be applicable to the 48 EFPY period of extended operation under first license renewal.

The associated ART values used to calculate the heatup and coo/down curves provided in WCAP-14177 are based on the Surry Unit 1 Intermediate to Lower Shell Circumferential Weld:

114-T, 228.4°F, and 3/4-T, 189. 5°F The heatup and coo/down curves for operation through 48 EFPY were based upon the K1r methodology.

These heatup and coo/down curves were subsequently evaluated using the K/c methodology for Subsequent License Renewal (SLR) at 68 EFPY in WCAP-18243-NP (Reference 3).

The limiting reactor vessel materials at 68 EFPY were determined to be the Surry Unit 1 Lower Shell Longitudinal Weld L2 at 1/4-T and the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld at 3/4-T.

The associated ART values calculated at 68 EFPY are:

114-T, 219.4 °F, and 3/4-T, 179. 8 °F The data points and the associated ART values used to generate the heatup and coo/down curves in TS Figures 3.1-1 and 3.1-2, respectively, are conservative based upon use of the K/c methodology. Therefore, the heatup and coo/down curves did not require revision as a result of SLR.

However, the fluence applicability is updated from 48 EFPY to 68 EFPY.

Page 5 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures The following two items are being added to the TS 3.1 Basis References list:

2. WCAP-14177, "Surry Units 1 and 2 Heatup and Coo/down Limit Curves for Normal Operation," (October 1994)

3. WCAP-18243-NP, Rev. 3, "Surry Units 1 and 2 Heatup and Coo/down Limit Curves for Normal Operation," (January 2019)

3.0 TECHNICAL EVALUATION

RCS heatup and cooldown limit curves are calculated using the most limiting value of RT NOT corresponding to the limiting material in the beltline region of the RV. The most limiting RT NOT of the material in the core region (beltline) of the RV is determined by using the unirradiated RV material fracture toughness properties and estimating the irradiation induced shift (LlRT NOT). RT NOT increases as the material is exposed to fast neutron irradiation; therefore, to find the most limiting core region (beltline) RT NOT at any time, LlRT NOT due to the neutron radiation exposure associated with that time must be added to the original unirradiated RT Nor Using the Adjusted Reference Temperature (ART) values, P-T limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G, as augmented by the ASME Code,Section XI, Appendix G.

The current P-T limits included in the Surry Units 1 and 2 TS are based on the Kia methodology and the latest fluence data through 48 EFPY.

K1a is the crack arrest fracture toughness, which is the critical value of the stress intensity factor (K1) for crack arrest as a function of temperature.

RV nozzle materials were evaluated in WCAP-18242-NP [Reference 6.3] at 48 EFPY and 68 EFPY. The evaluated nozzle forging materials are documented in SLRA Tables 4.2.4-1, 4.2.4-3, 4.2.4-5, and 4.2.4-7. The nozzle materials were assigned the fluence values at the postulated 1/4T flaw location for each specific nozzle in Table 4.2.1-1 and Table 4.2.1-2. Thus, Unit 1 Inlet Nozzle 1, Unit 2 Inlet Nozzle 1, and Outlet Nozzle 3 have neutron fluence values greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at 68 EFPY. To fully assess the Units 1 and 2 P-T limit curves applicability to 68 EFPY, a nozzle corner fracture mechanics analysis was completed for the nozzle materials. These nozzle P-T limit curves were generated and compared to the beltline P-T limit curves in TS to ensure the beltline curves are bounding. The detailed nozzle forging fracture mechanics evaluation and comparison to the applicable RV beltline P-T limit curves were documented in WCAP-18243-NP [Reference 6.4].

The current TS beltline P-T curves were confirmed to remain more limiting than the nozzle P-T curves through 68 EFPY.

Page 6 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures The development of the current P-T limit curves for normal heatup and cooldown of the RCS for Units 1 and 2 was documented in WCAP-14177 [Reference 6.2]. The existing P-T limit curves are based on the K1a methodology and the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material. The Units 1 and 2 P-T limit curves were developed by calculating ART values utilizing the vessel fluence at the clad/base metal interface corresponding to each RV material. Since the development of the curves, the applicability of the curves has been extended and the fluence values and initial material properties used to calculate ART values have been updated.

The K1c methodology was used to confirm the applicability of the P-T limit curves developed based on WCAP-14177. K1c is the plane strain fracture toughness, which is the material toughness property measured in terms of the stress intensity factor, Kl, that will lead to nonductile crack propagation.

The limiting RV material ART values with consideration of the updated 68 EFPY fluence values, revised Position 2.1 chemistry factor values, and updated initial RT NDT values, must be shown to be less than or equal to the limiting beltline material ART values used in development of the P-T limit curves contained in WCAP-14177 and the Units 1 and 2 TS. The Regulatory Guide 1.99 methodology [Reference 6.5] was used along with the surface fluence of Section 2 of WCAP-18242-NP to calculate ART values for the Units 1 and 2 RV materials at 48 EFPY and 68 EFPY.

Comparisons of the use of the K1c reference stress intensity factor instead of the older, more conservative K1a reference stress intensity factor were conducted to validate that the P-T limits for 48 EFPY are conservative for operation through the subsequent period of extended operation (i.e., 68 EFPY). The comparisons of the limiting ART values calculated as part of the RV integrity Time Limiting Aging Analysis (TLAA) evaluation, using updated fluence and initial material properties, to those used in calculation of the existing P-T limit curves are contained in Table 4.2.4-9 for Units 1 and 2.

With the consideration of fluence projections, the applicability of the P-T limit curves in WCAP-14177 may be extended to 68 EFPY for the Units 1 and 2 cylindrical shell materials.

Nozzle P-T limit curves were developed per WCAP-18243-NP and compared to the cylindrical shell beltline curves. ART values were generated without consideration of the methodology in TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels, U.S. NRC Technical Letter Report, Office of Nuclear Regulatory Research [RES]" [Reference 6.6].

The evaluations described in WCAP-18242 and WCAP-18243 have been used to prepare Section 4.2 of the Surry SLRA to support extension of the cumulative core burnup applicability limit from 48 EFPY to 68 EFPY for Surry Units 1 and 2. The discussions in WCAP-18242, WCAP-18243, and Section 4.2 of the Surry SLRA affirms the conservatism Page 7 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures of the existing RCS P-T Limits, LTOPS Setpoint, and LTOPS T-enable value for Surry Units 1 and 2 cumulative core burnups up to 68 EFPY.

The following subsections in Section 4.2 of the Surry SLRA (provided in Attachment 2) include detailed supporting information:

Neutron Fluence Projections (Section 4.2.1)

Upper-Shelf Energy (Section 4.2.2)

Pressurized Thermal Shock (Section 4.2.3)

Adjusted Reference Temperature (Section 4.2.4)

Pressure-Temperature Limits (Section 4.2.5)

Low Temperature Overpressure Protection Analyses (Section 4.2.6)

Per WCAP-18243-NP, the applicability of the RCS P-T limit curves may be extended through SLR because the current TS P-T limit curves bound the new P-T limit curves developed in WCAP-18243-NP regardless of the use of the TLR-RES/DE/CIB-2013-01 methodology. Per WCAP-18243-NP, the applicability of the current TS P-T limit curves may be extended through the subsequent period of extended operation.

In addition, the applicable RV flange and closure head initial RT NDT values are bounding and the P-T limit curves flange notch requires no change or further consideration. Finally, the lowest service temperature requirements are not applicable to Surry Units 1 and 2 because the plants areWestinghouse-designed per ASME Code, Section Ill, and utilize stainless steel reactor coolant system piping.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements Design Requirements: 10 CFR 50, Appendix A-General Design Criteria The regulations in Appendix A to Title 10 of the Code of Federal Regulations (10 CFR)

Part 50 establish minimum principal design criteria for water-cooled nuclear power plants, while 10 CFR 50 Appendix B and the licensee quality assurance programs establish quality assurance requirements for the design, manufacture, construction, and operation of structures, systems, and components. The current regulatory requirements of 10 CFR 50 Appendix A that are applicable to the Reactor Coolant Pressure Boundary (RCPB),

including the Reactor Vessel, include:

General Design Criteria (GDC) 14 (Reactor Coolant Pressure Boundary), GDC 31 (Fracture Prevention of Reactor Coolant Pressure Boundary), and GDC 32 (Inspection of Reactor Coolant Pressure Boundary).

During the initial plant licensing of Surry Units 1 and 2, it was demonstrated that the design of the RCPB, met the regulatory requirements in place at that time. The GDC included in Page 8 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures Appendix A to 10 CFR 50 did not become effective until May 21, 1971. The Construction Permits for Surry Units 1 and 2 were issued prior to May 21, 1971; consequently, Surry Units 1 and 2 were not subject to current GDC requirements (SECY-92-223, dated September 18, 1992). The following information demonstrates Surry Units 1 and 2 meet the intent of the GDC published in 1967 (Draft GDC). Specifically, Section 1.4 of the Surry Updated Final Safety Analysis Report (UFSAR) discusses Surry compliance with these criteria. The draft GDC associated with the RCPB are addressed below.

Quality Standards (Criterion 1 - draft)

Those systems and components of reactor facilities that are essential to the prevention of accidents which could affect the public health and safety or to the mitigation of their consequences shall be identified and then designed, fabricated, and erected in accordance with quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified.

Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. A showing of sufficiency and applicability of codes, standards, quality assurance programs, test procedures, and inspection acceptance levels used is required.

Design Conformance Structures, systems, and components important to safety are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

The Quality Assurance Program was established to provide assurance that safety-related structures, systems, and components satisfactorily perform their intended safety functions.

Reactor Coolant Pressure Boundary (Criterion 9 - draft)

The reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime.

Design Conformance The reactor coolant pressure boundary is designed, fabricated, and constructed to have an exceedingly low probability of gross rupture or significant uncontrolled leakage throughout its design lifetime. The reactor coolant system, in conjunction with its control and protective provisions, is designed to accommodate the system Page 9 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures pressures and temperatures attained under all expected modes of unit operation or anticipated system interactions, and to remain within the applicable code stress limits.

The fabrication of the components that constitute the pressure-retaining boundary of the reactor coolant system is carried out in strict accordance with the applicable codes:

In addition, there are areas where equipment specifications for reactor coolant system components are more restrictive than applicable codes.

The materials of construction of the pressure-retaining boundary of the reactor coolant system are protected by the control of coolant chemistry so as to prevent corrosion phenomena that might otherwise reduce the system structural integrity during its service lifetime.

Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention (Criterion 34 - draft)

The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagating type failures. Consideration shall be given (a) to the notch toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loadings, (c) to the quality control specified for materials and component fabrication to limit flaw sizes, and (d) to the provisions for control over service temperature and irradiation effects which may require operational restrictions.

Design Conformance The reactor coolant pressure boundary is designed and operated to reduce to an acceptable level the probability of a rapidly propagating failure. Consideration is given to (1) the provisions for control over service temperature and irradiation effects that may require operational restrictions, (2) the design and construction of the reactor pressure vessel in accordance with applicable codes, including those that establish the requirements for the absorption of energy within the elastic strain energy range and for the absorption of energy by plastic deformation, and (3) the design and construction of reactor coolant pressure boundary piping and equipment

• in accordance with applicable codes.

The reactor coolant pressure boundary is designed to reduce the probability of a rapidly propagating failure to an acceptable level. The fast neutron exposure of the core region of the reactor vessel changes the notch toughness of the vessel material.

This change is indicated by the increase in the nil ductility transition temperature and allowance for it is made in the operating procedures by ensuring that the vessel is not subjected to full operating pressure until its temperature exceeds the design transition temperature, defined to be the nil ductility transition temperature plus a 60°F margin.

The pressure during unit start-up and shutdown at temperatures below the nil ductility Page 10 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures transition temperature are maintained below the threshold of concern for safe operation.

The design transition temperature dictates the procedures to be followed in hydrostatic testing and in station operations to avoid excessive cold stress. The value of the design transition temperature is increased during the life of the station as required by the expected shift in the nil ductility transition temperature, which is confirmed by the experimental data obtained from irradiated specimens of reactor vessel materials during the unit lifetime.

All pressure-containing components of the reactor coolant system are designed, fabricated, inspected, and tested in conformance with the applicable codes.

Reactor Coolant Pressure Boundary Brittle Fracture Prevention (Criterion 35 - draft)

Under conditions where reactor coolant pressure boundary system components constructed of ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperature shall be at least 120° F above the nil ductility transition (NOT) temperature of the component material if the resulting energy release is expected to be absorbed by plastic deformation or 60° F above the NOT temperature of the component material if the resulting energy release is expected to be absorbed within the elastic strain range.

Design Conformance For conditions under which reactor coolant pressure boundary system components constructed of ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperatures shall be at least 120°F above the nil ductility transition temperature of the component material if the resulting energy

• release is expected to be absorbed by plastic deformation, or 60°F above the nil ductility temperature of the component material if the resulting energy release is expected to be absorbed within the elastic strain energy range.

Sufficient testing and analysis of materials used in reactor coolant system components are performed to ensure that the required nil ductility transition temperature limits specified in the criterion are met. Removable test capsules are installed in the reactor vessel and removed and tested at various times in the unit lifetime to determine the effects of the operation on system materials.

Reactor Coolant Pressure Boundary Surveillance (Criterion 36 - draft)

Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. For the reactor Page 11 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures vessel, a material surveillance program conforming with ASTM-E-185-66 shall be provided.

Design Conformance Reactor coolant pressure boundary components have provisions for the inspection, testing, and surveillance of critical areas by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance program conforming to current applicable codes is provided.

The design of the reactor vessel and its arrangement in the system permit accessibility during service life to all internal surfaces of the vessel and to certain external zones such as the areas of the nozzle-to-piping welds and the top and bottom heads. The reactor arrangement within the containment provides sufficient space for the inspection of the external surfaces of the reactor coolant piping, except for the area of pipe within the primary shielding concrete.

The monitoring of the nil ductility transition temperature properties of the core region plates, forgings, weldments, and associated heat-treated zones is performed in accordance with ASTM E 185, Recommended Practice for Swveil/ance Tests on Structural Materials in Nuclear Reactors. Samples of reactor vessel plate materials are retained and cataloged in case future engineering development shows the need for further testing.

The material properties surveillance program includes not only the conventional tensile tests, but also tests of fracture mechanics specimens. The fracture mechanics specimens are the wedge-opening-loading-type specimens. The observed irradiation shifts in the nil ductility transition temperature of the core region materials are used to confirm the calculated limits to start-up and shutdown transients.

Quality Assurance Quality assurance criteria provided in 10 CFR Part 50, Appendix B, that apply to the reactor coolant pressure boundary and reactor vessel include: Criteria Ill, V, XI, XVI, and XVII.

Criteria Ill and V require measures to ensure that applicable regulatory requirements and the design basis, as defined in 10 CFR 50.2, "Definitions," and as specified in the license application, are correctly translated into controlled specifications, drawings, procedures, and instructions. Criterion XI requires a test program to ensure that the subject systems will perform satisfactorily in service and requires that test results shall be documented and evaluated to ensure that test requirements have been satisfied.

Criterion XVI requires measures to ensure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected, and that significant conditions Page 12 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures adverse to quality are documented and reported to management. Criterion XVII requires maintenance of records of activities affecting quality.

4.2 No Significant Hazards Consideration Analysis Virginia Electric and Power Company (Dominion Energy Virginia) proposes a change to the Surry Power Station (Surry) Units 1 and 2 Technical Specifications (TS) to modify the Reactor Coolant System (RCS) Heatup and Cooldown Limitations, Figures 3.1-1 and 3.1-2, respectively. The proposed change revises the two TS figures to reflect a cumulative core burn up applicability limit of 68 EFPY for the RCS PressurefTemperature (P-T) Limits.

The cumulative core applicability limit is also increased to 68 EFPY for the Low Temperature Overpressure Protection System (L TOPS) Setpoint and L TOPS Enabling Temperature (T-enable). In addition, the material properties bases currently included on the TS figures will be relocated to the TS 3.1 Basis. In accordance with the criteria set forth in 10 CFR 50.92, Dominion Energy Virginia has evaluated the proposed TS change and determined that the change does not represent a significant hazards consideration.

The following is provided in support of this conclusion:

1. Does the change involve a significant increase in the probability of consequence of an accident previously evaluated?

Response: No The proposed change revises the Surry Units 1 and 2 TS RCS Heatup and Cooldown Limitations figures to reflect an increase in the cumulative core burnup applicability limit to 68 EFPY. The existing Surry TS RCS P-T Limits, L TOPS Setpoint, and T-enable value remain valid and conservative for cumulative core burnup up to 68 EFPY, thus increasing the cumulative core burnup applicability limit for RCS P-T Limits, L TOPS Setpoints and L TOPS T-enable to 68 EFPY has no bearing on the probability or consequences of an accident previously evaluated. These evaluations address the L TOPS design basis mass addition accident (inadvertent charging pump start), heat addition accident (Reactor Coolant Pump (RCP) start with a secondary-to-primary temperature difference of 50°F) and Pressurized Thermal Shock (PTS) events, the analysis of which is covered by 10 CFR 50.61.

The increased cumulative core burnup applicability is accomplished through application of improved analytical margins using the K1c reference stress intensity factor, instead of the older, more conservative K1a reference stress intensity factor. Dominion Energy Virginia assessed the effect of use of the analytical margins and determined that the existing TS limits (RCS P-T Limits, L TOPS Setpoints and L TOPS T-enable) governing reactor vessel integrity remain valid and conservative for cumulative core burnup to 68 EFPY. No changes to plant systems, structures or components are proposed, and no new operating modes are established. Therefore, there is no increase in the probability or consequences of any accident previously evaluated.

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Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No No changes to plant operating conditions, operating limits or setpoints are being proposed and no changes to plant systems, structures or components are being implemented.

The existing Surry TS RCS P-T Limits, L TOPS Setpoints, and L TOPS T-enable value remain valid and conservative for cumulative core burnups up to 68 EFPY. Analysis supporting the increased cumulative core burnup applicability limit was performed in accordance with applicable regulatory guidance and confirms that design functions (i.e., ensuring that combined pressure and thermal stresses under normal operating heatup and cooldown conditions and under design basis accident conditions at low temperature) are maintained. Therefore, the proposed change does not create the possibility of any accident or malfunction of a different type previously evaluated.

.. 3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

• The

• increased cumulative core burn up applicability limit is accomplished through application of improved analytical margins provided by using the K1c reference stress intensity factor, instead of the older, more conservative Kia reference stress intensity factor. Dominion Energy Virginia assessed the effect of the use of the analytical margins

, and determined that the existing TS P-T Limits, L TOPS Setpoint, and L TOPS T-enable value governing reactor vessel integrity remain valid and conservative for cumulative core burnups up to 68 EFPY. No changes to plant systems, structures or components are proposed, and no new operating modes are established. Furthermore, plant operating limits and setpoints are not being changed. Consequently, the TS P-T Limits, L TOPS Setpoint, and L TOPS T-enable value provide acceptable margin to vessel fracture under both normal operation and L TOPS design basis (mass addition and heat addition) accident conditions for cumulative core burnups up to 68 EFPY. Therefore, the proposed change does not result in a significant reduction in the margin of safety.

Based on the above, Dominion Energy Virginia concludes that the proposed change presents no significant hazards consideration under the standards set forth in 1 OCFR50.92(c), and accordingly, a finding of "no significant hazards considerations" is justified.

4.3 Conclusion Dominion Energy Virginia concludes, based on consideration discussed herein, that (1) there is reasonable assurance that the health and safety of the public will not be Page 14 of 15

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures endangered by the proposed change, (2) such activities will be conducted in accordance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL ASSESSMENT A review has determined that the proposed amendment would not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would not change an inspection or surveillance requirement.

As such, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a 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 amendment 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 amendment.

6.0 REFERENCES

6.1 Letter from Virginia Electric and Power C'ompany to the US Nuclear Regulatory Commission dated October 15, 2018 (Serial No.18-340), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Application for*Subsequent Renewed Operating Licenses." [ADAMS Accession No. ML18291A842]

6.2 WCAP-14177, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," October 1994.

6.3 WCAP-18242-NP, Revision 2, "Surry Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal," July 2018.

6.4 WCAP-18243-NP, Rev. 3, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," January 2019.

6.5 Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

6.6 TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels, U.S. NRC Technical Letter Report, Office of Nuclear Regulatory Research [RES]," November 14, 2014. [ADAMS Accession No. ML14318A177]

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Serial No.19-374 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures SUPPORTING TECHNICAL INFORMATION FROM SLR APPLICATION (SLRA),

SECTION 4.2, "REACTOR VESSEL NEUTRON EMBRITTLEMENT ANALYSIS"

(*Excerpted from SLRA submitted by Virginia Electric and Power Company to the US NRC by letter dated October 15, 2018 [ADAMS Accession No. ML18291A842])

Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

4.2 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses REACTOR VESSEL NEUTRON EMBRITTLEMENT ANALYSIS 1 O CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Reactors for Normal Operation," requires that all light water reactors meet the fracture toughness, P-T limits, and materials surveillance program requirements for the reactor coolant pressure boundary as set forth in 10 CFR 50, Appendices G and H. The materials included in the surveillance capsule program remain unchanged for the subsequent period of extended operation based upon the provisions outlined in earlier versions of ASTM E185, "Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels" (Reference 4.8-1) that existed at the time of initial plant construction. 1 O CFR 50.61 requires that all light water reactors meet the fracture toughness requirements for protection against pressurized thermal shock events. The Reactor Vessel Material Surveillance program is described in Section 82.1.19.

Inputs for reactor vessel (RV) integrity assessments are discussed in this section.

The best estimate copper (Cu) and nickel (Ni) chemical compositions for the Units 1 and 2 RV materials are presented in Table 4.2.2-1 through Table 4.2.2-4. The best estimate weight percent Cu and Ni values for the RV materials were reported in PWROG-16045-NP, "Determination of Un irradiated RT NOT and Upper-Shelf Energy Values of the Units 1 and 2 Reactor Vessel Materials" (Reference 4.8-2) and were included in RV integrity evaluations as part of this TLAA effort.

Prior to updating the RV integrity assessments for the subsequent period of extended operation both the fluence projections and material properties were reviewed and updated by WCAP-18028-NP, "Extended Beltline Pressure Vessel Fluence Evaluations Applicable to Surry Units 1 & 2" (Reference 4.8-3), WCAP-18242-NP, "Surry Units 1 and 2 Time Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal" (Reference 4.8-4) and PWROG-16045-NP. Revised initial material properties, including chemistry factors and fluence projections, through 68 EFPY are included in and Table 4.2.3-2 for Units 1 and 2 respectively.

The neutron fluence axial boundary of the 1.0 x 1017 n/cm2 fluence threshold is depicted in Figures 4.2.2-1 and 4.2.2-2 for Units 1 and 2 respectively. The configuration of the RVs, including the weld identification (ID) numbers, is illustrated in Figures 4.2.2-3 and 4.2.2-4 for Units 1 and 2, respectively.

Reactor vessel integrity assessments are performed for both the beltline region (identified in 1 O CFR 50, Appendix G) and extended beltline region (fluence values >1.0 x 1017 n/cm2,

E >1 MeV).

The beltline region is the region of the RV (shell material, including welds, heat-affected zones, and plate or forgings) that directly surrounds the effective height of the active core and the adjacent regions of the RV that are predicted to experience sufficient neutron irradiation damage to be Page4-10

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses considered in the selection of the most limiting material with regard to radiation damage during the licensed period.

The extended beltline means the region of the RV (shell material, including welds, heat-affected zones, and plate or forgings) adjacent to the beltline region that will have associated fluence values projected to exceed 1.0 x1017 n/cm2 during the subsequent period of extended operation.

The ferritic materials of the RV are subject to embrittlement due to high energy (E > 1.0 MeV) neutron exposure. Embrittlement means the material has lower toughness (i.e., will absorb less strain energy during crack propagation or rupture), thus allowing a crack to propagate more easily under thermal and pressure loading. Neutron embrittlement analyses account for the reduction in fracture toughness associated with the cumulative neutron fluence. Because these neutron embrittlement analyses use a fluence assumption based on the plant's current operating term, they are identified as time-limited aging analyses.

Fracture toughness (indirectly measured in foot-pounds of absorbed energy in a Charpy impact test) is temperature dependent in ferritic materials. An initial nil-ductility reference temperature (RT NOT) is associated with the transition from ductile to brittle behavior and is determined for vessel materials through a combination of Charpy and drop-weight testing. Toughness increases with temperature up to a maximum value called the "upper-shelf energy," or USE. Neutron embrittlement results in the USE decrease of RV steels. This means that RV materials may no longer behave in a ductile manner at postulated plant operating temperatures. For beltline materials the limit for initial USE is 75 ft-lbs. The limit for reduced USE of beltline materials following irradiation is 50 ft-lbs. The material outside the beltline was originally qualified using the requirements of the codes in effect at the time of the initial design and fabrication of the RVs for Units 1 and 2, which were a minimum Charpy impact energy value of 30 ft-lbs at 10°F as specified by ASTM E208, "Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels" (Reference 4.8-5) and required by ASME Code, Section Ill, "Rules for Construction of Nuclear Facility Components" (Reference 4.8-6).

To reduce the potential for brittle fracture during RV operation, changes in material toughness as a function of neutron radiation exposure (fluence) are accounted for during development of operating pressure temperature (P-T) limits that are included in the Technical Specifications. The P-T limits account for the decrease in material toughness of RV materials during plant operation._ Since the cumulative neutron fluence will increase during the subsequent period of extended operation, a review is needed to determine if additional components require evaluation for neutron em brittlement.

10 CFR 50.61 requirements for pressurized thermal shock events specify screening criteria of 270°F for plates, forgings, and axial welds and 300°F for circumferential welds. The RT PTS values have been projected through the subsequent period of extended operation.

Page4-11

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses USE and RT PTS calculations are performed for each beltline and extended beltline material to determine if the components will continue to have adequate fracture toughness with the reduction in toughness resulting from exposure to the predicted neutron fluence. While the decrease in USE for materials in the extended beltline approaches (but remains greater than) 50 ft-lbs, as a conservative measure, an equivalent margins analysis has been performed for the inlet and outlet nozzle welds.

The NRG has approved use of revised initial (unirradiated) RT NDT values and associated uncertainties for Linde 80 weld material. The NRG approved Topical Report BAW-2308 (Revision 1-A) in the "Final Safety Evaluation for Topical Report BAW-2308, Revision 1, 'Initial RT NDT of Linde 80 Weld Materials"' (Reference 4.8-7). Table 3 of the Topical Report contains the revised initial reference temperature (IRT To) and initial margin (I) values for Linde 80 weld materials that are approved by the NRG for the purpose of RV material property determination.

P-T limit curves are generated to provide minimum temperature limits that must be satisfied during operations. The P-T limit curves are based upon the RT NDT and 8RT NDT values computed for the licensed operating period along with appropriate margins.

The enabling temperature and LTOP setpoint are validated as they are impacted by fluence.

The RV material evaluations, calculated on the basis of neutron fluence, are part of the current licensing basis and support safety determinations. Therefore, these calculations have been identified as TLAAs.

The evaluations of TLAAs related to neutron embrittlement are described in the following subsections:

• Neutron Fluence Projections (Section 4.2.1)

• Upper-Shelf Energy (Section 4.2.2)

• Pressurized Thermal Shock (Section 4.2.3)

• Adjusted Reference Temperature (Section 4.2.4)

• Pressure-Temperature Limits (Section 4.2.5)

• Low Temperature Overpressure Protection Analyses (Section 4.2.6)

Page4-12

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses 4.2.1 NEUTRON FLUENCE PROJECTIONS TLAA

Description:

Neutron fluence is the term used to represent the cumulative number of neutrons per square centimeter that contact the RV shell. The fluence projections that quantify the number of neutrons that contact these surfaces have been used as inputs to the neutron embrittlement analyses that evaluate the reduction of fracture toughness aging effect resulting from neutron irradiation and will be treated as a TLAA.

TLAA Evaluation:

Per NUREG-1766, "Safety Evaluation Report Related to the License Renewal of North Anna Power Station, Units 1 and 2, and Surry Power Station, Units 1 and 2" (Reference 4.8-8), RV beltline neutron fluence values applicable to the 60-year period of operation were calculated using the NRC approved VEP-NAF-3-A, "Virginia Power Reactor Vessel Fluence Methodology Topical Report" (Reference 4.8-9). The methodology described in that report was developed in accordance with Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 4.8-10).

EFPY Projections EFPY values for Unit 1 and 2 as of January 5, 2017 are as follows:

Unit 1 Unit 2 33.78 EFPY 33.69 EFPY The first step in updating fluence projections for 80 years is to estimate the power history based upon actual unit operating history and a conservative capacity factor estimate for future cycles.

Units 1 and 2 are licensed for 60 years of operation; therefore, with a 20-year license renewal, the subsequent license renewal term is 80 years.

The EFPY projections through the end of the subsequent period of extended operation for a unit is the sum of the accumulated EFPY and the projected future EFPY. EFPY at the end of 60 years of operation was calculated to be 48 EFPY, assuming a 95% capacity factor for cycles beyond Cycle 19 for Unit 1, and for cycles beyond Cycle 18 for Unit 2. An estimate of the EFPY at the end of 80 years of operation can be made conservatively assuming a 100% capacity factor for the 20-year subsequent period of extended operation. Using this conservative approach the projected 80-year EFPY for both Units 1 and 2 is 68 EFPY.

Fluence Projections Reactor vessel integrity is assured by demonstrating that RV material fracture toughness will remain at levels that resist brittle fracture throughout the subsequent period of extended operation.

The first step in the analysis of vessel embrittlement is calculation of the neutron fluence that causes increased embrittlement.

Page4-13

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Fluence is projected for both beltline and extended beltline materials. The fluence methodology for beltline materials is approved by the NRG SER included in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Reference 4.8-11). NUREG-2191, X.M2, indicates the use of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (Reference 4.8-12) adherent methods to estimate neutron fluence for RV regions significantly above and below the active fuel region of the core and RVI components may require additional justification. Figures 4.2.2-1 and 4.2.2-2 depict the axial boundary of the 1.0 x 1017 n/cm2 fluence in the Z direction. The nozzle shell to intermediate shell circumferential weld is located close to the active fuel region and has historically been treated as beltline material.

The lower extent of the nozzle shell forging, connected to the nozzle shell to intermediate shell circumferential weld, is also beltline material. Fluence projections for these two materials satisfy Regulatory Guide 1.190. The inlet and outlet nozzles are located above the active fuel region.

Some of the inlet and outlet nozzles are projected to experience neutron fluence in excess of 1.0 x 1017 n/cm2. These inlet and outlet nozzles are treated as extended beltline material for subsequent license renewal.

While the fluence projections for the inlet and outlet nozzles may have greater uncertainty than other beltline materials, these fluence projections are acceptable for performing RV integrity assessments for the subsequent license renewal period. The basis for this determination is consistent with LTR-SDA-18-049, "Evaluation of Conservatisms and Margins Associated with Surry Units 1 and 2 Reactor Vessel Integrity Extended Beltline Evaluations for Subsequent License Renewal" (Reference 4.8-13) and LTR-REA-18-75, "Surry Extended Beltline Region Reactor Pressure Vessel Materials Fast Neutron Fluence Sensitivity Study on Material Mixture Above and Below the Active Core" (Reference 4.8-14), and is summarized as follows:

• The fluence at the inlet and outlet nozzles is significantly less than the fluence for the beltline materials, and the highest PTS and ART values are associated with the beltline materials,

• The projected fluence for the postulated flaw for the inlet and outlet nozzles assessed for the P-T Limit curves is based upon the lowest axial extent of the clad/ base metal interface on the inside radial surface of the RV without attenuation,

• Studies to date have shown that the DORT model calculates fluence in the Z direction above the core more conservatively than three-dimensional models such as RAPTOR-M3G,

• The controlling materials for PTS and P-T Limit curves continue to be the beltline materials,

• The fluence projections used in the SLR application conservatively utilized a constant material mixture of 90% water and 10% steel above and below the core. A sensitivity study was performed to show that this assumption was conservative compared to an analysis based upon more representative plant specific material mixture data above and below the C<Jre, Page4-14

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses

• The minimum fluence margin between the beltline materials and inlet and outlet nozzles is 178% (or 2.78 times) and represents the allowable increase in uncertainty that can be tolerated, and

• The SPS-specific minimum fluence margin of 178% (or 2.78 times) represents the increase in fluence uncertainty that is available relative to the customary use of +/-20% margin used in Regulatory Guide 1.190 for fluence projection.

Updated neutron fluence evaluations were performed and documented in WCAP-18028-NP. The fluence methodology used in WCAP-18028-NP is based on nuclear cross-section data derived from Evaluated Nuclear Data File/8 Version VI (ENDF/8-VI). Furthermore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190. The methods used to develop the calculated pressure vessel fluence are consistent with the NRG-approved methodology described in WCAP-14040-A and are documented in the UFSAR, Section 4.1.7.3, "Calculation of Integrated Fast Neutron (E Greater than 1.0 MeV) Flux at the Irradiation Samples." The final safety evaluation report for WCAP-14040, Revision 3, dated February 27, 2004, states that the proposed fluence methodology adheres to the guidance of Regulatory Guide 1.190 and is therefore acceptable. Updated neutron fluence evaluations were used as an input to the RV integrity evaluations in support of initial license renewal.

Consistent with Sections 3.1 and 4.2 of NUREG-2192, materials exceeding a fast neutron fluence (E > 1.0 MeV) of 1.0 x 1017 n/cm2 at the end of the subsequent period of extended operation are evaluated for changes in fracture toughness. Therefore, fast neutron fluence (E > 1.0 MeV) calculations were performed for the Units 1 and 2 RV circumferential welds (lower shell to lower vessel head, intermediate shell to lower shell, and nozzle shell to intermediate shell), inlet and outlet nozzle forging to vessel shell welds at the lowest extent, postulated 1/4T flaw location in the inlet and outlet nozzle, longitudinal welds (lower shell and intermediate shell), and plates (lower shell and intermediate shell), to determine if they will exceed a fast neutron fluence (E > 1.0 MeV) of 1.0 x 1017 n/cm2 at the end of the subsequent period of extended operation. The materials that exceed the 1.0 x 1017 n/cm2 fast neutron fluence (E > 1.0 MeV) threshold are evaluated to determine the effect of neutron irradiation embrittlement during the subsequent period of extended operation.

Table 4.2.1-1 and Table 4.2.1-2 summarize the results of the fluence projections to 68 EFPY for the Units 1 and 2 materials.

Table 4.2.1-1 indicates that some inlet and outlet nozzles have fast neutron fluence (E > 1.0 MeV) greater than 1.0 x 1017 n/cm2 at the nozzle forging to vessel shell weld and one inlet nozzle has fast neutron fluence (E > 1.0 MeV) greater than 1.0 x 1017 n/cm2 at the postulated 1/4T nozzle flaw location at 68 EFPY for Unit 1. Table 4.2.1-2, indicates that some inlet and outlet nozzles have fast neutron fluence (E > 1.0 MeV) greater than 1.0 x 1017 n/cm2 at the nozzle forging to vessel shell Page4-15

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses weld and one outlet and one inlet nozzle have fast neutron fluence (E > 1.0 MeV) greater than 1.0 x 1017 n/cm2 at the 1/4T nozzle flaw location at 68 EFPY for Unit 2. Table 4.2.1-1 and Table 4.2.1-2 indicate that the lower shell to lower vessel head circumferential weld will remain below 1.0 x 1017 n/cm2 through the subsequent period of extended operation for both Units 1 and 2.

TLAA Disposition: 10 CFR 54.21(c)(1)(ii)

The fluence analyses have been projected to the end of the subsequent period of extended operation. The results are to be used as inputs in the RV neutron embrittlement TLAA evaluations in Sections 4.2.2 through 4.2.6.

Page4-16

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.1-1 Unit 1 - Maximum Fast Neutron Fluence (E > 1.0 MeV) Experienced by the Pressure Vessel Materials in the Traditional Beltline and Extended Beltline Regions at 68 EFPY at the Clad/Base Metal Interface Fast Neutron Fluence (n/cm')

Material Region 68 EFPY Applicability Postulated 1/4T Flaw in Outlet Nozzle Nozzle 1 3.45E+16 N/A Nozzle 2 2.49E+16 N/A Nozzle 3 9.62E+16 N/A Postulated1/4T Flaw in Inlet Nozzle Nozzle 1 1.24E+17 Extended Nozzle 2 3.22E+16 N/A Nozzle 3 4.46E+16 N/A Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1 8.13E+16 N/A Nozzle 2 5.86E+16 N/A Nozzle 3 2.27E+17 Extended Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1 3.04E+17 Extended Nozzle 2 7.84E+16 N/A Nozzle 3 1.09E+17 Extended Nozzle Shell 7.54E+18 Traditional Nozzle Shell to Intermediate Shell Circumferential Weld 7.54E+18 Traditional Intermediate Shell Plate 1 6.29E+19 Traditional Plate 2 6.29E+19 Traditional Intermediate Shell Longitudinal Welds Weld 1 1.25E+19 Traditional Weld2 1.25E+19 Traditional Intermediate Shell to Lower Shell Circumferential Weld 6.31E+19 Traditional Plate 2 6.35E+19 Traditional Lower Shell Plate 1 6.35E+19 Traditional Plate 2 6.35E+19 Traditional Page4-17

Material Lower Shell Longitudinal Welds Weld 1 Weld2 Lower Shell to Lower Vessel Head Circumferential Weld Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Fast Neutron Fluence (n/cm')

Region 68 EFPY Applicability 1.26E+19 Traditional 1.26E+19 Traditional

<1E+17 other(a)

(a) Dominion used N/A for situations where it is possible during the life of the plant to reach 1.0 x 1017 n/cm2. EMAs were generated for these situations. The Lower Shell to Lower Vessel Head Circumferential Weld will never be in that category and are identified with a Region Applicability of "Other. EMAs were performed for all N/As.

Page4-18

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.1-2 Unit 2 - Maximum Fast Neutron Fluence (E > 1.0 MeV) Experienced by the Pressure Vessel Materials in the Traditional Beltline and Extended Beltline Regions at 68 EFPY at the Clad/Base Metal Interface Fast Neutron Fluence (n/cm")

Material Region 68 EFPY Applicability Postulated 114T Flaw in Outlet NozzletaJ Nozzle 1 3.38E+16 NIA Nozzle 2 2.48E+16 NIA Nozzle 3 1.07E+17 Extended Postulated 114T Flaw in Inlet Nozzle(a)

Nozzle 1 1.39E+17 Extended Nozzle 2 3.21E+16 NIA Nozzle 3 4.37E+16 NIA Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1 7.96E+16 NIA Nozzle 2 5.85E+16 NIA Nozzle 3 2.53E+17 Extended Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1 3.40E+17 Extended Nozzle 2 7.84E+16 NIA Nozzle 3 1.07E+17 Extended Nozzle Shell 8.65E+18 Traditional Nozzle Shell to Intermediate Shell Circumferential Weld 8.65E+18 Traditional Intermediate Shell Plate 1 7.20E+19 Traditional Plate 2 7.20E+19 Traditional Intermediate Shell Longitudinal Welds Weld 1 1.29E+19 Traditional Weld2 1.29E+19 Traditional Intermediate Shell to Lower Shell Circumferential Weld 7.22E+19 Traditional Lower Shell Plate 1 7.26E+19 Traditional Plate 2 7.26E+19 Traditional Page4-19

Material Lower Shell Longitudinal Welds Weld 1 Weld2 Lower Shell to Lower Vessel Head Circumferential Weld Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Fast Neutron Fluence (n/cm:.:)

Region 68 EFPY Applicability 1.30E+19 Traditional 1.30E+19 Traditional

<1E+17 Otherl0J (a) Nozzle 1/4T flaw maximum fluence valves are taken at the surface of the nozzle.

(b) Dominion used NIA for situations where it is possible during the life of the plant to reach 1.0 x 1017 n/cm2. EMAs were generated for these situations. The Lower Shell to Lower Vessel Head Circumferential Weld will never be in that category and are identified with a Region Applicability of "Other. EMAs were performed for all N/As.

4.2.2 UPPER-SHELF ENERGY TLAA

Description:

Upper-shelf energy (USE) is the parameter used to indicate the toughness of a material at elevated temperature. There are two sets of rules that govern USE acceptance criteria. 1 O CFR 50, Appendix G, Paragraph IV.A.1.a, states that RV beltline materials must have Charpy USE of no less than 75 ft-lb initially, and must maintain Charpy USE throughout the life of the vessel of no less than 50 ft-lb, unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation, that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," (Reference 4.8-15) Appendix G, "Fracture Toughness Criteria for Protection Against Failure." For materials outside the beltline, a minimum value of 30 ft-lbs at 10°F was specified by ASTM E208, and required by ASME Code, Section Ill, at the time of the design and fabrication of the RVs for Units 1 and 2.

The current licensing basis Charpy USE calculations were prepared for the Units 1 and 2 RV beltline materials for 48 EFPY. Since the USE value is a function of 48 EFPY fluence, associated with the 60-year licensed operating period, these USE calculations meet the criteria of 1 O CFR 54.3(a) and have been identified as TLAAs requiring evaluation for 80 years.

TLAA Evaluation:

Per Regulatory Guide 1.99,"Radiation Embrittlement of Reactor Vessel Materials,"

(Reference 4.8-16) the Charpy USE should be assumed to decrease as a function of fluence according to Figure 2 of the Regulatory Guide, which provides percent decrease in USE as a function of 1/4T fluence and the copper content for plates and welds, when credible surveillance data is not available. If credible surveillance data is available, the decrease in USE may be obtained by plotting the reduced plant surveillance data on Figure 2 of Regulatory Guide 1.99 and fitting the data with a line drawn parallel to the existing lines as the upper bound of all of the data. The 1/4T fluence at 68 EFPY is used to determine the reduction of the initial USE.

Page4-20

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses As documented in WCAP-18242-NP, the materials projected to exceed 1.0 x 1017 n/cm2 (E > 1.0 MeV) at 68 EFPY are evaluated to determine their impact on USE during the proposed subsequent period of extended operation. The forgings and welds corresponding to some inlet and outlet nozzles are predicted to experience neutron fluence greater than 1. 0 x 1017 n/cm2 at the end of the subsequent period of extended operation. However, for conservatism all of the inlet and outlet nozzle materials are considered part of the extended beltline in the USE evaluation. The Units 1 and 2 materials include three (3) inlet nozzles, three (3) outlet nozzles, three (3) inlet nozzle to upper-shell welds, and three (3) outlet nozzle to upper-shell welds per unit. (Note: nozzle-shell and upper-shell refer to the same component and are used interchangeably).

The identification of the RV plate and weld materials is shown in Figure 4.2.2-1 for Unit 1 and Figure 4.2.2-2 for Unit 2. The material property inputs used for the RV integrity evaluations are described in this section. The initial material properties were updated from previous RV integrity evaluations per PWROG-16045-NP and WCAP-18242-NP, Appendix E, and the fluence values were updated per WCAP-18028-NP and WCAP-18242-NP, Section 2. Additionally, initial USE values are supplied in Table 4.2.2-1 and Table 4.2.2-3.

The requirements on USE for beltline materials are included in 1 O CFR 50, Appendix G, which requires utilities to submit an analysis at least three years prior to the time that the USE of any RV material is predicted to drop below 50 ft-lb. Dominion has conservatively elected to perform equivalent margins analyses (EMAs) for inlet and outlet nozzle welds with Charpy USE near 50 ft-lb at the end of the subsequent period of extended operation.

Two methods can be used to predict the decrease in USE with irradiation, depending on the availability of credible surveillance capsule data as defined in Regulatory Guide 1.99. For vessel beltline materials that are not in the surveillance program or have non-credible data, the Charpy USE (Position 1.2) is assumed to decrease as a function of fluence and copper content, as indicated in Regulatory Guide 1.99. When two or more credible surveillance sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material. The surveillance data are then used in conjunction with Regulatory Guide 1.99 to predict the change in USE (Position 2.2) of the RV material due to irradiation. Per Regulatory Guide 1.99 (Revision 2),

when credible data exists the Position 2.2 projected USE value should be used in preference to the Position 2.1 projected USE value. Such cases exist in Table 4.2.2-5 wherein SLR USE values in the Position 1.2 section that fall below 50 ft-lbs are not an issue because corresponding values in the Position 2.2 section are above 50 ft-lbs when considering credible surveillance data.

The 68 EFPY Position 1.2 USE values of the vessel materials can be predicted using the corresponding fluence projections (1/4T for beltline materials and surface for inlet/outlet nozzles),

the copper content of the materials, and Figure 2 in Regulatory Guide 1.99.

Page4-21

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses The predicted Position 2.2 USE values are determined for the RV materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding fluence projection (1/4T for beltline materials and surface for inlet/outlet nozzles). The reduced plant surveillance data was obtained from Table 7-6 of BAW-2324, "Analysis of Capsule X Virginia Power Surry Unit No. 1, Reactor Vessel Material Surveillance Program" (Reference 4.8-17) for Unit 1. The reduced plant surveillance data was obtained from Table 5-12 of WCAP-16001, "Analysis of Capsule Y from Dominion Surry Unit 2 Reactor Vessel Radiation Surveillance Program" (Reference 4.8-18) for Unit 2. The surveillance data was plotted in Regulatory Guide 1.99, Figure 2 using the surveillance capsule fluence values documented in Table 2-1 of WCAP-18242-NP, for Unit 1 and Table 2-2 of WCAP-18242-NP, for Unit 2.

The projected USE values were calculated to determine if the values for Units 1 and 2 materials remain above the 50 ft-lb criterion at 68 EFPY. The projected USE values for the inlet and outlet nozzle forgings were conservatively calculated using the maximum fluence values corresponding to the lowest extent of the nozzle to shell welds. These calculations are summarized in Table 4.2.2-5 and Table 4.2.2-6.

Conclusion For Unit 1, the limiting USE value at 68 EFPY is 32 ft-lb (see Table 4.2.2-5); this value applies to the Intermediate to Lower Shell Circumferential Weld using Position 1.2. For Unit 2, the limiting USE value at 68 EFPY is 41 ft-lb (see Table 4.2.2-6); this value applies to the Upper to Intermediate Shell Circumferential Weld using Position 1.2.

The NRC has previously approved the use of the equivalent margins analysis (EMA) BAW-2494, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Surry Power Station Units 1 and 2 for Extended Life through 48 Effective Full Power Years" (Reference 4.8-19) to qualify all of the materials currently projected to drop below 50 ft-lb USE at 68 EFPY. These materials are identified by the notes in Table 4.2.2-1, Table 4.2.2-3, Table 4.2.2-5, Table 4.2.2-6 herein and are summarized below. The EMAs for these materials are updated for the subsequent period of extended operation under ANP-3679NP, "Low Upper-Shelf Toughness Fracture Mechanics Analysis for Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at 80 Years," (Reference 4.8-20) and ANP-3680NP, "Low Upper-Shelf Toughness Fracture Mechanics Analysis for Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at 80 Years" (Reference 4.8-21). The updated EMA is based upon the provisions outlined in ASME Code,Section XI, Appendix K. The selection of design transients for Levels C & D service loads are based on the guidance in Regulatory Guide 1.161, "Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 Ft-Lb." (Reference 4.8-22) and ASME Code,Section XI, Appendix K.

Page4-22

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses An EMA should be submitted three years before a material is projected to drop below 50 ft-lbs; however, no additional materials are projected to drop below 50 ft-lb USE during the subsequent period of extended operation.

The following Unit 1 and Unit 2 materials are addressed by EMAs for the subsequent period of extended operation:

Unit 1:

• Upper to Intermediate Shell Circumferential Weld, Heat# 25017 (J726)

• Intermediate Shell Longitudinal Welds L3 and L4, Heat# 8T1554

• Intermediate to Lower Shell Circumferential Weld, Heat# 72445

• Lower Shell Longitudinal Weld L 1, Heat# 8T1554

• Lower Shell Longitudinal Weld L2, Heat# 299L44

• Inlet Nozzle to Shell Welds, Heat# 299L44 and# 8T1762; (Projected USE > 50 ft-lbs at 68 EFPY)

• Outlet Nozzle to Shell Welds, Heat # 8T1762 and # 8T15548; (Projected USE > 50 ft-lbs at 68 EFPY)

Unit 2:

• Upper to Intermediate Shell Circumferential Weld, Heat# 4275 (J737)

• Intermediate Shell Longitudinal Welds L3 and L4, Heat# 72445

• Intermediate Shell Longitudinal Weld L4, Heat# 8T1762

• Intermediate to Lower Shell Circumferential Weld, Heat# 0227

• Lower Shell Longitudinal Weld L 1 and L2, Heat# 8T1762

• Inlet Nozzle to Shell Welds, Heat# 8T1762; (Projected USE not projected > 50 ft-lbs at 68 EFPY)

• Outlet Nozzle to Shell Welds, Rotterdam Weld; (Projected USE > 50 ft-lbs at 68 EFPY)

An EMA has been completed for the Unit 1 and Unit 2 Inlet and Outlet Nozzle to Shell Welds even though these materials are not projected to drop below 50 ft-lbs through 68 EFPY using the methods herein. The inlet and outlet nozzle welds are the only materials included in ANP-3679NP and ANP-3680NP that were not previously addressed by EMA. The EMA is applicable to the Page4-23

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Units 1 and 2 nozzle to shell welds which exceed the fluence criterion of 1.0 x 1017 n/cm2 before 68 EFPY. These materials include those listed below.

• Unit 1 Outlet Nozzle 1 to Upper Shell Weld

• Unit 1 Inlet Nozzle 1 to Upper Shell Weld

• Unit 1 Inlet Nozzle 3 to Upper Shell Weld

• Unit 2 Outlet Nozzle 1 to Upper Shell Weld

• Unit 2 Inlet Nozzle 1 to Upper Shell Weld

• Unit 2 Inlet Nozzle 3 to Upper Shell Weld For Unit 1, the limiting USE value for materials not requiring an EMA at 68 EFPY is 54 ft-lb (see Table 4.2.2-5); this value corresponds to the Inlet Nozzle to Upper Shell Welds (Heat# 299L44) using Position 2.2. For Unit 2, the limiting USE value for materials not requiring an EMA at 68 EFPY is also 54 ft-lb (see Table 4.2.2-6); this value corresponds to the Outlet Nozzle to Upper Shell Welds (Rotterdam) using Position 1.2. Except for the materials listed above, all of the beltline and extended beltline materials in the Units 1 and 2 RVs are projected to remain above the USE screening criterion value of 50 ft-lb (per 1 O CFR 50, Appendix G) through the subsequent period of extended operation (68 EFPY).

Equivalent Margins Analysis The ASME Code,Section XI, acceptance criteria for Levels A through D Service Loadings for all Units 1 and 2 RV beltline and extended beltline Linde 80 welds are satisfied and are reported in Framatome Reports BAW-2192, Supplement 1 (Revision 0), "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels A & B Service Loads Topical Report," (Reference 4.8-23) and BAW-2178, Supplement 1 (Revision 0), "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Levels C & D Service Loads Topical Report,"

(Reference 4.8-24) submitted to the NRG in December 2017. The Surry Power Plant specific versions of the EMA are documented in following reports:

• ANP-3679P, Revision 0, "Low Upper-Shelf Toughness Fracture Mechanics Analysis for Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at SO-Years" (Reference 4.8-25)

• ANP-3679NP, Revision 0, "Low Upper-Shelf Toughness Fracture Mechanics Analysis for Surry Units 1 and 2 Reactor Vessels for Levels A & B Service Loads at SO-Years"

• ANP-3680P, Revision 0, "Low Upper-Shelf Toughness Fracture Mechanics Analysis for Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at SO-Years" (Reference 4.8-26)

• ANP-3680NP, Revision 0, "Low Upper-Shelf Toughness Fracture Mechanics Analysis for Surry Units 1 and 2 Reactor Vessels for Levels C & D Service Loads at SO-Years" Page4-24

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses The plant-specific EMA reports contain the same information as in BAW-2192, Supplement 1 and BAW-2178, Supplement 1 except that the information for Oconee 1, 2, and 3 and Turkey Point 3 and 4 has been removed.

The 80-year clad/base metal fluence values reported in Table 3 -1 of BAW-2178, Supplement 1, and Table 3-1 of BAW-2192, Supplement 1 have been confirmed to bound the 68 EFPY fluence values reported in Table 4.2.1-1 and Table 4.2.1-2. The EMAs conservatively utilized 80-year fluence values shown in (e) of at least an order of magnitude higher than the 68 EFPY nozzle fluence reported in Table 4.2.1-1 and Table 4.2.1-2. In addition, the weld chemistry data reported in Table 3-1 of BAW-2178, Supplement 1, and Table 3-1 of BAW-2192, Supplement 1 is consistent with weld chemistry reported in Tables 4.2.2-1 through Table 4.2.2-4. The level C and D limiting design transients reported in Section 4.3.2 of BAW-2178P, Supplement 1, are applicable to Units 1 and 2 and are based on a review of the ASME Code, Section Ill, Reactor Vessel Design Specification transients and the UFSAR Chapter 14 events relative to transients that would result in the highest thermal stresses coupled with pressure stresses relative to the EMA analysis; this satisfies Regulatory Guide 1.161 with respect to Level C and D transient selection. The materials of construction, RV geometry, and range of explanatory variables for the J-R model (Section A.5 of BAW-2192, Supplement 1) reported in the topical reports are confirmed to be applicable to Linde 80 and Rotterdam beltline and extended beltline welds at Units 1 and 2.

As such, Units 1 and 2 are bounded by topical report submittals BAW-2178, Supplement 1, and BAW-2192, Supplement 1 relative to fluence, weld chemistry, geometry, materials of construction, design transients and the J-R model applicability. The results of the EMA for Units 1 and 2, as reported in BAW-2178 P/NP and BAW-2192 P/NP, are summarized below.

Levels A & 8 Service Loadings Reactor Vessel Shell Welds (Beltline)

The limiting RV shell weld is Unit 1 axial weld SA-1526.

• With factors of safety of 1.15 on pressure and 1.0 on thermal loading, the applied J-integral (J1) is less than the J-integral of the material at a ductile flaw extension of 0.10 in. (J0.1 ). The ratio J0.1/J1 is greater than the required value of 1.0.

• With a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect.

Page4-25

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Reactor Vessel Transition Welds and RV Nozzle Welds (Extended Beltline)

• The limiting weld for Units 1 and 2 considering RV transition welds (upper and lower) and the RV inlet and outlet nozzle-to-shell welds is the longitudinal weld SA-1585 near the base of the transition section.

• With factors of safety of 1.15 on pressure and 1.0 on thermal loading, the applied J-integral (J 1) is less than the J-integral of the material at a ductile flaw extension of 0.1 O in. (J0.1 ). The ratio J0.1/J1 is greater than the required value of 1.0.

• With a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect.

Levels C & D Service Loadings Reactor Vessel Shell Welds (Beltline)

The limiting weld among the RV shell welds is Unit 1 longitudinal weld SA-1526. The limiting transient for Level C & D service Loads is the SSDC 1.3 steam line break.

• With a factor of safety of 1.0 on loading, the applied J-integral (J1) for the limiting RV shell weld (Unit 1, SA-1526) is less than the lower bound J-integral of the material at a ductile flaw extension of 0.10 inch (J0.1) with a ratio J0.1/J1 that is greater than the required value of 1.0.

• With a factor of safety of 1.0 on loading, flaw extensions are ductile and stable for the limiting RV shell weld (SA-1526) since the slope of the applied J-integral curve is less than the slopes of both the lower bound and mean J-R curves at the points of intersection.

• For weld SA-1526 it was demonstrated that flaw growth is stable at much less than 75% of the vessel wall thickness. It has also been shown that the remaining ligament is sufficient to preclude tensile instability.

Page4-26

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Reactor Vessel Transition Welds and RV Nozzle Welds (Extended Beltline)

The upper transition weld and RV inlet and outlet nozzle-to-shell welds were evaluated for Levels C and D Service Loadings. The limiting transient for Level C & D service loads is the SSDC 1.3 steam line break.

• With a factor of safety of 1.0 on loading, the applied J-integral (J1) for the RV nozzle-to-shell welds and upper transition weld are less than the lower bound J-integral of the material at a ductile flaw extension of 0.1 O inch (J0.1). All ratios are greater than 1.0.

• With a factor of safety of 1.0 on loading, flaw extensions are ductile and stable for the limiting RV outlet nozzle-to-shell weld (i.e., limiting location considering RV nozzle-to-shell welds and upper transition weld).

• For the RV outlet nozzle-to-shell weld it was demonstrated that flaw growth is stable at much less than 75% of the vessel wall thickness. Tensile instability was not explicitly calculated but because this section of the RV is thicker compared to the RV shell welds, it is considered to be bounded by the RV shell location.

B&WOG J-R Model The original 8&WOG J-R Model 48 reported in 8AW-2192PA, Supplement 1, Appendix A, was used to obtain J material (i.e., J(0.1)) for the 80-year equivalent margins analyses reported in 8AW-2192, Supplement 1, and 8AW-2178, Supplement 1. Model 48 is based on fracture toughness data (1352 J delta-a data points) irradiated to a fluence ranging from 0.0 to 8.45 x 1018 n/cm2, which is less than the peak 1/1 OT 80-year fluence projected for Units 1 and 2. To further substantiate the use of the 8&WOG J-R model, the original J delta-a data used to generate the 8&WOG J-R model 48 was used to independently benchmark the original 8&WOG model using the R-project statistical tool. The benchmark is designated 8&WOG J-R Model 58. New J-R data (419 new J delta-a data points with fluence to 5.8 x 1019 n/cm2) were then added to the original population of welds (total population of 177 4 data points) and the fitting coefficients (assuming the same model form) were generated. The 8&WOG model that includes the total population of J delta-a data (1774) is designated Model 68. Model 68 is based on test data out to a fluence of 5.8 x 1019 n/cm2, which is greater than the peak 1/4T fluence of 8.16 x 1018 n/cm2 and 1/10T fluence of 1.083 x 1019 n/cm2 for Units 1 and 2 limiting weld SA-1526.

Page4-27

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Use of Model 68 for fluence values in excess of 5.8 x 1019 n/cm2 is considered to be a model extrapolation and the uncertainty may increase (i.e., -2SE). Fluence estimates at T/4 and T/10 are well below 5.8 x 1019 n/cm2 and the J-R Model is used well within the interpolation range (i.e., for weld SA-1526 fluence equals 8.16 x 1018 n/cm2 at 1/4T and 1.083 x 1019 n/cm2 at T/10). For Units 1 and 2, use of Model 68 (model extrapolation) increased the J(0.1)/J1 by approximately 6%

for Level A and 8, and 5% for Level C and D when compared to Model 48, and, all margins remain above the acceptance criterion of 1.0. In addition, the combination of Level C and D acceptance criteria applied to Level D transients provides additional conservatism in the equivalent margins analyses. The 8&WOG J-R models (including Models 48 and 68) are discussed in BAW-2192, Supplement 1, Appendix A.

TLAA Disposition: 10 CFR 54.21(c)(1)(ii)

The USE analyses have been projected to the end of the subsequent period of extended operation.

Page4-28

Figure 4.2.2-1 400 326.7 300 277.o * **

272.7****

261.9

• 2S6.2 **

200 182.88 (top of 100 I

C

.2 i 0

,! (core m

~

~

-100 e) lntermedrate 5heU (H*tNo.

C43.26-1) planeJ

- 182.88 (bottom o core)

-200 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Unit 1 - Axial Boundary of the 1.0E+17 n/cm2 Fluence Threshold in the +Z Direction (at 54 and 72 EFPY)

Nozzle 9ietl Forglrg (HeatNo.122V109VA1) 1E+17 n/ cm 2 t hreshold at 72 EFPY

-- 1E+17 n/ cm 2 th reshold at 54 EFPY Nozzle Sh ell to Intermediat e Shell Cirwmfenntiill Weld (Heat No. 25017)

~

~

lntermedare 3'lel (Heat No. C4326-2) lnterm edl.t t e She ll t o Lower Sh(!fl Circumferen tial W eld (Heat No. 72445 )

Lawer Shell (Heat No. C4--41~2)

Lower !hell (Heat No. C441~1) lntermedl'ate 9"lell (H*t No. C4326-1)

Lower Shell t o Lo wer V essel Head Circu mfere n th1I W e ld (Hea t Nos. 25017 & 721858 )

Lower 9iell (Heat No. C441~2)

-300.------------------...

-400 45 90 135 180 225 Azimuthal Location (desr**l O ut let Nozzle Forelnc to Vessel Shell Welds - Lowest Extent Inlet Nozzle Forelnc to Vessel Shell Welds - Lowest Extent 1/4 T Flaw Location in Outlet Nozzle

• * *

• 1/4 T Flaw Location In Inlet Nozzle Page 4-29 270 315 360

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Figure 4.2.2-2 Unit 2 - Axial Boundary of the 1.0E+17 n/cm2 Fluence Threshold in the +Z Direction (at 54 and 72 EFPY) 400 1E+l7 n/cm2 t h reshold at 7 2 EFPY

--1E+17 n/cm2 t h reshold at 54 EFPY 326.7 (Nozzle 300 277.o***

272.7****

261.9*

256.2**

200 182.88 (top of c re)

I C:

.2 100 lntermed'4rte Shell

( Heirt No.

C-4331-.2 )

0 l; ( core id plane) w Nonie Shel For,ging (He.al No. 12JV303VA1)

Nou:le Shell to Intermediate She ll Clrcumferent le l W e ld {Heat No. 4275) lntermedbte Shell (He at No. C4339-2)

J I ntermediate Shell to Lower Shell Circumferential Weld (Heat No. 0227)

-100

- 182.88 (bottom o core)

-200 Lower Shell (He.st No. C-4208-2) lower Shell (I-le~ No. C43l9-1) lnte m1edlate Shell (I-teat N o. C433l.-2)

-300 Lower Shell to Lower Vessel Head Circumferential Weld (Heat No. 0227)

-400 45 90 135 180 225 Azimutha l Location (degree)

Ou tlet Nozzle Fo rging t o Vessel Shell W e lds - l o w est Extent In le t N o zzle Fo rgin g t o Vessel She ll W e lds - Lo west Extent 1/4 T Fla w Lo catio n in Outlet No zzle

• 1/4 T Flaw Location in Inlet Nozzle Page4-30 270 315 Lawe, Shell (Heat No. C4 208-2) 360

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Figure 4.2.2-3 Reactor Vessel - Unit 1

~~K~""'.""---....:.__ NqzzletQ.. Sheil Wetd>lr 1-'-ir-1l.sfn,o"F** ---r~t""'----...;.;..:_.:_;__Jt26* (Rotterda"l) Weld

.CORE,:

I r..::::>--~ Weld SA.;;1494*

I

~- lnt~nne.diat~ ~hell (Pl~te)

---

1------ Weld sA:.-1494"

~-+---+----- W~ld-$A..;1526~

Lo'A'~r Shell (Pl~t~)

• Equivalent Margins* Arta1Sf.$is perform.ed fQr ttiese Linde 80 ar,d RoltfJrdam Welds~

Page4-31

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Figure 4.2.2-4 Reactor Vessel - Unit 2 L---L--~J....~..;;__.----.L737* (RotterdamrWetd 9*0"',*,

~t--J...;...._ ___.iiietd *s' 'k.;.15*5' **5*~-

' :coRE' yv,. -

.. rt.

r **.. * *. -

Weld SA-15i3ffOutslde.50o/o

'-1 ___ ___,---4-;...__--1---1---1-------"'*

... wi= -4' tnt;ide-50%

• 1M" * : ft.--19-'--;,,,..--*.------

~-'--------'--~ l11l~mietjJ~t.e*$heU (Pl1:1te) 1,--r--,--1--.--+--.-..1..-..,..,..__ R3008* (Rotterdam) Weld i-.--+-+------ Weld WF-4*.

.... 1 ------1-----1-----1---1-----l we1d WF,.. 4* inside: Sa%.

WF ~.s Qut$!qe at%

J LpWer Sh~II (Plate)

• Equiva1en,t Margins fm~ilysis.perf9rmed fQri'1~e Ling(f SQ and Rotterdam Weld$.

Page4-32

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-1 Best Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Unit 1 RV Beltline and Surveillance Materials RV Material Wt.%

Wt.%

RTNDT(U) 0'1 Initial USE Cu Ni (OF)

(OF)

(ft-lb)

RV Beltline Materials(a)

Upper Shell Forging 122V109VA1 0.11 0.74 40 0

114 Intermediate Shell Plate C4326-1 0.11 0.55 10 0

115 Intermediate Shell Plate C4326-2 0.11 0.55 11.4 0

94 Lower Shell Plate C4415-1 0.102 0.493 20 0

103 Lower Shell Plate C4415-2 0.11 0.5 4.6 0

82 Upper to Intermediate Shell 2: 54(b)

Circumferential Weld 0.33 0.1 0

20 (Heat# 25017)

Intermediate Shell Longitudinal Welds 0.16 0.57

-48.6 18 54(b)

L3 and L4 (Heat# 8T1554)

Intermediate to Lower Shell 54(b)

Circumferential Weld 0.22 0.54

-72.5 12 (Heat # 72445)

Lower Shell Longitudinal Weld L 1 0.16 0.57

-48.6 18 54(b)

(Heat# 8T1554)

Lower Shell Longitudinal Weld L2 0.34 0.68

-74.3 12.8 54(b)

(Heat# 299L44)

RV Surveillance Materials(c)

Lower Shell Plate C4415-1 0.102 0.493 20 0

103 Surveillance Weld (Heat# 299L44) 0.23 0.64 70 Notes:

(a)

All values were taken from Table 8 of PWROG-16045 NP, unless otherwise noted.

(b)

Per UFSAR (Tables 4.1-14 and 4.1-15), RV Equivalent Margins Analysis (EMA) report BAW-2494, was approved for these welds for 48 EFPY. The EMAs are updated for the subsequent period of extended operation. Linde 80 initial USE values are set to the generic value of 64 ft-lbs per BAW-2313, Supplement 1. Only limited Charpy test information is available for Heat# 25017.

Based on the average Charpy energy value of the weld qualification tests completed at 10°F, the USE for Heat# 25017 is at least 64 ft-lbs. This value of 64 ft-lbs is conservative compared to the generic Rotterdam-weld results documented in PWROG-17090-NP, "Generic Rotterdam Forging and Weld Initial Upper-Shelf Energy Determination." (Reference 4.8-27)

(c)

The surveillance plate data was taken to be the same as the vessel plate data. The surveillance weld data was obtained from BAW-2324.

Page4-33

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-2 Best Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Unit 1 RV Materials Wt.%

Wt.%

RTNDT(U)

Initial RV Material 0'1 USE Cu Ni

(°F)

(°F)

(ft-lb)

RV Extended Beltline Materials(a)

Inlet Nozzle 1 (Heat# 9-4787) 0.159 0.85 10.3 0

63 Inlet Nozzle 2 (Heat# 9-5078) 0.159 0.87 11.6 0

64 Inlet Nozzle 3 (Heat# 9-4819) 0.159 0.84

-47.2 0

68 Outlet Nozzle 1 (Heat# 9-4825-1) 0.159 0.85

-44.9 0

68 Outlet Nozzle 2 (Heat# 9-4762) 0.159 0.83

-87.5 0

82 Outlet Nozzle 3 (Heat# 9-4788) 0.159 0.84

-50.2 0

71 Inlet Nozzle to Upper Shell Heat# 299L44 0.34 0.68

-7 20.6 64 Welds Heat# 8T1762 0.19 0.57

-4.9 19.7 64 Heat # 8T1762 0.19 0.57

-4.9 19.7 64 Outlet Nozzle to Upper Shell Welds Heat # 8T1554 0.16 0.57

-4.9 19.7 64 B

Note:

(a) All values were taken from Table 8 of PWROG-16045-NP.

Page4-34

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-3 Best Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Unit 2 RV Beltline and Surveillance Materials RV Material Wt.%

Wt.%

RTNDT(U) 0'1 Initial USE Cu Ni

(°F)

(°F)

(ft-lb)

RV Beltline Materials1e1, Upper Shell Forging 123V303VA1 0.11 0.72 30 0

104 Intermediate Shell Plate C4331-2 0.12 0.6 15 0

84 Intermediate Shell Plate C4339-2 0.11 0.54 7.8 0

83 Lower Shell Plate C4208-2 0.15 0.55

-30 0

94 Lower Shell Plate C4339-1 0.107 0.53

-4.4 0

101 Upper to Intermediate Shell Circumferential 0.35 0.1 0

20

.:: 68<b)

Weld (Heat# 4275)

Intermediate Shell Longitudinal Welds 0.22 0.54

-72.5 12 64<b)

L3 and L4 (OD 50%) (Heat# 72445)

Intermediate Shell Longitudinal Weld 0.19 0.57

-48.6 18 64<b)

L4 (ID 50%) (Heat# 8T1762)

Intermediate to Lower Shell Circumferential 0.187 0.545 o(c) o(c) 82(c)

Weld (Heat # 0227)

Lower Shell Longitudinal Welds L 1 and L2 0.19 0.57

-48.6 18 64<b)

(Heat# 8T1762)

RV Surveillance Materialsl0J Lower Shell Plate C4339-1 0.107 0.53

-4.4 0

101 Surveillance Weld (Heat# 0227) 0.19 0.56 91 Notes:

(a)

All values were taken from Table 9 of PWROG-16045-NP, unless otherwise noted.

(b)

Per UFSAR (Tables 4.1-14 and 4.1-15), RV EMA report BAW-2494 was approved for these welds for 48 EFPY. The EMAs are updated for the subsequent period of extended operation. Linde 80 initial USE values are set to the generic value of 64 ft-lbs per BAW-2313, Supplement 1, "Supplement to B&W Fabricated Reactor Vessel Materials and Surveillance Data Information for Surry Unit 1 and Unit 2" (Reference 4.8-28). Only limited Charpy test information is available for Heat # 4275. Based on the average Charpy energy value of the weld qualification tests completed at 10°F, the USE for Heat# 4275 is at least 68 ft-lbs. This value of 64 ft-lbs is conservative compared to the generic Rotterdam weld results documented in PWROG-17090-NP.

(c)

Initial properties are established in Appendix 8 of WCAP-18242-NP. Since the initial RT NOT is based on measured data, au is equal to 0°F. Per UFSAR (Tables 4.1-14 and 4.1-15), RV EMA report BAW-2494 was approved for this weld for 48 EFPY. The EMA is updated for the subsequent period of extended operation.

(d)

The surveillance plate data was taken to be the same as the vessel plate data. The surveillance weld data was obtained from WCAP-16001.

Page4-35

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-4 Best Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Unit 2 RV Extended Beltline Materials Wt.%

Wt.%

RTNDT(U)

Initial RV Material 0"1 USE Cu Ni

(*F)

(°F)

(ft-lb)

RV Extended Beltline Materials<a)

Inlet Nozzle 1 (Heat# 9-5104) 0.159 0.84

-29.7 0

73 Inlet Nozzle 2 (Heat# 9-4815) 0.159 0.87 4.5 0

66 Inlet Nozzle 3 (Heat# 9-5205) 0.159 0.86 6.5 0

67 Outlet Nozzle 1 (Heat # 9-4825-2) 0.159 0.85

-58.1 0

73 Outlet Nozzle 2 (Heat# 9-5086-1) 0.159 0.86

-26.6 0

77 Outlet Nozzle 3 (Heat # 9-5086-2) 0.159 0.87

-33.8 0

71 Inlet Nozzle to Upper Shell Heat # 8T1762 0.19 0.57

-4.9 19.7 64 Welds Outlet Nozzle to Upper Shell Rotterdam 0.35 1

30 0

71(b)

Welds Notes:

(a)

All values were taken from Table 9 of PWROG-16045-NP. Associated a, values are also available from PWROG-16045-NP.

(b)

Per PWROG-16045-NP, this initial USE value is set equal to the USE value of the first tested capsule from WCAP-16001 (Reference 4.8-18). This methodology utilizes BTP 5-3 (Reference 4.8-29), Position 1.2 guidance, as no USE data is available from the supplier. The value used herein is conservative in comparison. In addition, Dominion conservatively elected to complete an EMA on this material.

Page4-36

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-5 Predicted USE Values at 68 EFPY for Unit 1 Wt.%

SLR 1/4T Initial USE{a) Projected USE SLR RV Material cu(a)

Fluence(b)

(ft-lb)

Decrease(c)

USE (x 1019 n/cm2)

(%)

(ft-lb)

Position 1.2 Upper Shell Forging 122V109VA1 0.11 0.465 114 17 95 Upper to Intermediate Shell Circumferential Weld(e) (Heat # 25017) 0.33 0.465 64 39 39(e)

Intermediate Shell Plate C4326-1 0.11 3.88 115 28 83 Intermediate Shell Plate C4326-2 0.11 3.88 94 28 68 ntermediate saell Longitudinal Welds L3 and L4 e) (Heat# 8T1554) 0.16 0.771 64 29 45(e)

Intermediate to (eower Shell Circumferential Weld e) (Heat# 72445) 0.22 3.89 64 50 32(e)

Lower Shell Plate C4415-1 0.102 3.92 103 27 75 Lower Shell Plate C4415-2 0.11 3.92 82 28.5 59 Lower Shell Longitudinal Weld L 1 (eJ 0.16 0.777 64 29 45(e)

(Heat# 8T1554)

Lower Shell Longitudinal Weld L2teJ 0.34 0.777 64 41 38(e)

(Heat# 299L44)

Inlet Nozzle 1 to Upper Shell Weld 0.34 0.0188 64 24 49(f)

(Heat# 299L44)

Inlet Nozzle 2 to Upper Shell Weld 0.34 0.00484 64 24 49(f)

(Heat# 299L44)

Inlet Nozzle.3 to Upper Shell Weld 0.34 0.00672 64 24 49(f)

(Heat# 299L44)

Inlet Nozzle 1 to Upper Shell Weld 0.19 0.0188 64 13 56 (Heat # 8T1762)

Inlet Nozzle 2 to Upper Shell Weld 0.19 0.00484 64 13 56 (Heat # 8T1762)

Inlet Nozzle 3 to Upper Shell Weld 0.19 0.00672 64 13 56 (Heat # 8T1762)

Outlet Nozzle 1 to Upper Shell Weld 0.19 0.00502 64 13 56 (Heat # 8T1762)

Outlet Nozzle 2 to Upper Shell Weld 0.19 0.00362 64 13 56 (Heat # 8T1762)

Outlet Nozzle 3 to Upper Shell Weld 0.19 0.0140 64 13 56 (Heat # 8T1762)

Outlet Nozzle 1 to Upper Shell Weld 0.16 0.00502 64 12 56 (Heat # 8T15548)

Outlet Nozzle 2 to Upper Shell Weld 0.16 0.00362 64 12 56 (Heat# 8T15548)

Page4-37

RV Material Outlet Nozzle 3 to Upper Shell Weld (Heat # 8T1554B)

Inlet Nozzle 1 (Heat# 9-4787)

Inlet Nozzle 2 (Heat# 9-5078)

Inlet Nozzle 3 (Heat# 9-4819)

Outlet Nozzle 1 (Heat# 9-4825-1)

Outlet Nozzle 2 (Heat# 9-4762)

Outlet Nozzle 3 (Heat# 9-4788)

Lower Shell Plate C4415-1 Lower Shell Plate C4415-2 Lower Shell Longitudinal Weld L2leJ (Heat# 299L44)

Inlet Nozzle 1 to Upper Shell Weld (Heat# 299L44)

Inlet Nozzle 2 to Upper Shell Weld (Heat# 299L44)

Inlet Nozzle 3 to Upper Shell Weld (Heat# 299L44)

Wt.%

cu(a) 0.16 0.159 0.159 0.159 0.159 0.159 0.159 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses SLR 1/4T Initial USE(a) Projected USE SLR Fluence(b)

(ft-lb)

Decrease(c)

USE (x 1019 n/cm2)

(%)

(ft-lb) 0.0140 64 12 56 0.0304 63 11 56 0.0078 64 10 58 0.0109 68 10 61 0.0081 68 10 61 0.0059 82 10 74 0.0227 71 10.5 64 Position 2.2(d) 0.102 3.9200 103 28 74 0.11 3.9200 82 28 59 0.34 0.7770 64 35 42(e) 0.34 0.0188 64 15 54 0.34 0.0048 64 15 54 0.34 0.0067 64 15 54 Page4-38

Notes:

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses (a)

Material data is from Table 4.2.2-1 and Table 4.2.2-2.

(b) The 1/4T fluence was calculated using the fluence data in Table 4.2.1-1, the Regulatory Guide 1.99 correlation, and the Units 1 and 2 RV wall thickness of 8.05 inches. The surface fluence at the lowest extent of the nozzle weld was used to represent the inlet and outlet nozzle forgings; this approach is conservative. Bounding material fluence values, only, are shown in Figure 5-1 of WCAP-18242-NP for the nozzle materials.

(c)

The Position 1.2 USE decrease values were calculated by plotting the 1/4T fluence values on Figure 2 of Regulatory Guide 1.99 and using the material specific Cu wt.

percent values.

(d) Surveillance data (deemed credible per Appendix A of WCAP-18242-NP) from Table 7-6 of BAW-2324 were used in the calculation of Unit 1 Position 2.2 USE projections.

Regulatory Guide 1.99, Position 2.2 indicates that an upper bound line drawn parallel to the existing lines (in Figure 2 of the Regulatory Guide) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.

(e) These weld materials were previously addressed by EMA Report BAW-2494 for 48 EFPY.

EMAs for these materials have been generated.

(f)

Per Regulatory Guide 1.99 (Revision 2), when credible data exists the Position 2.2 projected USE value should be used in preference to the Position 2.1 projected USE value.

Page4-39

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-6 Predicted USE Values at 68 EFPY for Unit 2 Wt.%

SLR 1/4T Initial Projected USE SLR RV Material Fluence(b)

USE(a)

USE Cu(a)

(x1019 n/cm2)

(ft-lb)

Decrease(c) (%)

(ft-lb)

Position 1.2 Upper Shell Forging 123V303VA1 0.11 0.5340 104 18 85 Upper to lntermed~ate Shell Circumferential Weld(e Heat# 4275 0.35 0.5340 68 39 41(e)

Intermediate Shell Plate C4331-2 0.12 4.4400 84 30 59 Intermediate Shell Plate C4339-2 0.11 4.4400 83 29 59 Intermediate Shell L~2gitudinal Welds L3 and L4 (OD 50%) e (Heat# 72445) 0.22 0.7960 64 34 42(e)

Intermediate Sh~II Longitudinal Weld L4 (ID 50%fr (Heat# 8T1762) 0.19 0.7960 64 32 44(e)

Intermediate to Lower Shell Circ. WeldleJ 0.187 4.4500 82 47 43(e)

(Heat # 0227)

Lower Shell Plate C4208-2 0.15 4.4800 94 35 61 Lower Shell Plate C4339-1 0.107 4.4800 101 29 72 Lower Shell Longitudinal Weld L 1 and L2 e) (Heat # 8T1762) 0.19 0.8020 64 33 43(e)

Inlet Nozzle 1 to Upper Shell Weld 0.19 0.0210 64 14 55 (Heat # 8T1762)

Inlet Nozzle 2 to Upper Shell Weld 0.19 0.0048 64 13.5 55 (Heat # 8T1762)

Inlet Nozzle 3 to Upper Shell Weld 0.19 0.0066 64 13.5 55 (Heat # 8T1762)

Outlet Nozzle 1 to Upper Shell Weld 0.35 0.0049 71 24 54 (Rotterdam)

Outlet Nozzle 2 to Upper Shell Weld 0.35 0.0036 71 24 54 (Rotterdam)

Outlet Nozzle 3 to Upper Shell Weld 0.35 0.0156 71 24 54 (Rotterdam)

Inlet Nozzle 1 (Heat# 9-5104) 0.159 0.0340 73 12.5 64 Inlet Nozzle 2 (Heat# 9-4815) 0.159 0.0078 66 10 59 Inlet Nozzle 3 (Heat# 9-5205) 0.159 0.0107 67 10 60 Outlet Nozzle 1 (Heat# 9-4825-2) 0.159 0.0080 73 10 66 Outlet Nozzle 2 (Heat# 9-5086-1) 0.159 0.0059 77 10 69 Outlet Nozzle 3 (Heat# 9-5086-2) 0.159 0.0253 71 10.5 64 Position 2.2t0J Lower Shell Plate C4339-1 0.107 4.4800 101 19 82 Intermediate Shell Plate C4339-2 0.11 4.4400 83 19 67 Intermediate to Lower Shell Circ. WeldleJ 0.187 4.4500 82 42 48(e)

(Heat# 0227)

Page4-40

Notes:

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses (a)

Material data is from Table 4.2.2-3 and Table 4.2.2-4.

(b) The 1/4T fluence was calculated using the fluence data in Table 4.2.1-2, the Regulatory Guide 1.99 correlation, and the Units 1 and 2 RV wall thickness of 8.05 inches. The surface fluence at the lowest extent of the nozzle weld was used to represent the inlet and outlet nozzle forgings; this approach is conservative. Bounding material fluence values, only, are shown in Figure 5-2 of WCAP-18242-NP for the nozzle materials.

(c)

The Position 1.2 USE decrease values were calculated by plotting the 1/4T fluence values on Figure 2 of Regulatory Guide 1.99 and using the material specific Cu wt.

percent values.

(d)

Surveillance data (deemed credible and non-credible per Appendix A of WCAP-18242-NP) from Table 5-12 of WCAP-16001 were used for Unit 2 Position 2.2 USE projections. Regulatory Guide 1.99, Position 2.2 indicates that an upper bound line drawn parallel to the existing lines (in Figure 2 of the Regulatory Guide) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE. Credibility Criterion 3 in the Discussion Section of Regulatory Guide 1.99 indicates that even if the surveillance data are not considered credible for determination oft.RT NDT* "they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82." Thus, the surveillance data may be used for Unit 2 USE projections.

(e) These weld materials were previously addressed by EMA Report BAW-2494 for 48 EFPY.

EMAs for these materials have been generated.

Page4-41

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-7 Reactor Vessel Weld Locations and 80-Year Fluence Projections (IS) Inside Wetted RV Material Material ID and /or Heat Surface Fluence or(*)

Number clad/base metal n/cm2 E> 1.0 MeV)

Unit 1, 80 Year Fluence (E > 1.0 MeV)

SA-1493 (IS) 1.50E+18 (Wire Ht. 8T1762)

Nozzle Shell to Outlet Nozzle Forging Welds SA-1494 (Wire Ht. 8T15548)

(IS) 1.50E+18 SA-1526 (IS) 1.50E+18 (Wire Ht. 299L44)

Nozzle Shell to Inlet Nozzle Forging Welds SA-1580 (Wire Ht. 8T1762)

(IS) 1.50E+18 Nozzle Shell to Intermediate Shell Circ. Weld J726

(*) 7.98E+18 (Wire Ht. 25017)

Intermediate Shell Long. Welds (Both)

SA-1494

(*)1.33E+19 (Wire Ht. 8T1554)

Intermediate Shell to Lower Shell Circ. Weld (ID 40%

SA-1585

(*)6.67E+19 (Wire Ht. 72445)

Intermediate Shell to Lower Shell Circ. Weld (OD SA-1650 NA 60%)

(Wire Ht. 72445)

Lower Shell Long. Weld (1)

SA-1494

(*)1.34E+19 (Wire Ht. 8T1554)

Lower Shell Long. Weld (2)

SA-1526

(*)1.34E+19 (Wire Ht. 299L44)

Unit 2, 80 Year Fluence (E > 1.0 MeV)

Nozzle Shell to Outlet Nozzle Forging Welds Rotterdam (IS) 1.50E+18 WF-4 (IS) 1.50E+18 (Wire Ht. 8T1762)

Nozzle Shell to Inlet Nozzle Forging Welds WF-8 (Wire Ht. 8T1762)

(IS) 1.50E+18 Nozzle Shell to Intermediate Shell Circ. Weld L737

(*) 9.21 E+18 (Wire Ht. 4275) ntermediate Shell Long. Weld (1), and (2) (100% and SA-1585

(*) 1.36E+19 OD 50%)

(Wire Ht. 72445)

Intermediate Shell Long. Weld (2) (ID 50%)

WF-4

(*) 1.36E+19 (Wire Ht. 8T1762)

Page4-42

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Table 4.2.2-7 Reactor Vessel Weld Locations and 80-Year Fluence Projections (IS) Inside Wetted RV Material Material ID and /or Heat Surface Fluence or(*)

Number clad/base metal n/cm2 E> 1.0 MeV)

Intermediate Shell to Lower Shell Circ. Weld R3008

(*) 7.67E+19 (Wire Ht. 0227)

Lower Shell Long. Weld (Both)

WF-4

(*) 1.37E+19 (Wire Ht. 8T1762)

Note: No Surry Unit 2 Outlet Nozzle to Upper Shell weld data is available. Generic chemistry values were taken from Regulatory Guide 1.99, Revision 2. The initial RT NOT value was determined using ASME Code, Section Ill, minimum criteria and BTP 5-3, Position 1.1 guidance. ASME Code, Section Ill, minimum criteria require measured data; thus, cru= 0°F. The initial USE value was determined using results from the first surveillance capsule removed and tested from the Surry Unit 2 RV and BTP 5-3, Position 2.1 guidance.

4.2.3 PRESSURIZED THERMAL SHOCK TLAA

Description:

A limiting condition on RV integrity known as Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a small-break loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the RV under the following conditions: severe overcooling of the inside surface of the vessel wall followed by repressurization, significant degradation of vessel material toughness caused by radiation embrittlement, and the presence of a critical-size defect anywhere within the vessel wall.

10 CFR 50.61(b)(1) (Reference 1.7-15) provides rules for protection against PTS events for pressurized water reactors. Licensees are required to perform an updated assessment of the projected values of the PTS reference temperature (RT PTs) whenever there is a signiffcant change in projected values of RT PTS or upon a request for a change in the expiration date for operation of the facility. The current analyses, evaluated for 48 EFPY fluence values predicted for 60 years of operation, are TLAAs requiring evaluation for 80 years since a change in the operating license term of the facility is being requested.

TLAA Evaluation:

10 CFR 50.61 (c) provides two methods for determining RT PTS* These methods are also described as Positions 1 and 2 in Regulatory Guide 1.99. Position 1 applies for material without credible surveillance data available and Position 2 is used for material with two or more credible surveillance data sets available. The RT PTS values are calculated for both Positions 1 and 2 by following the guidance in Regulatory Guide 1.99 (Sections 1.1 and 2.1, respectively), using the copper and Page4-43

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses nickel content of the Units 1 and 2 beltline materials, and subsequent period of extended operation fluence projections.

These accepted methods were used with the surface fluence values above to calculate the following RT PTS values for the Units 1 and 2 RV materials at 68 EFPY. The subsequent period of extended operation RT PTS calculations are summarized in Table 4.2.3-1 and Table 4.2.3-2 for Units 1 and 2, respectively.

PWROG-16045-NP, summarizes the results and methodologies used in the determination of the unirradiated nil ductility transition temperature (RT NOT) for the Unit 1 and Unit 2 RV materials.

Appendix E of WCAP-18242-NP provides the RT PTS calculations for the beltline and extended beltline materials.

10 CFR 50.61 (b)(2) establishes screening criteria for RT PTS as 270°F for plates, forgings, and longitudinal welds and 300°F for circumferential welds.

All of the beltline materials in the Unit 1 and Unit 2 RV are below the RT PTS screening criteria values of 270°F for base metal and longitudinal welds, and 300°F for circumferentially oriented welds through the subsequent period of extended operation (68 EFPY). It is recognized in SECY-82-465, "Pressurized Thermal Shock (PTS)" (Reference 4.8-30), Enclosure A, that the RT PTS screening criteria values of 270°F for base metal and longitudinal welds, and 300°F for circumferentially oriented welds are applicable to cylindrical beltline materials. The adjusted reference temperatures for all extended beltline materials are well below 270°F.

The Units 1 and 2 limiting RT PTS value for base metal or longitudinal weld materials at 68 EFPY is 253.2°F (see and Table 4.2.3-2), which applies to Unit 1 Lower Shell Longitudinal Weld L2 Heat # 299L44 (using credible surveillance data). The Units 1 and 2 limiting RT PTS value for circumferentially oriented welds at 68 EFPY is 229.8°F (see and Table 4.2.3-2), which applies to the Unit 1 Intermediate to Lower Shell Circumferential Weld Heat# 72445.

Appendix E of WCAP-18242-NP provides RT PTS calculations for the nozzle materials. The Units 1 and 2 materials remain below the 10 CFR 50.61 screening criteria.

TLAA Disposition: 10 CFR 54.21(c)(1)(ii)

The PTS analyses have been projected to the end of the subsequent period of extended operation.

Page4-44

Table 4.2.3-1 Calculation of Unit 1 RT PTS Values for 68 EFPY at the Clad/Base Metal Interface R.G.1.99, RV Material Rev.2 Position Upper Shell Forging 122V109VA1 1.1 Upper to Intermediate Shell Circumferential Weld 1.1 (Heat# 25017)

Intermediate Shell Plate C4326-1 1.1 Intermediate Shell Plate C4326-2 1.1 Intermediate Shell Longitudinal Welds L3 and L4 (Heat# 8T1554) 1.1 Intermediate to Lower Shell Circumferential Weld 1.1 (Heat # 72445)

Using credible surveillance data 2.1 Lower Shell Plate C4415-1 1.1 Using credible surveillance data 2.1 Lower Shell Plate C4415-2 1.1 Using credible surveillance data 2.1 Lower Shell Longitudinal Weld L 1 1.1 (Heat# 8T1554)

Lower Shell Longitudinal Weld L2 1.1 (Heat # 299L44)

Using credible surveillance data 2.1 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

Wt.%

cu(a)

Ni(a) 0.11 0.74 0.33 0.1 0.11 0.55 0.11 0.55 0.16 0.57 0.22 0.54 0.102 0.493 0.11 0.5 0.16 0.57 0.34 0.68 Surface CF(a)

Fluence<b)

Surface RT NDT(U)(c) aRTNDT(d)

(OF)

(x 1019 n/cm2, FF(b)

(OF)

(OF)

E > 1.0 MeV)

RV Beltline Materials 76.1 0.754 0.921 40 70.1 152 0.754 0.921 0

140 73.5 6.29 1.445 10 106.2 73.5 6.29 1.445 11.4 106.2 167 1.25 1.062

-48.6 177.4 167 6.31 1.445

-72.5 241.4 167 6.31 1.445

-72.5 241.4 66.6 6.35 1.447 20 96.3 83.1 6.35 1.447 20 120.2 73 6.35 1.447 4.6 105.6 83.1 6.35 1.447 4.6 120.2 167 1.26 1.064

-48.6 177.8 220.6 1.26 1.064

-74.3 234.8 249.8 1.26 1.064

-74.3 265.9 Page4-45 au<c)

O"A(e)

Margin RT PTS(f)

(OF)

(OF)

(OF)

(OF) 0 17 34 144.1 20 28 68.8 208.8 0

17 34 150.2 0

17 34 151.6 18 28 66.6 195..4 12 28 60.9 229.8 12 28 60.9 229.8 0

17 34 150.3 0

8.5 17 157.2 0

17 34 144.2 0

8.5 17 141.8 18 28 66.6 195.7 12.8 28 61.6 222.1 12.8 28 61.6 253.2

R.G.

RV Material 1.99, Rev. 2 Position Inlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 299L44)

Using credible suNeillance data 2.1 Inlet Nozzle 2 to Upper Shell Weld 1.1 (Heat # 299L44)

Using credible suNeillance data 2.1 Inlet Nozzle 3 to Upper Shell Weld 1.1 (Heat # 299L44)

Using credible suNeillance data 2.1 Inlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 8T1762)

Inlet Nozzle 2 to Upper Shell Weld 1.1 (Heat # 8T1762)

Inlet Nozzle 3 to Upper Shell Weld 1.1 (Heat# 8T1762)

Outlet Nozzle 1 to Upper Shell Weld 1.1 (Heat # 8T1762)

Outlet Nozzle 2 to Upper Shell Weld 1.1 (Heat # 8T1762)

Outlet Nozzle 3 to Upper Shell Weld 1.1 (Heat# 8T1762)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

cu(a) 0.34 0.34 0.34 0.19 0.19 0.19 0.19 0.19 0.19 Surface Wt.%

CF(a)

Fluence(b)

Surface Ni(a)

(OF)

(x 1019 nlcm2, FF(b)

E > 1.0 MeV)

RV Extended Beltline Materials 0.68 220.6 0.0304 0.221 249.8 0.0304 0.221 0.68 220.6 0.00784 0.093 249.8 0.00784 0.093 0.68 220.6 0.0109 0.116 249.8 0.0109 0.116 0.57 152.4 0.0304 0.221 0.57 152.4 0.00784 0.093 0.57 152.4 0.0109 0.116 0.57 152.4 0.00813 0.095 0.57 152.4 0.00586 0.075 0.57 152.4 0.0227 0.186 Page4-46 RT NDT(U)(c) ~RTNDT(d) au(c) 0"1::,.(e)

Margin RT PT(f)S (OF)

(OF)

(OF)

(OF)

(OF)

(OF)

-7 48.8 20.6 24.4 63.9 105.7

-7 55.3 20.6 14 49.8 98.1

-7 0.0 (20.4) 20.6 0

41.2 34.2

-7 0.0 (23.2) 20.6 0

41.2 34.2

-7 25.6 20.6 12.8 48.5 67.2

-7 29 20.6 14 49.8 71.8

-4.9 33.7 19.7 16.9 51.9 80.7

-4.9 0.0 (14.1) 19.7 0

39.4 34.5

-4.9 17.7 19.7 0

43.2 56.0

-4.9 0.0 (14.5) 19.7 0

39.4 34.5

-4.9 0.0(11.5) 19.7 0

39.4 34.5

-4.9 28.3 19.7 14.2 48.5 72

R.G.

RV Material 1.99, Rev. 2 Position Outlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 8T15548)

Outlet Nozzle 2 to Upper Shell Weld 1.1 (Heat# 8T15548)

Outlet Nozzle 3 to Upper Shell Weld 1.1 (Heat# 8T15548)

Inlet Nozzle 1 1.1 (Heat# 9-4787)

Inlet Nozzle 2 1.1 (Heat # 9-5078)

Inlet Nozzle 3 1.1 (Heat# 9-4819)

Outlet Nozzle 1 1.1 (Heat# 9-4825-1)

Outlet Nozzle 2 1.1 (Heat# 9-4762)

Outlet Nozzle 3 1.1 (Heat# 9-4788)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Surface Wt.%

Wt.%

CF(a)

Fluence(b) cu(a)

Ni(a)

(OF)

(x 1019 n/cm2, E > 1.0 MeV) 0.16 0.57 143.9 0.00813 0.16 0.57 143.9 0.00586 0.16 0.57 143.9 0.0227 0.159 0.85 123.5 0.0304 0.159 0.87 123.7 0.00784 0.159 0.84 123.4 0.0109 0.159 0.85 123.5 0.00813 0.159 0.83 123.3 0.00586 0.159 0.84 123.4 0.0227 Page4-47 Surface RT NDT(U)(c) ARTNDT(d) au(c)

O'a(e)

Margin RT PTS(f)

FF(b)

(OF)

(OF)

(OF)

(OF)

(OF)

(OF) 0.095

-4.9 0.0 (13.7) 19.7 0

39.4 34.5 0.075

-4.9 0.0 (10.8) 19.7 0

39.4 34.5 0.186

-4.9 26.8 19.7 13.4 47.6 69.5 0.221 10.3 27.3 0

13.7 27.3 65 0.093 11.6 0.0 (11.5) 0 0

0 11.6 0.116

-47.2 14.3 0

7.2 14.3

-18.5 0.095

-44.9 0.0 (11.7) 0 0

0

-44.9 0.075

-87.5 0.0 (9.3) 0 0

0

-87.5 0.186

-50.2 0.0 (22.9) 0 11.5 22.9

-4.3

Notes:

(a)

Chemical composition values taken from Table 4.2.2-1 and Table 4.2.2-2. Chemistry factor values taken from Table 3-1 O of WCAP-18242-NP.

(b)

Surface fluence values taken from Section 2 of WCAP-18242-NP. FF= fluence factor= f<0-2a-o.magcn>.

(c)

Initial RT NOT and au values are taken from Table 4.2.2-1 and Table 4.2.2-2.

(d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," (Reference 4.8-31 ), embrittlement effects may be neglected for materials with fluence values less than 1.0 x 1017 n/cm2(E > 1.0 MeV). These materials have fluence values at the clad/base metal interface surface less than 1.0 x 1017 n/cm2; therefore,

~RT NOT values for these materials are set equal to zero. Calculated ~RT NOT values are listed in parentheses for information purposes only.

(e)

Per Appendix A of WCAP-18242-NP, all Unit 1 surveillance data was deemed credible. Per the guidance of 1 O CFR 50.61, the base metal at,= 17°F for Position 1.1, and oA = 8.5°F for Position 2.1 with credible surveillance data. Also per 1 O CFR 50.61, the weld metal oA = 28°F for Position 1.1, and with credible surveillance data oA = 14°F for Position 2.1. However, oA need not exceed 0.5* ~RT NOT* For welds utilizing initial RT NOT values based on BAW-2308, oA = 28°F per BAW-2308, (Revision 1-A), SE and BAW-2308, (Revision 2-A) SE, "Final Safety Evaluation for Pressurized Water Reactors Owners Group (PWROG) Topical Report (TR) BAW-2308, (Revision 2), 'Initial RT NOT of Linde 80 Weld Materials'" (Reference 4.8-32).

(f)

RT PTS values calculated in accordance with 1 O CFR 50.61 methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-48

Table 4.2.3-2 Calculation of Unit 2 RT PTS Values for 68 EFPY at the Clad/Base Metal Interface R.G.

RV Material 1.99, Rev.2 Position Upper Shell Forging 123V303VA1 1.1 Upper to Intermediate Shell Circumferential Weld 1.1 (Heat# 4275)

Intermediate Shell Plate C4331-2 1.1 Intermediate Shell Plate C4339-2 1.1 Using non-credible surveillance 2.1 data Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%)

1.1 (Heat # 72445)

Using credible surveillance data 2.1 Intermediate Shell Longitudinal Weld L4 (ID 50%)

1.1 (Heat# 8T1762)

Intermediate to Lower Shell Circumferential Weld 1.1 (Heat # 0227)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

Wt.%

cu(a)

Ni(a) 0.11 0.72 0.35 0.1 0.12 0.6 0.11 0.54 0.22 0.54 0.19 0.57 0.187 0.545 Surface CF(a)

Fluence(b)

Surface RT NDT(U)(c) ARTNDT(d)

(OF)

(x 1019 n/cm2, FF(b)

(OF)

(OF)

E > 1.0 MeV)

RV Beltline Materials 75.8 0.865 0.959 30 72.7 160.5 0.865 0.959 0

154 83 7.2 1.467 15 121.8 73.4 7.2 1.467 7.8 107.7 75.7 7.2 1.467 7.8 111.1 167 1.29 1.071

-72.5 178.8 167 1.29 1.071

-72.5 178.8 167 1.29 1.071

-48.6 178.8 147.5 7.22 1.468 0

216.5 Page4-49 au(c) aa(e)

Margin RT PTS(f)

(OF)

(OF)

(OF)

(OF) 0 17 34 136.7 20 28 68.8 222.8 0

17 34 170.8 0

17 34 149.5 0

17 34 152.9 12 28 60.9 167.3 12 28 60.9 167.3 18 28 66.6 196.8 0

28 56 272.5

Table 4.2.3-2 Calculation of Unit 2 RT PTS Values for 68 EFPY at the Clad/Base Metal Interface R.G.

RV Material 1.99, Rev.2 Position Using credible surveillance data 2.1 Lower Shell Plate C4208-2 1.1 Lower Shell Plate C4339-1 1.1 Using non-credible surveillance 2.1 data Lower Shell Longitudinal Welds L 1 and L2 1.1 (Heat# 8T1762)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

Wt.%

cu(a)

Ni(a) 0.15 0.55 0.107 0.53 0.19 0.57 Surface CF(a)

Fluence(b)

Surface RT NDT(U)(c).aRTNDT(d)

(OF)

(x 1019 n/cm2, FF(bl (OF)

(OF)

E > 1.0 MeV)

RV Beltline Materials 132.5 7.22 1.468 0

194.5 107.3 7.26 1.469

-30 157.6 70.8 7.26 1.469

-4.4 104 75.7 7.26 1.469

-4.4 111.2 167 1.3 1.073

-48.6 179.2 Page4-50 au(c)

O"t:,.(e)

Margin RT PTS(f)

(OF)

(OF)

(OF)

(OF) 0 14 28 222.5 0

17 34 161.6 0

17 34 133.6 0

17 34 140.8 18 28 66.6 197.2

R.G.

RV Material 1.99, Rev.2 Position Inlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 8T1762)

Inlet Nozzle 2 to Upper Shell Weld 1.1 (Heat # 8T1762)

Inlet Nozzle 3 to Upper Shell Weld 1.1 (Heat # 8T1762 Outlet Nozzle 1 to Upper Shell 1.1 Weld (Rotterdam)

Outlet Nozzle 2 to Upper Shell 1.1 Weld (Rotterdam)

Outlet Nozzle 3 to Upper Shell 1.1 Weld (Rotterdam)

Inlet Nozzle 1 1.1 (Heat# 9-5104)

Inlet Nozzle 2 1.1 (Heat# 9-4815)

Inlet Nozzle 3 1.1 (Heat# 9-5205)

Outlet Nozzle 1 1.1 (Heat# 9-4825-2)

Outlet Nozzle 2 1.1 (Heat# 9-5086-1)

Outlet Nozzle 3 1.1 (Heat # 9-5086-2)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

cu(a) 0.19 0.19 0.19 0.35 0.35 0.35 0.159 0.159 0.159 0.159 0.159 0.159 Surface Wt.%

CF(a)

Fluence(b)

Surface Ni(a)

(OF)

(x 1019 n/cm2, FF(b)

E > 1.0 MeV)

RV Extended Beltline Materials 0.57 152.4 0.034 0.236 0.57 152.4 0.00784 0.093 0.57 152.4 0.0107 0.115 1

272 0.00796 0.094 1

272 0.00585 0.075 1

272 0.0253 0.199 0.84 123.4 0.034 0.236 0.87 123.7 0.00784 0.093 0.86 123.6 0.0107 0.115 0.85 123.5 0.00796 0.094 0.86 123.6 0.00585 0.075 0.87 123.7 0.0253 0.199 Page4-51 RT NDT(U)(c).6.RT NDT(d) au(c)

O"t:,.(e)

Margin RT PTS(f)

(OF)

(OF)

(OF)

(OF)

(OF)

(OF)

-4.9 36 19.7 18 53.4 84.4

-4.9 0.0(14.1) 19.7 0

39.4 34.5

-4.9 0.0 (17.5) 19.7 8.7 43.1 55.7 30 0.0 (25.5) 0 0.0 0.0 30.0 30 0.0 (20.5) 0 0

0 30 30 54 0

27 54 138

-29.7 29.1 0

14.6 29.1 28.6 4.5 0.0(11.5) 0 0

0 4.5 6.5 14.2 0

7.1 14.2 34.9

-58.1 0.0(11.6) 0 0

0

-58.1

-26.6 0.0 (9.3) 0 0

0

-26.6

-33.8 24.6 0

12.3 24.6 15.3

Notes:

(a)

Chemical composition values taken from Table 4.2.2-3 and Table 4.2.2-4. Chemistry factor values taken from Table 3-12 of WCAP-18242-NP.

(b)

Surface fluence values taken from Section 2 of WCAP-18242-NP. FF = fluence factor= f<0-20*0-10*10gcn.

(c)

Initial RT NOT and Ou values taken from Table 4.2.2-3 and Table 4.2.2-4.

(d)

Per NRG RIS 2014-11, embrittlement effects may be neglected for materials with fluence values less than 1.0 x 1017 n/cm2(E > 1.0 MeV).

These materials have fluence values at the clad/base metal interface surface less than 1.0 x 1017 n/cm2; therefore, i1RT NOT values for these materials are set equal to zero. Calculated i1RT NOT values are listed in parentheses for information purposes only.

(e)

Per Appendix A of WCAP-18242-NP, the surveillance plate data were deemed non-credible, whereas the surveillance data for the weld materials were deemed credible. Per the guidance of 10 CFR 50.61, the base metal au = 17°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. Per 10 CFR 50.61, the weld metal Ou= 28°F for Position 1.1, and with credible surveillance data cr1 = 14°F for Position 2.1. However, Ou need not exceed 0.5* b.RT NOT* For welds utilizing initial RT NOT values based on BAW-2308, (Revisions 1-A SE and 2-A SE), au= 28°F.

(f)

RT PTS values calculated in accordance with 10 CFR 50.61 methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-52

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses 4.2.4 ADJUSTED REFERENCE TEMPERATURE TLAA

Description:

The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T limit curves to account for irradiation effects. Regulatory Guide 1.99 provides the methodology for determining the ART of the limiting material. The initial nil-ductility reference temperature, RT NOT, is the temperature at which a non-irradiated metal (ferritic steel) changes in fracture characteristics from ductile to brittle behavior. Neutron fluence increases the RT NOT beyond its initial value.

RT NOT was evaluated in accordance with PWROG-16045-NP, which includes the generally accepted techniques outlined in:

• ASME Code, Section Ill, Paragraph NB 2331,

• Branch Technical Position 5-3,

• BWRVIP-173-A, "BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials" (Reference 4.8-33), and

• BAW-2313.

10 CFR 50, Appendix G, defines the fracture toughness requirements for the vessel. The shift in the initial RT NOT (~RT NOT) is evaluated as the difference in the 30 ft-lb index temperatures from the average Charpy curves measured before and after irradiation. This increase (~RT NOT) means that higher temperatures are required for the material to continue to act in a ductile manner. The ART is defined as the sum of the initial (unirradiated) reference temperature (Initial RT NOT), the mean value of the adjustment in reference temperature caused by irradiation (~RT NOT), and a margin (M) term.

Since the ~RT NOT value is a function of 48 EFPY fluence, associated with the 60 year licensed operating period, these ART calculations meet the criteria of 1 O CFR 54.3(a) and have been identified as TLMs requiring evaluation for 80 years.

TLAA Evaluation:

As described in Section 4.2.1, 68 EFPY fluence values were determined for the Units 1 and 2 RV beltline and extended beltline components. These 68 EFPY 1/4T fluence values were used to compute the ART values of Units 1 and 2, in accordance with Regulatory Guide 1.99.

Table 4.2.4-1 through 4.2.4-9 summarize the nozzle, and 1/4T ART calculations for Units 1 and 2 at 48 and 68 EFPY. The 3/4T ART values are included in WCAP-18242-NP. The limiting 48 EFPY and 68 EFPY ART values for Units 1 and 2 apply to the Unit 2 Intermediate to Lower Shell Circumferential Weld (using surveillance data).

Page4-53

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses The inlet and outlet nozzle forging ARTs are necessary to perform a nozzle corner fracture mechanics analysis. The nozzle forging ART calculations utilize the postulated nozzle forging surface 1/4T flaw fluence values in order to provide a conservative estimate of the fluence at the limiting nozzle corner location. The nozzle ART values are also considered herein because the nozzle fluence values for some nozzle materials exceed 1.0 x 1017 n/cm2 (E > 1.0 MeV), and thus all of the nozzle forgings are considered part of the extended beltline for conservatism. Since the surface fluence values are utilized for the ART calculations for the nozzle forging materials, the nozzle forgings are omitted from 1/4T ART calculations.

Table 4.2.4-9 compares the TLAA limiting ART values at 48 EFPY and 68 EFPY to the limiting ART values used in development of the existing 48 EFPY P-T limit curves documented in WCAP-14177, "Surry Power Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation" (Reference 4.8-34). The limiting ART values used to develop the existing P-T limit curves are summarized in Table 4.2.4-9. As shown in Table 4.2.4-9, the TLAA limiting ART values at 48 EFPY and 68 EFPY are less than the limiting ART values used to develop the existing P-T limit curves.

Appendix B of WCAP-18243-NP shows that the PT curves for the nozzles lie above and to the left of the PT curves for the beltline materials. Thus, the PT curves for the beltline materials are bounding through the subsequent period of extended operation.

TLAA Disposition: 10 CFR 54.21(c)(1)(ii).

The ART analyses have been projected to the end of the subsequent period of extended operation.

They may be used as inputs to 68 EFPY P-T limits for the subsequent period of extended operation.

Page4-54

Table 4.2.4-1 Calculation of the Unit 1 Nozzle ART Values at the Surface Location for 48 EFPY R.G.1.99, Surface Wt.%

Wt.%

CF(a)

Fluence(b)

Surface RT NDT(U)(c) aRTNDT(d) a}c)

CJ1:,.(e)

Margin ART(fl RV Material Rev.2 Position cu(a)

Ni(a)

(OF)

(x 1019 n/cm2, FF(b)

(OF)

(OF)

(OF)

(OF)

(OF)

(OF)

E > 1.0 MeV)

Inlet Nozzle 1 1.1 0.159 0.85 123.5 0.00870 0.100 10.3 0.0 (12.3) 0.0 0.0 0.0 10.3 (Heat # 9-4 787)

Inlet Nozzle 2 1.1 0.159 0.87 123.7 0.00219 0.035 11.6 0.0 (4.4) 0.0 0.0 0.0 11.6 (Heat# 9-5078)

Inlet Nozzle 3 1.1 0.159 0.84 123.4 0.00306 0.046

-47.2 0.0 (5.7) 0.0 0.0 0.0

-47.2 (Heat# 9-4819)

Outlet Nozzle 1 1.1 0.159 0.85 123.5 0.00237 0.038

-44.9 0.0 (4.6) 0.0 0.0 0.0

-44.9 (Heat# 9-4825-1)

Outlet Nozzle 2 1.1 0.159 0.83 123.3 0.0017 0.029

-87.5 0.0 (3.5) 0.0 0.0 0.0

-87.5 (Heat# 9-4762)

Outlet Nozzle 3 1.1 0.159 0.84 123.4 0.00672 0.083

-50.2 0.0 (10.3) 0.0 0.0 0.0

-50.2 (Heat# 9-4788)

Notes:

(a)

Chemical composition data taken from Table 4.2.2-1 and Table 4.2.2-2. Chemistry factor values taken from Table 3-1 O of WCAP-18242-NP.

(b)

Surface fluence values were from WCAP-18-242-NP. FF= fluence factor= f(o.2s-o.1o*1og(fl>.

(c)

Initial RT NOT values and au values are from Table 4.2.2-2 (d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1.0 x 1017 n/cm2(E > 1.0 MeV).

These materials have fluence values at the clad/base metal interface surface less than 1.0 x 1017 n/cm2; therefore, ~RT NOT values for these materials are set equal to zero. Calculated ~RT NOT values are listed in parentheses for information purposes only.

(e)

Per the guidance of Regulatory Guide 1.99, the base metal at,.= 17°F for Position 1.1. However, at,. need not exceed 0.5*b.RT NOT*

(f)

ART values calculated in accordance with Regulatory Guide 1.99, Revision 2 methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-55

Table 4.2.4-2 Calculation of the Unit 1 ART Values at the 1/4T Location for 48 EFPY R.G.

RV Material 1.99, Rev. 2 Position Upper Shell Forging 122V109VA1 1.1 Upper to Intermediate Shell Circumferential Weld 1.1 (Heat# 25017)

Intermediate Shell Plate C4326-1 1.1 Intermediate Shell Plate C4326-2 1.1 Intermediate Shell Longitudinal 1.1 Welds L3 and L4 (Heat# 8T1554)

Intermediate to Lower Shell Circumferential Weld 1.1 (Heat # 72445)

Using credible surveillance data 2.1 Lower Shell Plate C4415-1 1.1 Using credible surveillance data 2.1 Lower Shell Plate C4415-2 1.1 Using credible surveillance data 2.1 Lower Shell Longitudinal Weld L 1 1.1 (Heat# 8T1554)

Lower Shell Longitudinal Weld L2 1.1 (Heat# 299L44)

Using credible surveillance data 2.1 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

Wt.%

cu(a)

Ni(a) 0.11 0.74 0.33 0.10 0.11 0.55 0.11 0.55 0.16 0.57 0.22 0.54 0.102 0.493 0.11 0.5 0.16 0.57 0.34 0.68 114T Fluence(b)

CF(a)

(x 1019 nlcm2, 1/4T RT Nor(U) (c)

(OF)

E > 1.0 MeV)

FF(b)

(OF)

RV Beltline Materials 76.1 0.329 0.695 40 152 0.329 0.695 0

73.5 2.79 1.274 10 73.5 2.79 1.274 11.4 167 0.537 0.826

-48.6 167 2.81 1.275

-72.5 167 2.81 1.275

-72.5 66.6 2.82 1.276 20 83.1 2.82 1.276 20 73 2.82 1.276 4.6 83.1 2.82 1.276 4.6 167 0.542 0.829

-48.6 220.6 0.542 0.829

-74.3 249.8 0.542 0.829

-74.3 Page4-56 114T

~RTNDT(d) o}c)

C11:J,. (e)

Margin ART(fl (OF)

(OF)

(OF)

(OF)

(OF) 52.9 0.0 17.0 34.0 126.9 105.6 20.0 28.0 68.8 174.4 93.6 0.0 17.0 34.0 137.6 93.6 0.0 17.0 34.0 139 138 18.0 28.0 66.6 156.0 212.9 12.0 28.0 60.9 201.3 212.9 12.0 28.0 60.9 201.3 85 0.0 17.0 34.0 139 106.0 0.0 8.5 17.0 143.0 93.1 0.0 17.0 34.0 131.7 106.0 0.0 8.5 17.0 127.6 138.4 18.0 28.0 66.6 156.4 182.8 12.8 28.0 61.6 170.1 207.0 12.8 28.0 61.6 194.3

R.G.1.99, RV Material Rev. 2 Position Inlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 299L44)

Using credible surveillance data 2.1 Inlet Nozzle 2 to Upper Shell Weld 1.1 (Heat # 299L44)

Using credible surveillance data 2.1 Inlet Nozzle 3 to Upper Shell Weld 1.1 (Heat# 299L44)

Using credible surveillance data 2.1 Inlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 8T1762)

Inlet Nozzle 2 to Upper Shell Weld 1.1 (Heat # 8T1762)

Inlet Nozzle 3 to Upper Shell Weld 1.1 (Heat# 8T1762)

Outlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 8T1762)

Outlet Nozzle 2 to Upper Shell Weld 1.1 (Heat# 8T1762)

Outlet Nozzle 3 to Upper Shell Weld 1.1 (Heat# 8T1762)

Outlet Nozzle 1 to Upper Shell Weld 1.1 (Heat # 8T15548)

Outlet Nozzle 2 to Upper Shell Weld 1.1 (Heat# 8T15548)

Outlet Nozzle 3 to Upper Shell Weld 1.1 (Heat# 8T15548)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

cu(a) 0.34 0.34 0.34 0.19 0.19 0.19 0.19 0.19 0.19 0.16 0.16 0.16 CF(a) 1/4T Fluence\\01 Wt.%

(x 1019 n/cm2, 1/4T Ni(a)

(OF)

FF(b)

E > 1.0 MeV)

RV Extended Beltline Materials 0.68 220.6 0.0131 0.132 249.8 0.0131 0.132 0.68 220.6 0.00330 0.049 249.8 0.00330 0.049 0.68 220.6 0.00462 0.063 249.8 0.00462 0.063 0.57 152.4 0.0131 0.132 0.57 152.4 0.00330 0.049 0.57 152.4 0.00462 0.063 0.57 152.4 0.00345 0.051 0.57 152.4 0.00247 0.039 0.57 152.4 0.00981 0.108 0.57 143.9 0.00345 0.051 0.57 143.9 0.00247 0.039 0.57 143.9 0.00981 0.108 Page4-57 RT Nor(U)(c)

.6RTNDT(d)

O((c) aa(e)

Margin 1/4T ART(fl (OF)

(OF)

(OF)

(OF)

(OF)

(OF)

-7 29 20.6 14.5 50.4 72.4

-7 32.9 20.6 14.0 49.8 75.7

-7 0.0 (10.8) 20.6 0.0 41.2 34.2

-7 0.0 (12.2) 20.6 0.0 41.2 34.2

-7 0.0 (13.9) 20.6 0.0 41.2 34.2

-7 0.0 (15.8) 20.6 0.0 41.2 34.2

-4.9 20.1 19.7 0.0 44.2 59.4

-4.9 0.0 (7.5) 19.7 0.0 39.4 34.5

-4.9 0.0 (9.6) 19.7 0.0 39.4 34.5

-4.9 0.0 (7.7) 19.7 0.0 39.4 34.5

-4.9 0.0 (5.9) 19.7 0.0 39.4 34.5

-4.9 16.5 19.7 0.0 42.7 34.5

-4.9 0.0 (7.3) 19.7 0.0 39.4 34.5

-4.9 0.0 (5.6) 19.7 0.0 39.4 34.5

-4.9 15.6 19.7 7.8 42.4 53.0

Notes:

(a)

Chemical composition data taken from Table 4.2.2-1 and Table 4.2.2-2. Chemistry factor values taken from Table 3-10 of WCAP-18242-NP.

(b) 48 EFPY surface fluence values were from WCAP-18242-NP. The 1/4T fluence and 1/4T FF were calculated using the Regulatory Guide 1.99 correlations and the Unit 1 RV wall thickness of 8.05 inches.

(c)

Initial RT NOT values and cru values are from Table 4.2.2-1 and Table 4.2.2-2.

(d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1.0 x 1017 n/cm2(E > 1.0 MeV).

These materials have fluence values at the clad/base metal interface surface less than 1.0 x 1017 n/cm2; therefore,,iRT NOT values for these materials are set equal to zero. Calculated ~RT NOT values are listed in parentheses for information purposes only.

(e)

As summarized in Appendix A of WCAP-18242-NP, all surveillance data for Unit 1 were deemed credible. Per the guidance of Regulatory Guide 1.99, the base metal crt,. = 17°F for Position 1.1, and crt,. = 8.5°F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, the weld metal crt,. = 28°F for Position 1.1, and with credible surveillance data crt,. = 14°F for Position 2.1. However, crt,. need not exceed 0.5*.LiRT NOT* For welds utilizing initial RT NOT values based on BAW-2308, (Revisions 1 A SE and 2 A SE), crt,. = 28°F.

(f)

ART values calculated in accordance with Regulatory Guide 1.99, Revision 2 methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-58

Table 4.2.4-3 Calculation of the Unit 2 Nozzle ART Values at the Surface Location for 48 EFPY R.G.1.99, Surface Wt.%

Wt.%

CF(a)

Fluence(b)

Surface RT Nor(U) (c) ARTNDT(d) 0}C) cr fl (e)

Margin ART(fl RV Material Rev.2 Position cu(a)

Ni(a)

(OF)

(x 1019 n/cm2, FF(b)

(OF)

(OF)

(OF)

(OF)

(OF)

(OF)

E > 1.0 MeV)

Inlet Nozzle 1 1.1 0.159 0.84 123.4 0.00935 0.105

-29.7 0.0 (12.9) 0.0 0.0 0.0

-29.7 (Heat# 9-5104)

Inlet Nozzle 2 1.1 0.159 0.87 123.7 0.00223 0.036 4.5 0.0 (4.4) 0.0 0.0 0.0 4.5 (Heat# 9-4815)

Inlet Nozzle 3 1.1 0.159 0.86 123.6 0.00304 0.046 6.5 0.0 (5.7) 0.0 0.0 0.0 6.5 (Heat # 9-5205)

Outlet Nozzle 1 1.1 0.159 0.85 123.5 0.00235 0.037

-58.1 0.0 (4.6) 0.0 0.0 0.0

-58.1 (Heat # 9-4825-2)

Outlet Nozzle 2 1.1 0.159 0.86 123.6 0.00172 0.029

-26.6 0.0 (3.6) 0.0 0.0 0.0

-26.6 (Heat # 9-5086-1)

Outlet Nozzle 3 1.1 0.159 0.87 123.7 0.00723 0.088

-33.8 0.0 (10.8) 0.0 0.0 0.0

-33.8 (Heat # 9-5086-2)

Notes:

(a)

Chemical composition values taken from Table 4.2.2-3 and Table 4.2.2-4. Chemistry factor values taken from Table 3-12 of WCAP-18242-NP.

(b)

Surface fluence values were from WCAP-18242-NP. FF= fluence factor= f<0-2s-o.1a*1°g<n>.

(c)

Initial RT NOT values and cru values are from Table 4.2.2-4.

(d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1 x 1017 n/cm2(E > 1.0 MeV).

These materials have fluence values at the clad/base metal interface surface less than 1 x 1017 n/cm2; therefore, ~RT NOT values for these materials are set equal to zero. Calculated ~RT NOT values are listed in parentheses for information purposes only.

(e)

Per the guidance of Regulatory Guide 1.99, the base metal 0 11 = 17°F for Position 1. However, 0 11 need not exceed 0.5*bRT NOT*

(f)

ART values calculated in accordance with Regulatory Guide 1.99, Revision 2 methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-59

Table 4.2.4-4 Calculation of the Unit 2 ART Values at the 1/4T Location for 48 EFPY R.G.1.99, RV Material Rev.2 Position Upper Shell Forging 123V303VA1 1.1 Upper to Intermediate Shell 1.1 Circumferential Weld (Heat # 4275)

Intermediate Shell Plate C4331-2 1.1 Intermediate Shell Plate C4339-2 1.1 Using non-credible suNeillance data 2.1 Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%)

1.1 (Heat # 72445)

Using credible suNeillance data 2.1 Intermediate Shell Longitudinal Weld 1.1 L4 (ID 50%) (Heat# 8T1762)

Intermediate to Lower Shell 1.1 Circumferential Weld (Heat# 0227)

Using credible suNeillance data 2.1 Lower Shell Plate C4208-2 1.1 Lower Shell Plate C4339-1 1.1 Using non-credible suNeillance data 2.1 Lower Shell Longitudinal Welds L 1 1.1 and L2 (Heat# 8T1762)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

Wt.%

cu(a)

Ni(a) 0.11 0.72 0.35 0.10 0.12 0.6 0.11 0.54 0.22 0.54 0.19 0.57 0.187 0.545 0.15 0.55 0.107 0.53 0.19 0.57 CF(a) 1/4T Fluence(b) 1/4T RT Nor(U) (c)

(x 1019 n/cm2, (OF)

E > 1.0 MeV)

FF(b)

(OF)

RV Beltline Materials 75.8 0.362 0.719 30 160.5 0.362 0.719 0

83 3.07 1.296 15 73.4 3.07 1.296 7.8 75.7 3.07 1.296 7.8 167 0.563 0.839

-72.5 167 0.563 0.839

-72.5 167 0.563 0.839

-48.6 147.5 3.07 1.296 0

132.5 3.07 1.296 0

107.3 3.09 1.298

-30 70.8 3.09 1.298

-4.4 75.7 3.09 1.298

-4.4 167 0.567 0.841

-48.6 Page4-60

~RTNDT(d) o}c)

CJA (e)

Margin 1/4T ART(f)

(OF)

(OF)

(OF)

(OF)

(OF) 54.5 0.0 17.0 34.0 118.5 115.4 20.0 28.0 68.8 184.2 107.6 0.0 17.0 34.0 156.6 95.1 0.0 17.0 34.0 136.9 98.1 0.0 17.0 34.0 139.9 140.2 12.0 28.0 60.9 128.6 140.2 12.0 28.0 60.9 128.6 140.2 18.0 28.0 66.6 158.2 191.2 0.0 28.0 56.0 247.2 171.8 0.0 14.0 28.0 199.8 139.3 0.0 17.0 34.0 143.3 91.9 0.0 17.0 34.0 121.5 98.2 0.0 17.0 34.0 127.8 140.5 18.0 28.0 66.6 158.5

R.G.1.99, Wt.%

Wt.%

CF(a) 1/4T Fluence(bJ 1/4T RT NDT(U) (c) aRTNDT(d) 0}C) ab. (e)

Margin 1/4T RV Material Rev. 2 (x 1019 n/cm2, ART(f)

Position cu(a)

Ni(a)

(OF)

E > 1.0 MeV)

FF(b)

(OF)

(OF)

(OF)

(OF)

(OF)

(OF)

RV Extended Beltline Materials Inlet Nozzle 1 to Upper Shell 1.1 0.19 0.57 152.4 0.0141 0.138

-4.9 0.0 (21.0) 19.7 0.0 39.4 34.5 Weld (Heat # 8T1762)

Inlet Nozzle 2 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00336 0.050

-4.9 0.0 (7.6) 19.7 0.0 39.4 34.5 (Heat# 8T1762)

Inlet Nozzle 3 to Upper Shell 1.1 0.19 0.57 152.4 0.0059 0.063

-4.9 0.0 (9.6) 19.7 0.0 39.4 34.5 Weld (Heat # 8T1762)

Outlet Nozzle 1 to Upper Shell 1.1 0.35 1

272 0.00342 0.050 30 0.0 (13.7) 0.0 0.0 0.0 30 Weld (Rotterdam)

Outlet Nozzle 2 to Upper Shell 1.1 0.35 1

272 0.0025 0.039 30 0.0 (10.7) 0.0 0.0 0.0 30 Weld (Rotterdam)

Outlet Nozzle 3 to Upper Shell 1.1 0.35 1

272 0.0105 0.114 30 30.9 0.0 15.5 30.9 91.9 Weld (Rotterdam)

Notes:

(a)

Chemical composition values taken from Table 4.2.2-3 and Table 4.2.2-4. Chemistry factor values taken from Table 3-12 of WCAP-18242-NP.

(b) 48 EFPY surface fluence values were from WCAP-18242-NP. The 1/4T fluence and 1/4T FF were calculated using the Regulatory Guide 1.99, correlations and the Unit 2 RV wall thickness of 8.05 inches.

(c)

Initial RT NOT values and cru values are from Table 4.2.2-3 and Table 4.2.2-4.

(d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1 x 1017 n/cm2(E > 1.0 MeV). These materials have fluence values at the clad/base metal interface surface less than 1 x 1017 n/cm2; therefore, LlRT NOT values for these materials are set equal to zero.

Calculated LlRT NOT values are listed in parentheses for information purposes only.

(e)

Per Appendix A of WCAP-18242-NP, the surveillance plate data were deemed non credible, whereas the surveillance data for the weld materials were deemed credible. Per the guidance of Regulatory Guide 1.99, the base metal at,. = 17°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. Also per Regulatory Guide 1.99, the weld metal Ot,. = 28°F for Position 1.1, and with credible surveillance data at,. = 14 °F for Position 2.1. However, Ot,. need not exceed 0.5*.ilRT NOT* For welds utilizing initial RT NOT values based on BAW-2308, (Revisions 1-A SE and 2-A SE), at,.

= 28°F.

(f)

ART values calculated in accordance with Regulatory Guide 1.99, Revision 2 methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-61

Table 4.2.4-5 Calculation of the Unit 1 Nozzle ART Values at the Surface Location for 68 EFPY R.G.

Surface RV Material 1.99, Wt.%

Wt.%

CF(a)

Fluence(b)

Surface RT Nor(U) (c) ARTNDT(d) o}c) a A (e)

Margin Rev.2 cu(a)

Ni(a)

(OF)

(x 1019 n/cm2, FF(b)

(OF)

(OF)

(OF)

(OF)

(OF)

Position E > 1.0 MeV)

Inlet Nozzle 1 1.1 0.159 0.85 123.5 0.0124 0.127 10.3 15.6 0.0 7.8 15.6 (Heat# 9-4787)

Inlet Nozzle 2 1.1 0.159 0.87 123.7 0.00322 0.048 11.6 0.0 (5.9) 0.0 0.0 0.0 (Heat # 9-5078)

Inlet Nozzle 3 1.1 0.159 0.84 123.4 0.00446 0.062

-47.2 0.0 (7.6) 0.0 0.0 0.0 (Heat# 9-4819)

Outlet Nozzle 1 1.1 0.159 0.85 123.5 0.00345 0.051

-44.9 0.0 (6.3) 0.0 0.0 0.0 (Heat# 9-4825-1)

Outlet Nozzle 2 1.1 0.159 0.83 123.3 0.00249 0.039

-87.5 0.0 (4.8) 0.0 0.0 0.0 (Heat# 9-4762)

Outlet Nozzle 3 1.1 0.159 0.84 123.4 0.00962 0.107

-50.2 0.0 (13.2) 0.0 0.0 0.0 (Heat# 9-4788)

Notes:

(a)

Chemical composition data taken from Table 4.2.2-1 and Table 4.2.2-2. Chemistry factor values taken from Table 3-10 of WCAP-18242-NP.

(b)

Surface fluence values taken from Section 4.2.1. FF= fluence factor= t1°-2a-o.1o*iog(f))_

(c)

Initial RT NOT values and Ou values are from Table 4.2.2-2.

ART(f)(g)

(OF) 41.6 11.6

-47.2

-44.9

-87.5

-50.2 (d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1 x 1017 n/cm2(E > 1.0 MeV). These materials have fluence values at the clad/base metal interface surface less than 1 x 1017 n/cm2; therefore, ~RT NOT values for these materials are set equal to zero.

Calculated ~RT NOT values are listed in parentheses for information purposes only.

(e)

Per the guidance of Regulatory Guide 1.99, the base metal al).= 17°F for Position 1.1. However, cr !). need not exceed 0.5*.liRT NOT*

(f)

Nozzle materials are not limiting for P-T limit curves per WCAP-18243-NP, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation" (Reference 4.8-35).

(g)

ART values calculated in accordance with Regulatory Guide 1.99, Revision 2 methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-62

Table 4.2.4-6 Calculation of the Unit 1 ART Values at the 1/4T Location for 68 EFPY R.G.

RV Material 1.99, Rev.2 Position Upper Shell Forging 122V109VA1 1.1 Upper to Intermediate Shell Circumferential Weld 1.1 (Heat# 25017)

Intermediate Shell Plate C4326-1 1.1 Intermediate Shell Plate C4326-2 1.1 Intermediate Shell Longitudinal Welds L3 and L4 1.1 (Heat# 8T1554)

Intermediate to Lower Shell Circumferential Weld 1.1 (Heat # 72445)

Using credible surveillance data 2.1 Lower Shell Plate C4415-1 1.1 Using credible surveillance data 2.1 Lower Shell Plate C4415-2 1.1 Using credible surveillance data 2.1 Lower Shell Longitudinal Weld L 1 1.1 (Heat# 8T1554)

Lower Shell Longitudinal Weld L2 1.1 (Heat# 299L44)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

Wt.%

cu(a)

Ni(a) 0.11 0.74 0.33 0.10 0.11 0.55 0.11 0.55 0.16 0.57 0.22 0.54 0.102 0.493 0.11 0.50 0.16 0.57 0.34 0.68 1/4T Fluence(b)

CF(a)

(x 1019 n/cm2, 1/4T RT Nor(U) (c)

(OF)

E > 1.0 MeV)

FF(b)

(OF)

RV Beltline Materials 76.1 0.465 0.787 40 152 0.465 0.787 0

73.5 3.88 1.350 10 73.5 3.88 1.350 11.4 167 0.771 0.927

-48.6 167 3.89 1.350

-72.5 167 3.89 1.350

-72.5 66.6 3.92 1.352 20 83.1 3.92 1.352 20 73 3.92 1.352 4.6 83.1 3.92 1.352 4.6 167 0.777 0.929

-48.6 220.6 0.777 0.929

-74.3 Page4-63 1/4T b.RTNDT(d) a/cl CJ b. (e)

Margin ART(fl (OF)

(OF)

(OF)

(OF)

(OF) 59.9 0.0 17.0 34.0 133.9 119.6 20.0 28.0 68.8 188.4 99.2 0.0 17.0 34.0 143.2 99.2 0.0 17.0 34.0 144.6 154.8 18.0 28.0 66.6 172.8 225.5 12.0 28.0 60.9 213.9 225.5 12.0 28.0 60.9 213.9 90 0.0 17.0 34.0 144.0 112.3 0.0 8.5 17.0 149.3 98.7 0.0 17.0 34.0 137.3 112.3 0.0 8.5 17.0 133.9 155.2 18.0 28.0 66.6 173.2 205 12.8 28.0 61.6 192.3

R.G.

RV Material 1.99, Rev.2 Position Using credible surveillance data 2.1 Inlet Nozzle 1 to Upper Shell 1.1 Weld (Heat# 299L44)

Using credible surveillance data 2.1 Inlet Nozzle 2 to Upper Shell 1.1 Weld (Heat # 299L44)

Using credible surveillance data 2.1 Inlet Nozzle 3 to Upper Shell 1.1 Weld (Heat# 299L44)

Using credible surveillance data 2.1 Inlet Nozzle 1 to Upper Shell 1.1 Weld (Heat # 8T1762)

Inlet Nozzle 2 to Upper Shell 1.1 Weld (Heat# 8T1762)

Inlet Nozzle 3 to Upper Shell 1.1 Weld (Heat # 8T1762)

Outlet Nozzle 1 to Upper Shell 1.1 Weld (Heat# 8T1762)

Outlet Nozzle 2 to Upper Shell 1.1 Weld (Heat # 8T1762)

Outlet Nozzle 3 to Upper Shell 1.1 Weld (Heat # 8T1762)

Outlet Nozzle 1 to Upper Shell 1.1 Weld (Heat # 8T15548)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

cu(a) 0.34 0.34 0.34 0.19 0.19 0.19 0.19 0.19 0.19 0.16 114T Fluence(b)

Wt.%

CF(a)

{x 1019 nlcm2, 114T Ni(a)

(OF)

FF(b)

E > 1.0 MeV)

RV Beltline Materials 249.8 0.777 0.929 RV Extended Beltline Materials 0.68 220.6 0.0188 0.165 249.8 0.0188 0.165 0.68 220.6 0.00484 0.065 249.8 0.00484 0.065 0.68 220.6 0.00672 0.083 249.8 0.00672 0.083 0.57 152.4 0.0188 0.165 0.57 152.4 0.00484 0.065 0.57 152.4 0.00672 0.083 0.57 152.4 0.00502 0.067 0.57 152.4 0.00362 0.052 0.57 152.4 0.0140 0.137 0.57 143.9 0.00502 0.067 Page4-64 114T RTNor(U)

.aRTNDT(d) 0}C)

CJ ll. (e)

Margin ART(fl (c)(oF)

{OF)

(OF)

(OF)

(OF)

(OF)

-74.3 232.1 12.8 28.0 61.6 219.4

-7.0 36.5 20.6 18.2 55.0 84.5

-7.0 41.3 20.6 14.0 49.8 84.1

-7.0 0.0 (14.4) 20.6 0.0 41.2 34.2

-7.0 0.0 (16.3) 20.6 0.0 41.2 34.2

-7.0 18.3 20.6 9.2 45.1 56.4

-7.0 20.8 20.6 10.4 46.1 59.9

-4.9 25.2 19.7 12.6 46.8 67.1

-4.9 0.0 (10.0) 19.7 0.0 39.4 34.5

-4.9 12.7 19.7 6.3 41.4 49.2

-4.9 0.0 (10.2) 19.7 0.0 39.4 34.5

-4.9 0.0 (8.0) 19.7 0.0 39.4 34.5

-4.9 20.9 19.7 10.5 44.6 60.6

-4.9 0.0 (9.7) 19.7 0

39.4 34.5

R.G.

1/4T Fluence{b)

RT~or(U) 1/4T RV Material 1.99, Wt.%

Wt.%

CF(a)

(x 1019 n/cm2, 1/4T c) aRTNDT{d) o}c) a 11 (e)

Margin ART(fl Rev.2 cu(a)

Ni(a)

(OF)

FF{b)

(OF)

(OF)

(OF)

(OF)

Position E > 1.0 MeV)

(OF)

(OF)

RV Extended Beltline Materials Outlet Nozzle 2 to Upper Shell 1.1 0.16 0.57 143.9 0.00362 0.052

-4.9 0.0 (7.6) 19.7 0

39.4 34.5 Weld (Heat# 8T1554B)

Outlet Nozzle 3 to Upper Shell 1.1 0.16 0.57 143.9 0.0140 0.137

-4.9 19.7 19.7 9.9 44.1 58.9 Weld (Heat# 8T1554B)

Notes:

(a)

Chemical composition data taken from Table 4.2.2-1 and Table 4.2.2-2. Chemistry factor values taken from Table 3.-1 O of WCAP-18242-NP.

(b)

The 1/4T fluence and 1/4T FF were taken from Table 5-1 of WCAP-18243-NP.

(c)

Initial RT NOT values and au values are from Table 4.2.2-1 and Table 4.2.2-2.

(d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1 x 1017 n/cm2(E > 1.0 MeV).

These materials have fluence values at the clad/base metal interface surface less than 1 x 1017 n/cm2; therefore, ~RT NOT values for these materials are set equal to zero. Calculated ~RT NOT values are listed in parentheses for information purposes only.

(e)

As summarized in Appendix G of WCAP-18343-NP, all surveillance data for Unit 1 were deemed credible. Per the guidance of Regulatory Guide 1.99 (Revision 2), the base metal a,, = 17°F for Position 1.1, and a,,= 8.5°F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99 (Revision 2), the weld metal a,,= 28°F for Position 1.1, and with credible surveillance data a,,= 14°F for Position 2.1.

However, a,, need not exceed 0.5*.b.RT NOT* For welds utilizing initial RT NOT values based on BAW-2308, a,,= 28°F (f)

ART values calculated in accordance with Regulatory Guide 1.99 (Revision 2) methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-65

Table 4.2.4-7 Calculation of the Unit 2 ART Nozzle Values at the Surface Location for 68 EFPY R.G.

Surface ART(fl(

RV Material 1.99, Wt.%

Wt.%

CF(a)

Fluence(b)

Surface RT Nor(U) (c) ARTNDT(d) c,/C)

C1A (e)

Margin g)

Rev. 2 cu(a)

Ni(a)

(OF)

(x 1019 n/cm2, FF(b)

(OF)

(OF)

(OF)

(OF)

(OF)

(OF)

Position E > 1.0 MeV)

Inlet Nozzle 1 1.1 0.159 0.84 123.4 0.0139 0.137

-29.7 16.8 0.0 8.4 16.8 4.0 Heat# 9-5104)

Inlet Nozzle 2 1.1 0.159 0.87 123.7 0.00321 0.048 4.5 0.0 (5.9) 0.0 0.0 0.0 4.5 (Heat# 9-4815)

Inlet Nozzle 3 1.1 0.159 0.86 123.6 0.00437 0.061 6.5 0.0 (7.5) 0.0 0.0 0.0 6.5 (Heat# 9-5205)

Outlet Nozzle 1 1.1 0.159 0.85 123.5 0.00338 0.05

-58.1 0.0 (6.2) 0.0 0.0 0.0

-58.1 (Heat# 9-4825-2)

Outlet Nozzle 2 1.1 0.159 0.86 123.6 0.00248 0.039

-26.6 0.0 (4.8) 0.0 0.0 0.0

-26.6 (Heat# 9-5086-1)

Outlet Nozzle 3 1.1 0.159 0.87 123.7 0.0107 0.115

-33.8 14.2 0.0 7.1 14.2

-5.4 (Heat # 9-5086-2)

Notes:

(a)

Chemical composition values taken from Table 4.2.2-3 and Table 4.2.2-4. Chemistry factor values taken from Table 3-12 of WCAP-18242-NP.

(b)

Surface fluence values taken from Section 4.2.1. FF = fluence factor= fco.2s-o.10*1°gc~>.

(c)

Initial RT NDT values and Ou values are from Table 4.2.2-4.

(d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1 x 1017 n/cm2(E > 1.0 MeV).

These materials have fluence values at the clad/base metal interface surface less than 1 x 1017 n/cm2; therefore, ~RT NDT values for these materials are set equal to zero. Calculated ~RT NDT values are listed in parentheses for information purposes only.

(e)

Per the guidance of Regulatory Guide 1.99, the base metal a"= 17°F for Position 1.1. However, a" need not exceed 0.5*b.RT NDT*

(f)

Nozzle materials are not limiting for P-T limit curves per WCAP-18243-NP.

(g)

ART values calculated in accordance with Regulatory Guide 1.99 (Revision 2) methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-66

Table 4.2.4-8 Calculation of the Unit 2 ART Values at the 1/4T Location for 68 EFPY R.G.

RV Material 1.99, Rev.2 Position Upper Shell Forging 123V303VA1 1.1 Upper to Intermediate Shell Circumferential Weld 1.1 (Heat# 4275)

Intermediate Shell Plate C4331-1.1 Intermediate Shell Plate C4339-2 1.1 Using non-credible surveillance 2.1 data Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%)

1.1 (Heat # 72445)

Using credible surveillance data 2.1 Intermediate Shell Longitudinal Weld L4 (ID 50%)

1.1 (Heat# 8T1762)

Intermediate to Lower Shell Circumferential Weld 1.1 (Heat # 0227)

Using credible surveillance data 2.1 Lower Shell Plate C4208-2 1.1 Lower Shell Plate C4339-1 1.1 Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

Wt.%

cu(a)

Ni(a) 0.11 0.72 0.35 0.10 0.12 0.60 0.11 0.54 0.22 0.54 0.19 0.57 0.187 0.545 0.15 0.55 0.107 0.53 1/4T Fluence(b)

CF(a)

(x 1019 n/cm2, 1/4T RT Nor(U) (c)

(OF)

E > 1.0 MeV)

FF(b)

(OF)

RV Beltline Materials 75.8 0.534 0.825 30 160.5 0.534 0.825 0

83.0 4.44 1.378 15 73.4 4.44 1.378 7.8 75.7 4.44 1.378 7.8 167.0 0.796 0.936

-72.5 167.0 0.796 0.936

-72.5 167.0 0.796 0.936

-48.6 147.5 4.45 1.379 0

132.5 4.45 1.379 0

107.3 4.48 1.380

-30 70.8 4.48 1.380

-4.4 Page4-67 1/4T ARTNDT(d) 0}C) aa (e)

Margin ART(fl (OF)

(OF)

(OF)

(OF)

(OF) 62.5 0.0 17.0 34.0 126.5 132.3 20.0 28.0 68.8 201.2 114.4 0.0 17.0 34.0 163.4 101.2 0.0 17.0 34.0 143.0 104.3 0.0 17.0 34.0 146.1 156.3 12.0 28.0 60.9 144.7 156.3 12.0 28.0 60.9 144.7 156.3 18.0 28.0 66.6 174.3 203.4 0.0 28.0 56.0 259.4 182.7 0.0 14.0 28.0 210.7 148.1 0.0 17.0 34.0 152.1 97.7 0.0 17.0 34.0 127.3

R.G.

1.99, RV Material Rev.2 Position Using non-credible surveillance 2.1 data Lower Shell Longitudinal Welds 1.1 L 1 and L2 (Heat# 8T1762)

Inlet Nozzle 1 to Upper Shell 1.1 Weld (Heat# 8T1762)

Inlet Nozzle 2 to Upper Shell Weld 1.1 (Heat# 8T1762)

Inlet Nozzle 3 to Upper Shell 1.1 Weld (Heat # 8T1762)

Outlet Nozzle 1 to Upper Shell 1.1 Weld (Rotterdam)

Outlet Nozzle 2 to Upper Shell 1.1 Weld (Rotterdam)

Outlet Nozzle 3 to Upper Shell 1.1 Weld (Rotterdam)

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Wt.%

cu(a) 0.19 0.19 0.19 0.19 0.35 0.35 0.35 1/4T Fluence<b)

Wt.%

CF(a)

(x 1019 n/cm2, 1/4T Ni(a)

(OF)

FF(b)

E > 1.0 MeV) 75.7 4.48 1.380 0.57 167.0 0.802 0.938 RV Extended Beltline Materials 0.57 152.4 0.0210 0.177 0.57 152.4 0.00484 0.065 0.57 152.4 0.00660 0.082 1.0 272.0 0.00491 0.066 1.0 272.0 0.00361 0.052 1.0 272.0 0.0156 0.147 Page4-68 RT NDT(U) (c).dRTNDT(d) o}c) crA (e) 1/4T Margin ART (f)

(OF)

(OF)

(OF)

(OF)

(OF)

(OF)

-4.4 104.5 0.0 17.0 34.0 134.1

-48.6 156.7 18.0 28.0 66.6 174.6

-4.9 27 19.7 13.5 47.8 69.9

-4.9 0.0 (10.0) 19.7 0.0 39.4 34.5

-4.9 12.5 19.7 6.3 41.3 48.9 30 0.0 (18.0) 0.0 0.0 0.0 30.0 30 0.0 (14.3) 0.0 0.0 0.0 30.0 30 40.0 0.0 20.0 40.0 110.0

Notes:

(a)

Chemical composition values taken from Table 4.2.2-3 and Table 4.2.2-4. Chemistry factor values taken from Table 3-12 of WCAP-18243-NP.

(b) The 1/4T fluence and 1/4T FF were taken from Table 5-2 of WCAP-18343-NP.

(c)

Initial RT NOT values and cru values are from Table 4.2.2-3 and Table 4.2.2-4.

(d)

Per NRC RIS 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," embrittlement effects may be neglected for materials with fluence values less than 1 x 1017 n/cm2(E > 1.0 MeV).

These materials have fluence values at the clad/base metal interface surface less than 1 x 1017 n/cm2; therefore, ~RT NDT values for these materials are set equal to zero. Calculated ~RT NDT values are listed in parentheses for information purposes only.

(e) As summarized in Appendix G of WCAP-18243-NP, the surveillance plate data were deemed non-credible, whereas the surveillance data for the weld materials were deemed credible. Per the guidance of Regulatory Guide 1.99 (Revision 2), the base metal cra = 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data. Also per Regulatory Guide 1.99 (Revision 2), the weld metal cra = 28°F for Position 1.1, and with credible surveillance data cra = 14°F for Position 2.1. However, cra need not exceed 0.5*b.RT NOT* For welds utilizing initial RT NDT values based on BAW-2308, cra = 28°F.

(f)

ART values calculated in accordance with Regulatory Guide 1.99 (Revision 2) methodology.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-69

Table 4.2.4-9 Summary of the Units 1 and 2 Limiting ART Values Used in the Applicability Evaluation of the Reactor Vessel Heatup and Cooldown Curves 1/4T Limiting ART (°F) 3/4T Limiting ART (°F)

Plant Limiting Material Existing 48 EFPY TLAA TLAA Existing 48 EFPY TLAA TLAA Curves Documented Evaluation Evaluation Curves Documented Evaluation at Evaluation at in WCAP-14177, in WCAP-14177, Rev. o(a) at48 EFPY at 68 EFPY Rev. o(a) 48 EFPY 68 EFPY (Circ Flaw) Circ. Weld:

Intermediate to Lower Shell Circ. Weld, 201.3 213.9 158.5 173.6 Heat# 72445 SPS Unit 1 (Axial Flaw) Long. Weld:

Lower Shell Long. Weld L2 194.3 219.4 131.3 153.8 Heat# 299L44 (Position 2.1)

(Circ Flaw) Circ. Weld:

228.4 189.5 Intermediate to Lower Shell Circ. Weld, 199.8 210.7 166.3 179.8 Heat# 0227 (Position 2.1)

SPS Unit 2 (Axial Flaw) Plate:

155_5(b) 153.4(c) 135.6 144.0 Intermediate Shell Plate C4331-2 Axial Flaw) Weld:

116.id) 130.7'e)

Lower Shell Longitudinal Weld L 1 and L2 158.5 174.6 Heat # 8T1762 Notes: Limiting values depicted as bold and underlined.

(a) The limiting 48 EFPY 1/4T and 3/4T ART values in the Technical Specifications correspond to the Unit 1 Intermediate to Lower Shell Circumferential Weld (Heat# 72445). The basis for the P-T limit curves is contained in WCAP-14177; however, the applicability was extended to 48 EFPY in a later analysis. See Appendix C of WCAP-18242-NP for details.

(b) Value from Table 6.1-5 of WCAP-18242-NP.

(c)

Value from Table 6.1-11 of WCAP-18242-NP.

(d) Value from Table 6.1-6 of WCAP-18242-NP.

(e) Value from Table 6.1-12 of WCAP-18242-NP.

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses Page4-70

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses 4.2.5 PRESSURE-TEMPERATURE LIMITS TLAA

Description:

1 O CFR 50 Appendix G requires that the RV be maintained within established pressure-temperature (P-T) limits, including heatup and cooldown operations. These limits specify the maximum allowable pressure as a function of reactor coolant temperature. As the RV is exposed to increased neutron irradiation, its fracture toughness is reduced. The P-T limits must account for the anticipated RV fluence.

The current P-T limits are based upon fluence projections for 60 years of plant operation. Because they were based upon a fluence assumption of 60 years of operation, the P-T limits analyses meet the definition of 10 CFR 54.3(a) (Reference 1. 7-2) and have been identified as TLAAs.

TLAA Evaluation:

Heatup and cooldown limit curves are calculated using the most limiting value of RT NOT corresponding to the limiting material in the beltline region of the RV. The most limiting RT NOT of the material in the core region (beltline) of the RV is determined by using the unirradiated RV material fracture toughness properties and estimating the irradiation induced shift (.6.RT NoT)-

RT NOT increases as the material is exposed to fast neutron irradiation; therefore, to find the most limiting core region (beltline) RT NOT at any time,.6.RT NOT due to the neutron radiation exposure associated with that time must be added to the original unirradiated RT NOT* Using the ART values, P-T limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G, as augmented by ASME Code,Section XI, Appendix G.

The current P-T limits for Units 1 and 2 are based on the K18 methodology and the latest fluence data through 48 EFPY and are maintained in the Technical Specifications.

According to NUREG-2192, Section 4.2.2.1.4, the P-T limits for the subsequent period of extended operation need not be submitted as part of the SLRA since the P-T limits are required to be updated through the 1 O CFR 50.90 licensing process when necessary for P-T limits that are located in the Technical Specifications. The current licensing basis will ensure that the P-T limits for the subsequent period of extended operation will be updated prior to exceeding the EFPY for which they remain valid.

Nozzle materials were evaluated in WCAP-18242-NP at 48 EFPY and 68 EFPY; the nozzle forging materials evaluated are documented in Tables 4.2.4-1, 4.2.4-3, 4.2.2-5, and e. All nozzle materials were assigned the fluence values at the postulated 1/4T flaw location for each specific nozzle in Table 4.2.1-1 and Table 4.2.1-2. Thus, Unit 1 Inlet Nozzle 1 and Unit 2 Inlet Nozzle 1 and Outlet Nozzle 3 have neutron fluence values greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at 68 EFPY. In order to fully assess the Units 1 and 2 P-T limit curves applicability to 68 EFPY, a nozzle corner fracture mechanics analysis was completed for all nozzle materials. These nozzle P-T limit curves Page4-71

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses were generated and compared to the beltline P-T limit curves to ensure that the beltline curves are bounding. The detailed nozzle forging fracture mechanics evaluation and comparison to the applicable RV beltline P-T limit curves were documented in WCAP-18243-NP. The current beltline curves were confirmed to remain more limiting than the nozzle curves through 68 EFPY.

The development of the current P-T limit curves for normal heatup and cooldown of the primary reactor coolant system for Units 1 and 2 was documented in WCAP-14177. The existing P-T limit curves are based on the Kie methodology and the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material. The Units 1 and 2 P-T limit curves were developed by calculating ART values utilizing the vessel fluence at the clad/base metal interface corresponding to each RV material. Since the development of the curves, the applicability of the curves has been extended and the fluence values and initial material properties used to calculate ART values have been updated.

The Kie methodology was used to confirm the applicability of the P-T limit curves developed based on WCAP-14177. The limiting RV material ART values with consideration of the updated 68 EFPY fluence values, revised Position 2.1 chemistry factor values, and updated initial RT NOT values must be shown to be less than or equal to the limiting beltline material ART values used in development of the P-T limit curves contained in WCAP-14177 and the Units 1 and 2 Technical Specifications.

The Regulatory Guide 1.99 methodology was used along with the surface fluence of Section 2 of WCAP-18242-NP to calculate ART values for the Units 1 and 2 RV materials at 48 EFPY and 68 EFPY.

Comparisons of the use of the Kie reference stress intensity factor, instead of the older, more conservative Kia reference stress intensity factor were conducted to validate that the PT limits for 48 EFPY are conservative for operation through the subsequent period of extended operation. The comparisons of the limiting ART values calculated as part of this RV integrity TLAA evaluation, using updated fluence and initial material properties, to those used in calculation of the existing P-T limit curves are contained in Table 4.2.4-9 for Units 1 and 2. With the consideration of TLAA fluence projections, the applicability of the P-T limit curves in WCAP-14177 may be extended to 68 EFPY for the Units 1 and 2 cylindrical shell materials. Nozzle P-T limit curves were developed per WCAP-18243-NP and compared to the cylindrical shell beltline curves. ART values were generated without the consideration of the methodology in TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels, U.S. NRC Technical Letter Report, Office of Nuclear Regulatory Research [RES]" (Reference 4.8-36). Per WCAP-18243-NP, the applicability of the P-T limit curves may be extended through SLR, because the current Technical Specifications P-T limit curves bound the new P-T limit curves developed in WCAP-18243-NP regardless of the use of the TLR-RES/DE/CIB-2013-01 methodology. Per WCAP-18243-NP, the applicability of the P-T limit curves may be extended through the subsequent period of extended operation.

Page4-72

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses In addition, the applicable RV flange and closure head initial RT NOT values are bounding and the P-T limit curves flange notch requires no change or further consideration. Finally, the lowest service temperature requirements are not applicable to Units 1 and 2, because the plants are Westinghouse-designed per ASME Code, Section Ill, and utilize stainless steel reactor coolant system piping.

TLAA Disposition: 10 CFR 54.21(c)(1)(iii)

Since the P-T limits will be updated through the 10 CFR 50.90 process at a later, appropriate date, the effects of aging on the intended function(s) of the RVs will be adequately managed for the subsequent period of extended operation. The Reactor Vessel Material Surveillance program (82.1.19) and plant Technical Specifications will ensure that updated P-T limits based upon updated ART values will be submitted to the NRG for approval prior to exceeding the period of applicability for Units 1 and 2.

4.2.6 LOW TEMPERATURE OVERPRESSURE PROTECTION TLAA

Description:

Low temperature overpressure protection (LTOP) system (sometimes referred to as the Reactor Coolant System Overpressure Mitigating System, or the RV Overpressure Mitigating System) at Unit 1 and Unit 2 is required by Technical Specification Limited Condition for Operation 3.1.G. Two pressurizer power operated relief valves (PORV) provide the automatic relief capability during the design basis mass input and the design basis heat input transients to automatically prevent the reactor coolant system pressure from exceeding the P-T limit curves based on 10 CFR 50, Appendix G.

LTOP system setpoints are based on the P-T limits calculation which is a TLAA.

TLAA Evaluation:

The LTOP enabling temperature has been determined for 68 EFPY as discussed in Appendix D of WCAP-18243-NP. Using Code Case N-514, the LTOP enabling temperature is 283°F. The Surry Technical Specification 3.1.G.1.c.(4) specifies an arming temperature of 350°F which is conservative and remains valid for the subsequent period of extended operation.

In WCAP-18242-NP the maximum allowable LTOP system PORV setpoint was calculated to be 399.6 psig for the Units 1 and 2 subsequent period of extended operation. The calculation was performed in accordance with the WCAP-14040-A methodology using LTOP input parameters and the limiting axial flaw steady state ASME Code,Section XI, Appendix G limits calculated for the subsequent period of extended operation at 68 EFPY.

Page4-73

Surry Power Station, Units 1 and 2 Application for Subsequent License Renewal Time-Limited Aging Analyses The evaluation showed that the current Technical Specification value of:;; 390.0 psig is bounding and will remain valid for the subsequent period of extended operation. Since the maximum allowable PORV setpoint was determined using the methodology in WCAP-14040-A, this demonstrates that the current licensing basis PORV setpoint that was developed using K18 ASME Code,Section XI, Appendix G limits without applying uncertainties was sufficiently conservative.

TLAA Disposition: 10 CFR 54.21(c)(1)(ii)

The L TOP system setpoint and enabling temperature have been projected to the end of the subsequent period of extended operation.

Page4-74

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures MARKED-UP TECHNICAL SPECIFICATIONS AND BASES PAGES Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

TS 3.1-9 06 26 15 Heatup and cooldown limit curves are calculated using Htt~ffis-t-ttffH-ttH-2' value of the nil-ductility reference temperature, RT NOT, at the end of (EFPY) for Units 1 and 2. The heatup and cooldown limit curves were previously calculated using the most limiting value of RT NOT (228.4°F) which occurred at the 1/4-T, 0° azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld.

Subsequently, the reactor vessel material property basis was amended based upon new data,vhich showed that the most limiting value of RTr~D'f (222.5°F) at 4 8 EFPY occurs at the 1/4 T, 0° azimuthal location in the Unit 2 intermediate to lower shell circumferential weld. The revised limiting material property (i.e., Unit 2 RTND'f of 222.5°F) justified continued use of the existing heatup and cooldown limit curves (based on the Unit 1 RTND'f of 228.4°F) to 48 EFPY for Units 1 and 2. The limiting RTND'f at the 1/4 T location in the core region is greater than the R.Tl</D'f of the limiting unirradiated material.

This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RT NOT; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RT NOT* Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RT NOT at the end of Units 1 and 2 (as well as adjustments for location of the pressure sensing instrument).

Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the t-.RT NOT determined from the surveillance capsule exceeds the calculated (1;) t-.RT NOT for the equivalent capsule radiation exposure, or when the service period exceeds

~

EFPY for Units 1 and 2 prior to a scheduled refueling outage.

Amendment Nos. 285 and 285

TS 3.1-10 06 29 06 Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix Gin Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of one and one half Tis assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RT NDT, is used and this includes the radiation-induced shift, ~RT NDT, corresponding to the end of the period for which heatup and cooldown curves are generated. *n the 1986 Edition of the ASME Code ating the allowable limit curves for various heatup and cooldown rates ecifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Km, for the metal temperature at that time. KrR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The Km curve is given by the equation:

K1R = 26.78 + 1.223 exp [0.0145(T - RT NDT + 160)]

(1) where Km is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RT NDT* Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

(1) ~

where, K1M is the stress intensity factor caused by membrance (pressure) stress.

Amendment Nos. 248 and 247

Kit is the stress intensity factor caused by the thermal gradients TS 3.1-11 05 31 11 KiR is provided by the code as a function of temperature relative to the RT NDT of the material.

C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RT NDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, Kit, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

The heatup limit curve, Figure 3.1-1, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60°F per hour. The cooldown limit curves of Figure 3.1-2 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The cooldown limit curves are valid for cooldown rates up to 100°F/hr.

The heatup and cooldown curves were prepared based upon the most limit' predicted adjusted reference temperature at the end of 4&

adjusted reference temperature was calculated using materials properties data from the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) documented in the most recent revision to BA W-1543 and reactor vessel neutron fluence data obtained from plant-specific analyses.

Amendment Nos. 274 and 274

TS 3.1-12 06 26 15 The reactor boltup temperature is defined in 10 CFR 50, Appendix G as "The highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload." The reactor vessel may be bolted up at a temperature greater than the initial RT DT of the material stressed by the boltup (e.g., the vessel flange). As noted on Figures 3.1-1 and 3.1-2, the limiting boltup temperature is 10°F. An administrative minimum boltup temperature limit greater than 10°F is imposed in station procedures to ensure the Reactor Coolant System temperatures are sufficiently high to prevent damage to the reactor vessel closure head/vessel flange during the removal or installation of reactor vessel head bolts. The limiting boltup temperature and the administrative minimum boltup temperature limit are in effect when the reactor vessel head bolts are under tension.

References WCAP-14177, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (October 1994)

(3)

WCAP-18243, Rev. 2, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (July 2018)

Amendment Nos. 285 aAd 285

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so 100 150 200 250 300 350 400 450 50 0 550 600 650 Indicat ed Cold Leg Temperature (Deg. F)

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EFPY w

Amendment Nos. 285 1rnd 285

TS 3.1 BASIS INSERTS INSERT 1 The limiting RT NDT at the 1/4-T location in the core region is greater that the RT NDT of the limiting unirradiated material.

INSERT 2 The technical basis for the data points and the associated RT NDT values used to generate the heatup and cooldown curves is provided in WCAP-14177 (Reference 2) and were determined to be applicable to the 48 EFPY period of extended operation under first license renewal. The associated RT NDT values used to calculate the heatup and cooldown curves provided in WCAP-14177 (Revision 2) are based upon the Surry Unit 1 Intermediate to Lower Shell Gire Weld:

1/4-T, 228.4°F and 3/4-T, 189.5°F The heatup and cooldown curves for operation through 48 EFPY were based upon the K1r methodology. These heatup and cooldown curves were subsequently evaluated using the Kie methodology for Subsequent License Renewal (SLR) at 68 EFPY in WCAP-18243-NP (Reference 3).

The limiting reactor vessel materials at 68 EFPY were determined to be the Surry Unit 1 Lower Shell Longitudinal Weld L2 at 1/4-T and the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld at 3/4-T. The associated RT NDT values calculated at 68 EFPY are:

1/4-T, 219.4 °F and 3/4-T, 179.8 °F The data points and the associated RT NDT values used to generate the heatup and cooldown curves in TS Figures 3.1-1 and 3.1-2, respectively, are conservative based upon use of the Kie methodology. Therefore, the heatup and cooldown curves did not require revision as a result of SLR. However, the fluence applicability is updated from 48 EFPY to 68 EFPY.

__j

Serial No. 19-37 4 Docket Nos. 50-280/281 LAR - SLR Revision of HU/CD Limits Figures PROPOSED TECHNICAL SPECIFICATIONS AND BASES PAGES Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

TS 3.1-9 Heatup and cooldown limit curves are calculated using a bounding value of the nil-ductility reference temperature, RT NDT* at the end of 68 Effective Full Power Years (EFPY) for Units 1 and 2. The heatup and cooldown limit curves were calculated using the most limiting value of RT NDT (228.4 °F) which occurred at the 1/4-T, 0° azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld. The limiting RT NDT at the 1/4-T location in the core region is greater than the RT NDT of the limiting unirradiated material. This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RT NDT; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RT NDT* Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RT NDT at the end of 68 EFPY for Units 1 and 2 (as well as adjustments for location of the pressure sensing instrument).

Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the ~RT NDT determined from the surveillance capsule exceeds the calculated

~RT NDT for the equivalent capsule radiation exposure, or when the service period exceeds 68 EFPY for Units 1 and 2 prior to a scheduled refueling outage.

Amendment Nos.

TS 3.1-10 Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of one and one half Tis assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RT NDT* is used and this includes the radiation-induced shift, ~RT NDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

The approach for calculating the allowable limit curves for various heatup and cooldown rates in the 1986 Edition of the ASME Code specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1R, for the metal temperature at that time. K1R is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K1R curve is given by the equation:

K1R = 26.78 + 1.223 exp [0.0145(T-RTNDT + 160)]

(1) where K1R is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RT NDT* Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

(1) where, K1M is the stress intensity factor caused by membrance (pressure) stress.

Amendment Nos.

TS 3.1-11 Kit is the stress intensity factor caused by the thermal gradients K1R is provided by the code as a function of temperature relative to the RT NDT of the material.

C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, K1R is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RT NDT* and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K1t, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

The heatup limit curve, Figure 3.1-1, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60°F per hour. The cooldown limit curves of Figure 3.1-2 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The cooldown limit curves are valid for cooldown rates up to 100°F/hr.

The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 68 EFPY for Units 1 and 2. The adjusted reference temperature was calculated using materials properties data from the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) documented in the most recent revision to BA W-1543 and reactor vessel neutron fluence data obtained from plant-specific analyses.

Amendment Nos.

r TS 3.1-lla The technical basis for the data points and the associated RT NDT values used to generate the heatup and cooldown curves is provided in WCAP-14177 (Reference 2) and were determined to be applicable to the 48 EFPY period of extended operation under first license renewal. The associated RT NDT values used to calculate the heatup and cooldown curves provided in WCAP-14177 (Revision 2) are based upon the Surry Unit 1 Intermediate to Lower Shell Circ Weld:

1/4-T, 228.4°F and 3/4-T, 189.5°F The heatup and cooldown curves for operation through 48 EFPY were based upon the K1r methodology. These heatup and cool down curves were subsequently evaluated using the K1c methodology for Subsequent License Renewal (SLR) at 68 EFPY in WCAP-18243-NP (Reference 3).

The limiting reactor vessel materials at 68 EFPY were determined to be the Surry Unit 1 Lower Shell Longitudinal Weld L2 at 1/4-T and the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld at 3/4-T. The associated RT NDT values calculated at 68 EFPY are:

1/4-T, 219.4°F and 3/4-T, 179.8°F The data points and the associated RT NDT values used to generate the heatup and cooldown curves in TS Figures 3.1-1 and 3.1-2, respectively, are conservative based upon use of the K1c methodology. Therefore, the heatup and cooldown curves did not require revision as a result of SLR. However, the fluence applicability is updated from 48 EFPY to 68 EFPY.

Amendment Nos.

TS 3.1-12 The reactor boltup temperature is defined in 10 CFR 50, Appendix G as "The highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload." The reactor vessel may be bolted up at a temperature greater than the initial RT NDT of the material stressed by the boltup (e.g., the vessel flange). As noted on Figures 3.1-1 and 3.1-2, the limiting boltup temperature is 10°F. An administrative minimum boltup temperature limit greater than 10°F is imposed in station procedures to ensure the Reactor Coolant System temperatures are sufficiently high to prevent damage to the reactor vessel closure head/vessel flange during the removal or installation of reactor vessel head bolts. The limiting boltup temperature and the administrative minimum boltup temperature limit are in effect when the reactor vessel head bolts are under tension.

References (1)

UFSAR, Section 4.1, Design Bases (2)

WCAP-14177, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (October 1994)

(3)

WCAP-18243, Rev. 2, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (July 2018)

Amendment Nos.

bD

~

GJ 2500.00 2000.00

i 1500.00 GJ..

a.

GJ r:o Ill a:

Ill "ti

~ 1000.00 "ti Ill B

"ti 500.00

-14.70 Figure 3.1-1 Surry Units 1 and 2 Reactor Coolant System Heatup Limitations Limiting Boltup Temperature Surry 1 Initial RT Nor Closure Flange Region: 10°F 0

50 100 150 200 250 300 350 400 450 500 550 600 650 Indicated Cold Leg Temperature {Deg. FJ Figure 3.1-1 : Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°F/hr) Applicable for 68 EFPY Amendment Nos.

iiii Cl) 2500.00 2000.00

i 1500.00 Cl)..

Q.

GI IIO C

Ill a:

Cl).,,

~ 1000.00 IP...

Ill u ;;

.E 500.00

-14.70 Figure 3.1-2 Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations Limiting Boltup Temperature Surry 1 Initial RT Nor Closure Flange Region: 10°F 0

50 100 150 200 250

.3,00 3,50 400 450 500 550 600 650 Indicated Cold Leg Temperature {Deg. F}

Figure 3.1-2 : Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 68 EFPY Amendment Nos.

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