Text

WCAP-18242-NP, Rev. 0, Surry, Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal
	ML18085A161
Person / Time
Site:	Surry
Issue date:	10/31/2017
From:	Alpan A, Lynch B; Westinghouse
To:	; Office of Nuclear Reactor Regulation
References
	WCAP-18242-NP, Rev 0
	Download: ML18085A161 (126)
	v • d • e

Attachment 3 Serial Number 18-098 Docket Nos. 50-280/281 WCAP-18242-NP, REV. 0, SURRY UNITS 1 AND 2 TIME-LIMITED AGING ANALYSIS ON REACTOR VESSEL INTEGRITY FOR SUBSEQUENT LICENSE RENEWAL OCTOBER 2017 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

SURRY POWER STATION UNITS 1 AND 2

WCAP-18242-NP Revision 0 Westinghouse Non-Proprietary Class 3 Surry Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal Westinghouse October 2017

Westinghouse Non-Proprietary Class 3 WCAP-18242-NP Revision 0 Surry Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal Brett Lynch*

Structural Design & Analysis Ill Arzu Alpan*

Radiation Engineering & Analysis October 2017 Reviewers: Benjamin E. Mays*

Structural Design & Analysis Ill Gregory A. Fischer*

Radiation Engineering & Analysis Approved:

Lynn A. Patterson*, Manager Structural Design & Analysis III Laurent P. Houssay*, Manager Radiation Engineering & Analysis

Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

Revision 0:

Original Issue WCAP-18242-NP Westinghouse Non-Proprietary Class 3 RECORD OF REVISION II October 2017 Revision 0

Westinghouse Non-Proprietary Class 3

)II TABLE OF CONTENTS LIST OF TABLES....................................................................................................................................... iv LIST OF FIGURES.................................................................................................................................... vii ACRONYMS............................................................................................................................................. viii EXECUTIVE

SUMMARY

........................................................................................................................... X 1

TIME-LIMITED AGING ANALYSIS......................................................................................... 1-l 2

CALCULATED FLUENCE......................................................................................................... 2-l 3

MATERIAL PROPERTY INPUT................................................................................................. 3-1 4

PRESSURIZED THERMAL SHOCK.......................................................................................... 4-l 5

UPPER-SHELF ENERGY........................................................................................................... 5-1 6

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES....................... 6-1 6.1 ADJUSTED REFERENCE TEMPERATURES AND P-T LIMIT CURVES APPLICABILITY............................................................................................................ 6-3 7

SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES.................................................. 7-l 8

REFERENCES............................................................................................................................. 8-1 APPENDIX A CREDIBILITY EVALUATION OF THE SURRY UNITS 1 AND 2 SURVEILLANCE DATA........................................................................................................................... A-1 APPENDIX B WELD MATERIAL HEAT # 0227 INITIAL RT NDT AND UPPER-SHELF ENERGY DETERMINATION........................................................................................................ B-1 APPENDIX C

SUMMARY

OF THE APPLICABILITY OF P-T LIMIT CURVES FOR SURRY UNITS 1 AND 2.......................................................................................................................... C-1 APPENDIX D EMERGENCY RESPONSE GUIDELINES.................................................................. D-1 WCAP-18242-NP October 2017 Revision 0

Table 1-1 Table 2-1 Table 2-2 Table 2-3 Table 2-4 Table 3-1 Table 3-2 Table 3-3 Table 3-4 Table 3-5 Table 3-6 Table 3-7 Table 3-8 Table 3-9 Table 3-10 Table 3-11 Table 3-12 Table 4-1 Table 4-2 Westinghouse Non-Proprietary Class 3 IV LIST OF TABLES Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3............... 1-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center for Surry Unit 1..................................................................................................................... 2-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center for Surry Unit 2..................................................................................................................... 2-3 Surry Unit 1 - Maximum Fast Neutron Fluence (E > 1.0 MeV) Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline Regions....................... 2-4 Surry Unit 2 - Maximum Fast Neutron Fluence (E > 1.0 MeV) Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline Regions....................... 2-5 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NOT Values, and Initial USE Values for the Surry Unit 1 RPV Beltline and Surveillance Materials............................ 3-9 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NOT Values, and Initial USE Values for the Surry Unit 1 RPV Extended Beltline Materials...................................... 3-10 Best-Estimate Cu and Ni Weight Percent Values, Initial RTNoT Values, and Initial USE Values for the Surry Unit 2 RPV Beltline and Surveillance Materials.......................... 3-11 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NOT Values, and Initial USE Values for the Surry Unit 2 RPV Extended Beltline Materials...................................... 3-12 Initial RT NOT Values for the Surry Unit 1 Replacement Reactor Vessel Closure Head and Vessel Flange Materials................................................................................................. 3-13 Initial RT NOT Values for the Surry Unit 2 Replacement Reactor Vessel Closure Head and Vessel Flange Materials................................................................................................. 3-13 Surveillance Data for Weld Wire Heat# 299L44.......................................................... 3-l 4 Surveillance Data for Weld Wire Heat# 72445............................................................. 3-15 Calculation of Position 2.1 CF Values for Surry Unit l................................................. 3-l 6 Summary of the Surry Unit 1 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1

....................................................................................................................................... 3-17 Calculation of Position 2.1 CF Values for Surry Unit 2................................................. 3-18 Summary of the Surry Unit 2 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1

....................................................................................................................................... 3-19 Calculation of Surry Unit 1 RT PTs Values for 68 EFPY (SLR) at the Clad/Base Metal Interface........................................................................................................................... 4-3 Calculation of Surry Unit 2 RT PTS Values for 68 EFPY (SLR) at the Clad/Base Metal Interface........................................................................................................................... 4-6 WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 V

Table 5-1 Predicted USE Values at 68 EFPY (SLR) for Surry Unit 1............................................. 5-4 Table 5-2 Predicted USE Values at 68 EFPY (SLR) for Surry Unit 2............................................. 5-6 Table 6.1-1 Calculation of the Surry Unit 1 Nozzle ART Values at the Surface Location for 48 EFPY

......................................................................................................................................... 6-5 Table 6.1-2 Calculation of the Surry Unit 1 ART Values at the l/4T Location for 48 EFPY............. 6-6 Table 6.1-3 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 48 EFPY............. 6-9 Table 6.1-4 Calculation of the Surry Unit 2 Nozzle ART Values at the Surface Location for 48 EFPY

....................................................................................................................................... 6-12 Table 6.1-5 Calculation of the Surry Unit 2 ART Values at the l/4T Location for 48 EFPY........... 6-13 Table 6.1-6 Calculation of the Surry Unit 2 ART Values at the 3/4T Location for 48 EFPY........... 6-15 Table 6.1-7 Calculation of the Surry Unit 1 Nozzle ART Values at the Surface Location for 68 EFPY

....................................................................................................................................... 6-17 Table 6.1-8 Calculation of the Surry Unit 1 ART Values at the l/4T Location for 68 EFPY........... 6-18 Table 6.1-9 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 68 EFPY........... 6-21 Table 6.1-10 Calculation of the Surry Unit 2 ART Nozzle Values at the Surface Location for 68 EFPY

....................................................................................................................................... 6-24 Table 6.1-11 Calculation of the Surry Unit 2 ART Values at the l/4T Location for 68 EFPY........... 6-25 Table 6.1-12 Calculation of the Surry Unit 2 ART Values at the 3/4T Location for 68 EFPY........... 6-27 Table 6.1-13 Summary of the Surry Units 1 and 2 Limiting ART Values Used in the Applicability Evaluation of the Reactor Vessel Heatup and Cooldown Curves.................................. 6-29 Table 7-1 Surry Unit 1 Surveillance Capsule Withdrawal Schedule............................................... 7-2 Table 7-2 Surry Unit 2 Surveillance Capsule Withdrawal Schedule............................................... 7-3 Table A-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation for Surry Unit 1

........................................................................................................................................ A-5 Table A-2 Calculation of Interim Chemistry Factors for the Credibility Evaluation for Surry Unit 2

........................................................................................................................................ A-6 Table A-3 Mean Chemical Composition and Temperature for Weld Heat# 299L44...................... A-7 Table A-4 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat# 299L44................................................................................................................ A-8 Table A-5 Mean Chemical Composition and Temperature for Weld Heat# 72445........................ A-9 Table A-6 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat# 72445................................................................................................................. A-10 Table A-7 Surry Unit 1 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line... A-11 Table A-8 Surry Unit 2 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line... A-12 WCAP-18242-NP October 2017 Revision 0

TableA-9 Table B-1 Table B-2 Table B-3 Table C-1 Table C-2 Table C-3 Table D-1 Westinghouse Non-Proprietary Class 3 vi Calculation of Residual vs. Fast Fluence for Surry Units 1 and 2............................... A-14 Weld Material Qualification Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227).......................................................... B-1 Supplemental Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227)............................................................................... B-2 Charpy V-Notch Test Data for Surry Unit 2 Surveillance Weld (Heat # 0227).............. B-3 Surry Units 1 and 2 P-T Limit Curve Applicability History........................................... C-1 Data Points for Surry Units 1 and 2 Current Technical Specifications Heatup P-T Limit Curves............................................................................................................................. C-5 Data Points for Surry Units 1 and 2 Current Technical Specifications Cooldown P-T Limit Curves................................................................................................................... C-7 Evaluation of Surry Units 1 and 2 ERG Limit Category................................................ D-1 WCAP-18242-NP October 2017 Revision 0

Figure 3-1 Figure 3-2 Figure 3-3 Figure 3-4 Figure 5-1 Figure 5-2 Figure C-1 Figure C-2 Westinghouse Non-Proprietary Class 3 Vil LIST OF FIGURES RPV Base Metal Material Identifications for Surry Unit 1.............................................. 3-5 RPV Weld Identifications for Surry Unit 1...................................................................... 3-6 RPV Base Metal Material Identifications for Surry Unit 2.............................................. 3-7 RPV Weld Identifications for Surry Unit 2...................................................................... 3-8 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Surry Unit 1 at 68 EFPY........................................ 5-8 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Surry Unit 2 at 68 EFPY........................................ 5-9 Surry Units 1 and 2 Heatup P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications................................................................................................. C-3 Surry Units 1 and 2 Cooldown P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications................................................................................................. C-4 WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 1/4T - one-quarter thickness 3/4T - three-quarter thickness 3D - three dimensional ACRONYMS ADAMS - Agencywide Documents Access and Management System AN0-1 -Arkansas Nuclear One Unit 1 ART - Adjusted Reference Temperature ASME -American Society of Mechanical Engineers ASTM - American Society for Testing and Materials B&PV - Boiler and Pressure Vessel BTP - Branch Technical Position CF - Chemistry Factor CFR - Code of Federal Regulations CLB - Current Licensing Basis CMTR-Certified Material Test Report(s)

COMS - Cold Overpressure Mitigating System CR Crystal River Unit 3 Cu - Copper EFPY - Effective Full Power Years EMA-Equivalent Margins Analysis EOC - End of Cycle EOL-End of License ERG - Emergency Response Guideline(s)

EOLE - End of License Extension FF - Fluence Factor FSAR - Final Safety Analysis Report ft-lb - Foot-Pound GALL-Generic Aging Lessons Learned GE - General Electric HSST - Heavy Section Steel Technology ID - Inner Diameter IS - Intermediate Shell LOCA-Loss-of-Coolant Accident LS - Lower Shell LST - Lowest Service Temperature LTOPS - Low Temperature Overpressure Protection system Ni-Nickel WCAP-18242-NP VIII October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 ACRONYMS - Continued NRC - Nuclear Regulation Commission NUREG - NRC technical report designation ~clear Regulatory ~ommission)

OD - Outer Diameter ONS Oconee Nuclear Station Unit 3 ORNL-Oak Ridge National Laboratory P-T - Pressure-Temperature PB Point Beach Unit 1 psi - pounds per square inch PTS - Pressurized Thermal Shock PWR - Pressurized Water Reactor PWROG - Pressurized Water Reactor Owners Group RlS - Regulatory Issue Summary RPV - Reactor Pressure Vessel RV - Reactor Vessel

~T NDT -Change in Reference Nil-Ductility Transition Temperature RT NDT - Reference Nil-Ductility Transition Temperature RT NDT(U) - Initial ( or Unirradiated) Reference Nil-Ductility Transition Temperature RT PTS - Reference Temperature for Pressurized Thermal Shock SE - Safety Evaluation SLR - Subsequent ( or Second) License Renewal SPEO - Subsequent Period of Extended Operation SRM - Standard Reference Material SSC - Systems, Structures, and Components TLAA-Time-Limited Aging Analysis TMil (orTMI-1)-Three Mile Island Unit 1 TMI2 (or TMI-2)-Three Mile Island Unit 2 T NOT - Nil-Ductility Transition Temperature US - Upper Shell USE - Upper-Shelf Energy WOG - Westinghouse Owners Group Wt.%- Weight Percent WCAP-18242-NP ix October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 X

EXECUTIVE

SUMMARY

This report presents the Time-Limited Aging Analyses (TLAA) for the Surry Units 1 and 2 reactor pressure vessels (RPVs) in accordance with the requirements of the License Renewal Rule, 10 CFR Part

54. TLAAs are calculations that address safety-related aspects of the RPV within the bounds of the current 60-year license. These calculations must also be evaluated to account for an extended period of operation (80 years) also termed Subsequent (or Second) License Renewal (SLR) period or Subsequent Period of Extended Operation (SPEO).

Surry Units l and 2 are currently licensed through 60 years of operation; therefore, with a 20-year license extension, the subsequent license renewal term is applicable through 80 years of operation.

The evaluations in this report for 60 years of operation are applicable through 48 effective full-power years (EFPY), which is deemed end of license extension (EOLE).

Similarly, evaluations in this report performed at 80 years of operation are applicable through 68 EFPY, which is deemed end of subsequent (or second) license renewal. Updated neutron fluence evaluations were performed and documented in WCAP-18028-NP, Revision 0, as well as in in Section 2 of this report. Updated neutron fluence evaluations were used to identify the Surry Units 1 and 2 extended beltline materials and as input to the reactor vessel (RV) integrity evaluations in support of license renewal.

In addition to the RV integrity TLAA evaluations, the Surry Units l and 2 surveillance data credibility evaluation is contained in Appendix A of this report. Appendix B contains an updated evaluation of weld Heat# 0227 initial material properties, and Appendix C contains a brief history of the Surry Units 1 and 2 Pressure-Temperature (P-T) limit curves. Appendix D provides an Emergency Response Guidelines (ERG) assessment for Surry Units l and 2.

A summary of results for the Surry Units l and 2 TLAA is provided below. Based on the results of this TLAA evaluation, it is concluded that the Surry Units 1 and 2 RV will continue to meet regulatory requirements through the SLR period of operation.

Fluence The RV beltline and extended beltline neutron fluence values applicable to a postulated 20-year license renewal period were calculated for the Surry Units 1 and 2 materials. The analysis methodologies used to calculate the Surry Units 1 and 2 vessel fluence values satisfy the requirements set forth in Regulatory Guide 1.190. See Section 2 for more details.

SLR Pressurized Thermal Shock All of the beltline and extended beltline materials in the Surry Units l and 2 RVs are projected to remain below the RTPTs screening criteria values of270°F for base metal and/or longitudinal welds and 300°F for circumferentially oriented welds (per 10 CFR 50.61) through SLR (68 EFPY). See Section 4 for more details.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 XI SLR Upper-Shelf Energy Many of the beltline and extended beltline materials in the Surry Units 1 and 2 RV are projected to remain above the upper-shelf energy (USE) screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G),

through SLR (68 EFPY). However, multiple materials must take advantage of a generic or material-specific Equivalent Margins Analysis (EMA). See Section 5 for more details.

EOLE and SLR Adjusted Reference Temperatures Adjusted Reference Temperatures (ARTs) are calculated at 48 and 68 EFPY. The 48 and 68 EFPY ART values are used to perform an applicability check on the existing pressure-temperature (P-T) limit curves for Surry Units 1 and 2. With the consideration of TLAA fluence projections, revised Position 2.1 chemistry factor values, and recalculated initial RT NDT values, the applicability of the Surry Units 1 and 2 cylindrical shell P-T limit curves may be extended to 68 EFPY. Nozzle P-T limit curves were developed in order to satisfy NRC Regulatory Issue Summary (RIS) 2014-11 guidance through 68 EFPY, as documented in WCAP-18243-NP. As documented in WCAP-18243-NP, the existing P-T limit curves (48 EFPY) bound the nozzle curves developed for 68 EFPY. See Section 6 for more details.

Surveillance Capsule Withdrawal Schedule The surveillance capsule withdrawal schedules reflect asset management objectives of identifying when capsules will reach approximate 80, 100, and 120 years of plant operation equivalent peak reactor vessel fluence values. 100 and 120 years of operation are approximated by multiplying 80 year fluence by 1.25 and by doubling 60 year fluence, respectively. With consideration of a 20-year license renewal to 80 years of operation (68 EFPY), an additional capsule withdrawal is recommended for both Surry Units 1 and 2.

A fifth capsule from Surry Unit 1 and a sixth capsule from Surry Unit 2 are recommended to be withdrawn to continue to satisfy ASTM E185-82. It is recommended that Capsule Z be withdrawn from Surry Unit 1 at approximately 44 EFPY (projected to occur in 2027) to provide surveillance capsule data in consideration of a 20-year license renewal to 80 years (68 EFPY) of plant operation. The remaining stand-by capsules (U, S, and Y) should remain in the Surry Unit 1 reactor vessel and continue to accrue irradiation. Capsules U, S, and Y are not required to support the current licensing basis. Capsules U and S may remain in the Surry Unit 1 reactor vessel through 68 EFPY, but Capsule Y should be removed prior to 58 EFPY (projected to occur in 2041).

For Surry Unit 2, it is recommended that Capsule U be withdrawn at approximately 49 EFPY (projected to occur in 2032) to provide surveillance capsule data in consideration of a 20-year license renewal to 80 years (68 EFPY) of plant operation. Capsules T and Z should remain in the Surry Unit 2 reactor vessel and continue to accrue irradiation. Capsules T and Z are not required to support the current licensing basis. Capsule Z may remain in the Surry Unit 2 reactor vessel through 68 EFPY, but Capsule T should be removed prior to 60 EFPY (projected to occur in 2044).

See Section 7 for more details.

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Westinghouse Non-Proprietary Class 3 1

TIME-LIMITED AGING ANALYSIS Time-Limited Aging Analyses (TLAAs) are those licensee calculations that:

1. Consider the effects of aging 1-1

2. Involve time-limited assumptions defined by the current operating term (e.g., 60 years)

3. Involve systems, structures, and components (SSCs) within the scope of license renewal

4. Involve conclusions or provide the basis for conclusions related to the capability of the SSC to perform its intended functions

5. Were determined to be relevant by the licensee in making a safety determination

6. Are contained or incorporated by reference in the current licensing basis (CLB)

The potential TLAAs for the reactor pressure vessel (RPV) are identified in Table 1-1 along with indication of whether or not they meet the six criteria of 10 CFR 54.3 (Reference 1) for TLAAs.

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Westinghouse Non-Proprietary Class 3 1-2 Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 Pressure-Low Temperature Pressurized Temperature Time-Limited Aging Analysis Calculated Thermal Upper-Shelf Limits for Overpressure Fluence Shock<*>

Energy Heatup and Protection System Cooldown Setpoints<h>

Considers the Effects of Aging YES YES YES YES YES Involves Time-Limited Assumptions Defined by the YES YES YES YES YES Current Operating Term Involves SSC Within the YES YES YES YES YES Scope of License Renewal Involves Conclusions or Provides the Basis for Conclusions Related to the YES YES YES YES YES Capability of SSC to Perform Its Intended Function Determined to be Relevant by the Licensee in Making a YES YES YES YES YES Safety Determination Contained or Incorporated by YES YES YES YES YES Reference in the CLB Notes:

(a) The limiting Pressurized Thermal Shock (PTS) values are used to determine the appropriate Emergency Response Guideline (ERG) Limits category for Surry Units 1 and 2 through the end of the potential subsequent license extension period. However, ERG limits are outside the scope of 10 CFR Part 54.3. ERG limits are discussed in Appendix D.

(b) Low Temperature Overpressure Protection System (LTOPS), which is also referred to as Cold Overpressure Mitigating System (COMS), setpoints were not explicitly addressed in this report; however, as concluded in Section 6, no changes were necessary to the applicability term of the current 48 EFPY P-T limit curves at Surry Units 1 and 2.

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Westinghouse Non-Proprietary Class 3 2-1 2

CALCULATEDFLUENCE For the initial 60-year EOLE term, the Surry Units 1 and 2 fracture toughness properties provide adequate margins of safety against vessel failure. However, as the reactor operates, neutron irradiation (fluence) reduces material fracture toughness. RPV integrity is assured by demonstrating that RPV material fracture toughness will remain at levels that resist brittle fracture throughout the period of SLR operation. The first step in the analysis of vessel embrittlement is calculation of the neutron fluence that causes increased embrittlement.

Estimated RPV beltline and extended beltline fast neutron fluences (E > 1.0 MeV) at the end of 80 years of operation were calculated for Surry Units l and 2. The analyses methodologies used to calculate the Surry Units 1 and 2 RPV fluences satisfy the guidance set forth in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 2). These methodologies have been approved by the U.S. NRC and are described in detail in Reference 3.

In accordance with Sections 3.1 and 4.2 of NUREG-2192 (Reference 4), materials exceeding a fast neutron fluence (E > 1.0 MeV) of 1.0 x 1017 n/cm2 at the end of the SLR period are evaluated for changes in fracture toughness. RPV materials that are not traditionally plant-limiting because of low levels of neutron radiation must now be evaluated to determine the accumulated fluence at SLR. Therefore, fast neutron fluence (E > 1.0 MeV) calculations were performed for the Surry Units 1 and 2 RPV 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, 1/4T flaw location in the inlet and outlet nozzle (References 5 and 6), 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 SLR. The materials that exceed the 1.0 x 1017 n/cm2 fast neutron fluence (E > 1.0 MeV) threshold, and were not evaluated in past analyses ofrecord as part of the traditional beltline, are referred to as extended beltline materials in this report and are evaluated to determine the effect of neutron irradiation embrittlement during the SLR period.

In performing the fast neutron exposure evaluations for the Surry Units l and 2 reactor vessels, a series of fuel-cycle-specific forward transport calculations were carried out using the following two-dimensional/one-dimensional fluence rate synthesis technique:

rp(r, e, z) = rp(r, B) X rp(r, z) rp(r) where rp(r,B, z) is the synthesized 3D neutron fluence rate distribution, rp(r,B\\s the transport solution in r,0 geometry, rp(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and rp(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at Surry Units 1 and 2.

All of the transport calculations were carried out using the DORT discrete ordinates code (Reference 7) with the BUGLE-96 cross-section library (Reference 8). The BUGLE-96 library provides a coupled 47-neutron-, 20-gamma-ray-group cross-section data set produced specifically for light water reactor WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-2 applications. In these analyses, anisotropic scattering was treated with a P5 Legendre expansion and the angular discretization was modeled with an S16 order of angular quadrature. Energy-and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.

The calculations for fuel Cycles I through 26 for Surry Unit 1 and fuel Cycles 1 through 25 for Surry Unit 2 determine the neutron exposure of the pressure vessel and surveillance capsules based on completed fuel cycles. For Surry Unit 1, projections for Cycle 27 and beyond were based on Cycle 26.

For Surry Unit 2, projections for Cycle 26 and beyond were based on Cycle 25. Projected results (Cycle 27 and beyond for Surry Unit 1 and Cycle 26 and beyond for Surry Unit 2) will remain valid as long as future plant operation is consistent with these assumptions.

Table 2-1 gives the Surry Unit 1 calculated fast neutron fluences (E > 1.0 MeV) for all withdrawn surveillance capsules (Capsules T, W, V, and X). Table 2-2 gives the Surry Unit 2 calculated fast neutron fluences (E > 1.0 MeV) for all withdrawn surveillance capsules (Capsules X, W, V, S, W-1, and Y). The EFPY and fast neutron fluences (E > 1.0 MeV) in Tables 2-1 and 2-2 were obtained from calculations performed to support the Measurement Uncertainty Recapture (MUR) power uprate. These fast neutron fluences (E > 1.0 MeV) were calculated using methodologies that follow the guidance of Regulatory Guide 1.190.

Table 2-1 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center for Surry Unit 1 Azimuthal Cumulative Fast Neutron Location Capsule from Core Irradiation Irradiation Fluence ID Cardinal Cycle(s)

Time (E > 1.0 MeV)

Axis(°)

(EFPY)

(n/cm2)

T 15 l

1.1 2.71E+l8 w

35 1-4 3.4 3.68E+18 V

15 1-8 8.0 1.80E+l9 X

25 1-12 16.1 2.llE+19 15 13-14 WCAP-18242-NP October 2017 Revision 0

Table 2-2 Westinghouse Non-Proprietary Class 3 2-3 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center for Surry Unit 2 Azimuthal Cumulative Fast Neutron Location Capsule from Core Irradiation Irradiation Fluence ID Cardinal Cycle(s)

Time (E > 1.0 MeV)

Axis(°)

(EFPY)

(n/cm 2

)

X 15 1

1.2 2.97E+l8 w

25 1-4 3.8 6.36E+ 18 V

15 1-8 8.4 l.89E+l9 s

45 1-13 15.0 l.07E+l9 W-1 15 11-14 5.3 7.80E+l8 y

25 1-12 20.3 2.72E+l9 15 13-17 Selected results for the pressure vessel from the neutron transport analyses are provided in Tables 2-3 and 2-4 for Surry Units 1 and 2, respectively. Calculated fast neutron fluences (E > 1.0 MeV) for reactor vessel materials, on the pressure vessel clad/base metal interface, is provided for the nominal EOC 26 for Surry Unit 1 (32.5 EFPY) and nominal EOC 25 for Surry Unit 2 (31.3 EFPY). Surry Units 1 and 2 80-year plant life corresponds to 68 EFPY.

From Table 2-3 it is observed that one outlet nozzle and two inlet 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 l.O x 1017 n/cm2 at the l/4T nozzle flaw location at 68 EFPY for Surry Unit l. From Table 2-4, it is observed that one outlet nozzle and two inlet 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 outlet and one inlet nozzle have fast neutron fluence (E > 1.0 MeV) greater than 1.0 x 1017 n/cm 2

at the l/4T nozzle flaw location at 68 EFPY for Surry Unit 2. Tables 2-3 and 2-4 indicate that the lower shell to lower vessel head circumferential weld will remain below 1.0 x 1017 n/cm2 through SLR for both Surry Units 1 and 2.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-4 Table 2-3 Surry Unit 1-Maximum Fast Neutron Fluence (E > 1.0 MeV) Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline Regions Material Fast Neutron Fluence (n/crn

32.5 EFPY 54 EFPY 68EFPY 1/4T Flaw in Outlet Nozzle Nozzle 1 1.53E+l6 2.69E+l6 3.45E+ l6 Nozzle 2 1.08E+ 16 l.93E+ l6 2.49E+ l6 Nozzle 3l*I 4.48E+l6 7.59E+l6 9.62E+ l6 1/4T Flaw in Inlet Nozzle Nozzle tl01 5.80E+ l6 9.82E+l6 l.24E+ l7 Nozzle 2 l.40E+16 2.50E+l6 3.22E+16 Nozzle 3 l.98E+l6 3.48E+16 4.46E+16 Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1 3.62E+l6 6.35E+ l6 8.13E+ 16 Nozzle 2 2.55E+ l6 4.55E+ l6 5.86E+16 Nozzle 3\\cJ l.06E+l7 l.79E+ 17 2.27E+17 Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1\\d) 1.42E+l7 2.40E+ l7 3.04E+17 Nozzle 2 3.43E+ l6 6.10E+16 7.84E+16 Nozzle 3\\eJ 4.85E+ l6 8.5 1E+16 l.09E+17 Nozzle Shell 3.64E+ 18 6.00E+18 7.54E+ l8 Nozzle Shell to Intermediate Shell Circumferential Weld 3.64E+ l8 6.00E+ l8 7.54E+ l8 Intermediate Shell Plate 1

3. 17E+ l9 5.06E+l9 6.29E+l9 Plate2

3. 17E+ l9 5.06E+ l9 6.29E+ l9 Intermediate Shell Longitudinal Welds Weld 1 5.75E+l8 9.85E+l8 l.25E+ l9 Weld 2 5.75E+I8 9.85E+18 1.25E+l9 Intermediate Shell to Lower Shell Circumferential Weld 3.18E+ l9 5.08E+l9 6.3 1E+19 Lower Shell Plate l 3.20E+ l9 5.l 1E+l9 6.35E+ l9 Plate 2 3.20E+ l9 5.11E+l9 6.35E+ l9 Lower Shell Longitudinal Welds Weld 1 5.80E+18 9.94E+18 1.26E+ l9 Weld2 5.80E+ l8 9.94E+18 l.26E+ l9 Lower Shell to Lower Vessel Head Circumferential Weld

<1E+17

<IE+l7

< IE+ l7 Notes:

(a) l/4T flaw in Outlet Nozzle 3 is projected to reach I.OE+ 17 n/cm2 at approximately 70.7 EFPY.

(b) l/4T flaw in Inlet Nozzle I is projected to reach l.OE+l 7 n/cm2 at approximately 55.0 EFPY.

(c) Outlet Nozzle 3 forging to vessel shell weld reached I.OE+ 17 n/cm2 at approximately 30.8 EFPY.

(d) Inlet Nozzle I forging to vessel shell weld reached I.OE+ 17 n/cm2 at approximately 23.2 EFPY.

72 EFPY 3.67E+16 2.65E+l6 l.02E+l7 1.32E+l7 3.42E+I6 4.74E+l6 8.63E+l6 6.23E+l6 2.40E+ l7 3.22E+ l 7 8.34E+ l6 l.1 6E+ l7 7.98E+l8 7.98E+18 6.65E+l9 6.65E+l9 l.33E+l9 l.33E+l9 6.67E+l9 6.70E+l9 6.70E+l9 l.34E+19 1.34E+l9

<IE+l7 (e) Inlet Nozzle 3 forging to vessel shell weld is projected to reach l.OE+l 7 n/cm2 at approximately 62.8 EFPY.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-5 Table 2-4 Surry Unit 2 - Maximum Fast Neutron Fluence (E > 1.0 MeV) Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline Regions Material Neutron Fluence [n/cm2]

31.3 EFPY 54 EFPY 68EFPY 1/4T Flaw in Outlet Nozzle Nozzle 1 l.49E+16 2.66E+l6 3.38E+16 Nozzle 2 l.09E+l6 1.95E+16 2.48E+l6 Nozzle 3l*J 4.29E+l6 8.28E+ l6 l.07E+ l7 l/4T Flaw in Inlet Nozzle Nozzle Jl01 5.55E+ l6 l.07E+l7 l.39E+17 Nozzle 2 l.41E+l6 2.52E+l6 3.21E+l6 Nozzle 3 l.93E+l6 3.44E+l6 4.37E+16 Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle I 3.52E+l6 6.27E+l6 7.96E+ l6 Nozzle 2 2.57E+l6 4.60E+l6 5.85E+ l6 Nozzle 3l<J l.01E+17 l.95E+l7 2.53E+17 Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle Jl"J l.36E+17 2.62E+ l7 3.40E+17 Nozzle 2 3.45E+ 16 6.17E+ l6 7.84E+ l6 Nozzle 3leJ 4.73E+16 8.41E+16 l.07E+l7 Nozzle Shell 3.52E+l8 6.70E+ l8 8.65E+18 Nozzle Shell to Intermediate Shell Circumferential Weld 3.52E+ l8 6.70E+ l8 8.65E+ l8 Intermediate Shell Plate I 3.10E+l9 5.64E+ l9 7.20E+19 Plate2 3.IOE+19 5.64E+ l9 7.20E+ l9 Intermediate Shell Longitudinal Welds Weld 1 5.98E+l8 l.03E+19 1.29E+ l9 Weld2 5.98E+l8 1.03E+ 19 l.29E+ l9 Intermediate Shell to Lower Shell Circumferential Weld 3.l lE+l9 5.66E+19 7.22E+l9 Lower Shell Plate 1 3.12E+l9 5.68E+ l9 7.26E+l9 Plate2 3.12E+19 5.68E+l9 7.26E+19 Lower Shell Longitudinal Welds Weld I 6.03E+l8 l.03E+l9 1.30E+19 Weld2 6.03E+l8 1.03E+ 19 1.30E+ l9 Lower Shell to Lower Vessel Head Circumferential Weld

<IE+l7

<1E+l7

<IE+ l7 Notes:

(a) l/4T flaw in Outlet Nozzle 3 is projected to reach I.OE+ 17 n/cm2 at approximately 63.8 EFPY.

(b) l/4T flaw in Inlet Nozzle 1 is projected to reach l.OE+ 17 n/cm2 at approximately 50.9 EFPY.

( c) Outlet Nozzle 3 forging to vessel shell weld reached 1.0E+ 17 n/cm2 at approximately 31.0 EFPY.

( d) Inlet Nozzle I forging to vessel shell weld reached I.OE+ 17 n/cm2 at approximately 23.5 EFPY.

72 EFPY 3.58E+l6 2.63E+16 1.15E+ l7 l.48E+17 3.40E+l6 4.63E+l6 8.45E+ l6 6.20E+16 2.70E+l7 3.62E+l7 8.32E+ l6 l.13E+l7 9.21E+l8 9.21E+l8 7.65E+ l9 7.65E+l9 l.36E+ l9 l.36E+ l9 7.67E+ l9 7.71E+19 7.71E+ l9 1.37E+l9 1.37E+l9

< IE+l7

( e) Inlet Nozzle 3 forging to vessel shell weld is projected to reach I.OE+ 17 n/cm2 at approximately 63.9 EFPY.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-1 3

MATERIAL PROPERTY INPUT The Surry Unit 1 beltline materials consist of two (2) Intermediate Shell (IS) Plates, two (2) Lower Shell (LS) Plates, one (1) Upper Shell (US) Forging (also termed nozzle shell forging), two (2) IS Longitudinal Welds, two (2) LS Longitudinal Welds, and two (2) circumferential welds: the IS to LS Circumferential Weld and the US to IS Circumferential Weld. The Surry Unit 1 surveillance plate material was made from reactor vessel Lower Shell Plate C4415-l. Since Lower Shell Plate C4415-l shares a heat number with Lower Shell Plate C4415-2, the surveillance plate results also apply to Lower Shell Plate C4415-2. The Surry Unit l reactor vessel beltline LS Longitudinal weld (L2) was fabricated using weld wire Heat #

299L44, Linde 80 Flux Type, Lot Number 8596. The weld material in the Surry Unit 1 surveillance program was fabricated with the same material heat, flux type, and lot number as reactor vessel beltline Longitudinal Weld L2. Weld material Heat # 299L44 was included in the surveillance programs of other plants, as summarized in Table 3-7. The US to IS Circumferential Weld (W06) was fabricated with weld wire Heat # 25017, SAF 89 Flux Type, Flux Lot Number 1197. The IS to LS Circumferential Weld (W05) was fabricated with weld wire Heat# 72445, Linde 80 Flux Type, Flux Lot Number 8597 (40%) and Flux Lot Number 8623 (60%). Surveillance data does not exist for Heat# 25017 or Heat# 72445 in the Surry Unit 1 reactor vessel surveillance program; however weld wire Heat # 72445 was included in the surveillance programs of other plants, as summarized in Table 3-8. The LS Longitudinal Weld (LI) and both IS Longitudinal Welds (L3 and L4) were fabricated using weld wire Heat# 8Tl554, Linde 80 Flux Type, Flux Lot Number 8579. Surveillance data does not exist for Heat # 8Tl554.

The Surry Unit 2 beltline materials consist of two (2) Intermediate Shell (IS) Plates, two (2) Lower Shell (LS) Plates, one (1) Upper Shell (US) Forging, two (2) IS Longitudinal Welds, two (2) LS Longitudinal Welds, and two (2) circumferential welds: the IS to LS Circumferential Weld and US to IS Circumferential Weld. The Surry Unit 2 surveillance plate material was made from reactor vessel Lower Shell Plate C4339-l. Since Lower Shell Plate C4339-l shares a heat number with Intermediate Shell Plate C4339-2, the surveillance plate results also apply to Intermediate Shell Plate C4339-2. The Surry Unit 2 reactor vessel beltline IS to LS Circumferential Weld (W05) was fabricated using weld wire Heat # 0227, Grau Lo Flux Type, Flux Lot Number LW320. The weld material in the Surry Unit 2 surveillance program was fabricated with the same material heat, flux type, and lot number as the IS to LS Circumferential Weld. The US to IS circumferential weld (W06) was fabricated with weld wire Heat #

4275, SAF 89 Flux Type, Flux Lot Number 02275. Weld material Heat # 0227 and Heat# 4275 are not included in the surveillance programs of other plants. The IS Longitudinal Weld L3 and 50% of IS Longitudinal Weld L4 were fabricated with weld wire Heat # 72445, Linde 80 Flux Type, Flux Lot Number 8597. Data does not exist for Heat # 72445 in the Surry Unit 2 reactor vessel surveillance program; however, weld wire Heat# 72445 was included in the surveillance programs of other plants, as summarized in Table 3-8. The remaining 50% of IS Longitudinal Weld L4, LS Longitudinal Weld LI, and 63% of LS Longitudinal Weld L2 were fabricated from weld wire Heat # 8T 1762, Linde 80 Flux Type, Flux Lot Number 8597. The remaining 37% of LS Longitudinal Weld L2 was fabricated from weld wire Heat# 8Tl762, Linde 80 Flux Type, Flux Lot Number 8632. Surveillance data does not exist for Heat#

8Tl762.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-2 Based on the results of Section 2 of this report, the materials that exceeded the l x 1017 n/cm2 (E > 1.0 MeV) threshold at 68 EFPY are considered to be the Surry Units 1 and 2 extended beltline materials and are evaluated to determine their impact on the proposed SLR period of operation. The forgings and welds corresponding to the Surry Units 1 and 2 Inlet Nozzles 1, Inlet Nozzles 3, and Outlet Nozzles 3 are predicted to experience neutron fluence greater than 1.0 x 1017 n/cm2 at SLR. However, for conservatism all of the Surry Units 1 and 2 inlet and outlet nozzle materials are considered part of the extended beltline.

The Surry Units 1 and 2 extended beltline materials consist of three (3) Inlet Nozzles, three (3) Outlet Nozzles, three (3) Inlet Nozzle to US Welds, and three (3) Outlet Nozzle to US Welds per Unit. The Surry Unit 1 Inlet Nozzle to Upper Shell Welds were fabricated using Heat #s 299L44 and 8T 1762, Linde 80 Flux Type, Lot Number 8596. The Surry Unit 1 Outlet Nozzle to Upper Shell Welds were fabricated using Heat# 8Tl 762, Linde 80 Flux Type, Lot Number 8578 and Heat# 8T1554B, Linde 80 Flux Type, Flux Lot Number 8579. The Surry Unit 2 lnlet Nozzle to Upper Shell Welds were fabricated using Heat#

8Tl 762, Linde 80 Flux Type, and Lot Numbers 8597 and 8632. The materials constituting the Surry Unit 2 Outlet Nozzle to Upper Shell Welds could not be determined; however, these welds were completed at Rotterdam per BAW-2313, Revision 7, Supplement 1, Revision l (Reference 9). Surveillance data from Surry Unit 1 and additional plant surveillance programs exists, as previously described, for Heat #

299L44. No additional surveillance data exists for any of the materials in the Surry Units l and 2 extended beltline. The data supporting this materials summary was gathered primarily from Reference

10.

The identification of the reactor vessel (RV) beltline and extended beltline plate and weld materials are included in Figures 3-1 and 3-2 for Surry Unit 1 and Figures 3-3 and 3-4 Surry Unit 2 per PWROG-16045-NP, Revision O (Reference 10). The material property inputs used for the subsequent RV integrity evaluations contained in this report are described in this section. Note that some of the beltline material initial properties were updated from previous RV integrity evaluations per PWROG-16045-NP, Revision O and Appendix E herein, and the fluence values were updated per WCAP-18028-NP, Revision O (Reference 11) and Section 2 herein. Additionally, initial USE values are supplied in Table 3-1 and Table 3-3 for certain welds, which had an initial USE value designated as "EMA" in PWROG-16045-NP, Revision 0. The sources and methods used in the determination of the chemistry factors and the fracture toughness properties are summarized below.

Chemical Compositions The best-estimate copper (Cu) and nickel (Ni) chemical compositions for the Surry Units 1 and 2 beltline and extended beltline materials are presented in Tables 3-1 through 3-4. The best-estimate weight percent copper and nickel values for the beltline and extended beltline materials were previously reported in PWROG-16045-NP, Revision O and were included in RV integrity evaluations as part of this TLAA effort.

Fracture Toughness Properties The fracture toughness properties (initial RT NDT and initial USE) of most of beltline plate materials were originally determined using NUREG-0800, BTP 5-3 Position 1.1 (Reference 12) methodology, with three exceptions. Surry Unit 1 IS Plate C4326-l, Surry Unit 1 LS Plate C4415-l, and Surry Unit 2 LS Plate WCAP-18242-NP October 201 7 Revision 0

Westinghouse Non-Proprietary Class 3 3-3 C4339-l were determined using the ASME Code,Section III (Reference 13). Many of the beltline and extended beltline fracture toughness properties were updated per ASME Section III, the GE Method (Reference 14), and NUREG-0800, BTP 5-3 Position 1.1 methodologies, as described in PWROG-16045-NP, Revision O (Reference 10). The initial RT NOT values for Surry Unit 1 Longitudinal Welds Ll, L2, L3, and L4 and Intermediate to Lower Shell Circumferential Weld Heat # 72445 were determined using the "Master Curve" method (RT NDT = T0 + 35°F) per BAW-2308 Revision 1-A Safety Evaluation (SE) and Revision 2-A SE (References 15 and 16). The initial RT NOT values for Surry Unit 2 Longitudinal Welds Ll, L2, L3, and L4 were also determined using this method. Chemistry factor (CF) values and margin terms require evaluation when using "Master Curve"-generated initial RT NDT values to calculate adjusted reference temperature (ART) and reference temperature for pressurized thermal shock (RT PTs) values.

When using these "Master Curve"-generated initial RT NDT values, the CF and margin terms will be adjusted to a minimum of l 67°F and 28°F, respectively. However, if the material-specific CF value or margin term is greater than 167°F or 28°F, respectively, the material-specific value(s) will be used. The most up-to-date initial RTNoT and initial USE values are documented in PWROG-16045-NP, Revision 0 for Surry Units 1 and 2 with the exception of the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227), which was updated in Appendix E herein. Table 8 of PWROG-16045-NP contains the Surry Unit I initial properties, and Table 9 of PWROG-16045-NP contains the Surry Unit 2 initial properties. The Surry Unit 2 IS to LS Circumferential Weld initial material properties are updated in Appendix B herein. The beltline and extended beltline material properties of the Surry Units 1 and 2 reactor vessels are presented in Tables 3-1 through 3-4 herein.

The initial RT NOT values of the reactor vessel flange and closure head serve as input to the P-T limit curves "flange-notch" per 10 CFR 50, Appendix G (Reference 17) and were confirmed to be acceptable.

Since Surry Units 1 and 2 share P-T Limit curves for operation, materials for both plants must be considered concerning input acceptability. The closure heads at both Surry Units I and 2 have been replaced, and the initial RT NOT values of the Surry Units 1 and 2 flange materials were updated in PWROG-16045-NP, Revision O (Reference I 0). The Surry Unit 1 replacement closure head has an initial RT NOT value of -67°F, determined per ASME Code Section Ill, NB-2300. The Surry Unit 1 reactor vessel flange bas an initial RT NDT of -l l 4.6°F, calculated using the GE methodology.

The Surry Unit 2 replacement bead bas an initial RT NOT value of -60°F, determined per ASME Code Section III, NB-2300.

The Surry Unit 2 reactor vessel flange has an initial RTNoT of -156.3°F, calculated using the GE methodology. Since the basis for the flange region of the current Surry Units 1 and 2 P-T Limit curves is 10°F, all of the updated initial RT NDT values are bounded by the value used to establish the current 48 EFPY P-T Limit curves flange limitations. Thus, the reactor vessel flange and closure bead initial RT NDT values currently in use remain conservative, and the P-T limit curves flange requirements need not change. See Tables 3-5 and 3-6 for a summary of the initial RT NOT values for these two components at each plant.

Chemistry Factor Values The chemistry factor (CF) values were calculated using Positions 1.1 and 2.1 of Regulatory Guide 1.99, Revision 2 (Reference 18). Position 1.1 uses Tables 1 and 2 from the Regulatory Guide along with the best-estimate copper and nickel weight percent values (contained in Tables 3-1 through 3-4, and Tables 3-7 and 3-8). Position 2.1 uses the surveillance capsule data from all capsules tested to date and surveillance data from other plants, as applicable. A credibility evaluation of the surveillance data is provided in Appendix A. The calculated capsule fluence values are provided in Tables 2-1 and 2-2 and are WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-4 used to determine the Position 2.1 CFs as shown in Tables 3-9 and 3-11 for Surry Units I and 2, respectively. Tables 3-10 and 3-12 summarize the Positions I.I and 2.1 CF values determined for the Surry Units 1 and 2 RPV beltline and extended beltline materials, respectively.

WCAP-18242-NP October 201 7 Revision 0

Westinghouse Non-Proprietary Class 3 vent pipe (MK-23}*-----..

control rod,nechaniam*

~~-housing (HK-A2 thru Al~)

slwoud $\\1pport lifting lug--~~--..

closure stud---.....

(Kl:<-6'2) closure nut-.~~~ -t (Ml<-63}

,---cloaure head cap closure-head flange.

spnerical washers (t1K-6~ and 65)

inner 0-ring gasket

~t,11-----........---------'1¥.t.::Ei (MK-71.)

vessel flange leak monitoring tube (MK-60 and outlet nozzle 3x(HK-59)

Vessel *suppo~tj -*

pad

. ~*..

-.. ~

~..

~

instr:.1mentation nozzles (HK-B91 thru B113)

Figure 3-1 upper she[l interm:di te course

- lower. she l core sup4rt (MK-70)

--out-ex- 0-ring gasket (MK-72)

Lrefueli~g seal ledge CHK-66) t-'...;~:.::s:sn

(

head cap RPV Base Metal Material Identifications for Surry Unit 1

Note: Figure may not be representative of the replacement RV closure head at Surry Unit 1.

3-5 WCAP-18242-NP October 2017 Revision 0

9.0" 19.. 7*

48.3..

WCAP-18242-NP Westinghouse Non-Proprietary Class 3 C1RCUMf(RfN1'Jlt. SUHS VnlJCAt S_lutS 210*

... i,

-l

e#

C -

-j

... I OS Figure 3-2 RPV Weld Identifications for Surry Unit 1 3-6 CA326,..1 October 2017 Revision 0

SPARE CROM HEAD AOAPlER CLOSURE HEAD LIFTING LUG REAClOR VESSEL CLOSUR( Hf.AO Westinghouse Non-Proprietary Class 3 THERMAL SLEEVE ClOS£0 HEAD AOAPTER MOUNTING LUC CLOSURE STUO(MK-62).

NUT{MK-63),SPHE:RICAL WASHERS(MK-§~ond MK-65 l

INTtRt.lE.Ol A TE SHELL C0URS£ CORE SUPPORT CUJDE (MK-70)

LOWER HEAD RING I.

' I LOWER SHELL COURSE I

I Figure 3-3 RPV Base Metal Material Identifications for Surry Unit 2

Note: Figure may not be representative of the replacement RV closure head at Surry Unit 2.

3-7 WCAP-18242-NP October 2017 Revision 0

9.0" COflt 144" CL 19.7" 48.3" WCAP-18242-NP Westinghouse Non-Proprietary Class 3 Ct~tRMtAl S[lf1S VERTlCAL SCAMS t70-

-W06 1eo*

... --* -i go*

'¥ I -

-wos 210*

C4339*-1 90° Figure 3-4 RPV Weld Identifications for Surry Unit 2 3-8 October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-9 Table 3-1 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Surry Unit l RPV Beltline and Surveillance Materials RPV Material Wt.%

Wt.%

RTNDT(U) a, Initial USE Cu Ni (OF)

(OF)

(ft-lb)

Reactor Vessel Beltline Materiali 0J Upper Shell Forging 122Vl09VA1 0.11 0.74 40 0

114 Intermediate Shell Plate C4326-l 0.11 0.55 10 0

115 Intermediate Shell Plate C4326-2 0.11 0.55 11.4 0

94 Lower Shell Plate C4415-l 0.102 0.493 20 0

103 Lower Shell Plate C4415-2 0.11 0.50 4.6 0

82 Upper to Intermediate Shell Circumferential Weld 0.33 0.10 0

20.0 2:64(b)

(Heat # 25017)

Intermediate Shell Longitudinal Welds 0.16 0.57

-48.6 18.0 64(b)

L3 and L4 (Heat # 8Tl554)

Intermediate to Lower Shell Circumferential Weld 0.22 0.54

-72.5 12.0 64(b)

(Heat # 72445)

Lower Shell Longitudinal Weld L1 0.16 0.57

-48.6 18.0 64(b)

(Heat # 8Tl 554)

Lower Shell Longitudinal Weld L2 0.34 0.68

-74.3 12.8 64(b)

(Heat # 299L44)

Reactor Vessel Surveillance MaterialicJ 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, Revision O (Reference 10), unless otherwise noted.

(b)

Per Surry Power Station UFSAR (Reference 19), reactor vessel Equivalent Margins Analysis (EMA) report BAW-2494, Revision I (Reference 20) has been approved for these welds. The EMAs are updated for SLR under PWROG P A-MSC-1481. Linde 80 initial USE values are set to the generic value of 64 ft-lbs per BA W-2313, Revision 7, Supplement 1, Revision 1 (Reference 9). 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.

(c)

The surveillance plate data was taken to be the same as the vessel plate data. The surveillance weld data was obtained from BA W-2324 (Reference 21 ).

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-10 Table 3-2 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Surry Unit 1 RPV Extended Beltline Materials Wt.%

Wt.%

RTNDT(U)

RPV Material Cu Ni (OF)

Reactor Vessel Extended Beltline Materiali 0>

Inlet Nozzle 1 (Heat # 9-4787) 0.159 0.85 10.3 Inlet Nozzle 2 (Heat # 9-5078) 0.159 0.87 11.6 Inlet Nozzle 3 (Heat # 9-48 19) 0.159 0.84

-47.2 Outlet Nozzle 1 (Heat # 9-4825-1) 0.159 0.85

-44.9 Outlet Nozzle 2 (Heat # 9-4762) 0.159 0.83

-87.5 Outlet Nozzle 3 (Heat # 9-4788) 0.159 0.84

-50.2 Inlet Nozzle to Upper Shell Heat # 299L44 0.34 0.68

-7.0 Welds Heat # 8T 1762 0.19 0.57

-4.9 Outlet Nozzle to Upper Shell Heat # 8Tl 762 0.19 0.57

-4.9 Welds Heat # 8Tl554B 0.16 0.57

-4.9 Note:

(a) All values were taken from Table 8 of PWROG-16045-NP, Revision O (Reference 10).

WCAP-18242-NP a,

(OF) 0 0

0 0

20.6 19.7 19.7 19.7 Initial USE (ft-lb) 63 64 68 68 82 71 64 64 64 64 October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-11 Table 3-3 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Surry Unit 2 RPV Beltline and Surveillance Materials RPV Material Wt.%

Wt.%

RTNDT(U) a, Initial USE Cu Ni (OF)

(OF)

(ft-lb)

Reactor Vessel Beltline Material/a)

Upper Shell Forging 123 V303V A 1 0.11 0.72 30 0

104 Intermediate Shell Plate C433 !-2 0.12 0.60 15.0 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-l 0.107 0.53

-4.4 0

101 Upper to Intermediate Shell Circumferential Weld 0.35 0.10 (Heat# 4275) 0 20.0 2'.68(b)

Intermediate Shell Longitudinal Welds 0.22 0.54

-72.5 12.0 64(b)

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

Intermediate Shell Longitudinal Weld 0.19 0.57

-48.6 18.0 64(b)

L4 (ID 50%) (Heat# 8Tl 762)

Intermediate to Lower Shell Circumferential Weld 0.187 0.545 o(c) o<c) 8i{c)

(Heat# 0227)

Lower Shell Longitudinal Welds LI and L2 0.19 0.57

-48.6 18.0 64(b)

(Heat# 8Tl 762)

Reactor Vessel Surveillance Material/dJ Lower Shell Plate C4339-l 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 ofPWROG-16045-NP, Revision O (Reference 10), unless otherwise noted.

(b) Per Surry Power Station UFSAR (Reference 19), reactor vessel EMA report BA W-2494, Revision 1 (Reference

20) has been approved for these welds. The EMAs are updated for SLR under PWROG PA-MSC-1481. Linde 80 initial USE values are set to the generic value of 64 ft-lbs per BAW-2313, Revision 7, Supplement I, Revision 1 (Reference 9). 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.

( c) Initial properties are established in Appendix B. Since the initial RT NDT is based on measured data, cr1 is equal to 0°F. Per Surry Power Station UFSAR (Reference 19), reactor vessel EMA report BAW-2494, Revision 1 (Reference 20) has been approved for this weld. The EMA is updated for SLR under PWROG PA-MSC-1481.

(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 (Reference 22).

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-12 Table 3-4 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Surry Unit 2 RPV Extended Beltline Materials Wt.%

Wt.%

RT DT(U)

<J1 Initial RPV Material USE Cu Ni (OF)

(OF)

(ft-lb)

Reactor Vessel Extended Be/t/ine Materia/s(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 # 8Tl 762 0.19 0.57

-4.9 19.7 64 Welds Outlet Nozzle to Upper Shell Welds Rotterdam 0.35 1.0 30 0

71 (b)

Notes:

(a) All values were taken from Table 9 of PWROG-16045-NP, Revision O (Reference 10). Associated cr1 values are also available from PWROG-16045-NP, Revision 0.

(b) Per PWROG-16045-NP, Revision O (Reference 10), this initial USE value is set equal to the USE value of the first tested capsule from WCAP-16001 (Reference 22). This methodology utilizes BTP 5-3 (Reference 12),

Position 1.2 guidance, as no USE data is available from the supplier.

WCAP-1 8242-NP October 2017 Revision 0

Table 3-5 Table 3-6 Westinghouse Non-Proprietary Class 3 3-13 Initial RT NOT Values for the Surry Unit 1 Replacement Reactor Vessel Closure Head and Vessel Flange Materials RPV Material Initial RT NOT (°F)

Replacement Closure Head

-6i*>

E4381/E4382 Vessel Flange FV-1870

-144.6(b)

Notes:

(a) Value taken from Table 8 of PWROG-16045-NP, Revision O (Reference 10). This value is based on ASME Code Section III, NB-2300 criteria. Note that the original Closure Head Flange initial RT NDTwas 10°F per WCAP-14177 (Reference 23).

(b) Value taken from Table 8 of PWROG-16045-NP, Revision 0. This value is based on the GE Methodology. Note that the Vessel Flange Initial RT NUT used in previous reactor vessel integrity calculations was 10°F as documented in WCAP-14177.

Initial RT NOT Values for the Surry Unit 2 Replacement Reactor Vessel Closure Head and Vessel Flange Materials RPV Material Initial RT NDT (°F)

Replacement Closure Head

-60<*)

02Wl-l-l-l Vessel Flange FV-2542

-156.3(b)

Notes:

(a) Value taken from Table 9 of PWROG-16045-NP, Revision O (Reference 10). This value is based on ASME Code Section III, NB-2300 criteria. Note that the original Closure Head Flange initial RT NUT was 10°F per WCAP-14177 (Reference 23).

(b) Value taken from Table 9 of PWROG-16045-NP, Revision 0. This value is based on the GE Methodology. Note that the Vessel Flange Initial RT NDT used in previous reactor vessel integrity calculations was -65°F as documented in WCAP-14177.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 Table 3-7 Surveillance Data for Weld Wire Heat# 299L44 Capsule Cu Ni CF Fluence ART DT Irradiation Capsule Designation<*>

wt.%

(OF)

(x 1019 n/cm2, Temperature (OF)

(OF)

E > 1.0 MeV)

TM12-LG1 (CR-3ibl 0.37 0.70 234.0 0.830 216 556 Wl(CR-3i°l 0.37 0.70 234.0 0.780 262 545 TMil-E 0.33 0.67 215.2 0.107 74 556 TMI1-C 0.33 0.67 215.2 0.882 166 556 TMI2-LG 1 (TMI-2t>

0.33 0.67 215.2 0.968 226 556 CR3-LG 1 (ONS-3) 0.36 0.70 230.5 0.779 202 556 A5(d) 0.23 0.64 175.8 2.75 246.6 556 Surry Unit 1: Capsule T 0.23 0.64 175.8 0.271 171 537 Surry Unit I: Capsule V 0.23 0.64 175.8 1.80 250 539 Surry Unit 1: Capsule X 0.23 0.64 175.8 2.11 234 542 Notes:

(a) Data was obtained from ANP-2650 (Reference 24), unless otherwise noted. Material source is indicated in parentheses.

(b) Material is from different sources, irradiated in the same capsule.

(c) Capsule Wl was irradiated in Surry Unit 2. The fluence value is updated from ANP-2650 per MCOE-LTR-17-26, Revision 1 (Reference 25) and Section 2. The irradiation temperature value is the time-weighted average Tcold considering the cycles that Wl was inside the Surry Unit 2 reactor vessel.

(d) Data taken from AREVA-17-01417 (Reference 26).

3-14 WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-15 Table 3-8 Surveillance Data for Weld Wire Heat# 72445 Capsule Fluence ARTNDT Irradiation Cu wt.

Niwt.

CF Capsule Designation<*>

(OF)

(x 1019 n/cm2, (OF)

Temperature E > 1.0 MeV)

(OF)

CR3-LG1 (AN0-1) 0.22 0.59 165.5 0.510 139 556 CR3-LG2 (AN0-1) 0.22 0.59 165.5 1.670 164 556 Wl (ANO-lib>

0.22 0.59 165.5 0.780 138 545 Point Beach Unit 1: Capsule V 0.23 0.62 172.4 0.634 107 542 Point Beach Unit 1: Capsule S 0.23 0.62 172.4 0.829 165 542 Point Beach Unit l: Capsule R 0.23 0.62 172.4 2.190 155 541.6 Point Beach Unit 1: Capsule T 0.23 0.62 172.4 2.230 181 533.4 Notes:

(a)

Data was obtained from ANP-2650 (Reference 24), unless otherwise noted. Material source is inilicated in parentheses.

(b)

Capsule Wl was irradiated in Surry Unit 2. The fluence value is updated from ANP-2650 per MCOE-LTR-17-26, Revision 1 (Reference 25) and Section 2. The irradiation temperature value is the time-weighted average Tcold considering the cycles that WI was inside the Surry Unit 2 reactor vessel.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-16 Table 3-9 Calculation of Position 2.1 CF Values for Surry Unit 1<a>

Capsule Adjusted FF* Adjusted RPV Material Capsule Fluence FF<c>

ARTNDT ARTNDT (d)

ARTNDT FF 2

(x 1019 n/cm2, (OF)

E > 1.0 MeV)

(OF)

T 0.271 0.644 50 50 32.21 0.415 Lower Shell V

1.80 1.1 61 113 113 131.23 1.349 Plate C44 l 5-l (b)

X 2.11 1.203 86 86 103.46 1.447 (Longitudinal)

SUM:

266.91 3.211 CF rMI5-I = I:(FF

L'i.RTNDT '..,. I:(FF2) = (266.91)..,. (3.211) = 83.1°F T

0.271 0.644 171 208 133.69 0.415 V

1.80 1.161 250 309 358.56 1.349 X

2.11 1.203 234 293 351.89 1.447 TMI2-LGI 0.830 0.948 216 230 217.98 0.898 Surveillance WI 0.780 0.930 262 265 246.53 0.865 Weld Material TMI1-E 0.107 0.431 74 91 39.02 0.185 (Heat # 299L44)

TMI1-C 0.882 0.965 166 185 178.87 0.93 1 TMI2-LGI 0.968 0.991 226 247 244.95 0.982 CR3-LGI 0.779 0.930 202 216 200.87 0.865 A5 2.75 1.270 246.6 326 413.61 1.612 SUM:

2385.98 9.550 CF Heat# 29QL44 = I:(FF

L'i.RTNnT)..,. I:(FF2) = (2385.98)..,. (9.550) = 249.8°F CR3-LG1 0.510 0.812 139 153 124.24 0.659 CR3-LG2 1.67 1.141 164 178 203.15 1.303 Surveillance Wl 0.780 0.930 138 141 131.17 0.865 Weld Material PB-1 : V 0.634 0.872 107 107 93.34 0.761 (Heat # 72445)

PB-1 : S 0.829 0.947 165 165 156.32 0.898 PB-I : R 2.19 1.213 155 155 187.48 1.471 PB-I : T 2.23 1.217 181 172 209.86 1.482 SUM:

1105.56 7.438 CF Heat#72445 = I:(FF

L'i.RT NDT) -;- I:(FF2) = (1105.56) -;- (7.438) = 148.6°F Notes:

(a)

Fluence and ti.RT NDT data taken from Tables 3-7 and 3-8, unless otherwise noted.

(b)

Surry Unit I Lower Shell Plate C4415-l fluence values obtained from Section 2. ti.RT NDT values obtained from BAW-2324 (Reference 21).

(c)

FF= fluence factor = t<0*2s-o.,o*Iog(f)J_

(d)

The surveillance weld ti.RT NDT values have been adjusted, as applicable, first by adding the temperature adjustment, then by multiplying by a ratio detem1ined using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry. Pre-adjusted values are listed in the ti.RT NDT column. Temperature adjustment = l.O*(Tcapsule - Tplant), where Tplant = 542°F for Surry Unit 1 and Tcapsule is the irradiation temperature in Table 3-7 or 3-8. The temperature adjustment procedure is not utilized when plant-specific capsules are analyzed alone. The ratio procedure is applicable only to surveillance welds and the ratio applied = CFvessel weld I CFsurv. Weld* If the ratio procedure yields a ratio less than 1, a ratio of 1.00 is utilized; this approach is conservative.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-17 Table 3-10 Summary of the Surry Unit 1 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Chemistry Factor RPV Material Position 1.1 Position 2.1 (OF)

(OF)

Reactor Vessel Beltline Materials Upper Shell Forging 122Vl09VA1 76.1 Intermediate Shell Plate C4326-l 73.5 Intermediate Shell Plate C4326-2 73.5 Lower Shell Plate C4415-l 66.6 83.l(a)

Lower Shell Plate C4415-2 73.0 83.l(a)

Upper to Intermediate Shell Circumferential Weld (Heat# 25017) 152.0 Intermediate Shell Longitudinal Welds 143.9(c)

L3 and L4 (Heat# 8T1554)

Intermediate to Lower Shell Circumferential Weld (Heat# 72445) 158.o(c) 148.6(c)

Lower Shell Longitudinal Weld L1 143.9(c)

(Heat# 8Tl554)

Lower Shell Longitudinal Weld L2 220.6(c) 249.8(c)

(Heat# 299L44)

Reactor Vessel Extended Beltline Materials

Inlet Nozzle 1 (Heat# 9-4787) 123.5 Inlet Nozzle 2 (Heat# 9-5078) 123.7 Inlet Nozzle 3 (Heat # 9-4819) 123.4 Outlet Nozzle 1 (Heat# 9-4825-1) 123.5 Outlet Nozzle 2 (Heat # 9-4762) 123.3 Outlet Nozzle 3 (Heat # 9-4788) 123.4 Inlet Nozzle to Upper Heat# 299L44 220.6 249.8 Shell Welds Heat# 8Tl 762 152.4 Outlet Nozzle to Upper Heat# 8Tl 762 152.4 Shell Welds Heat# 8Tl554B 143.9 Reactor Vessel Surveillance Materials Lower Shell Plate C44 l 5-l 66.6 Surveillance Weld (Heat# 299L44) 175.8 Notes:

(a) Since Lower Shell Plate C44 l 5-1 shares a heat number with Lower Shell Plate C4415-2, the surveillance plate results also apply to Lower Shell Plate C44 l 5-2.

(b) The nozzle forging Cu wt.% values were conservatively rounded up to 0.16 for the purposes of CF determination.

(c) Linde 80 weld wire initial RTNDT values were established using master curve data. Per BAW-2308 Revision 1-A Safety Evaluation (SE) and Revision 2-A SE (References 15 and 16) Chemistry Factors must be adjusted to a minimum of 167°F when used in ART and RT PTS calculations. If the Position I.I CF is greater than l 67°F, it is used in calculations.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 Table 3-11 Calculation of Position 2.1 CF Values for Surry Unit zea>

Capsule Adjusted FF* Adjusted RPV Fluence dRTNDT dRTNoT (d)

Material Capsule (x 1019 n/cm2, FF<c>

(OF)

(OF) dRTNDT E > 1.0 MeV)

(OF)

Lower Shell X

0.297 0.668 59.08 59.08 39.45 Plate C4339-l V

1.89 1.174 79.12 79.12 92.91 (Longitudinal) y 2.72 1.267 114.22 114.22 144.72 X

0.297 0.668 48.67 48.67 32.50 Lower Shell V

1.89 1.174 63.60 63.60 74.68 Plate C4339-l y

2.72 1.267 106.81 106.81 135.33 (Transverse)

SUM:

519.59 CF C4339-1 = r(FF * ~RTNDT)..;- r(FF2) = (519.59)..;- (6.860, = 1s.1°F X

0.297 0.668 95.65 95.65 63.86 Surveillance V

1.89 1.174 140.21 140.21 164.64 Weld Material (Heat # 0227) y 2.72 1.267 178.32 178.32 225.94 SUM:

454.45 CF Heat# 0227 = L(FF * ~RT NDT) 7 L(FF2) = (454.45) 7 (3.430) = 132.5°F CR3-LGI 0.510 0.812 139 152 123.43 CR3-LG2 1.67 1.141 164 177 202.01 Surveillance WI 0.780 0.930 138 140 130.24 Weld PB-1: V 0.634 0.872 107 106 92.46 Material Cb)

(Heat # 72445)

PB-I: S 0.829 0.947 165 164 155.37 PB-I : R 2.19 1.213 155 154 186.26 PB-I : T 2.23 1.217 181 171 208.64 SUM:

1098.42 CF Hent#72445 = L(FF * ~RT NOT)~ L(FF2) = (1098.42)..;- (7.438) = 147.7°F Notes:

(a)

Fluence and ~RT NOT data are from WCAP-16001, Revision O (Reference 22), unless otherwise noted.

(b)

Fluence and ~RT NOT data are from Table 3-8.

(c)

FF= fluence factor = t<0*28-0*10*1og(f)) _

3-1 8 FF2 0.446 1.379 1.605 0.446 1.379 1.605 6.860 0.446 1.379 1.605 3.430 0.659 1.303 0.865 0.761 0.898 1.471 1.482 7.438 (d)

The surveillance weld ~RT NOT values have been adjusted, as applicable, first by adding the temperature adjustment, then by multiplying by a ratio determined using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry. Pre-adjusted values are listed in the

~RTNoT column. Temperature adjustment = l.O*(Tcapsu1e-Tp1an,), where Tplant = 543°F for Surry Unit 2 and Tcapsule is the irradiation temperature in Table 3-8. The temperature adjustment procedure is not utilized when plant-specific capsules are analyzed alone. The ratio procedure is applicable only to surveillance welds and the ratio applied = CF vessel weld / CFsurv. Weld* If the ratio procedure yields a ratio less than 1, a ratio of 1.00 is utilized; this approach is conservative.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-19 Table 3-12 Summary of the Surry Unit 2 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Chemistry Factor RPV Material Position 1.1 Position 2.1 (OF)

(OF)

Reactor Vessel Be/tline Materials Upper Shell Forging 123V303VA1 75.8 Intermediate Shell Plate C433 l-2 83.0 Intermediate Shell Plate C4339-2 73.4 75.i*)

Lower Shell Plate C4208-2 107.3 Lower Shell Plate C4339-l 70.8 75_7<*)

Upper to Intermediate Shell Circumferential 160.5 Weld (Heat# 4275)

Intermediate Shell Longitudinal Welds 158.o(c) 147.ic)

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

Intermediate Shell Longitudinal Weld 152.4<c)

L4 (ID 50%) (Heat# 8Tl 762)

Intermediate to Lower Shell Circumferential 147.5 132.5 Weld (Heat # 0227)

Lower Shell Longitudinal Welds Ll and L2 152.4<c)

(Heat# 8Tl 762)

Reactor Vessel Extended Beltline Material/bl Inlet Nozzle 1 (Heat# 9-5104) 123.4 Inlet Nozzle 2 (Heat# 9-4815) 123.7 Inlet Nozzle 3 (Heat# 9-5205) 123.6 Outlet Nozzle l (Heat# 9-4825-2) 123.5 Outlet Nozzle 2 (Heat# 9-5086-1) 123.6 Outlet Nozzle 3 (Heat# 9-5086-2) 123.7 Inlet Nozzle to Upper Heat# 8Tl 762 152.4 Shell Welds Outlet Nozzle to Upper Rotterdam 272.0 Shell Welds Reactor Vessel Surveillance Materials Lower Shell Plate C4339-1 70.8 Surveillance Weld (Heat # 0227) 150.8 Notes:

(a) Since Lower Shell Plate C4339-l shares a heat number with Intermediate Shell Plate C4339-2, the surveillance plate results also apply to Intermediate Shell Plate C4339-2.

(b) The nozzle forging Cu wt. % values were conservatively rounded up to 0.16 for the purposes of CF determination.

(c) Linde 80 weld wire initial RTNoT values were established using master curve data. Per BAW-2308 Revision 1-A SE and Revision 2-A SE (References 15 and 16) Chemistry Factors must be adjusted to a minimum of 167°F when used in ART and RT PTS calculations. If the Position I. I CF is greater than 167°F, it is used in calculations.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 4-1 4

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

In 1985, the U.S. NRC issued a formal ruling on PTS (10 CFR 50.61 [Reference 27)) that established screening criteria on pressurized water reactor (PWR) vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTs-RTrTs screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end of license. The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RT PTs) values consistent with the methods given in Regulatory Guide 1.99, Revision 2 (Reference 18).

These accepted methods were used with the surface fluence values of Section 2 to calculate the following RT PTS values for the Surry Units 1 and 2 RPV materials at 68 EFPY. The SLR RT PTS calculations are summarized in Tables 4-1 and 4-2 for Surry Units 1 and 2, respectively.

PTS Conclusion The Surry Units 1 and 2 limiting RT PTs value for base metal or longitudinal weld materials at 68 EFPY is 253.2°F (see Table 4-1 and Table 4-2), which corresponds to Surry Unit 1 Lower Shell Longitudinal Weld L2 Heat# 299L44 (using credible surveillance data). The Surry Units 1 and 2 limiting RT PTs value for circumferentially oriented welds at 68 EFPY is 229.8°F (see Table 4-1 and Table 4-2), which corresponds to the Surry Unit 1 Intermediate to Lower Shell Circumferential Weld Heat # 72445. The Surry Units 1 and 2 limiting RT PTS material for base metal or longitudinal weld materials under the previous 48 EFPY analysis was Surry Unit 1 Lower Shell Longitudinal Weld L2 Heat# 299L44. This material remains the limiting base metal or longitudinal weld material at 68 EFPY. The Surry Units 1 and 2 limiting RTrTs material for circumferentially oriented welds at under the previous 48 EFPY analysis was the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld. This material is no longer the limiting circumferentially oriented weld material at 68 EFPY. Note that the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227) is not the limiting circumferentially oriented weld, because this material takes advantage of credible surveillance data. Limiting fluence values corresponding to the lowest extent of the nozzle welds was used to conservatively calculate the RT PTS values for both the nozzle welds and nozzle forgings.

Therefore, all of the beltline and extended beltline materials in the Surry Units 1 and 2 reactor vessel are below the RT PTs screening criteria values of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds through SLR (68 EFPY).

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 4-2 The Alternate PTS Rule (10 CFR 50.61a [Reference 28)) was published in the Federal Register by the NRC in 2010. This alternate rule is less restrictive than the PTS Rule (10 CFR 50.61) and is intended to be used for situations in which the IO CFR 50.61 criteria cannot be met. Surry Units 1 and 2 meet the criteria for the PTS Rule through SLR and therefore do not utilize the Alternate PTS Rule at this time.

WCAP-18242-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 Table 4-1 Calculation of Surry Unit 1 RT PTs Values for 68 EFPY (SLR) at the Clad/Base Metal Interface R.G.

Surface 1.99, Wt.

Wt.

CF<*>

Fluence<h>

Surface RTNDT(U)

(c)

RPV Material Rev. 2 cu<*>

Ni<*>

(OF)

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

Position E > 1.0 MeV)

Reactor Vessel Belt/ine Materials Upper Shell Forging 122V I 09V A I 1.1 0.11 0.74 76.1 0.754 0.921 40 Upper to Intermediate Shell I. I 0.33 0.1 0 152.0 0.754 0.921 0

Circumferential Weld (Heat # 25017)

Intermediate Shell Plate C4326-l 1.1 0.11 0.55 73.5 6.29 1.445 10 Intermediate Shell Plate C4326-2 1.1 0.11 0.55 73.5 6.29 1.445 11.4 Intermediate Shell Longitudinal Welds 1.1 0.16 0.57 167.0 1.25 1.062

-48.6 L3 and L4 (Heat # 8T 1554)

Intermediate to Lower Shell I. I 0.22 0.54 167.0

-72.5 6.31 1.445

_ Circumferential Weld_(Heat # 72445) _ ---------------- --------- -------- --------------------- ---------------

Using credible surveillance data 2.1 167.0 6.31 1.445

-72.5 Lower Shell Plate C44 l 5-l I. I 0.102 0.493 66.6 6.35 1.447 20 Using credible surveillance data 2.1 83.1 6.35 1.447 20 Lower Shell Plate C44 l 5-2 1.1 0.11 0.50 73.0 6.35 1.447 4.6 Using credible surveillance data 2.1 83.1 6.35 1.447 4.6 Lower Shell Longitudinal Weld LI I. I 0.16 0.57 167.0 1.26 1.064

-48.6 (Heat # 8Tl 554)

Lower Shell Longitudinal Weld L2 1.1 0.34 0.68 220.6 1.26 1.064

-74.3

__________ @eat# 299L44)_ ________________ ----------------- ----------- --------- ----------------------- --------------e------------

Using credible surveillance data 2.1 249.8 1.26 1.064

-74.3 WCAP-18242-NP 4-3 ARTNDT (d) au (c)

(OF)

(OF) 70.1 0.0 140.0 20.0 106.2 0.0 106.2 0.0 177.4 18.0 241.4 12.0

~---- --

241.4 12.0 96.3 0.0 120.2 0.0 105.6 0.0 120.2 0.0 177.8 18.0 234.8 12.8 265.9 12.8
'1t,. (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 28.0 66.6 28.0 60.9
28. 0 60.9 17.0 34.0 8.5 17.0 17.0 34.0 8.5 17.0 28.0 66.6 28.0 61.6 28.0 61.6 October 2017 Revision 0 RTrTs (OF) 144.l 208.8 150.2 151.6 195.4 229.8 229.8 150.3 157.2 144.2 141.8 195.7 222.l 253.2
Westinghouse Non-Proprietary Class 3 Table 4-1 Calculation of Surry Unit 1 RT PTs Values for 68 EFPY (SLR) at the Clad/Base Metal Interface R.G.
Surface 1.99, Wt.
Wt.
CF<*>
Fluence(bJ Surface RTNDT(U)
(c)
RPV Material O/o (OF)
(x 10 19 n/cm2, FF(bl (OF)
Rev. 2 cu<*)
Ni<*>
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle 1 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.0304 0.221
-7.0 (Heat# 299L4'!) ------------
~- ~------ ---------- --------------------- -----------~----------
~--
Using credible surveillance data 2.1 249.8 0.0304 0.221
-7.0 Inlet Nozzle 2 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.00784 0.093
-7.0
__________ _(Heat# 299L44) __________________ ----------------~--- ----------- ----------- ------------------------- ---------------e--*---------
Using credible surveillance data 2.1 249.8 0.00784 0.093
-7.0 Inlet Nozzle 3 to Upper Shell Weld I. I 0.34 0.68 220.6 0.0109 0.116
-7.0
_______ (Heat# 299L4~-------------- ----------

---------- ------------------- --- -----------~----------

Using credible surveillance data 2.1 249.8 0.0109 0.116
-7.0 Inlet Nozzle 1 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.0304 0.221
-4.9 (Heat # 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld
1. 1 0.19 0.57 152.4 0.00784 0.093
-4.9 (Heat# 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.0109 0.116
-4.9 (Heat# 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld I.I 0.19 0.57 152.4 0.00813 0.095
-4.9 (Heat# 8Tl 762)
Outlet Nozzle 2 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00586 0.075
-4.9 (Heat# 8Tl 762)
Outlet Nozzle 3 to Upper Shell Weld I.I 0.19 0.57 152.4 0.0227 0.186
-4.9 (Heat# 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld 1.1 0.16 0.57 143.9 0.00813 0.095
-4.9 (Heat# 8Tl554B)
WCAP-1 8242-NP 4-4
~TNDT (d)
G u (c)
(OF)
(OF) 48.8 20.6 55.3 20.6 0.0 (20.4) 20.6 0.0 (23.2) 20.6 25.6 20.6 29.0 20.6 33.7 19.7 0.0 (14.1) 19.7 0.0 (17.7) 19.7 0.0 (14.5) 19.7 0.0 ( 11.5) 19.7 28.3 19.7 0.0 (13.7) 19.7 Gt, (c)
Margin (OF)
(OF) 24.4 63.9 14.0 49.8 0.0 41.2
~------- -----
0.0 41.2 12.8 48.5 14.0 49.8 16.9 51.9 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 14.2 48.5 0.0 39.4 October 2017 Revision 0 RTPTs (OF) 105.7 98.1 34.2 34.2 67.2 71.8 80.7 34.5 34.5 34.5 34.5 72.0 34.5
Westinghouse Non-Proprietary Class 3 4-5 Table 4-1 Calculation of Surry Unit 1 RT PTS Values for 68 EFPY (SLR) at the Clad/Base Metal Interface R.G.
Surface 1.99, Wt.
Wt.
CF<*>
Fluence<bl Surface RTNDT(U)
(c) aRTNDT (d)
<Ju(c) a,.,. (c)
Margin RPV Material O/o O/o (x 1019 n/cm2, FF(bl
{°F)
(OF)
Rev. 2 cu<*>
Ni<*l (OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Outlet Nozzle 2 to Upper Shell Weld 1.1 0.16 0.57 143.9 0.00586 0.075
-4.9 0.0 (10.8) 19.7 0.0 39.4 (Heat# 8Tl 554B)
Outlet Nozzle 3 to Upper Shell Weld (Heat# 8Tl 554B) 1.1 0.16 0.57 143.9 0.0227 0.186
-4.9 26.8 19.7 13.4 47.6 Inlet Nozzle 1 (Heat# 9-4787) 1.1 0.159 0.85 123.5 0.0304 0.221 10.3 27.3 0.0 13.7 27.3 Inlet Nozzle 2 (Heat# 9-5078) 1.1 0.159 0.87 123.7 0.00784 0.093 11.6 0.0(11.5) 0.0 0.0 0.0 Inlet Nozzle 3 (Heat# 9-4819) 1.1 0.159 0.84 123.4 0.0109 0.116
-47.2 0.0 (14.3) 0.0 0.0 0.0 Outlet Nozzle I (Heat# 9-4825-1) 1.1 0.159 0.85 123.5 0.00813 0.095
-44.9 0.0 (11.7) 0.0 0.0 0.0 Outlet Nozzle 2 (Heat# 9-4762) 1.1 0.159 0.83 123.3 0.00586 0.075
-87.5 0.0 (9.3) 0.0 0.0 0.0 Outlet Nozzle 3 (Heat# 9-4788)
I. I 0.159 0.84 123.4 0.0227 0.186
-50.2 0.0 (22.9) 0.0 0.0
0.0 Notes
(a)
Chemical composition values taken from Tables 3-1 and 3-2 of this report. Chemistry Factor values taken from Table 3-10 of this report.
(b)
Surface fluence values taken from Section 2 of this report. FF= fluence factor= t<0*2s-o.io*Jog(f)l.
(c)
Initial RT NDT and a1 values are taken from Tables 3-1 and 3-2.
(d)
Calculated L\\RT NDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29). Actual calculated L\\RT NDT values are listed in parentheses for these materials.
(e)
Per Appendix A of this report, all Surry Unit I surveillance data was deemed credible. Per the guidance of 10 CFR 50.61 (Reference 27), the base metal a,.,. = l 7°F for Position 1.1, and a,.,. = 8.5°F for Position 2.1 with credible surveillance data. Also per 10 CFR 50.61, the weld metal a,.,. =
28°F for Position I.I, and with credible surveillance data a,.,. = 14°F for Position 2.1. However, a,.,. need not exceed 0.5*1'.\\RTNDT* For welds utilizing initial RT NDT values based on BA W-2308, a6 = 28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 RTrTs (OF) 34.5 69.5 65.0 11.6
-47.2
-44.9
-87.5
-50.2
Westinghouse Non-Proprietary Class 3 Table 4-2 Calculation of Surry Unit 2 RT PTs Values for 68 EFPY (SLR) at the Clad/Base Metal Interface R.G.
Surface 1.99, Wt.
Wt.
CF<*>
Fluence<b)
Surface RTNDT(U)
(c)
RPV Material O/o O/o Rev. 2 cu<*>
Ni<*>
(OF)
(x 1019 n/cm2, FF(b>
(OF)
Position E > 1.0 MeV)
Reactor Vessel B eltline Materials Upper Shell Forging 123V303VA1 1.1 0.11 0.72 75.8 0.865 0.959 30 Upper to Intermediate Shell 1.1 0.35 0.10 160.5 0.865 0.959 0
Circumferential Weld (Heat # 4275)
Intermediate Shell Plate C433 l -2 1.1 0.12 0.60 83.0 7.20 1.467 15.0 Intermediate Shell Plate C4339-2 1.1 0.11 0.54 73.4 7.20 1.467 7.8 Using non-credible surveillance data 2.1 75.7 7.20 1.467 7.8 Intermediate Shell Longitudinal Welds 1.1 0.22 0.54 167.0 1.29 1.071
-72.5 L3 and L4 (OD 50%) (Heat # 72445)
Using credible surveillance data 2.1 167.0 1.29 1.071
-72.5 Intem1ediate Shell Longitudinal Weld L4 1.1 0.19 0.57 167.0 1.29 1.071
-48.6 (ID 50%) (Heat # 8Tl 762)
Intermediate to Lower Shell 1.1 0.187 0.545 147.5 0
Circumferential Weld (Heat # 0227) 7.22 1.468 Using credible surveillance data 2.1 132.5 7.22 1.468 0
Lower Shell Plate C4208-2 1.1 0.15 0.55 107.3 7.26 1.469
-30 Lower Shell Plate C4339-l 1.1 0.107 0.53 70.8 7.26 l.469
-4.4
~------------------------------------------- ------------- -------- ---- ---------- ---------------------- ------------------------
Using non-credible surveillance data 2.1 75.7 7.26 1.469
-4.4 Lower Shell Longitudinal Welds LI and 1.1 0.19 0.57 167.0 1.30 1.073
-48.6 L2 (Heat # 8T 1762)
Reactor Vessel Extended Beltline Materials Inlet Nozzle 1 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.0340 0.236
-4.9 (Heat# 8Tl 762)
WCAP-18242-NP 4-6
.6.RTNDT (d) a/>
(OF)
(OF) 72.7 0.0 154.0 20.0 121.8 0.0 107.7 0.0 111.1 0.0 178.8 12.0 178.8 12.0 178.8 18.0 216.5 0.0 194.5 0.0 157.6 0.0 104.0 0.0 111.2 0.0 179.2 18.0 36.0 19.7 Ge,. (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 17.0 34.0 28.0 60.9 28.0 60.9 28.0 66.6 28.0 56.0 14.0 28.0 17.0 34.0 17.0 34.0
17. 0 34.0 28.0 66.6 18.0 53.4 October 2017 Revision 0 RTrTs (OF) 136.7 222.8 170.8 149.5 152.9 167.3 167.3 196.8 272.5 222.5 161.6 133.6 140.8 197.2 84.4
Westinghouse Non-Proprietary Class 3 4-7 Table 4-2 Calculation of Surry Unit 2 RT PTs Values for 68 EFPY (SLR) at the Clad/Base Metal Interface R.G.
Surface 1.99, Wt.
Wt.
CF<*>
Fluence<hl Surface RTNDT(U)
(c)
~TNDT (d)
<Ju(c)
GA (c)
Margin RPV Material (x 10 19 n/cm2, FF(bl (OF)
(OF)
(OF)
(OF)
Rev. 2 cu<*)
Ni<*>
(OF)
(OF)
Position E > 1.0 MeV)
Inlet Nozzle 2 to Upper Shell Weld 1.1 0.19 (Heat# 8T I 762) 0.57 152.4 0.00784 0.093
-4.9 0.0 (14. 1) 19.7 0.0 39.4 Inlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.0107 (Heat# 8Tl 762)
0. I 15
-4.9 0.0 (17.5) 19.7 0.0 39.4 Outlet Nozzle 1 to Upper Shell Weld
1. 1 0.35 (Rotterdam) 1.0 272.0 0.00796 0.094 30.0 25.5 0.0 12.7 25.5 Outlet Nozzle 2 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.00585 0.075 30.0 0.0 (20.5) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.0253 0.199 30.0 54.0 0.0 27.0 54.0 (Rotterdam)
Inlet Nozzle 1 (Heat# 9-5 104)
I.I 0.159 0.84 123.4 0.0340 0.236
-29.7 29.1 0.0 14.6 29.1 Inlet Nozzle 2 (Heat# 9-48 I 5) 1.1
0. 159 0.87 123.7 0.00784 0.093 4.5 0.0 (I I.5) 0.0 0.0 0.0 Inlet Nozzle 3 (Heat# 9-5205)
I. I 0.1 59 0.86 123.6 0.0107
0. I 15 6.5 0.0 (14.2) 0.0 0.0 0.0 Outlet Nozzle 1 (Heat# 9-4825-2) 1.1 0.159 0.85 123.5 0.00796 0.094
-58.1 0.0 (1 1.6) 0.0 0.0 0.0 Outlet Nozzle 2 (Heat# 9-5086-1 )
1.1 0.159 0.86 123.6 0.00585 0.075
-26.6 0.0 (9.3) 0.0 0.0 0.0 Outlet Nozzle 3 (Heat# 9-5086-2) 1.1 0.159 0.87 123.7 0.0253 0.199
-33.8 0.0 (24.6) 0.0 0.0
0.0 Notes
(a)
Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistrl Factor values taken from Table 3-12 of this report.
(b)
Surface tluence values taken from Section 2 of this report. FF= tluence factor= t<02 -o.wiog(f)l.
(c)
Initial RT NOT and G1 values taken from Tables 3-3 and 3-4.
(d)
Calculated t1RTNoT values less than 25°F have been reduced to zero perTLR-RES/DE/CIB-2013-01 (Reference 29). Actual calculated t1RTNoT values are listed in parentheses for these materials.
(e)
Per Appendix A of this report, the surveillance plate data were deemed non-credible, whereas the surveillance data for the weld materials were deemed credible. Per the guidance of IO CFR 50.61 (Reference 27), the base metal Gr, = I 7°F for Position 1.1 and for Position 2. 1 with non-credible surveillance data. Per 10 CFR 50.61, the weld metal Gr, = 28°F for Position 1.1, and with credible surveillance data Gr, = 14°F for Position 2.1. However, Gr, need not exceed 0.5*t1RT NDT* For welds utilizing initial RT NDT values based on BA W-2308, Gr, = 28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 RTPTS (OF) 34.5 34.5 81.0 30.0 138.0 28.6 4.5 6.5
-58.1
-26.6
-33.8
Westinghouse Non-Proprietary Class 3 5-1 5
UPPER-SHELF ENERGY The decrease in Charpy upper-shelf energy (USE) is associated with the determination of acceptable RPV toughness during the license renewal period when the vessel is exposed to additional irradiation.
The requirements on USE are included in 10 CFR 50, Appendix G (Reference 17).
10 CFR 50, Appendix G requires utilities to submit an analysis at least three years prior to the time that the USE of any RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.
There are two methods that 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, Revision 2 (Reference 18). 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, Revision 2. 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 the Regulatory Guide to predict the change in USE (Position 2.2) of the RPV material due to irradiation.
The 68 EFPY Position 1.2 USE values of the vessel materials can be predicted using the corresponding l/4T fluence projections, the copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2.
The predicted Position 2.2 USE values are determined for the reactor vessel materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding 1/4T fluence projection. The reduced plant surveillance data was obtained from Table 7-6 ofBAW-2324 (Reference 21) for Surry Unit 1. The reduced plant surveillance data was obtained from Table 5-12 of WCAP-16001, Revision O (Reference 22) for Surry Unit 2. The surveillance data was plotted in Regulatory Guide 1.99, Revision 2, Figure 2 (see Figures 5-1 and 5-2 of this report) using the surveillance capsule fluence values documented in Table 2-1 of this report for Surry Unit 1 and Table 2-2 of this report for Surry Unit 2. Bounding material fluence values, only, are shown in Figures 5-1 and 5-2 for some materials. This data was fitted by drawing a line parallel to the existing lines as the upper bound of all the surveillance data. These reduced lines were used instead of the existing lines to determine the Position 2.2 SLR USE values.
The projected USE values were calculated to determine if the Surry Units 1 and 2 beltline and extended beltline materials remain above the 50 ft-lb criterion at 68 EFPY (SLR). These calculations are summarized in Tables 5-1 and 5-2. Fluence values corresponding to the lowest extent of the nozzle welds at the surface were used to conservatively calculate the projected USE values for the nozzle forgings.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-2 USE Conclusion For Surry Unit 1, the limiting USE value at 68 EFPY is 32 ft-lb (see Table 5-1); this value corresponds to the Intermediate to Lower Shell Circumferential Weld (Heat# 72445) using Position 1.2. For Surry Unit 2, the limiting USE value at 68 EFPY is 41 ft-lb (see Table 5-2); this value corresponds to the Upper to Intermediate Shell Circumferential Weld (Heat# 4275) using Position 1.2.
The NRC has previously approved the use of the equivalent margins analysis (EMA) BAW-2494, Revision 1 (Reference 20) 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 Tables 3-1, 3-3, 5-1 and 5-2 herein and are summarized below. The EMAs for these materials are updated for SLR under PWROG PA-MSC-1481. 1 An EMA should be submitted 3 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 SLR period of operation.
The following Surry Units 1 and 2 materials are addressed by EMAs in PA-MSC-1481 for SLR.
Surry Unit 1:
Upper to Intermediate Shell Circumferential Weld, Heat# 25017 Intermediate Shell Longitudinal Welds L3 and L4,Heat # 8Tl554 Intermediate to Lower Shell Circumferential Weld, Heat# 72445 Lower Shell Longitudinal Weld Ll, Heat# 8Tl554 Lower Shell Longitudinal Weld L2,Heat # 299L44 Inlet Nozzle to Shell Welds, Heat# 299L44 and# 8Tl762; (Projected USE > 50 ft-lbs at 68 EFPY)
Outlet Nozzle to Shell Welds, Heat# 8Tl 762 and# 8Tl554B; (Projected USE > 50 ft-lbs at 68 EFPY)
Surry Unit 2:
Upper to Intermediate Shell Circumferential Weld, Heat# 4275 Intermediate Shell Longitudinal Welds L3 and L4, Heat# 72445 Intermediate Shell Longitudinal Weld L4, Heat# 8T 1762 Intermediate to Lower Shell Circumferential Weld, Heat # 0227 Lower Shell Longitudinal Weld Ll and L2, Heat# 8Tl 762 Inlet Nozzle to Shell Welds, Heat# 8Tl 762; (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)
Note that Dominion has conservatively elected to complete an EMA for the Surry Units 1 and 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 PA-MSC-1481 that were not previously addressed by EMA. The EMA would be applicable to the Surry Units 1 and 2 nozzle to shell welds which exceed the fluence criterion of 1 x 10 17 n/cm2 before 68 EFPY. These materials include those listed below.
A generic EMA for Westinghouse plants (such as Surry Units I and 2) was completed and documented in WCAP-13587, Revision 1 (Reference 30). However, this generic EMA will not be utilized for Surry Units 1 and 2, because EMAs in PA-MSC-1481 address specific Surry Units 1 and 2 materials for SLR.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-3 Suny Unit 1 Outlet Nozzle l to Upper Shell Weld Surry Unit 1 Inlet Nozzle 1 to Upper Shell Weld Surry Unit 1 Inlet Nozzle 3 to Upper Shell Weld Surry Unit 2 Outlet Nozzle l to Upper Shell Weld Suny Unit 2 Inlet Nozzle l to Upper Shell Weld Suny Unit 2 Inlet Nozzle 3 to Upper Shell Weld For Surry Unit 1, the limiting USE value for materials not requiring an EMA at 68 EFPY is 54 ft-lb (see Table 5-1 ); this value corresponds to the Inlet Nozzle to Upper Shell Welds (Heat# 299L44) using Position 2.2. For Suny Unit 2, the limiting USE value for materials not requiring an EMA at 68 EFPY is also 54 ft-lb (see Table 5-2); this value corresponds to the Outlet Nozzle to Upper Shell Welds using Position 1.2. Except for the materials listed above, all of the beltline and extended beltline materials in the Suny Units 1 and 2 reactor vessels are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G) through SLR (68 EFPY).
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 Table 5-1 Predicted USE Values at 68 EFPY (SLR) for Surry Unit 1 SLR 1/4T Initial USE<a>
Wt.%
Fluence<bJ RPV Material cu<*)
(x 1019 n/cm2)
(ft-lb)
Position 1.2 Upper Shell Forging l 22Vl 09V Al 0.11 0.465 114 Upper to Intermediate Shell 0.33 0.465 64 Circumferential Weld(eJ (Heat # 25017)
Intermediate Shell Plate C4326-l 0.11 3.88 115 Intermediate Shell Plate C4326-2 0.11 3.88 94 Intermediate Shell Longitudinal Welds 0.16 0.771 64 L3 and L4(eJ (Heat # 8Tl554)
Intermediate to Lower Shell Circumferential Weld(eJ (Heat# 72445) 0.22 3.89 64 Lower Shell Plate C44 l 5-l 0.102 3.92 103 Lower Shell Plate C4415-2 0.11 3.92 82 Lower Shell Longitudinal Weld L1 (e) 0.16 0.777 64 (Heat# 8Tl554)
Lower Shell Longitudinal Weld Li 0>
0.34 0.777 64 (Heat # 299L44)
Inlet Nozzle 1 to Upper Shell Weld 0.34 0.0188 64 (Heat # 299L44)
Inlet Nozzle 2 to Upper Shell Weld 0.34 0.00484 64 (Heat # 299L44)
Inlet Nozzle 3 to Upper Shell Weld 0.34 0.00672 64 (Heat # 299L44)
Inlet Nozzle 1 to Upper Shell Weld 0.19 0.0188 64 (Heat# 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld 0.19 0.00484 64 (Heat # 8Tl762)
Inlet Nozzle 3 to Upper Shell Weld 0.19 0.00672 64 (Heat # 8Tl 762)
Outlet Nozzle I to Upper Shell Weld 0.19 0.00502 64 (Heat# 8Tl 762)
Outlet Nozzle 2 to Upper Shell Weld 0.19 0.00362 64 (Heat # 8Tl 762)
Outlet Nozzle 3 to Upper Shell Weld 0.19 0.0140 64 (Heat# 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld 0.16 0.00502 64 (Heat # 8Tl554B)
Outlet Nozzle 2 to Upper Shell Weld 0.16 0.00362 64 (Heat # 8Tl554B)
Outlet Nozzle 3 to Upper Shell Weld 0.16 0.0140 64 (Heat# 8Tl554B)
WCAP-18242-NP 5-4 Projected USE SLR USE Decrease<cJ (%)
(ft-lb) 17 95 39 39(e) 28 83 28 68 29 45(e) 50 3i{e) 27 75 28.5 59 29 45(e) 41 33(*)
24 49 24 49 24 49 13 56 13 56 13 56 13 56 13 56 13 56 12 56 12 56 12 56 October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-5 Table 5-1 Predicted USE Values at 68 EFPY (SLR) for Surry Unit 1 SLR l/4T Initial USE<a>
Projected USE SLR Wt.%
Fluence<bJ USE RPV Material cu<a)
Decrease<cJ (%)
(x 10 19 n/cm 2
)
(ft-lb)
(ft-lb)
Inlet Nozzle 1 (Heat # 9-4787) 0.159 0.0304 63 11 56 Inlet Nozzle 2 (Heat # 9-5078) 0.159 0.00784 64 10 58 Inlet Nozzle 3 (Heat # 9-4819) 0.159 0.0109 68 10 61 Outlet Nozzle 1 (Heat # 9-4825-1) 0.159 0.00813 68 10 61 Outlet Nozzle 2 (Heat # 9-4762) 0.159 0.00586 82 10 74 Outlet Nozzle 3 (Heat# 9-4788) 0.159 0.0227 71 10.5 64 Position 2.idl Lower Shell Plate C4415-l 0.102 3.92 103 28 74 Lower Shell Plate C44 l 5-2 0.11 3.92 82 28 59 Lower Shell Longitudinal Weld L2<*l (Heat # 299L44) 0.34 0.777 64 35 4i(*)
Inlet Nozzle 1 to Upper Shell Weld 0.34 0.0188 64 15 54 (Heat # 299L44)
Inlet Nozzle 2 to Upper Shell Weld 0.34 0.00484 64 15 54 (Heat# 299L44)
Inlet Nozzle 3 to Upper Shell Weld 0.34 0.00672 64 15 54 (Heat # 299L44)
Notes:
(a)
Material data is from Tables 3-1 and 3-2 of this report.
(b)
The l/4T fluence was calculated using the fluence data in Table 2-3, the Regulatory Guide 1.99, Revision 2 (Reference 18) correlation, and the Surry Units 1 and 2 reactor vessel 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 for the nozzle materials.
(c)
The Position 1.2 USE decrease values were calculated by plotting the l/4T fluence values on Figure 2 of Regulatory Guide 1.99, Revision 2 and using the material-specific Cu wt. % values.
(d)
Surveillance data (deemed credible per Appendix A) from Table 7-6 ofBAW-2324 (Reference 21) were used in the calculation of Surry Unit 1 Position 2.2 USE projections. Regulatory Guide 1.99, Revision 2, Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of the 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, Revision 1 (Reference 20) and are included herein to establish a baseline for SLR evaluation. EMAs for these materials are addressed under PA-MSC-1481.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 Table 5-2 Predicted USE Values at 68 EFPY (SLR) for Surry Unit 2 SLR l/4T Initial USE<a>
Wt.%
Fluence<hl RPV Material Cu(a)
(x 1019 n/cm2)
(ft-lb)
Position 1.2 Uooer Shell Forging 123V303VAI 0.11 0.534 104 Upper to Intermediate Shell Circumferential 0.35 0.534 68 Weld(e) Heat # 4275 Intermediate Shell Plate C433 l-2 0.12 4.44 84 Intermediate Shell Plate C4339-2 0.11 4.44 83 Intermediate Shell Longitudinal Welds L3 0.22 0.796 64 and U (OD 50%l e) (Heat # 72445)
Intermediate Shell Longitudinal Weld L4 0.19 0.796 64 (ID 50%ie) (Heat # 8T 1762)
Intermediate to Lower Shell Circ. Weld(e) 0.187 4.45 82 (Heat # 0227)
Lower Shell Plate C4208-2 0.15 4.48 94 Lower Shell Plate C4339-l 0.107 4.48 101 Lower Shell Longitudinal Weld L1 and 0.19 0.802 64 L2(e) {Heat# 8T 1762)
Inlet Nozzle I to Upper Shell Weld 0.19 (Heat # 8T 17 62) 0.0210 64 Inlet Nozzle 2 to Upper Shell Weld 0.19 0.00484 64 (Heat # 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld (Heat # 8Tl 762) 0.19 0.00660 64 Outlet Nozzle 1 to Upper Shell Weld (Rotterdam) 0.35 0.00491 71 Outlet Nozzle 2 to Upper Shell Weld (Rotterdam) 0.35 0.00361 71 Outlet Nozzle 3 to Upper Shell Weld (Rotterdam) 0.35 0.0156 71 Inlet Nozzle 1 (Heat # 9-5104) 0.159 0.0340 73 Inlet Nozzle 2 (Heat# 9-4815) 0.159 0.00784 66 Inlet Nozzle 3 (Heat# 9-5205) 0.159 0.0107 67 Outlet Nozzle 1 (Heat # 9-4825-2) 0.159 0.00796 73 Outlet Nozzle 2 (Heat # 9-5086-1) 0.159 0.00585 77 Outlet Nozzle 3 (Heat # 9-5086-2) 0.159 0.0253 71 Position 2_id)
Lower Shell Plate C4339-l 0.107 4.48 101 Intermediate Shell Plate C4339-2 0.11 4.44 83 Intermediate to Lower Shell Circ. Weld(e)
(Heat # 0227) 0.187 4.45 82 Notes on the following page.
WCAP-18242-NP 5-6 Projected USE SLR USE Decrease<c> (%)
(ft-lb) 18 85 39 41 (e) 30 59 29 59 34 4i{e) 32 44(e) 47 43(e) 35 61 29 72 33 43(e) 14 55 13.5 55 13.5 55 24 54 24 54 24 54 12.5 64 10 59 10 60 10 66 10 69 10.5 64 19 82 19 67 42 48(e)
October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-7 Notes:
(a) Material data is from Tables 3-3 and 3-4 of this report.
(b) The l/4T fluence was calculated using the fluence data in Table 2-4, the Regulatory Guide 1.99, Revision 2 (Reference 18) correlation, and the Surry Units I and 2 reactor vessel 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 for the nozzle materials.
(c) The Position 1.2 USE decrease values were calculated by plotting the l/4T fluence values on Figure 2 of Regulatory Guide 1.99, Revision 2 and using the material-specific Cu wt. % values.
(d) Surveillance data (deemed credible and non-credible per Appendix A) from Table 5-12 of WCAP-16001,
Revision O (Reference 22) were used for Surry Unit 2 Position 2.2 USE projections. Regulatory Guide 1.99, Revision 2, Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of the 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, Revision 2, 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 Surry Unit 2 USE projections.
(e) These weld materials were previously addressed by EMA Report BAW-2494, Revision I (Reference 20) and are included herein to establish a baseline for SLR evaluation. EMAs for these materials are addressed under PA-MSC-1481.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 100
% Copper Weld Line Base Metal Weld 0.35 0.30 Limiting Weld Percent USE 0.30 I
0.25 I Upper Limit Decrease is 44% from Capsule X 0.25 I
0.20 0.20 0.15
- i:;.- -i-0.15 0.10
)
~ -
0.10 0.05 i.--i-i-
w (I')
... --~
i.-........
i.---- --
i.----
~
~
~
C:
C.
0...
C 10 a,
Cl C1) -
C:
a,
(.)...
a,
a.
.... :.. - ~ -
t --~
~
I i.---
~
i.-
Plate Line
~
i.--
~
~
i.-
Limiting Plate Percent USE i.--
- ~
Decrease is 24% from Capsule X I-(longitudinal-orientation)
Longitudinal Shell Inlet Nozzle (Surface)
Welds L 1, L2, L3 and I
i---
Fluence = 3.04 x 1017 n/cm2 L4 Bounding Fluence I I
-1 I
= 7. 77 x 101s n/cm2 I- ! 1/4T Lower Shell Plates P..11 Outlet Nozzle (Surface)
Fluence = 3.92 x 1019 n/cm2
.......... Fluence - 2.27 x 1017 n/cm2 1/4T Intermediate Shell 1/4T Inlet Nozzle to Upper Shell Weld I Plates and Intermediate to -
V Fluence = 1.88 x 1017 n/cm2 I
Lower Shell Circ. Weld Bounding Fluence = 3.89
~k""
1/4T Outlet Nozzle to Upper Shell Weld I x10 19 n/cm2 V
Fluence = 1.40 x 1017 n/cm2 I
~v 1/4T Upper Shell Forging and Upper to V
~I' Intermediate Shell Circ. Weld Fluence = 4.65 x 101s n/cm2 1
1 E+17 1E+18 1E+19 1E+20 Surveillance Material: LS Plate C4415-1 Neutron Fluence, n/cm2 (E > 1 MeV)
Surveillance Material: Weld Heat # 299L44 Figure 5-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Surry Unit 1 at SLR (68 EFPY) 5-8 WCAP-18242-NP October20!7 Revision 0
w
(/'J C:
Q.
0...
C Q)
Cl
~ -
C:
Q)
(J...
Q)
Q..
100
% Copper Base Metal 0.35 0.30 0.25 0.20 0.15 0.10
~ --
i--
~
10 --_
~
1 1E+17 I
I I
I I
i,,.i,.,,
~
~
I"'("""
Weld 0.30 0.25 0.20 0.15 0.10 0.05
~ --
-i.---
~
Westinghouse Non-Proprietary Class 3 I
I Plate Line I
I Limiting Plate Percent USE Weld Line Decrease is 10% from Capsule X - -
Limiting Weld Percent USE ---
(transverse-orientation)
Decrease is 22% from Capsule X H Upper Limit
\\
i-
~ -
I'""
~
\\
\\
~
~
i--
i--*... i--
i,,,oi-
\\
~
-~ -~ --
...-... i-- iii" - ~
i,... -
~~
~
.... i,,,oi-
- i--
~ -:t: -----
~
~-
i,... --~
_..-i-
- i---
1/4T Lower Shell Plates Fluence = 4.48 x 1019 n/cm2 -...
I I
-1 I
I Inlet Nozzle (Surface) 11 Longitudinal Shell Welds r-1/4T Intermediate to Lower IFluence = 3.40 x 1017 n/cm2 L 1, L2, L3 and L4 ff Shell Circumferential Weld...
Bounding Fluence = 8.02 Heat # 0227
..J Outlet Nozzle (Surface) r x 1018 n/cm2 Fluence = 4.45 x 1019 n/cm2 I Flue nee = 2.53 x 1011 n/cm2 1/4T Intermediate Shell Plates _ ~
1/4T Inlet Nozzle to Upper Shell Weld 1 Fluence = 4.44 x1 019 n/cm2 1
Fluence = 2.10 x 1017 n/cm2 I
1/4T Upper Shell Forging and Upper
.. r-,.
J 1/4T Outlet Nozzle to Upper Shell Weld I I
Flue nee = 1.56 x 1017 n/cm2 1E+18 Neutron Fluence, n/cm2 (E > 1 MeV) to Intermediate Shell Circ. Weld Fluence = 5.34 x 101s n/cm2 1E+19 a surveillance Material: LS Plate C4339-1
+ Surveillance Weld Data : Heat# 0227 1E+20 Figure 5-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Surry Unit 2 at SLR (68 EFPY) 5-9 WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-1 6
HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Heatup and cooldown limit curves are calculated using the most limiting value of RT NDT (reference nil-ductility transition temperature) corresponding to the limiting material in the beltline region of the RPV.
The most limiting RT NDT of the material in the core (beltline) region of the RPV is determined by using the unirradiated RPV material fracture toughness properties and estimating the irradiation-induced shift
(~RTNDT)-
RT NDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RT NDT at any time period in the reactor's life, ~T NDT due to the radiation exposure associated with that time period must be added to the original unirradiated RT NDT* Using the adjusted reference temperature (ART) values, pressure-temperature (P-T) limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G (Reference 17), as augmented by Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code (Reference 31 ).
According to subsection 4.2.2.1.4 of NUREG-2192 (Reference 4), P-T limit curves for SLR (68 EFPY) do not need to be submitted as part of the Surry Units 1 and 2 License Renewal Application since P-T limit curves are available through the current license period ( 48 EFPY). However, new P-T limit curve development or an extension of the applicability of the current curves must be completed prior to the expiration of the current curves as specified in the Surry Units 1 and 2 licensing basis.
Nozzle materials were evaluated in this report at 48 EFPY and 68 EFPY as part of the extended beltline; the nozzle forging materials evaluated are documented in Tables 6.1-1, 6.1-4, 6.1-7, and 6.1-10. All nozzle materials were assigned the fluence values at the 1/4T flaw location for each specific nozzle in Tables 2-3 and 2-4. Thus, Surry Unit 1 Inlet Nozzle 1 and Surry Unit 2 Inlet Nozzle 1 and Outlet Nozzle 3 have neutron fluence values greater than 1 x 10 17 n/cm2 (E > 1.0 MeV) at 68 EFPY. In order to fully assess the Surry Units 1 and 2 P-T limit curves applicability to 68 EFPY, a nozzle comer fracture mechanics analysis was completed to satisfy NRC RIS 2014-11 (Reference 32). These nozzle P-T limit curves 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 was documented in WCAP-18243-NP (Reference 33). The current beltline curves were confirmed to be bounding.
The P-T limit curves for normal heatup and cooldown of the primary reactor coolant system for Surry Units 1 and 2 were previously developed in WCAP-14177, Revision O (Reference 23). The existing P-T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material. The Surry 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 RPV 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. A summary of the applicability of the P-T Limit curves and the references is provided in Appendix C.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-2 To confirm the applicability of the P-T limit curves developed based on WCAP-14177, Revision 0 (Reference 23) in the Surry Units 1 and 2 Technical Specifications, the limiting reactor vessel material ART values with consideration of the updated TLAA 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 existing P-T limit curves contained in Reference 23 and the Surry Units 1 and 2 Technical Specifications. The Regulatory Guide 1.99, Revision 2 (Reference 18) methodology was used along with the surface fluence of Section 2 to calculate ART values for the Surry Units l and 2 reactor vessel materials at 48 EFPY and 68 EFPY. The ART calculations are summarized in Tables 6.1-1 through 6.1-12 for Surry Units 1 and 2.
Existing P-T Limit Curves Applicability Conclusions 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 6.1-13 for Surry Units 1 and 2.
With the consideration of TLAA fluence projections, the applicability of the Surry Units 1 and 2 P-T limit curves remains valid to 48 EFPY.
Additionally, the applicability of the P-T limit curves in WCAP-14177, Revision O (Reference 23) may be extended to 68 EFPY for the Surry Units 1 and 2 cylindrical shell materials. 1n order to satisfy NRC RlS 2014-11 (Reference 32) guidance, nozzle P-T limit curves were developed per WCAP-18243-NP (Reference 33) and compared to the cylindrical shell beltline curves.
Per WCAP-18243-NP, the applicability of the P-T limit curves may be extended through SLR. For more detailed conclusions, see Section 6.1 for Surry Units 1 and 2.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6.1 ADJUSTED REFERENCE TEMPERATURES AND P-T LIMIT CURVES APPLICABILITY 6-3 Tables 6.1-1 through 6.1-12 summarize the nozzle, l/4T, and 3/4T ART calculations for Surry Units 1 and 2 at 48 and 68 EFPY.
The limiting 48 EFPY and 68 EFPY ART values for Surry Units 1 and 2 correspond to the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (using surveillance data). In addition, as described in Section 3, the applicable reactor vessel 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 (LST) requirements are not applicable to Surry Units 1 and 2, because the plants are Westinghouse-designed and utilize stainless steel reactor coolant system piping.
The inlet and outlet nozzle forging ARTs are necessary to perform a nozzle comer fracture mechanics analysis. The nozzle forging ART calculations utilize the nozzle forging surface l/4T flaw fluence values in order to provide a conservative estimate of the fluence at the limiting nozzle comer location. The nozzle ART values are also considered herein because the nozzle fluence values for some nozzle materials exceed 1 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 l/4T flaw fluence values are utilized for the ART calculations for the nozzle forging materials, the nozzle forgings are omitted from l/4T and 3/4T ART calculations.
The following two conclusions from Section 4 of TLR-RES/DE/CIB-2013-01 (Reference 29) will be utilized, as appropriate, in the ART evaluations documented in Tables 6.1-1 through 6.1-12.
l.
The beltline is defined as the region of the RPV adjacent to the reactor core that is projected to receive a neutronjluence level of 1x1017 n/cm2 (E > 1.0 MeV) or higher at the end of the licensed operating period.
2.
Embrittlement effects may be neglected for any region of the RPV if either of the following conditions are met: (1) neutronfluence is less than lx1017 n/cm2 (E > 1.0 MeV) at EOL or (2) the mean value of jj_T30 estimated using an ETC acceptable to the staff is less than 25°F at EOL. The estimate of l).T30 at EOL shall be made using best-estimate chemistry values.
Therefore, embrittlement of reactor vessel (RV) materials with jj_T30 (which is equivalent to ~T NDT) values less than 25°F will not be considered in the subsequent ART calculations. The use of these conclusions results in the nozzle forging material ART values at 48 EFPY being identical to those at 68 EFPY for Surry Unit 1 (see Tables 6.1-1 and 6.1-7) and Unit 2 (see Tables 6.1-4 and 6.1-10).
Table 6.1-13 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, Revision O (Reference 23). The limiting ART values used to develop the existing P-T limit curves are summarized in Table 6.1-13. As shown in Table 6.1-13, 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.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-4 P-T Limits Applicability Conclusion For Surry Units 1 and 2, it is concluded that the existing P-T limit curves in the Surry Units I and 2 Technical Specifications, developed under WCAP-14177, Revision O (Reference 23) remain valid through 48 EFPY. Additionally, the applicability of the P-T limit curves in WCAP-14177, Revision O may be extended to 68 EFPY for the Surry Units I and 2 cylindrical shell materials. To satisfy NRC RlS 2014-11 (Reference 32) guidance, nozzle P-T limit curves were developed and compared to the cylindrical shell beltline curves per WCAP-18243-NP (Reference 33). The applicability of the P-T limit curves may be extended per WCAP-18243-NP.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-5 Table 6.1-1 Calculation of the Surry Unit 1 Nozzle ART Values at the Surface Location for 48 EFPY R.G.
Surface 1.99, Wt.%
Wt.%
CF<*>
Fluence<h)
Surface RTNDT(U)
(c) aRTNoT (d)
CJ.'°)
CJ/) Margin RPV Material cu<*)
Ni<*>
(x 10 19 n/cm2, FF
Rev. 2 (OF)
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Inlet Nozzle 1 (Heat# 9-4787) 1.1 0.159 0.85 123.5 0.00868 0.099 10.3 0.0 (12.3) 0.0 0.0 0.0 Inlet Nozzle 2 (Heat# 9-5078)
I. I 0.159 0.87 123.7 0.00219 0.035 11.6 0.0 (4.4) 0.0 0.0 0.0 Inlet Nozzle 3 (Heat# 9-4819) 1.1 0.159 0.84 123.4 0.00306 0.046
-47.2 0.0 (5.7) 0.0 0.0 0.0 Outlet Nozzle 1 (Heat# 9-4825-1) 1.1
0. 159 0.85 123.5 0.00237 0.038
-44.9 0.0 (4.7) 0.0 0.0 o.o*
Outlet Nozzle 2 (Heat# 9-4762) 1.1 0.159 0.83 123.3 0.00170 0.029
-87.5 0.0 (3.5) 0.0 0.0 0.0 Outlet Nozzle 3 (Heat# 9-4788) 1.1 0.159 0.84 123.4 0.00672 0.083
-50.2 0.0 (I 0.3) 0.0 0.0
0.0 Notes
(a)
Chemical composition data taken from Tables 3-1 and 3-2 of this report. Chemistry factor values taken from Table 3-10 of this report.
(b)
Surface fluence values were obtained by interpolation of conesponding Section 2 fluence values. FF= fluence factor = t<0 2s-o.io*Jog(f)l.
(c)
Initial RT NOT values and cr1 values are from Table 3-2 of this report.
(d)
Calculated ~RTNoTvalues less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated ~RTNoT values are listed in parentheses for these materials.
(e)
Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 18), the base metal cr11 = l 7°F for Position 1.1. However, cr11 need not exceed 0.5* ~RT NOT*
WCAP-18242-NP October 2017 Revision 0 ART (OF) 10.3 11.6
-47.2
-44.9
-87.5
-50.2
Westinghouse Non-Proprietary Class 3 Table 6.1-2 Calculation of the Surry Unit 1 ART Values at the 1/4T Location for 48 EFPY R.G.
1/4T 1.99, CF<*>
Fluence<hl 1/4T RPV Material Wt. %
Wt. %
Rev. 2 c u<*>
Ni<*>
(OF)
(x 10 19 n/cm2, FF(bl Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 122Vl09VAI I.I 0.11 0.74 76.1 0.329 0.695 Upper to Intennediate Shell I.I 0.33 0.10 152.0 0.329 0.695 Circumferential Weld (Heat # 25017)
Intermediate Shell Plate C4326-I I.I 0.11 0.55 73.5 2.79 1.274 Intermediate Shell Plate C4326-2 1.1 0.11 0.55 73.5 2.79 1.274 Intermediate Shell Longitudinal Welds 1.1 0.16 0.57 167.0 0.537 0.826 L3 and L4 (Heat # 8Tl 554)
Intermediate to Lower Shell 1.1 0.22 0.54 167.0 2.81 1.275
___ Circumferential Weld_(Heat # 72445)_ _ ------------------------- ------- ------ ------------------------------
Using credible surveillance data 2.1 167.0 2.81 1.275 Lower Shell Plate C44 I 5-l 1.1 0.102 0.493 66.6 2.83 1.276 Using credible surveillance data 2.1 83.1 2.83 1.276 Lower Shell Plate C4415-2 1.1 0.11 0.50 73.0 2.83 1.276

--------------------- - -----------~--- ------- ---

Using credible surveillance data 2.1 83.1 2.83 1.276 Lower Shell Longitudinal Weld Ll I. I 0.16 0.57 167.0 0.541 0.828 (Heat # 8TI 554)
Lower Shell Longitudinal Weld L2 1.1 0.34 0.68 220.6 0.54 1 0.828
_________.{_Heat # 299L44) ----------------- ----------------
Using credible surveillance data 2.1 249.8 0.541 0.828 WCAP-1 8242-NP RTNDT(U)
(c) aRTNDT (d) a/cl (OF)
(OF)
(OF) 40 52.9 0.0 0
105.6 20.0 10 93.6 0.0 11.4 93.6 0.0
-48.6 138.0 18.0
-72.5 212.9 12.0 f--
-72. 5 212.9 12.0 20 85.0 0.0 20 106.1 0.0 4.6 93.2 0.0 4.6 106. 1 0.0
-48.6 138.3 18.0
-74.3 182.7 12.8
-74.3 206.9 12.8 6-6 a,.,. (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 28.0 66.6 28.0 60.9
28. 0 60.9 17.0 34.0
8. 5
17. 0 17.0 34.0 8.5 17.0 28.0 66.6 28.0 61.6 28.0 61.6 October 2017 Revision 0 1/4T ART (OF) 126.9 174.4 137.6 139.0 155.9 201.3 201.3 139.0 143.1 131.8 127.7 156.3 170.0 194.2
Westinghouse Non-Proprietary Class 3 Table 6.1-2 Calculation of the Surry Unit 1 ART Values at the l/4T Location for 48 EFPY R.G.
l/4T 1.99, CF<*>
Fluence<b) l/4T RTNDT(U)
(c)
Wt.%
Wt.%
RPV Material cu<*)
Ni<*>
(x 1019 n/cm2, FF<bl Rev. 2 (OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle 1 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.0131 0.132
-7.0
___________ (Heat # 299L44) _________________ ----------------'--*------ ------- ------
Using credible surveillance data 2.1 249.8 0.0131 0.132
-7.0 Inlet Nozzle 2 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.00331 0.049
-7.0
________ (Heat# 299L44) - ------------- ---------------

--------- f-------------

Using credible surveillance data 2.1 249.8 0.00331 0.049
-7. 0 Inlet Nozzle 3 to Upper Shell Weld l.l 0.34 0.68 220.6 0.00462 0.063
-7.0 e-
____ (Heat # 299L44)_ ________________ -------------------- ------- ---

~---------- f--------------

Using credible surveillance data 2.1 249.8 0.00462 0.063
-7.0 Inlet Nozzle I to Upper Shell Weld 1.1 0.19 0.57 (Heat # 8Tl 762) 152.4 0.013 1 0.132
-4.9 Inlet Nozzle 2 to Upper Shell Weld l.l 0.19 0.57 (Heat # 8T 17 62) 152.4 0.00331 0.049
-4.9 Inlet Nozzle 3 to Upper Shell Weld 1.1 (Heat # 8Tl 762) 0.19 0.57 152.4 0.00462 0.063
-4.9 Outlet Nozzle I to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00345 0.051
-4.9 (Heat # 8Tl 762)
Outlet Nozzle 2 to Upper Shell Weld 1.1
0. 19 0.57 152.4 0.00247 0.039
-4.9 (Heat# 8Tl 762)
Outlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.0098 1 0.108
-4.9 (Heat # 8T l 762)
Outlet Nozzle I to Upper Shell Weld 1.1 0.16 0.57 143.9 0.00345 0.051
-4.9 (Heat # 8Tl554B)
WCAP-18242-NP
.:iRTNDT (d) a/cl (OF)
(OF) 29.0 20.6 32.9 20.6 0.0 (10.8) 20.6 0.0 (12.2) 20.6 0.0 (13.9) 20.6 0.0(15.8) 20.6 0.0 (20.1) 19.7 0.0 (7.5) 19.7 0.0 (9.6) 19.7 0.0 (7.7) 19.7 0.0 (5.9) 19.7 0.0 (16.5) 19.7 0.0 (7.3) 19.7 6-7
<J,1 (c)
Margin (OF)
(OF) 14.5 50.4 14.0 49.8 0.0 41.2 0.0 41.2 0.0 41.2 0.0 41.2 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 October 2017 Revision 0 l/4T ART (OF) 72.4 75.7 34.2 34.2 34.2 34.2 34.5 34.5 34.5 34.5 34.5 34.5 34.5
Westinghouse Non-Proprietary Class 3 6-8 Table 6.1-2 Calculation of the Surry Unit 1 ART Values at the 1/4T Location for 48 EFPY R.G.
1/4T CF<*l (c)
(d) 1.99, Wt.%
Wt.%
Fluence<hl 1/4T RTNDT(U)
~TNDT a/°l CJ& (c)
Margin RPV Material cu<*)
Ni<*>
(x 10 19 n/cm2, FF(bJ
{°F)
(OF)
(OF)
Rev. 2 (OF)
(OF)
(OF)
Position E > 1.0 MeV)
Outlet Nozzle 2 to Upper Shell Weld (Heat# 8Tl 554B) 1.1 0.16 0.57 143.9 0.00247 0.039
-4.9 0.0 (5.6) 19.7 0.0 39.4 Outlet Nozzle 3 to Upper Shell Weld (Heat# 8Tl 554B) 1.1 0.16 0.57 143.9 0.00981 0.108
-4.9 0.0 (15.6) 19.7 0.0 39.4 Notes:
(a)
Chemical composition data taken from Tables 3-1 and 3-2 of this report. Chemistry factor values taken from Table 3-10 of this report.
(b) 48 EFPY surface fluence values were obtained by interpolation of corresponding Section 2 fluence values. The 1/4T fluence and l/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 18) correlations and the Surry Unit 1 reactor vessel wall thickness of 8.05 inches.
( c)
Initial RT NDT values and cr1 values are from Tables 3-1 and 3-2 of this report.
(d)
Calculated t.RT NDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-0 I (Reference 29); actual calculated t.RT NDT values are listed in parentheses for these materials.
(e)
As summarized in Appendix A of this report, all surveillance data for Surry Unit 1 were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal a&= l 7°F for Position I.I, and O'& = 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 I.I, and with credible surveillance data a6 = 14°F for Position 2.1. However, a6 need not exceed 0.5* t.RT NDT* For welds utilizing initial RT NDT values based on BA W-2308, a6 = 28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 1/4T ART (OF) 34.5 34.5
Westinghouse Non-Proprietary Class 3 Table 6.1-3 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 48 EFPY R.G.
3/4T 1.99, Wt.%
Wt.%
CF<a>
Fluence(b>
3/4T RPV Material Rev. 2 cu<*>
Ni<a>
(OF)
(x 10 19 n/cm2, FF(bl Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 122V I 09V A I I. I 0.11 0.74 76.1 0.125 0.464 Upper to Intermediate Shell 1.1 0.33 Circumferential Weld (Heat# 25017) 0.10 152.0 0.125 0.464 Intermediate Shell Plate C4326-l 1.1 0.11 0.55 73.5 1.06 1.017 Intennediate Shell Plate C4326-2 1.1 0.11 0.55 73.5 1.06 1.017 Intermediate Shell Longitudinal Welds 1.1 0.16 L3 and L4 (Heat# 8Tl554) 0.57 167.0 0.204 0.574 Intermediate to Lower Shell 1.1 0.22 0.54 167.0 1.07 1.018
___ Circumferential Weld_{Heat # 72445) ____ ---------- --------- --------- ------------ ---------------------- -------------
Using credible surveillance data 2.1 167.0 1.07 1.018 Lower Shell Plate C44 l 5-l I. I 0.102 0.493 66.6 1.08 1.020 L..__ ___________
Using credible surveillance data 2.1 83.1 1.08 1.020 Lower Shell Plate C44 l 5-2 1.1 0.11 0.50 73.0 1.08 1.020 Using credible surveillance data 2.1 83.1 1.08 1.020 Lower Shell Longitudinal Weld LI 1.1 0.16 0.57 167.0 0.206 0.576 (Heat# 8T l 554)
Lower Shell Longitudinal Weld L2 1.1 0.34 0.68 220.6 0.206 0.576
__________________ (Heat# 299L44)
Using credible surveillance data 2.1 249.8 0.206 0.576 WCAP-18242-NP RTNDT(U)
(c)
~RTNDT (d) a/cl (OF)
(OF)
(OF) 40 35.3 0.0 0
70.5 20.0 10 74.8 0.0 11.4 74.8 0.0
-48.6 95.9 18.0
-72.5 170.1 12.0
-72.5 170.1 12.0 20 68.0 0.0 20 84.8 0.0 4.6 74.5 0.0 4.6 84.8 0.0
-48.6 96.3 18.0
-74.3 127.2 12.8
-74.3 144.0 12.8 6-9
<J11 (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 28.0 66.6 28.0 60.9 28.0 60.9 17.0 34.0 8.5 17.0 17.0 34.0 8.5 17.0 28.0 66.6 28.0 61.6 28.0 61.6 October 2017 Revision 0 3/4T ART (OF) 109.3 139.3 118.8 120.2 113.9 158.5 158.5 122.0 121.8 113.1 106.4 114.2 114.4 131.3
Westinghouse Non-Proprietary Class 3 Table 6.1-3 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 48 EFPY R.G.
3/4T 1.99, Wt.%
Wt. %
CF<*l Fluence<hl 3/4T RPV Material cu<*)
Ni<*>
(x 10 19 n/cm2, FFCbJ Rev. 2 (OF)
Position E > 1.0 MeV)
Reactor Vessel Extended B eltline Materials Inlet Nozzle I to Upper Shell Weld I.I 0.34 0.68 220.6 0.00500 0.067
____________ (Heat# 299L44)

1----------------- --------- ------------ ------------------- -------------

Using credible surveillance data 2.1 249.8 0.00500 0.067 Inlet Nozzle 2 to Upper Shell Weld I. I 0.34 0.68 220.6 0.00126 0.022
________________ {_Heat# 299Lj_i)__ _____________ --------------- -------- ------------ ------------------- -
Using credible surveillance data 2.1 249.8 0.00126 0.022 Inlet Nozzle 3 to Upper Shell Weld I. I 0.34 0.68 220.6 0.00176 0.029
______________ {_Heat# 299L44)
Using credible surveillance data 2.1 249.8 0.00176 0.029 Inlet Nozzle I to Upper Shell Weld I. I 0.19 0.57 152.4 0.00500 0.067 (Heat# 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00126 0.022 (Heat# 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00176 0.029 (Heat# 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00131 0.023 (Heat # 8T 1762)
Outlet Nozzle 2 to Upper Shell Weld I.I 0.19 0.57 152.4 0.000939 0.017 (Heat# 8Tl 762)
Outlet Nozzle 3 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00373 0.054 (Heat # 8Tl 762)
Outlet Nozzle I to Upper Shell Weld I. I 0.16 0.57 143.9 0.00131 0.023 (Heat# 8T1554B)
WCAP-18242-NP RTNDT(U) (c)
~TNDT (d) a/cl (OF)
(OF)
(OF)
-7.0 0.0 (14.8) 20.6
-7.0 0.0(16.7) 20.6
-7.0 0.0 (4.9) 20.6
-7.0 0.0 (5.6) 20.6
-7.0 0.0 (6.5) 20.6
-7.0 0.0 (7.4) 20.6
-4.9 0.0 (10.2) 19.7
-4.9 0.0 (3.4) 19.7
-4.9 0.0 (4.5) 19.7
-4.9 0.0 (3.5) 19.7
-4.9 0.0 (2.6) 19.7
-4.9 0.0 (8.2) 19.7
-4.9 0.0 (3.3) 19.7 6-10
<J&(c)
Margin (OF)
(OF) 0.0 41.2 0.0 41.2 0.0 41.2 0.0 41.2 0.0 41.2 0.0 41.2 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 October 2017 Revision 0 3/4T ART (OF) 34.2 34.2 34.2 34.2 34.2 34.2 34.5 34.5 34.5 34.5 34.5 34.5 34.5
Westinghouse Non-Proprietary Class 3 6-11 Table 6.1-3 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 48 EFPY R.G.
3/4T CF<*>
Fluence(b) 3/4T (c)
(d) 1.99, Wt.%
Wt.%
RTNDT(U)
~RTNDT (Jt>
(JA (c)
Margin RPV Material cu<*)
N1<*>
(x 1019 n/cm2, FF
Rev. 2 (OF)
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Outlet Nozzle 2 to Upper Shell Weld I. I 0.16 (Heat # 8TJ554B) 0.57 143.9 0.000939 0.017
-4.9 0.0 (2.5) 19.7 0.0 39.4 Outlet Nozzle 3 to Upper Shell Weld I.]
0.16 0.57 143.9 0.00373 0.054
-4.9 0.0(7.7) 19.7 0.0 39.4 (Heat# 8Tl554B)
Notes:
(a)
Chemical composition data taken from Tables 3-1 and 3-2 of this report. Chemistry factor values taken from Table 3-10 of this report.
(b) 48 EFPY surface fluence values were obtained by interpolation of corresponding Section 2 fluence values. The 3/4T fluence and 3/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 18) correlations and the Surry Unit 1 reactor vessel wall thickness of 8.05 inches.
( c)
Initial RT or values and cr1 values are from Tables 3-1 and 3-2 of this report.
(d)
Calculated L'.RTNor values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-0l(Reference 29); actual calculated L'.RTNor values are listed in parentheses for these materials.
(e)
As summarized in Appendix A of this report, all surveillance data for Surry Unit l were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal cr6 = l 7°F for Position l.l, and cr6 = 8.5°F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, Revision 2, the weld metal cr6 = 28°F for Position I. I, and with credible surveillance data cr6 = 14°F for Position 2.1. However, cr6 need not exceed 0.5*1'.'.RT NDT* For welds utilizing initial RT NDT values based on BA W-2308, cr6 = 28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 3/4T ART (OF) 34.5 34.5
Westinghouse Non-Proprietary Class 3 6-12 Table 6.1-4 Calculation of the Surry Unit 2 Nozzle ART Values at the Surface Location for 48 EFPY R.G.
Surface 1.99, Wt.%
Wt. %
CF<*>
Fluence(b>
Surface RT
<c>
ART NOT (d)
<11 (c)
<111 (c)
Margin RPV Material NDT(U)
Rev.2 c u<*>
Ni<a>
(OF)
(x 1019 n/cm2, FF<h>
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Inlet Nozzle 1 (Heat# 9-5104)
I.I 0.159 0.84 123.4 0.00935 0.105
-29.7 0.0 (12.9) 0.0 0.0 0.0 Inlet Nozzle 2 (Heat# 9-4815)
I.I 0.159 0.87 123.7 0.00223 0.035 4.5 0.0 (4.4) 0.0 0.0 0.0 Inlet Nozzle 3 (Heat# 9-5205)
I.I 0.159 0.86 123.6 0.00304 0.046 6.5 0.0 (5.7) 0.0 0.0 0.0 Outlet Nozzle 1 (Heat # 9-4825-2)
I.I 0.159 0.85 123.5 0.00235 0.037
-58.1 0.0 (4.6) 0.0 0.0 0.0 Outlet Nozzle 2 (Heat # 9-5086-1) 1.1 0.159 0.86 123.6 0.00172 0.029
-26.6 0.0 (3.6) 0.0 0.0 0.0 Outlet Nozzle 3 (Heat# 9-5086-2) 1.1 0.159 0.87 123.7 0.00721 0.087
-33.8 0.0 (10.8) 0.0 0.0
0.0 Notes
(a)
Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistry Factor values taken from Table 3-12 of this report.
(b)
Surface fluence values were obtained by interpolation of corresponding Section 2 fluence values. FF= fluence factor= t<0*2s-o.w iog(f)J.
(c)
Initial RT NDT values and cr1 values are from Table 3-4 of this report.
(d)
Calculated ~RT NDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated ~RT NDT values are listed in parentheses for these materials.
(e)
Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 18), the base metal cr6 = l 7°F for Position I. However, cr6 need not exceed 0.5*~RTNDT*
WCAP-18242-NP October 2017 Revision 0 ART (OF)
-29.7 4.5 6.5
-58.1
-26.6
-33.8
Westinghouse Non-Proprietary Class 3 Table 6.1-5 Calculation of the Surry Unit 2 ART Values at the 1/4T Location for 48 EFPY R.G.
1/4T 1.99, CF<*>
Fluence<h) l/4T Wt. %
Wt. %
RPV Material Rev.2 cu<*>
Ni<*>
(OF)
(x 10 19 n/cm2, FF(b>
Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging I 23V303VAI I. I 0.1 1 0.72 75.8 0.361 0.719 Upper to Intermediate Shell I.I 0.35
0. 10 160.5 0.361 0.71 9 Circumferential Weld (Heat # 4275)
Intermediate Shell Plate C433 I -2 I.I 0.12 0.60 83.0 3.07 1.296 Intermediate Shell Plate C4339-2 I.I 0.11 0.54 73.4 3.07 1.296 Using non-credible surveillance data 2.1 75.7 3.07 1.296 Intermediate Shell Longitudinal Welds I.I 0.22 L3 and L4 (OD 50%) (Heat # 72445) 0.54 167.0 0.563 0.839 Using credible surveillance data 2.1 167. 0 0.563
0. 839 Intennediate Shell Longitudinal Weld L4 I.I 0.19 0.57 167.0 0.563 0.839 (ID 50%) (Heat # 8T I 762)
Intermediate to Lower Shell I.I 0.187 0.545 147.5 Circwnferential Weld (Heat # 0227) 3.07 1.296 Using credible surveillance data 2.1 132.5 3.07 1.296 Lower Shell Plate C4208-2 I.I 0.15 0.55 107.3 3.08 1.297 Lower Shell Plate C4339-l I.I 0.107 0.53 70.8 3.08 1.297 Using non-credible surveillance data 2.1 75.7 3.08 1.297 Lower Shell Longitudinal Welds LI and I.I 0.1 9 0.57 167.0 0.568 0.842 L2 (Heat # 8T I 762)
WCAP-18242-NP RTNDT(U)
(c) tiRTNDT (d)
(OF)
(OF) 30 54.5 0
115.3 15.0 107.6 7.8 95.1 7.8
98. 1
-72.5 140.2
-72.5 140.2
-48.6 140.2 0
191.2 0
171.8
-30 139.2
-4.4 91.8
-4.4 98.2
-48.6 140.5 CJ/cl (OF) 0.0 20.0 0.0 0.0 0.0 12.0 12.0 18.0 0.0 0.0 0.0 0.0 0.0 18.0 6-1 3
<Ja(c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 17.0 34.0 28.0 60.9
28. 0 60.9 28.0 66.6 28.0 56.0 14.0
28. 0 17.0 34.0 17.0 34.0 17.0 34.0 28.0 66.6 October 2017 Revision 0 l/4T ART (OF) 118.5 184.2 156.6 136.9 139.9 128.6 128. 6 158.2 247.2 199.8 143.2 121.4 127.8 158.5
Westinghouse Non-Proprietary Class 3 6-14 Table 6.1-5 Calculation of the Surry Unit 2 ART Values at the 1/4T Location for 48 EFPY R.G.
1/4T 1.99, Wt.%
Wt.%
CF<a>
Fluence<bl 1/4T RTNDT(U)
(c)
.tlRTNoT (d)
<J1 (c)
<Jt,. (c)
Margin RPV Material Rev.2 cu<a)
Ni<al (OF)
(x 10 19 n/cm2, FF(bl (OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle 1 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.0141 0.138
-4.9 0.0 (21.0) 19.7 0.0 39.4 (Heat# 8Tl 762)
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 (Heat# 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00460 0.063
-4.9 0.0 (9.6) 19.7 0.0 39.4 (Heat# 8Tl762)
Outlet Nozzle 1 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.00342 0.050 30 0.0 (13.7) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 2 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.00250 0.039 30 0.0 (10.7) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.0105 0.113 30 30.8 0.0 15.4 30.8 (Rotterdam)
Notes:
(a)
Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistry Factor values taken from Table 3-12 of this report.
(b) 48 EFPY surface fluence values were obtained by interpolation of corresponding Section 2 fluence values. The l/4T fluence and l/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 18) correlations and the Surry Unit 2 reactor vessel wall thickness of 8.05 inches.
( c)
Initial RT NDT values and cr1 values are from Tables 3-3 and 3-4 of this report.
( d)
Calculated t.RT NDT values Jess than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated t.RT NDT values are listed in parentheses for these materials.
(e)
Per Appendix A of this report, 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 cr8 = l 7°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. Also per Regulatory Guide 1.99, Revision 2, the weld metal cr8 = 28°F for Position 1.1, and with credible surveillance data cr8 = 14°F for Position 2.1. However, cr8 need not exceed 0.5*t.RTNDT* For welds utilizing initial RTNDT values based on BAW-2308, cr8 =
28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 1/4T ART (OF) 34.5 34.5 34.5 30.0 30.0 91.6
Westinghouse Non-Proprieta1y Class 3 Table 6.1-6 Calculation of the Surry Unit 2 ART Values at the 3/4T Location for 48 EFPY R.G.
3/4T 1.99, Wt. %
Wt. %
CF<*>
Fluence<hl 3/4T RPV Material cu<*)
Ni<*>
(x 10 19 n/cm2, FFCbl Rev. 2 (OF)
Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 123V303VAI 1.1 0.11 0.72 75.8 0.137 0.483 Upper to Intermediate Shell I. I 0.35 0.10 160.5 0.137 0.483 Circumferential Weld (Heat # 4275)
Intermediate Shell Plate C433 l -2 1.1 0.12 0.60 83.0 1.17 1.043 Intermediate Shell Plate C4339-2 I. I 0.11 0.54 73.4 1.17 1.043 Using non-credible surveillance data 2.1
75. 7 1.17 1.043 Intermediate Shell Longitudinal Welds I. I 0.22 0.54 167.0 0.214 0.586 L3 and L4 (OD 50%) (Heat # 72445)

1------------

Using credible surveillance data 2.1 167.0 0.214 0.586 Intermediate Shell Longitudinal Weld L4 I.I 0.19 0.57 167.0 0.214 0.586 (ID 50%) (Heat # 8Tl 762)
Intermediate to Lower Shell 1.1
0. 187 0.545 147.5 Circumferential Weld (Heat # 0227) 1.1 7 1.044

----------------~

Using credible surveillance data
2. 1 132.5 1.17 1.044 Lower Shell Plate C4208-2 I. I 0.15 0.55 107.3 1.17 1.045 Lower Shell Plate C4339-l 1.1 0.1 07 0.53 70.8 1.17 1.045

--------------r-------- ------- ------

----- ~ ----------

Using non-credible surveillance data 2.1 75.7 1.17 1.045 Lower Shell Longitudinal Welds Ll and I.I 0.19 0.57 167.0 0.216 0.588 L2 (Heat # 8Tl 762)
WCAP-1 8242-NP RTNDT(U)
(c) aRTNDT (d) a,<c)
(OF)
(OF)
(OF) 30 36.6 0.0 0
77.6 20.0 15.0 86.6 0.0 7.8 76.6 0.0 7.8 79.0 0.0
-72.5 97.9 12.0
-72. 5
97. 9 12.0
-48.6 97.9 18.0 0
153.9 0.0 0
138.3 0.0
-30 112.1 0.0
-4.4 74.0 0.0 1---------------- -------------- ----------
-4.4 79.1 0.0
-48.6 98.2 18.0 6-15 a,... (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0

~

17.0 34.0 28.0 60.9
28. 0 60.9 28.0 66.6 28.0 56.0 14.0
28. 0 17.0 34.0 17.0 34.0 17.0 34.0 28.0 66.6 October 201 7 Revision 0 3/4T ART (OF) 100.6 146.4 135.6 118.4 120.8 86.3 86.3 115.9 209.9 166.3 116.1 103.6 108.7 116.2
Westinghouse Non-Proprietary Class 3 6-16 Table 6.1-6 Calculation of the Surry Unit 2 ART Values at the 3/4T Location for 48 EFPY R.G.
3/4T 1.99, Wt. %
Wt. %
CF<*>
Fluence<bl 3/4T RTNDT(U)
(c)
.iRTNDT (d) a?>
0 11 (c)
Margin RPV Material Rev. 2 c u<*)
Ni<*>
(OF)
(x 10 19 n/cm2, FF<bl (OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle I to Upper Shell Weld I. I 0.19 0.57 152.4 0.00538 0.071
-4.9 0.0 (10.8) 19.7 0.0 39.4 (Heat # 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00128 0.023
-4.9 0.0 (3.4) 19.7 0.0 39.4 (Heat# 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld 1.1
0. 19 0.57 152.4 0.00175 0.029
-4.9 0.0 (4.5) 19.7 0.0 39.4 (Heat# 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld I. I 0.35 1.0 272.0 0.00130 0.023 30 0.0 (6.2) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 2 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.000953 0.017 30 0.0 (4.7) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld I. I 0.35 1.0 272.0 0.00399 0.057 30 0.0 (15.4) 0.0 0.0 0.0 (Rotterdam)
Notes:
(a)
Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistry Factor values taken from Table 3-12 of this report.
(b) 48 EFPY surface tluence values were obtained by interpolation of corresponding Section 2 tluence values. The 3/4T tluence and 3/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 18) correlations and the Surry Unit 2 reactor vessel wall thickness of 8.05 inches.
(c)
Initial RT NOT values and CJ1 values are from Tables 3-3 and 3-4 of this report.
(d)
Calculated t.RT NDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated t.RT NOT values are listed in parentheses for these materials.
(e)
Per Appendix A of this report, 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 (Reference 18), the base metal Ga = l 7°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. Also per Regulatory Guide 1.99, Revision 2, the weld metal CJ11 = 28°F for Position 1.1, and with credible surveillance data CJ11 = 14°F for Position 2.1. However, CJ11 need not exceed 0.5*t.RT NOT* For welds utilizing initial RT NOT values based on BA W-2308, CJ11 = 28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 3/4T ART (OF) 34.5 34.5 34.5 30.0 30.0 30.0
Westinghouse Non-Proprietary Class 3 6-17 Table 6.1-7 Calculation of the Surry Unit 1 Nozzle ART Values at the Surface Location for 68 EFPY R.G.
Surface 1.99, Wt. %
Wt. %
CF<*>
Fluence<h)
Surface RTNDT(U)
(c)
~TNDT (d) u /cl Ue,. (c)
Margin RPV Material cu<*)
Ni<*>
(x 1019 n/cm2, FF<h>
Rev. 2 (OF)
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Inlet Nozzle 1 (Heat# 9-4 787) 1.1 0.159 0.85 123.5 0.0124 0.127 10.3 0.0 (15.6) 0.0 0.0 0.0 Inlet Nozzle 2 (Heat# 9-5078) 1.1 0.159 0.87 123.7 0.00322 0.048 11.6 0.0 (5.9) 0.0 0.0 0.0 Inlet Nozzle 3 (Heat# 9-4819)
I. I 0.159 0.84 123.4 0.00446 0.062
-47.2 0.0 (7.6) 0.0 0.0 0.0 Outlet Nozzle I (Heat# 9-4825-1)
I. I 0.159 0.85 123.5 0.00345 0.051
-44.9 0.0 (6.3) 0.0 0.0 0.0 Outlet Nozzle 2 (Heat# 9-4762) 1.1 0.159 0.83 123.3 0.00249 0.039
-87.5 0.0 (4.8) 0.0 0.0 0.0 Outlet Nozzle 3 (Heat# 9-4788)
I. I 0.159 0.84 123.4 0.00962 0.1 07
-50.2 0.0 (13.2) 0.0 0.0
0.0 Notes
(a)
Chemical composition data taken from Tables 3-1 and 3-2 of this report. Chemistry factor values taken from Table 3-10 of this report.
(b)
Surface fluence values taken from Section 2 of this report. FF = fluence factor= t<0*28-0,o* iog(t)l.
( c)
Initial RT NOT values and cr1 values are from Table 3-2 of this report.
(d)
Calculated t.RTNor values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated t.RT or values are listed in parentheses for these materials.
( e)
Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 18), the base metal cr,.. = l 7°F for Position I. I. However, cr,.. need not exceed 0.5*t.RT NOT*
(t)
Nozzle materials are not limiting for P-T limit curves per WCAP-18243-NP (Reference 33).
WCAP-18242-NP October 2017 Revision 0 ART<ll (OF) 10.3 11.6
-47.2
-44.9
-87.5
-50.2
Westinghouse Non-Proprietary Class 3 Table 6.1-8 Calculation of the Surry Unit 1 ART Values at the l/4T Location for 68 EFPY R.G.
l/4T 1.99, Wt.%
Wt.%
CF<*>
Fluence<bl l/4T RPV Material cu<*)
Ni<*>
(x 10 19 n/cm2, FF(bl Rev. 2 (OF)
Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 122Vl09VAI I. I 0.11 0.74 76.1 0.465 0.787 Upper to Intennediate Shell I. I 0.33 0.10 152.0 0.465 0.787 Circumferential Weld (Heat # 25017)
Intermediate Shell Plate C4326-l I. I 0.11 0.55 73.5 3.88 1.350 Intermediate Shell Plate C4326-2 1.1 0.11 0.55 73.5 3.88 1.350 Intermediate Shell Longitudinal Welds 1.1 0.16 0.57 167.0 0.771 0.927 L3 and L4 (Heat # 8T 1554)
Intermediate to Lower Shell 1.1 0.22 0.54 167.0 3.89 1.350
~ Circumferential Weld_@eat # 72445) __ -------------
L...__ ______
Using credible surveillance data 2.1 167.0 3.89 1.350 Lower Shell Plate C441 5-l 1.1 0.102 0.493 66.6 3.92 1.352

---------------~---- ------ -----

l-----------

Using credible surveillance data 2.1 83.1 3.92 1.352 Lower Shell Plate C44 15-2 1.1 0.11 0.50 73.0 3.92 1.352 Using credible surveillance data 2.1
83. 1 3.92 1.352 Lower Shell Longitudinal Weld L1 1.1 0.16 0.57 167.0 0.777 0.929 (Heat # 8Tl 554)
Lower Shell Longitudinal Weld L2 1.1 0.34 0.68 220.6 0.777 0.929
___________ (Heat # 299L44) ----------------- ---------------------- ------- ------- ------------------ - --------
Using credible surveillance data 2.1 249.8 0.777 0.929 WCAP-18242-NP RTNDT(U) (c)
ART NOT (d) u/c>
(OF)
(OF)
(OF) 40 59.9 0.0 0
119.6 20.0 10 99.2 0.0 11.4 99.2 0.0
-48.6 154.8 18.0
-72.5 225.5 12.0
-72.5 225.5 12.0 20 90.0 0.0 20 112.3 0.0 4.6 98.7 0.0 4.6 112.3 0.0
-48.6 155.2 18.0
-74.3 205.0 12.8
-74.3 232.1 12.8 6-18 Ut,. (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 28.0 66.6 28.0 60.9 28.0 60.9 17.0 34.0 8.5 17.0 17.0 34.0 8.5 17.0 28.0 66.6 28.0 61.6 28.0 61.6 October 2017 Revision 0 l/4T ART (OF) 133.9 188.4 143.2 144.6 172.8 213.9 213.9 144.0 149.3 137.3 133.9 173.2 192.3 219.4
Westinghouse Non-Proprietary Class 3 Table 6.1-8 Calculation of the Surry Unit 1 ART Values at the 1/4T Location for 68 EFPY R.G.
114T 1.99, Wt. %
CF<*l Fluence<hl 1/4T RTNDT(U)
(c)
Wt. %
RPV Material cu<*)
Ni ">
(x 10 19 n/cm2, FF<h>
Rev. 2 (OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle I to Upper Shell Weld I. I 0.34 0.68 220.6 0.0188 0.165
-7.0
___________ (Heat # 299L4~--------- _ ------------------------ -------
L...-_ _ ______ ----------------
Using credible surveillance data 2.1 249.8 0.0188 0.165
-7.0 Inlet Nozzle 2 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.00484 0.065
-7.0
_ {_Heat # 299L4'!)

----- ~-------

Using credible surveillance data 2.1 249.8 0.00484 0.065
-7. 0 Inlet Nozzle 3 to Upper Shell Weld I. I 0.34 0.68 220.6 0.00672 0.083
-7.0 e- _____ (Heat # 299L44) ____
_ ~------------ -------

~---------

Using credible surveillance data 2.1 249.8 0.00672 0.083
-7.0 Inlet Nozzle I to Upper Shell Weld I. I 0.19 0.57 152.4 0.0188 0.165
-4.9 (Heat # 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00484 0.065
-4.9 (Heat # 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld I.I 0.19 0.57 152.4 0.00672 0.083
-4.9 (Heat # 8T 1762)
Outlet Nozzle I to Upper Shell Weld I.I 0.19 0.57 152.4 0.00502 0.067
-4.9 (Heat # 8Tl 762)
Outlet Nozzle 2 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00362 0.052
-4.9 (Heat # 8Tl 762)
Outlet Nozzle 3 to Upper Shell Weld I.I 0.19 0.57 (Heat # 8Tl 762) 152.4 0.0140 0.137
-4.9 Outlet Nozzle I to Upper Shell Weld I.I 0.16 0.57 143.9 0.00502 0.067
-4.9 (Heat # 8Tl554B)
WCAP-1 8242-NP aRTNDT (d)
<J/cl (OF)
(OF) 36.5 20.6 41.3 20.6 0.0 (1 4.4) 20.6

~-----

0.0 (16.3) 20.6 0.0 (1 8.3) 20.6
~*----- ~*---
0.0 (20.8) 20.6 25.2 19.7 0.0 (I 0.0) 19.7 0.0 (12.7) 19.7 0.0 (10.2) 19.7 0.0 (8.0) 19.7 0.0 (20.9) 19.7 0.0 (9.7) 19.7 6-1 9 (J6. (c)
Margin (OF)
(OF) 18.2 55.0 14.0 49.8 0.0 41.2 0.0 41.2 0.0 41.2 0.0 41.2 12.6 46.8 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 October 2017 Revision 0 114T ART (OF) 84.5 84.1 34.2 34.2 34.2 34.2 67.1 34.5 34.5 34.5 34.5 34.5 34.5
Westinghouse Non-Proprietary Class 3 6-20 Table 6.1-8 Calculation of the Surry Unit 1 ART Values at the 1/4T Location for 68 EFPY R.G.
1/4T CF<*l (c)
(d) 1.99, Wt.%
Wt.%
Fluence<bl l/4T RTNDT(U)
-6.RTNDT u/cl Ue,.(c)
Margin RPV Material cu<*)
N{*l (x 10 19 n/cm2, FF(bl Rev. 2 (OF)
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Outlet Nozzle 2 to Upper Shell Weld I.I 0.16 0.57 143.9 0.00362 0.052
-4.9 0.0 (7.6) 19.7 0.0 39.4 (Heat # 8T1554B)
Outlet Nozzle 3 to Upper Shell Weld 1.1 0.16 0.57 143.9 0.0140 0.137
-4.9 0.0 (19.7) 19.7 0.0 39.4 (Heat# 8Tl554B)
Notes:
(a)
Chemical composition data taken from Tables 3-1 and 3-2 of this report. Chemistry factor values taken from Table 3-10 of this report.
(b)
The l/4T fluence and l/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations (Reference 18) and the Surry Unit I reactor vessel wall thickness of 8.05 inches.
(c)
Initial RT NDT values and cr1 values are from Tables 3-1 and 3-2 of this report.
( d)
Calculated ~RT NDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated ~RT NDT values are listed in parentheses for these materials.
(f)
As summarized in Appendix A of this report, all surveillance data for Surry Unit I were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 18), the base metal cr6 = 17°F for Position 1.1, and cr6 = 8.5°F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, Revision 2, the weld metal cr6 = 28°F for Position 1.1, and with credible surveillance data cr6 = 14°F for Position 2.1.
However, cr6 need not exceed 0.5*~RT NDT* For welds utilizing initial RT NDT values based on BA W-2308, cr6 = 28°F per References 15 and 16.
WCAP-1 8242-NP October 2017 Revision 0 l/4T ART (OF) 34.5 34.5
Westinghouse Non-Proprietary Class 3 Table 6.1-9 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 68 EFPY R.G.
3/4T 1.99, Wt. %
Wt. %
CF<*>
Fluence
3/4T RPV Material Rev. 2 cu<*>
Nf *>
(OF)
(x 1019 n/cm2, FFCbl Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 122V!09VA!
1.1 0.11 0.74 76.1
0. 177 0.541 Upper to Intermediate Shell 1.1 0.33 Circumferential Weld (Heat# 25017)
0. 10 152.0
0. 177 0.54 1 Intermediate Shell Plate C4326-l 1.1 0.1 1 0.55 73.5 1.48 1.108 Intermediate Shell Plate C4326-2 1.1
0. 11 0.55 73.5 1.48 1.108 Intermediate Shell Longitudinal Welds I. I
0. 16 0.57 167.0 0.294 0.665 L3 and L4 (Heat # 8Tl 554)
Intermediate to Lower Shell 1.1 0.22 0.54 167.0 1.48 1.109
__ Circumferential Weld (Heat# 72445) _ ------------._ ________ ----- ~---

1-------------

Using credible surveillance data 2.1 167.0 1.48 1.109 Lower Shell Plate C4415-l 1.1 0.102 0.493 66.6 1.49 1.1 11

---------- ---~

i....__ _ ___ ____
Using credible surveillance data 2.1 83.1 1.49 1.111 Lower Shell Plate C4415-2 1.1 0.11 0.50 73.0 l.49 1.11 1

----------------~----- ---- - --- ------- ------------------- i...------------

Using credible surveillance data 2.1
83. 1 1.49 1.111 Lower Shell Longitudinal Weld Ll 1.1 0.16 0.57 167.0 0.296 0.667 (Heat # 8T!554)
Lower Shell Longitudinal Weld L2 1.1 0.34 0.68 220.6 0.296 0.667
__________ (Heat# 299L44)_ _____________ ------------1--------- -

-------L----------

Using credible surveillance data 2.1 249.8 0.296 0.667 WCAP-18242-NP RTNDT(U)
(c) aRTNDT (d)
(OF)
(OF) 40 41.1 0
82.2 10 81.4 11.4 81.4
-48.6 111.0
-72.5 185.2 f--------------- -------- -----
-72.5 185.2 20 74.0 1--------------- ------------
20 92.3 4.6 81.1 1--------------- --------------*-
4.6 92.3
-48.6 11 l.3
-74.3 147.1 f------------- --------------
-74.3 166.5 er.<°>
(OF) 0.0 20.0 0.0 0.0 18.0 12.0 12.0 0.0 0.0 0.0 0.0 18.0 12.8 12.8 6-2 1 ff t,. (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 28.0 66.6 28.0 60.9 28.0 60.9 17.0 34.0 8.5 17.0 17.0 34.0 8.5 17.0 28.0 66.6 28.0 61.6 1------------ ------
28.0 61.6 October 2017 Revision 0 3/4T ART (OF) 115.1 151.0 125.4 126.8 129.0 173.6 173.6 128.0 129.3 119.7 113.9 129.3 134.3 153.8
Westinghouse Non-Proprietary Class 3 Table 6.1-9 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 68 EFPY R.G.
3/4T 1.99, Wt.%
Wt.%
cF<*J Fluence<bl 3/4T RTNDT(U)
(c)
RPV Material cu<*)
Ni<*>
(x 10 19 n/cm2, FF(b)
Rev. 2 (OF)
(OF)
Position E > 1.0 MeV)
R eactor Vessel Extended B eltline Materials Inlet Nozzle I to Upper Shell Weld 1.1 0.34 0.68 220.6 0.00714 0.087
-7.0
__________ (Heat# 299L44) ______________ --------------*------- ------ ------- ~---------------- ~--------e--------------
Using credible surveillance data
2. 1 249.8 0.00714 0.087
-7.0 Inlet Nozzle 2 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.00184 0.031
-7.0
___________ (Heat # 299L44) _______________ ------------------- ------- -----

L-----------1----------------

Using credible surveillance data
2. 1 249.8 0.00184 0.031
-7.0 Inlet Nozzle 3 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.00256 0.040
-7.0
________ (Heat # 299L44) _______________ ------------
,__ ________________ L---------1-----------
Using credible surveillance data 2.1 249.8 0.00256 0.040
-7.0 Inlet Nozzle 1 to Upper Shell Weld 1.1 0.19 0.57 (Heat # 8T l 762) 152.4 0.00714 0.087
-4.9 Inlet Nozzle 2 to Upper Shell Weld 1.1 (Heat # 8Tl 762) 0.19 0.57 152.4 0.00184 0.031
-4.9 Inlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 (Heat# 8Tl 762) 152.4 0.00256 0.040
-4.9 Outlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 8T 1762) 0.19 0.57 152.4 0.00191 0.032
-4.9 Outlet Nozzle 2 to Upper Shell Weld I.I 0.19 0.57 152.4 0.00138 0.024
-4.9 (Heat # 8Tl 762)
Outlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00533 0.070
-4.9 (Heat# 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld 1.1 (Heat# 8Tl554B) 0.16 0.57 143.9 0.00191 0.032
-4.9 WCAP-18242-NP LlRTNDT (d) o.CC>
(OF)
(OF) 0.0 (19. 1) 20.6 0.0 (21. 7) 20.6 0.0 (6.8) 20.6 0.0 (7. 7) 20.6 0.0 (8.8) 20.6 0.0 (JO.OJ 20.6 0.0 (13.2) 19.7 0.0 (4.7) 19.7 0.0 (6.1) 19.7 0.0 (4.8) 19.7 0.0 (3.7) 19.7 0.0 (10.7) 19.7 0.0 (4.5) 19.7 6-22 0 ~ (c)
Margin (OF)
(OF) 0.0 41.2 0.0 41.2 0.0 41.2 0.0 41.2 0.0 41.2

L-------

0.0 41.2 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 0.0 39.4 October 2017 Revision 0 3/4T ART (OF) 34.2 34.2 34.2 34.2 34.2
~---------
34.2 34.5 34.5 34.5 34.5 34.5 34.5 34.5
Westinghouse Non-Proprietary Class 3 6-23 Table 6.1-9 Calculation of the Surry Unit 1 ART Values at the 3/4T Location for 68 EFPY R.G.
3/4T CF<*>
(c) 1.99, Wt. %
Wt.%
Fluence<h) 3/4T RTNDT(U)
~RTNDT(d) o}c) 0'4 (c)
Margin RPV Material c u<*)
Ni<*>
(x 1019 n/cm2, FF<h>
Rev. 2 (OF)
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Outlet Nozzle 2 to Upper Shell Weld 1.1 0.16 0.57 143.9 0.00138 0.024
-4.9 0.0 (3.5) 19.7 0.0 39.4 (Heat# 8Tl554B)
Outlet Nozzle 3 to Upper Shell Weld I. I 0.16 0.57 143.9 0.00533 0.070
-4.9 0.0 (10.1) 19.7 0.0 39.4 (Heat# 8Tl554B)
Notes:
(a)
Chemical composition data taken from Tables 3-1 and 3-2 of this report. Chemistry factor values taken from Table 3-10 of this report.
(b)
The 3/4T fluence and 3/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 18) correlations and the Surry Unit I reactor vessel wall thickness of 8.05 inches.
( c)
Initial RT NOT values and cr1 values are from Tables 3-1 and 3-2 of this report.
(d)
Calculated L'>RTNDT values Jess than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated L'>RTNoT values are listed in parentheses for these materials.
(g)
As summarized in Appendix A of this report, all surveillance data for Surry Unit I were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 18), the base metal cr6 = l 7°F for Position 1.1, and cr6 = 8.5°F for Position 2.1 with credible surveillance data. Also per Regulatory Guide 1.99, Revision 2, the weld metal cr6 = 28°F for Position 1.1, and with credible surveillance data cr6 = 14°F for Position 2.1.
However, cr6 need not exceed 0.5*1'.'1RT NOT* For welds utilizing initial RT NOT values based on BA W-2308, cr6 = 28°F per References 15 and I 6.
WCAP-18242-NP October 2017 Revision 0 3/4T ART (OF) 34.5 34.5
Westinghouse Non-Proprietary Class 3 6-24 Table 6.1-10 Calculation of the Surry Unit 2 ART Nozzle Values at the Surface Location for 68 EFPY R.G.
Surface 1.99, Wt.%
Wt.%
CF<*>
Fluence<h)
Surface RTNDT(U)
(c)
.6.RTNDT (d) a.CO>
Oe,.(c)
Margin RPV Material Rev. 2 cu<*)
Ni<*>
(OF)
(x 1019 n/cm2, FF
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Inlet Nozzle l (Heat # 9-5104) 1.1 0.159 0.84 123.4 0.0139 0.137
-29.7 0.0 (16.8) 0.0 0.0 0.0 Inlet Nozzle 2 (Heat # 9-4815) 1.1 0.159 0.87 123.7 0.00321 0.048 4.5 0.0 (5.9) 0.0 0.0 0.0 Inlet Nozzle 3 (Heat # 9-5205) 1.1 0.159 0.86 123.6 0.00437 0.061 6.5 0.0 (7.5) 0.0 0.0 0.0 Outlet Nozzle I (Heat# 9-4825-2) 1.1 0.159 0.85 123.5 0.00338 0.050
-58.1 0.0 (6.2) 0.0 0.0 0.0 Outlet Nozzle 2 (Heat# 9-5086-1 )
I. I 0.159 0.86 123.6 0.00248 0.039
-26.6 0.0 (4.8) 0.0 0.0 0.0 Outlet Nozzle 3 (Heat # 9-5086-2)
I. I 0.159 0.87 123.7 0.0107 0.115
-33.8 0.0 (14.2) 0.0 0.0
0.0 Notes
(a)
Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistry Factor values taken from Table 3-12 of this report.
(b)
Surface fluence values taken from Section 2 of this report. FF= fluence factor = r0*28-0*10' 10g(Ol.
(c)
Initial RT NDT values and a 1 values are from Table 3-4 of this report.
(d)
Calculated LlRT NDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated LlRT NDT values are listed in parentheses for these materials.
( e)
Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 18), the base metal aA = l 7°F for Position I. I. However, aA need not exceed 0.5*1'.lRTNDT*
(f)
Nozzle materials are not limiting for P-T limit curves per WCAP-18243-NP (Reference 33).
WCAP-18242-NP October 2017 Revision 0 ART<f)
(OF)
-29.7 4.5 6.5
-58.1
-26.6
-33.8
Westinghouse Non-Proprietary Class 3 Table 6.1-11 Calculation of the Surry Unit 2 ART Values at the 1/4T Location for 68 EFPY R.G.
1/4T 1.99, Wt.%
Wt.%
CF<*>
Fluence<bl 1/4T RPV Material Rev. 2 cu<*)
Ni<*>
(x 1019 n/cm2, FF(bl (OF)
Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 123V303VAI I. I 0.11 0.72 75.8 0.534 0.825 Upper to Intermediate Shell 1.1 0.35 0.1 0 160.5 0.534 0.825 Circumferential Weld (Heat# 4275)
Intermediate Shell Plate C433 l-2 I.I 0.12 0.60 83.0 4.44 1.378 Intermediate Shell Plate C4339-2 I. I 0.11 0.54 73.4 4.44 1.378

L...__------

Using non-credible surveillance data 2.1 75.7 4.44 1.378 Intem1ediate Shell Longitudinal Welds 1.1 0.22 0.54 167.0 0.796 0.936 L3 and L4 (OD 50%) (Heat# 72445)

1------------

Using credible surveillance data 2.1 167.0 0.796 0.936 Intem1ediate Shell Longitudinal Weld L4 I.I 0.19 (JD 50%) (Heat# 8Tl 762) 0.57 167.0 0.796 0.936 Intermediate to Lower Shell I. I 0.187 0.545 147.5 Circumferential Weld (Heat# 0227) 4.45 1.379 Using credible surveillance data 2.1 132.5 4.45 1.379 Lower Shell Plate C4208-2 1.1 0.15 0.55 107.3 4.48 1.380 Lower Shell Plate C4339-l 1.1 0.107 0.53 70.8 4.48 1.380 Using non-credible surveillance data 2.1 75.7 4.48 1.380 Lower Shell Longitudinal Welds L1 and 1.1 0.19 0.57 167.0 0.802 0.938 L2 (Heat # 8Tl 762)
WCAP-18242-NP RTNDT(U)
(c)
LiRTNDT (d) o?>
(OF)
(OF)
(OF) 30 62.5 0.0 0
132.3 20.0 15.0 114.4 0.0 7.8 101.2 0.0
~------- ------------~---
7.8 104.3 0.0
-72.5 156.3 12.0 f---------------- -------------- ------
-72.5 156.3 12.0
-48.6 156.3 18.0 0
203.4 0.0 0
182.7 0.0
-30 148.1 0.0
-4.4 97.7 0.0
-4.4 104.5 0.0
-48.6 156.7 18.0 6-25 0 1!,. (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 17.0 34.0 28.0 60.9 28.0 60.9 28.0 66.6 28.0 56.0 14.0 28.0 17.0 34.0 17.0 34.0 17.0 34.0 28.0 66.6 October 2017 Revision 0 1/4T ART (OF) 126.5 201.2 163.4 143.0 146.1 144.7 144.7 174.3 259.4 210.7 152.1 127.3 134.1 174.6
Westinghouse Non-Proprietary Class 3 6-26 Table 6.1-11 Calculation of the Surry Unit 2 ART Values at the 1/4T Location for 68 EFPY R.G.
1/4T cp<a>
Fluence(b)
RT Cc>
(d) cs?>
Gt. (c)
RPV Material 1.99, Wt.%
Wt.%
1/4T NDT(U)
ART 'DT Margin Rev. 2 cu<*)
Ni<*>
(OF)
(x 10 19 n/cm2, pp(b)
(OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle I to Upper Shell Weld I.I 0.19 0.57 152.4 0.0210 0.177
-4.9 27.0 19.7 13.5 47.8 (Heat# 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00484 0.065
-4.9 0.0 (10.0) 19.7 0.0 39.4 (Heat# 8TI 762)
Inlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00660 0.082
-4.9 0.0 (12.5) 19.7 0.0 39.4 (Heat# 8T 1762)
Outlet Nozzle I to Upper Shell Weld I. I 0.35 1.0 272.0 0.00491 0.066 30 0.0 (18.0) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 2 to Upper Shell Weld I. I 0.35 1.0 272.0 0.00361 0.052 30 0.0 (14.3) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld I. I 0.35 1.0 272.0 0.0156 0.147 30 40.0 0.0 20.0 40.0 (Rotterdam)
Notes:
(a)
Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistry Factor values taken from Table 3-12 of this report.
(b)
The l/4T fluence and l/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 18) correlations and the Surry Unit 2 reactor vessel wall thickness of 8.05 inches.
(c)
Initial RT NDT values and cr1 values are from Tables 3-3 and 3-4 of this report.
(d)
Calculated ~RTNDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated ~RTNDT values are listed in parentheses for these materials.
(e)
Per Appendix A of this report, 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 (Reference 18), the base metal a,.. = l 7°F for Position 1.1 and for Position 2.1 with non-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,.. = l 4°F for Position 2.1. However, a,.. need not exceed 0.5*~RT NDT* For welds utilizing initial RT NDT values based on BA W-2308, a,.. = 28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 1/4T ART (OF) 69.9 34.5 34.5 30.0 30.0 110.0
Westinghouse Non-Proprietary Class 3 Table 6.1-12 Calculation of the Surry Unit 2 ART Values at the 3/4T Location for 68 EFPY R.G.
3/4T 1.99, Wt.%
Wt.%
CF<*>
Fluence<hl 3/4T RPV Material Rev. 2 cu<*)
Ni<*l (OF)
(x 10 19 n/cm2, FFCbJ Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging l 23V303V A I I. I 0.11 0.72 75.8 0.203 0.573 Upper to Intermediate Shell 1.1 0.35 Circumferential Weld (Heat # 4275) 0.10 160.5 0.203 0.573 Intermediate Shell Plate C433 l-2
1. 1 0.12 0.60 83.0 1.69 l.145 Intermediate Shell Plate C4339-2 l.l 0.11 0.54 73.4 1.69 1.145 L....-______
Using non-credible surveillance data 2.1 75.7 1.69 1.145 Intermediate Shell Longitudinal Welds I. 1 0.22 0.54 167.0 0.303 0.673 L3 and L4 (OD 50%) (Heat# 72445)
Using credible surveillance data 2.1 167.0 0.303 0.673 Intermediate Shell Longitudinal Weld L4 I. I 0.19 0.57 167.0 0.303 0.673 (ID 50%) (Heat # 8Tl 762)
Intermediate to Lower Shell 1.1 0.187 0.545 147.5 Circumferential Weld (Heat # 0227) 1.70
1. 145 Using credible surveillance data 2.1 132.5 1.70 1.145 Lower Shell Plate C4208-2 I.I 0.15 0.55 107.3 1.70 1.147 Lower Shell Plate C4339-l 1.1 0.107 0.53 70.8 1.70 1.147 Using non-credible surveillance data
2. 1 75.7 1.70 1.147 Lower Shell Longitudinal Welds L1 and I. I 0.19 0.57 167.0 0.305 0.675 L2 (Heat # 8Tl 762)
WCAP-18242-NP RTNDT(U)
(c) dRTNDT (d) o,<c)
(OF)
(OF)
(OF) 30 43.4 0.0 0
92.0 20.0 15.0 95.0 0.0 7.8 84.0 0.0 7.8 86.6 0.0
-72.5 112.4 12.0

~----

-72.5 112.4 12.0
-48.6 112.4 18.0 0
168.9 0.0 0
151.8 0.0
-30 123.1 0.0
-4.4 81.2 0.0
-4.4 86.8 0.0
-48.6 112.7 18.0 6-27 Ge,. (c)
Margin (OF)
(OF) 17.0 34.0 28.0 68.8 17.0 34.0 17.0 34.0 17.0 34.0 28.0 60.9 28.0 60.9 28.0 66.6 28.0 56.0 14.0 28.0 17.0 34.0 17.0 34.0 17.0 34.0 28.0 66.6 October 201 7 Revision 0 3/4T ART (OF) 107.4 160.8 144.0 125.8 128.4 100.8 100.8 130.3 224.9 179.8 127.1 110.8 116.4 130.7
Westinghouse Non-Proprietary Class 3 6-28 Table 6.1-12 Calculation of the Surry Unit 2 ART Values at the 3/4T Location for 68 EFPY R.G.
3/4T 1.99, Wt.%
Wt.%
CF<*l Fluence<b) 3/4T RTNDT(U)
(c) aRTNDT (d)
(c)
G!J. (c)
Margin RPV Material 01 Rev. 2 cu<*l Ni<*l (OF)
(x 10 19 n/cm2, FF(bl (OF)
(OF)
(OF)
(OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle I to Upper Shell Weld I. I 0.19 0.57 152.4 0.00798 0.094
-4.9 0.0 (14.3) 19.7 0.0 39.4 (Heat# 8T l 762)
Inlet Nozzle 2 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00184 0.031
-4.9 0.0 (4.7) 19.7 0.0 39.4 (Heat # 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00251 0.039
-4.9 0.0 (6.0) 19.7 0.0 39.4 (Heat # 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld I.I 0.35 1.0 272.0 0.00187 0.031 30 0.0 (8.4) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 2 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.00137 0.024 30 0.0 (6.5) 0.0 0.0 0.0 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld 1.1 0.35 1.0 272.0 0.00594 0.076 30 0.0 (20.7) 0.0 0.0 0.0 (Rotterdam)
Notes:
(a)
Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistry Factor values taken from Table 3-12 of this report.
(b)
The 3/4T fluence and 3/4T FF were calculated using the Regulatory Guide 1.99, Revision 2 (Reference 18) correlations and the Surry Unit 2 reactor vessel wall thickness of 8.05 inches.
(c)
Initial RT NOT values and cr1 values are from Tables 3-3 and 3-4 of this report.
(d)
Calculated ~RTNoT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 (Reference 29); actual calculated ~RTNoT values are listed in parentheses for these materials.
(e)
Per Appendix A of this report, 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 (Reference 18), the base metal cr6 = l 7°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. Also per Regulatory Guide 1.99, Revision 2, the weld metal cr6 = 28°F for Position 1.1, and with credible surveillance data cr6 = 14°F for Position 2.1. However, cr6 need not exceed 0.5* ~RT NOT* For welds utilizing initial RT NOT values based on BA W-2308, cr6 = 28°F per References 15 and 16.
WCAP-18242-NP October 2017 Revision 0 3/4T ART (OF) 34.5 34.5 34.5 30.0 30.0 30.0
Westinghouse Non-Proprietary Class 3 6-29 Table 6.1-13 Summary of the Surry Units 1 and 2 Limiting ART Values Used in the Applicability Evaluation of the Reactor Vessel Heatup and Cooldown Curves Plant Surry Unit 1 Sun-y Unit 2 1/4T Limiting ART (°F) 3/4T Limiting ART (°F)
Limiting Material Existing 48 EFPY Existing 48 EFPY Curves TLAA TLAA Curves TLAA TLAA Documented in Evaluation at Evaluation at Documented in Evaluation at Evaluation at WCAP-14177, 48 EFPY 68 EFPY WCAP-14177, 48 EFPY 68 EFPY Revision o<*J Revision o<*J
{Circ Flaw) Circ. Weld:
Intermediate to Lower Shell 201.3 213.9 158.5 173.6 Circ. Weld, Heat# 72445
{Axial Flaw) Lon2. Weld:
Lower Shell Long. Weld L2 194.2 219.4 131.3 153.8 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)
{Axial Flaw) Plate:
Not Limiting Not Limiting 135.6 144.0 Intermediate Shell Plate C433 l-2 Axial Flaw) Weld: Lower Shell Longitudinal Weld LI and L2 158.5 174.6 Not Limiting Not Limiting Heat# 8Tl 762 Note:
(a) The limiting 48 EFPY l/4T and 3/4T ART values in the Technical Specifications correspond to the Surry Unit I Intermediate to Lower Shell Circumferential Weld (Heat# 72445). The basis for the P-T limit curves is contained in WCAP-14177, Revision O (Reference 23); however, the applicability was extended to 48 EFPY in a later analysis. See Appendix C for details.
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Westinghouse Non-Proprietary Class 3 7-1 7
SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES Tables 7-1 and 7-2 contain the Surry Units 1 and 2 recommended surveillance capsule withdrawal schedules, originally presented in MCOE-LTR-17-26, Revision 1 (Reference 25). These schedules meet the recommendations of ASTM El85-82 (Reference 34) as required by 10 CFR 50, Appendix H (Reference 35). To support a 20-year Subsequent License Renewal (SLR) at Surry Units 1 and 2, it is recommended that additional capsules are withdrawn as described herein. The recommendations for the next capsule withdrawal at each Unit are consistent with Dominion's withdrawal schedule change request (Reference 36), which has been accepted for NRC review (Reference 37).
For Surry Unit 1, it is recommended that Capsule Z be withdrawn and tested at approximately 44 EFPY (projected to occur in 2027), which is when fluence value of the capsule is projected to reach the projected 80-year (68 EFPY) peak vessel fluence (6.35 x 1019 n/cm2) (Reference 25). This fluence value is also below twice the 60-year RV peak fluence to support the current 60-year license ( 48 EFPY), and therefore also satisfies the existing license requirement for surveillance capsule withdrawal and testing.
Standby Capsules U, S, and Y should remain in the reactor. For asset management considerations, if additional metallurgical data or dosimetry measurements are desired, withdrawal and testing of one of these capsules should be considered. Removal of these capsules from the Surry Unit 1 reactor is recommended before fluence values exceed 1.25 times the 80-year peak RV fluence. Capsules U and S may remain in the RV through 68 EFPY, but Capsule Y should be removed prior to 58 EFPY (projected to occur in 204 l ).
For Surry Unit 2, it is recommended that Capsule U be withdrawn and tested at approximately 49 EFPY (projected to occur in 2032), which is when the fluence value of the capsule is projected to reach the projected 80-year (68 EFPY) peak vessel fluence (7.26 x 1019 n/cm2) (Reference 25). This fluence value is also below twice the 60-year RV peak fluence to support the current 60-year license ( 48 EFPY), and therefore also satisfies the existing license requirement for surveillance capsule withdrawal and testing.
Standby Capsules T and Z should remain in the reactor.
For asset management considerations, if additional metallurgical data or dosimetry measurements are desired, withdrawal and testing of one of these capsules should be considered. Removal of these capsules from the Surry Unit 2 reactor is recommended before fluence values exceed 1.25 times the 80-year peak RV fluence. Capsule Z may remain in the RV through 68 EFPY, but Capsule T should be removed prior to 60 EFPY (projected to occur in 2044).
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Westinghouse Non-Proprietary Class 3 7-2 Table 7-1 Surry Unit 1 Surveillance Capsule Withdrawal Schedule Withdrawal Projected Lead Factor Capsule Fluence Capsule ID Status at Withdrawal (n/cm2, E > 1.0 MeV)
EFPY T<a>
Withdrawn 1974 1.1 2.71 X 1018 (EOC 1) w (a,b)
Withdrawn 1978 3.4 3.68 X 1018 (EOC 4) y (a)
Withdrawn 1986 8.0 1.80 X 1019 (EOC 8) x<a,c)
Withdrawn 1997 16.1 2.11 X 1019 (EOC 14) z<d>
In Reactor 285° 44 1.52 6.41 X 1019 u<e,f)
In Reactor 65° Standby 1.07
< 7.94 X 1019 s <e)
In Reactor 295° Standby 1.20
< 7.94 X 1019 y (e,g)
In Reactor 165° Standby 1.43
< 7.94 X 1019 Notes:
(a)
EFPY and fluence values were determined per Westinghouse calculations performed considering the Measurement Uncertainty Recapture (MUR) uprate. Fluence values (Section 2 of this report) were calculated using methodologies which follow the guidance and meet the requirements of Regulatory Guide 1.190 (Reference 2).
(b)
Only dosimetry was measured.
(c)
Fluence value is significantly different from UFSAR (Reference 19). Capsule X was moved from the 65° location to the 165° location at End of Cycle (EOC) 12.
(d)
Capsule Z is to be withdrawn following 44 EFPY of plant operation (projected to occur in 2027), which is when the fluence on the capsule will have reached the projected 80-year (68 EFPY) peak vessel fluence (6.35 x 1019 n/cm2). This fluence value is also below twice the 60-year license RV peak fluence, and therefore also satisfies this recommendation. Capsule Z was moved from the 245° location to the 285° location at EOC 12.
( e)
Capsules U, S, and Y will remain in the reactor accruing fluence. If additional metallurgical data or dosimetry measurements are desired, withdrawal and testing of one of these capsules should be considered. Removal of the capsules is recommended before fluence values exceed the 80-year RV fluence times 1.25. Capsules U and S may remain in the RV through 68 EFPY, but Capsule Y should be removed prior to 58 EFPY (projected to occur in 2041 ).
(f)
Capsule U was moved from the 45° location to the 65° location at EOC 12.
(g)
Capsule Y was moved from the 305° location to the 165° location at EOC 14.
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Westinghouse Non-Proprietary Class 3 7-3 Table 7-2 Surry Unit 2 Surveillance Capsule Withdrawal Schedule Capsule ID Withdrawal Projected Lead Factor Capsule Fluence Status (n/cm2, E > 1.0 MeV)
EFPY at Withdrawal x<*>
Withdrawn 1975 1.2 2.97 X 1018 (EOC 1) w<*,c)
Withdrawn 1979 3.8 6.36 X 1018 (EOC 4)
W-1 ca,bJ Withdrawn 1997 5.3 7.80 X 10 18 (EOC 14) s<*,c)
Withdrawn 1996 15.0 1.07 X 10 19 (EOC 13) y(a)
Withdrawn 1986 8.4 1.89 X 10 19 (EOC 8) y (a,d)
Withdrawn 2002 20.3 2.72 X 1019 (EOC 17) u<*>
In Reactor 285° 49 1.43 7.31 X 1019 T<tl In Reactor 165° Standby 1.43
< 9.08 X 1019 z<tl In Reactor 245° Standby 1.13
< 9.08 X 1019 Notes:
(a)
EFPY and fluence values were determined per Westinghouse calculations performed considering the Measurement Uncertainty Recapture (MUR) uprate. Fluence values were calculated (Section 2) using methodologies which follow the guidance and meet the requirements of Regulatory Guide 1.190 (Reference 2).
(b)
Master Integrated Reactor Vessel Materials Surveillance Program capsule also called Capsule Wl. EFPY value corresponds to total EFPY accrued. Inserted at EOC I 0.
(c)
Only dosimetry was measured.
(d)
Capsule Y was moved from the 295° location to the 165° location at EOC 12.
(e)
Capsule U is to be withdrawn following 49 EFPY of plant operation (projected to occur in 2032), which is when the fluence on the capsule will have reached the projected 80-year (68 EFPY) peak vessel fluence (7.26 x 1019 n/cm2). This fluence value is also below twice the 60-year RV peak fluence, and therefore also satisfies this recommendation. Capsule U was moved from the 65° location to the 285° location at EOC 22.
(t)
Capsules T and Z will remain in the reactor accruing fluence. If additional metallurgical data or dosimetry measurements are desired, withdrawal and testing of one of these capsules should be considered. Removal of these capsules is recommended before fluence values exceed the 80-year RV fluence times 1.25. Capsule Z may remain in the RV through 68 EFPY, but Capsule T should be removed prior to 60 EFPY (projected to occur in 2044). Capsule Twas moved from the 55° location to the 165° location at EOC 17. Capsule Z was moved from the 305° location to the 245° location at EOC 12.
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REFERENCES
1. Code of Federal Regulations, 10 CFR Part 54.3, "Definitions," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 72, dated August 28, 2007.
2. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,"
March 2001.
3. Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.
4. NUREG-2192, "Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, July 2017. [Agencywide Documents Access and Management System (ADAMS) Accession Number MLI 7 J 88AJ 58].
5. Virginia Electric and Power Company Letter to the U.S. Nuclear Regulatory Commission (Serial No.15-023), "Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed License Amendment Request Clarification of Reactor Coolant System Heatup and Cooldown Limitations Technical Specification Figures Response to Additional Information," dated February 4, 2015. [ADAMS Accession Number ML1504JA 720].
6.
Virginia Electric and Power Company Letter to the U.S. Nuclear Regulatory Commission (Serial No. 15-023B), "Virginia Electric and Power Company Surry Power Stations Units 1 and 2 Final Update Regarding Fluence Assessment for Reactor Vessel Inlet and Outlet Nozzles," dated October 26, 2015. [ADAMS Accession Number MLI 5302A340).
7. RSICC Computer Code Collection CCC-650, "DOORS 3.2, One, Two, and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998.
8. RSICC Data Library Collection DLC-185, "BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," July 1999.
9. AREVA NP, Inc. Report BAW-2313, Revision 7, Supplement 1, Revision 1, "Supplement to B&W Fabricated Reactor Vessel Materials and Surveillance Data Information for Surry Unit l and Unit 2," AREVA Document No. 77-2313S-007-001, February 2017.
10. Pressurized Water Reactor Owners Group (PWROG) Report PWROG-16045-NP, Revision 0, "Determination of Unirradiated RT NOT and Upper-Shelf Energy Values of the Surry Units 1 and 2 Reactor Vessel Materials," March 2017.
11. Westinghouse Report WCAP-18028-NP, Revision 0, "Extended Beltline Pressure Vessel Fluence Evaluations Applicable to Surry Units 1 & 2," September 2015.
12. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 of LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements," Revision 2, U.S. Nuclear Regulatory Commission, March 2007.
13. ASME Boiler and Pressure Vessel (B&PV) Code, Section Ill, Division 1, Subarticle NB-2300, "Fracture Toughness Requirements for Material."
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14. BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.
15. Framatome ANP Report BAW-2308, Revision 1-A Safety Evaluation (SE), "Final Safety Evaluation for Topical Report BAW-2308, Revision 1, 'Initial RTNoT of Linde 80 Weld Materials'," August 2005. [ ADAMS Accession Number ML052070408}.
16. Framatome ANP Report 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 NDT of Linde 80 Weld Materials'," March 2008. [ ADAMS Accession Number ML080770349}.
17. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
18. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
19. Surry Power Station Updated Final Safety Analysis Report, Revision 48, September 2016.
20. Framatome ANP Report BAW-2494, Revision 1, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Surry Units 1 and 2 for Extended Life through 48 Effective Full Power Years," September 2005.
21. Framatome ANP Report BAW-2324, Revision 0, "Analysis of Capsule X, Virginia Power Surry Unit No. 1, Reactor Vessel Material Surveillance Program," April 1998.
22. Westinghouse Report WCAP-16001, Revision 0, "Analysis of Capsule Y from Dominion Surry Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003.
23. Westinghouse Report WCAP-14177, Revision 0, "Surry Units l and 2 Heatup and Cooldown Limit Curves for Normal Operation," October 1994.
24. PWROG Report ANP-2650, "Updated Results for Request for Additional Information Regarding Reactor Pressure Vessel Integrity," July 2007.
25. Westinghouse Letter MCOE-LTR-17-26, Revision 1, "Surry Units 1 & 2 Surveillance Capsule Withdrawal Schedule for SLR," August 2017.
26. AREVA NP, Inc. Report AREVA-17-01417, "Transmittal of Information for Surry Specific Weld Wire Heat 299L44 (Capsule A5) from BAW-2313 Revision 7, PA-MSC-1200RO Task 1," May 2017.
27. Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.
28. Code of Federal Regulations, IO CFR 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 75, No. 1, dated January 4, 2010, with corrections dated February 3, 2010 (No. 22), March 8, 2010 (No. 44), and November 26, 2010 (No. 227).
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29. U.S. NRC Technical Letter Report TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels," Office of Nuclear Regulatory Research [RES],
dated November 14, 2014. [ADAMS Accession Number MLJ4318AJ77}.
30. Westinghouse Report WCAP-13587, Revision 1, "Reactor Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors," September 1993.
31. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."
32. U.S. NRC Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components,"
U.S. Nuclear Regulatory Commission, dated October 14, 2014.
33. Westinghouse Report WCAP-18243-NP, Revision 0, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," October 2017.
34. ASTM El 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.
35. Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
36. Virginia Electric and Power Company (Dominion Energy Virginia) Letter to the U.S. Nuclear Regulatory Commission (Serial No.17-243), "Virginia Electric and Power Company (Dominion Energy Virginia) Surry Power Stations Units 1 and 2 Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules," dated July 28, 2017.
37. U.S. Nuclear Regulatory Commission Response to Dominion Energy Virginia (Serial No.17-380), "Acceptance Review Re: Surveillance Capsule Withdrawal Schedule - Surry Power Stations Units 1 and 2," dated September 18, 2017.
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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A CREDIBILITY EVALUATION OF THE SURRY UNITS 1 AND 2 SURVEILLANCE DATA INTRODUCTION Regulatory Guide 1.99, Revision 2 (Reference A-1) describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels.
Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describe the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
To date, there have been four surveillance capsules removed from the Surry Unit 1 reactor vessel; three were tested to provide Charpy data. Five plant-specific surveillance capsules were removed from the Surry Unit 2 reactor vessel; three were tested to provide Charpy data. Additional weld surveillance data will also be evaluated from other plants. To use the surveillance data, the data must be shown to be credible. In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.
The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Surry Units 1 and 2 reactor vessel surveillance data to determine if the surveillance data is credible.
EVALUATION Criterion 1:
Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.
The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements" (Reference A-2), as follows:
"the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage. "
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Westinghouse Non-Proprietary Class 3 A-2 The Surry Unit 1 reactor vessel beltline region traditionally consists of the following materials:
1.
Intermediate Shell Plates C4326-1 and C4326-2
2.
Lower Shell Plates C4415-1 and C44 l 5-2
3.
Upper Shell Forging 122Vl09VA1
4.
Upper to Intermediate Shell Circumferential Weld Seam (Heat # 25017, SAF 89 Flux Type, Flux Lot Number 1197).
5.
Intermediate to Lower Shell Circumferential Weld Seam (Heat # 72445, Linde 80 Flux Type, (40%) Flux Lot Number 8597 and (60%) Flux Lot Number 8623)
6.
Intermediate Shell Plate Longitudinal Weld Seams L3 and L4 (Heat# 8Tl554, Linde 80 Flux Type, Flux Lot Number 8579)
7.
Lower Shell Longitudinal Weld Seams Ll (Heat # 8Tl554, Linde 80 Flux Type, Lot 8579) and L2 (Heat # 299L44, Linde 80 Flux Type, Lot 8596).
The Surry Unit 2 reactor vessel beltline region traditionally consists of the following materials:
1.
Intermediate Shell Plates C4331-2 and C4339-2
2.
Lower Shell Plates C4208-2 and C4339-l
3.
Upper Shell Forging 123V303VA1
4.
Upper to Intermediate Shell Circumferential Weld Seam (Heat # 4275, SAF 89 Flux Type, Flux Lot Number 02275)
5.
Intermediate to Lower Shell Circumferential Weld Seam (Heat # 0227, Grau Lo Flux Type, Lot LW320)
6.
Intermediate Shell Plate Longitudinal Weld Seams L3 (Heat # 72445, Linde 80 Flux Type, Flux Lot Number 8597) and L4 (50% - Heat # 72445, Linde 80 Flux Type, Flux Lot Number 8597 and 50% - Heat # 8Tl 762, Linde 80 Flux Type, Flux Lot Number 8597)
7.
Lower Shell Longitudinal Weld Seams Ll (Heat# 8Tl 762, Linde 80 Flux Type, Flux Lot Number 8597) and L2 (Heat # 8Tl 762, Linde 80 Flux Type, (63%) Flux Lot Number 8597 and (37%) Flux Lot Number 8632).
Per WCAP-7723, Revision O (Reference A-3) and WCAP-8085 Revision O (Reference A-4), the Surry Units 1 and 2 respective surveillance programs were developed to the requirements of ASTM El85.
WCAP-8085 specifically refers to the 1970 edition of ASTM El85 which states that the surveillance materials must be representative of materials in the highest flux region of the reactor.
Table 3-1 provides the initial material properties of the Surry Unit 1 reactor vessel beltline materials. Each of the beltline base metal materials has similar chemical properties. Lower Shell Plate C44 l 5-l has the highest initial RT NOT value ( other than the Upper Shell Forging), and is also representative of Lower Shell Plate C4415-2 which shares the same material heat number. Since this material is also in the high flux region of the reactor, this material meets the intent of Criterion 1. Per Table 3-1, each of the Surry Unit 1 beltline weld materials has similar USE and low initial RT NOT values. Since Heat # 299L44 has the WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-3 highest Cu wt. % value, and is located in the high flux region of the reactor, this material meets the intent of Criterion 1.
Table 3-3 provides the initial material properties of the Surry Unit 2 reactor vessel beltline materials. Each of the beltline base metal materials has similar chemical properties. Intermediate Shell Plate C4339-2 has the lowest initial USE value, and Upper Shell Forging 123V303VA1 has the highest initial RT NDT value.
Since Lower Shell Plate C4339-l is also representative of Intermediate Shell Plate C4339-2, which shares the same material heat number, and this material is in the high flux region of the reactor, this material meets the intent of Criterion 1. Per Table 3-3, each of the Surry Unit 1 beltline weld materials has similar chemical properties and low USE values. Since Heat # 0227 has the highest initial RT NDT value and is located in the high flux region of the reactor, this material meets the intent of Criterion 1.
Based on the discussion above, Criterion l is met for the Surry Units 1 and 2 surveillance programs.
Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.
Based on engineering judgment, the scatter in the data presented in the plots documented in BAW-2324 (Reference A-5) and WCAP-16001 (Reference A-6) is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Surry Units 1 and 2 surveillance materials unambiguously.
Hence, the Surry Units 1 and 2 surveillance programs meet this criterion.
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Westinghouse Non-Proprietary Class 3 Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of LlRT NOT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and l 7°F for base metal. Even if the fluence range is large ( two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, 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 E185-82 (Reference A-7).
A-4 The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these LlRT NOT values about this line is less than 28°F for welds and less than l 7°F for the plates.
Following is the calculation of the best-fit lines. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 (Reference A-8). At this meeting, the NRC presented five cases. Of the five cases, three Cases will be used to represent the Surry Units 1 and 2 Surveillance Material:
Case 1:
"Surveillance Data from Plant and No Other Source" Surry Unit 1 Lower Shell Plate C4415-1 Surry Unit 2 Lower Shell Plate C4339-l Surry Unit 2 Weld Material Heat# 0227 - Intermediate to Lower Shell Circ. Weld Case 4:
"Surveillance Data from Plant and Other Sources" Weld Material Heat # 299L44 - Surry Unit 1 Lower Shell Longitudinal Weld L2 and Inlet Nozzle to Upper Shell Welds.
Case 5:
"Surveillance Data from Other Sources Only" Weld Material Heat# 72445 from other Sources - Surry Unit 1 Intermediate to Lower Shell Circ. Weld and Surry Unit 2 Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%).
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Westinghouse Non-Proprietary Class 3 Credibility Assessment Case 1: Lower Shell Plate C44 l 5-l, Lower Shell Plate C4339-l, and Weld Heat # 0227 A-5 In accordance with the NRC guidelines, the plant-specific data from only Surry Units 1 and 2 will be analyzed first (Case 1). Case 1 interim chemistry factors are determined for both Surry Units l and 2 as summarized in Tables A-1 and A-2. Note that when evaluating the credibility of the surveillance weld data, the measured f1RT NOT values for the surveillance weld material do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld material measured shift values. In addition, only plant-specific (Surry Unit 1 or Surry Unit 2) data is being considered; therefore, no temperature adjustment is required.
The Surry Unit 1 Lower Shell Plate C4415-1 surveillance material data and credibility conclusions pertain to the Lower Shell Plate C44 l 5-l and to Lower Shell Plate C44 l 5-2 (same material heat). The Surry Unit 1, Case 1, chemistry factor is summarized in Table A-1.
TableA-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation for Surry Unit 1 Capsule Fluence<*>
ARTNDT (c)
Material Capsule (x 1019 FF
FF*.ART NDT(°F)
FF 2
n/cm2, E >
(OF) 1.0 MeV)
Lower Shell Plate T
0.271 0.644 50 32.21 0.415 C4415-l V
1.80 1.161 113 131.23 1.349 (Longitudinal)
X 2.11 1.203 86 103.46 1.447 SUM:
266.91 3.211 CF c441s-1= L(FF*~RTNDT)-;- L(FF2) = (266.9 1) -;- (3.211)= 83.1°F Notes:
(a) Capsule fluence values taken from Section 2.
(b) FF = fluence factor = £<0*28
o. lO*log I).
(c) !'.RT NOT values obtained from Table 7-6 ofBAW-2324 (Reference A-5).
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Westinghouse Non-Proprietary Class 3 A-6 Suny Unit 2 Lower Shell Plate C4339-l surveillance material data and credibility conclusions pertain to the Lower Shell Plate C4339-l and Intermediate Shell Plate C4339-2 (same material heat). Surry Unit 2 Weld Material Heat # 0227 surveillance data and credibility conclusions only pertain to Suny Unit 2 Intermediate to Lower Shell Circumferential Weld. Surry Unit 2, Case 1, chemistry factors are summarized in Table A-2.
TableA-2 Calculation of Interim Chemistry Factors for the Credibility Evaluation for Surry Unit 2 Capsule Fluence<a>
~TNDT
{c)
FF*~TNDT Material Capsule (x 10 19 FF{b>
FF 2
n/cm2, E >
(OF)
(OF) 1.0 MeV)
Lower Shell Plate X
0.297 0.668 59.08 39.45 0.446 C4339-l V
1.89 1.174 79.12 92.91 1.379 (Longitudinal) y 2.72 1.267 114.22 144.72 1.605 Lower Shell Plate X
0.297 0.668 48.67 32.50 0.446 C4339-l V
1.89 1.174 63.60 74.68 1.379 (Transverse) y 2.72 1.267 106.81 135.33 1.605 SUM:
519.59 6.860 CF C4339-I = I:(FF * ~RT NOT) + I(FF2) = (519.59) + (6.860) = 75.7°F Surveillance Weld X
0.297 0.668 95.65 63.86 Material V
1.89 1.174 140.21 164.64 (Heat # 0227) y 2.72 1.267 178.32 225.94 SUM:
454.45 CF Heat #0227 = I:(FF * ~RTNoT) + I:(FF2) = (454.45) + (3.430) =
Notes:
(a) Capsule fluence values taken from Section 2.
(b) FF = fluence factor = ~o.is -o.,o*1oso.
(c) MTNDT values obtained from Table 5-12 ofWCAP-16001 (Reference A-6).
WCAP-18242-NP 0.446 1.379 1.605 3.430 132.5°F October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-7 Credibility Assessment Case 4: Weld Heat# 299L44 (Surry Unit 1 and other sources)
Case 4 ("Surveillance Data from Plant and Other Sources") most closely represents the situation for the Surry Unit 1 Lower Shell Longitudinal Weld L2 and Inlet Nozzle to Upper Shell Welds (Heat# 299L44).
In accordance with the NRC Case 4 guidelines, the data from Surry Unit 1 and all Capsules listed in Table 3-7 containing Weld Heat# 299L44 will be analyzed together. Data is adjusted to the mean chemical composition and operating temperature of the surveillance capsules. Table A-3 provides the chemistry and temperature adjustment for Weld Heat# 299L44 data from all sources. The average chemistry and temperature are used to calculate Adjusted tt.RT NDT values and the interim CF for weld Heat # 299L44 data from all sources, as shown in Table A-4.
TableA-3 Mean Chemical Composition and Temperature for Weld Heat# 299L44<*>
Cu Ni Material Capsule Wt.%
Wt.%
Weld Metal Heat T
299144 V
0.23 0.64 (Surry Unit l Data)
X TMI2-LG1 (CR-3) 0.37 0.70 Wl(CR-3)
Weld Metal Heat TMil-E
299144 (Other Plant TMil-C 0.33 0.67 Data)
TMI2-LG I (TMI-2)
CR3-LG I (ONS-3) 0.36 0.70 AS 0.23 0.64 MEAN 0.30 0.67 Note:
(a) Data obtained from Table 3-7 or calculated herein.
WCAP-18242-NP Inlet Temperature during Period of Irradiation (°F) 537 539 542 556 545 556 556 556 556 556 550 Temperature Adjustment (°F)
-13
-11
-8 6
-5 6
6 6
6 6
October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-8 TableA-4 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat # 299L44 Capsule Chemistry Fluence ARTNDT Adjusted FF* Adjusted Capsule Factor (x 10 19 FF<*>
ARTNDT (b)
ART DT FF2 Position 1.1 n/cm 2
, E >
(OF)
(OF)
(OF) 1.0 MeV)
T 175.8 0.271 0.644 171 184.9 119. IO 0.415 V
175.8 1.80 1.161 250 279.6 324.75 1.349 X
175.8 2.11 1.203 234 264.4 318. 11 1.447 TMI2-LGl(CR-3) 234.0 0.830 0.948 216 195.4 185.15 0.898 Wl 234.0 0.780 0.930 262 226.2 210.40 0.865 TMil-E 215.2 0.107 0.431 74 76.0 32.72 0.185 TMI1-C 215.2 0.882 0.965 166 163.4 157.65 0.931 TMI2-LGl(TMl-2) 215.2 0.968 0.991 226 220.4 218.39 0.982 CR3-LG1 230.5 0.779 0.930 202 185.1 172.15 0.865 AS 175.8 2.75 1.270 246.6 295.5 375.26 1.612 SUM:
2113.67 9.550 CF Heat #299L44= L{Ff * ~RT NDT)..;- L(FF2) = (2 11 3.67)..;- (9.550) = 221.3°F Notes:
(a) FF= fluence factor = f0-2s-o.io*1og Q_
(b) Adjusted ~RT NDT values are ~T NDT values adjusted first to the mean operating temperature using the temperature adjustments in Table A-3, then to the mean chemical composition using the ratio procedure.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-9 Credibility Assessment Case 5: Weld Heat# 72445 (other sources only)
Case 5 ("Surveillance Data from Other Sources Only") most closely represents the situation for the Surry Units 1 and 2 reactor vessels use of Weld Heat # 72445. Surry Unit l Intermediate to Lower Shell Circumferential Weld and Surry Unit 2 Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%) are fabricated from Weld Heat # 72445, but neither plant included this weld metal heat in their original surveillance programs.
In accordance with the NRC Case 5 guidelines, the data from all capsules listed in Table 3-8 containing Weld Heat # 72445 will be analyzed together. Data is adjusted to the mean chemical composition and operating temperature of the surveillance capsules. Table A-4 provides the chemistry and temperature adjustment for Weld Heat # 72445 data from all sources. The average chemistry and temperature will be used to calculate Adjusted ~T NDT values and the interim CF for Weld Heat # 72445 data from all sources, as shown in Table A-6.
TableA-5 Mean Chemical Composition and Temperature for Weld Heat# 72445<a>
Material Capsule Cu Wt.%
CR3-LGI 0.22 CR3-LG2 0.22 Weld Metal Heat WI 0.22
72445 0.23 (Other Plant Point Beach Unit 1: Capsule V Data)
Point Beach Unit I: Capsule S 0.23 Point Beach Unit I : Capsule R 0.23 Point Beach Unit 1: Capsule T 0.23 MEAN 0.23 Note:
(a) Data obtained from Table 3-8 or calculated herein.
WCAP-18242-NP Ni Inlet Temperature during Period of Wt.%
Irradiation (°F) 0.59 556 0.59 556 0.59 545 0.62 542 0.62 542 0.62 541.6 0.62 533.4 0.61 545 Temperature Adj ustment(°F) 11 11 0
-3
-3
-3.4
-11.6 October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-10 TableA-6 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat# 72445 Chemistry Capsule Fluence Adjusted FF* Adjusted Capsule Factor (x 1019 FF<a>
ARTNDT ARTNoT
{b)
ARTNDT FF 2
Position (OF) n/cm2, E >
(OF)
{°F) 1.1 1.0MeV)
CR3-LG1 165.5 0.510 0.812 139 154.5 125.46 0.659 CR3-LG2 165.5 1.67 1.141 164 180.3 205.72 1.303 Wl 165.5 0.780 0.930 138 142.1 132.23 0.865 PB-I: Capsule V 172.4 0.634 0.872 107 103.0 89.81 0.761 PB-I: Capsule S 172.4 0.829 0.947 165 160.4 151.94 0.898 PB-I: Capsule R 172.4 2.19 1.213 155 150.1 182.00 1.471 PB-I : Capsule T 172.4 2.23 1.217 181 167.7 204.15 1.482 SUM:
1091.31 7.438 CF Heat #72445= L(FF * ~RTNoT) 7 L{FF 2
) = (1091.31) 7 (7.438) = 146.7°F Notes:
(a) FF= fluence factor = {0*2s-o.,o*1og 0_
(b) Adjusted 6RTNoT values are 6RTNoT values adjusted first to the mean operating temperature using the temperature adjustments in Table A-5, then to the mean chemical composition using the ratio procedure.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-II The scatter of ~RT NDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 (Reference A-1) is presented in Table A-7 for Surry Unit 1 and in A-8 for Surry Unit 2.
Table A-7 Surry Unit 1 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line CF Capsule Fluence Measured Material Capsule (Slope best-fit)
(x 10 19 n/cm2, FJ?'*l dRT "DT (OF)
E>LOMeV)
(OF)
Lower Shell Plate T
83.1 0.27 1 0.644 50 C44 15-I V
83.1 1.80 1.161 113 (Longitudinal)
X 83.1 2.11 1.203 86 T
22 1.3 0.271 0.644 171 V
221.3 1.80 1.161 250 X
221.3 2.11 1.203 234 TMI2-LG I 221.3 0.830 0.948 216 Surveillance W I 221.3 0.780 0.930 262 Weld Material (Heat# 299L44)
TMTI -E 221.3 0.107 0.431 74 TMII-C 221.3 0.882 0.965 166 TMI2-LGI 22 1.3 0.968 0.991 226 CR3-LG I 221.3 0.779 0.930 202 AS 221.3 2.75 1.270 246.6 CR3-LG I 146.7 0.510 0.812 139 CR3-LG2 146.7 1.67 1.141 164 Surveillance W I 146.7 0.780 0.930 138 Weld Material PB-1 :V 146.7 0.634 0.872 107 (Heat# 72445)
PB-I : S 146.7 0.829 0.947 165 PB-1 : R 146.7 2.19 1.213 155 PB-I : T 146.7 2.23 1.217 181 Notes:
(a) FF= fluence factor= f0*2s-o.,o*iogf).
(b) Adjusted to mean temperature and chemistry, as applicable.
(c) Scatter ~RT Nm= Absolute Value [Predicted ~RTNDT - Adjusted ~RTNoTl-WCAP-18242-NP Adjusted<bJ Predicted dRT 'DT dRTNDT
{°F)
(OF) 50.00 53.54 11 3.00 96.51 86.00 99.97 184.86 142.58 279.63 257.00 264.42 266.23 195.36 209.73 226. 16 205.87 76.00 95.28 163.40 213.5 1 220.40 219.28 185.12 205.80 295.54 280.99 154.50 119.12 180.25 167.43 142.14 136.47 102.96 127.97 160.38 138.98 150.08 177.90 167.71 178.58
<17°F Scatter (Base dRTNDT
(<)
Metal)
(OF)
<28°F (Weld) 3.54 Yes 16.49 Yes 13.97 Yes 42.28 No 22.63 Yes 1.81 Yes 14.37 Yes 20.29 Yes 19.28 Yes 50.11 No 1.12 Yes 20.68 Yes 14.55 Yes 35.38 No 12.82 Yes 5.67 Yes 25.01 Yes 21.40 Yes 27.81 Yes 10.87 Yes October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-12 TableA-8 Surry Unit 2 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line
<17°F CF Capsule Fluence Measured Adjusted<b)
Predicted Scatter (Base Material Capsule (Slope best-m)
(x 10 19 n/cm2, Ff<*>
ARTNDT ARTNDT ARTNDT ARTNDT
(<)
Metal)
(OF)
E> 1.0MeV)
(OF)
(OF)
(OF)
(OF)
<28°F
<Weld)
Lower Shell Plate X
75.7 0.297 0.668 59.08 59.08 50.54 8.54 C4339-I V
75.7 1.89 1.174 79.12 79.12 88.89 9.77 (Longitudinal) y 75.7 2.72 l.267 l 14.22 114.22 95.92 18.30 Lower Shell Plate X
75.7 0.297 0.668 48.67 48.67 50.54 1.87 C4339-1 V
75.7 1.89 1.174 63.60 63.60 88.89 25.29 (Transverse) y 75.7 2.72 1.267 106.81 106.81 95.92 10.89 Surveillance X
132.5 0.297 0.668 95.65 95.65 88.47 7.18 Weld Material V
132.5 1.89 1.174 140.21 140.2 1 155.59 15.38 (Heat # 0227) y 132.5 2.72 1.267 178.32 178.32 167.88 10.44 CR3-LG1 146.7 0.510 0.812 139 154.50 l 19.12 35.38 CR3-LG2 146.7 1.67 1.14]
164 180.25 167.43 12.82 Surveillance WI 146.7 0.780 0.930 138 142.14 136.47 5.67 Weld Material PB-l: V 146.7 0.634 0.872 107 102.96 127.97 25.0l (Heat# 72445)
PB-l : S 146.7 0.829 0.947 165 160.38 138.98 21.40 PB-1 : R 146.7 2.19 1.213 155 150.08 177.90 27.81 PB-I: T 146.7 2.23 1.217 181 167.71 178.58 10.87 Notes:
(a) FF = fluence factor= t<0*28 - O. JO'log I)_
(b) Adjusted to mean temperature and chemistry, as applicable.
(c) Scatter ARTNDT = Absolute Value [Predicted ARTNoT - Adjusted ARTNoTl-The data is deemed credible if all points in a data set fall within a+/- lcr scatter band. Statistically,+/- lcr would be expected to encompass 68% of the data. Tables A-7 and A-8 indicate that plate C4415-l, weld Heat# 299L44, weld Heat# 0227, and weld Heat# 72445 surveillance data falls inside the +/- lcr scatter band, and plate C4339-l surveillance data does not fall within the+/- lcr scatter band. Therefore, the plate C4415-l, weld Heat# 299L44, weld Heat# 0227 data, and weld Heat# 72445 are deemed "credible",
and C4339-l is deemed "non-credible" per the third criterion.
WCAP-18242-NP October 2017 Revision 0 Yes Yes No Yes No Yes Yes Yes Yes No Yes Yes Yes Yes Yes Yes
Westinghouse Non-Proprietary Class 3 A-13 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.
The capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite to the center of the core. The test capsules are contained in baskets attached to the thermal shield (References A-3 and A-4). The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F.
Hence, Criterion 4 is met for the Surry Units 1 and 2 surveillance programs.
Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.
The Surry Units 1 and 2 surveillance programs contain Standard Reference Material (SRM). The material was obtained from an A533 Grade B, Class 1 plate (HSST Plate 02). NUREG/CR-6413, ORNL/TM-13133 (Reference A-9) contains a plot of Residual vs. Fast Fluence for the SRM (Figure 11 in the report).
This Figure shows a 2a uncertainty of 50°F. The data used for this plot is contained in Table 14 in the NU REG report. However, the NUREG report does not consider the most up-to-date fluence and L'iRT NOT values for Surry surveillance capsules. Thus, Table A-9 contains an updated calculation of Residual vs.
Fast Fluence, considering the updated capsule fluence and L'iRT NOT values for the Surry surveillance capsules.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-14 TableA-9 Calculation of Residual vs. Fast Fluence for Surry Units 1 and 2 Capsule fluence Measured RG 1.99, Residua1<c>
Capsule (x 10 19 n/cm2, FF Shut<*>
Rev. z
(°F)
E > 1.0 MeV)
(°F)
Shift (°F)
Surry Unit 1 Capsule T 0.271 0.644 72 78.54 6.54 Surry Unit 1 Capsule V 1.80 1.161 142 141.57 0.43 Surry Unit 1 Capsule X 2.11 1.203 142 146.65 4.65 Surry Unit 2 Capsule X 0.297 0.668 62.19 81.39 19.20 Surry Unit 2 Capsule V 1.89 1.174 116.55 143.14 26.59 Surry Unit 2 Capsule Y 2.72 1.267 148.02 154.45 6.43 Notes:
(a) Measured 6T30 values for the SRM were taken from Table 7-6 of BAW-2324 (Reference A-5) for Surry Unit I and Table 5-12 ofWCAP-16001 (Reference A-6) for Surry Unit 2.
(b) Per NUREG/CR-64 13, ORNUfM-13133, the Cu and Ni values for the SRM (HSST Plate 02) are 0.17 and 0.64, respectively. This equates to a chemistry factor value of l 2 l.9"F based on Regulatory Guide 1.99, Revision 2, Position 1.1. The calculated shift is thus equal to CF* FF.
(c) Residual = Absolute Value [Measured Shift - RG 1.99 Shift].
The residual is less than 50°F (the allowable scatter in NUREG/CR-4613, ORNL/TM-13133) for all capsules.
Hence, Criterion 5 is met for the Surry Units 1 and 2 surveillance programs.
CONCLUSION Based on the preceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B:
The Surry Unit 1 surveillance plate data are deemed "credible" The Surry Unit 2 surveillance plate data are deemed "non-credible" The Weld Heat # 0227 data are deemed "credible" The Weld Heat # 2991A4 data are deemed "credible" The Weld Heat # 72445 data are deemed "credible" WCAP-18242-NP October 201 7 Revision 0
Westinghouse Non-Proprietary Class 3 A-15 APPENDIX A REFERENCES A-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
A-2 Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
A-3 Westinghouse Report WCAP-7723, Revision 0, "Virginia Electric and Power Co. Surry Unit No.
1 Reactor Vessel Radiation Surveillance Program," July 1971.
A-4 Westinghouse Report WCAP-8085, Revision 0, "Virginia Electric & Power Co. Surry Unit No. 2 Reactor Vessel Radiation Surveillance Program," June 1973.
A-5 Framatome ANP Report BAW-2324, Revision 0, "Analysis of Capsule X, Virginia Power Surry Unit No. 1, Reactor Vessel Material Surveillance Program," April 1998.
A-6 Westinghouse Report WCAP-16001, Revision 0, "Analysis of Capsule Y from Dominion Surry Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003.
A-7 ASTM El 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.
A-8 K. Wichman, M. Mitchell, and A. Hiser, US NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, "NRC/Industry Workshop on RPV Integrity Issues," February 12, 1998.
[ADAMS Accession Number MLl 10070570].
A-9 NUREG/CR-6413; ORNL/TM-13133, "Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials," April 1996.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B WELD MATERIAL HEAT# 0227 INITIAL RT NDT AND UPPER-SHELF ENERGY DETERMINATION Charpy V-notch data exists from multiple sources for the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227, Grau Lo flux LW320). Table B-1 provides Charpy V-notch test data taken from the Record of Weld Material Qualification for Heat# 0227, Grau Lo flux L W320 per Certified Material Test Reports (CMTRs). Table B-2 provides supplemental Charpy V-notch test data also obtained from CMTRs. Table B-3 provides the Charpy V-notch test data taken from Reference B-1 for the Surry Unit 2 surveillance weld, which was fabricated using the same weld Heat and flux type as the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld. Since the surveillance weld test data provides the most complete record of Charpy V-notch test information, it is appropriate to include this data for determination of the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld initial material properties.
Table B-1 Weld Material Qualification Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227i">
Temperature Temperature<bl Energy Energy<b)
(OC)
(OF)
(kgm/cm2)
(ft-lb)
-12 10 11.4 66
-12 10 8.8 51
-12 10 8.0 46 Notes:
(a) Data obtained from CMTRs.
(b) Converted value. Energy values were converted from kgm/cm2 to ft-lb utilizing the formula below. Note that 0.315 inch and 0.394 inch are the nominal dimensions of the Charpy specimen cross section per WCAP-8085 (Reference B-1 ).
2 14.223 lb*cm 2 3.28 ft Energy (ft-lbs) = Energy (kgm/cm )
kg*inz
-m- * (0.3 15 m.
0.394 m.)
WCAP-18242-NP October 2017 Revision 0
Table B-2 Westinghouse Non-Proprietary Class 3 Supplemental Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0221t>
Temperature Temperature<h>
Energy Energy
(OC)
(OF)
(kgm/cm2)
(ft-lb)
-1 2 10 8.1 47
-12 10 5.5 32
-12 10 6.6 38
-12 10 8.8 51
-12 10 7.5 43
-12 10 6.6 38
-12 10 11.4 66
-12 10 8.8 51
-12 10 8.8 51
-12 10 10.5 61
-12 10 10.4 60
-12 10 8.5 49
-12 10 9.5 55
-12 10 10.2 59
-12 10 10.0 58 Notes:
(a) Data obtained from CMTRs.
(b) Converted value. Energy values were converted from kgm/cm2 to ft-lb utilizing the formula below. Note that 0.315 inch and 0.394 inch are the nominal dimensions of the Charpy specimen cross section per WCAP-8085 (Reference B-1 ).
E (ft-lb ) = E (k m/
2)
14.223 lb *cm2
3.28 ft * (0 315.
0 394. )
nergy s
nergy g
cm kg*in 2 m
m.
m.
B-2 WCAP-1 8242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-3 Table B-3 Charpy V-Notch Test Data for Surry Unit 2 Surveillance Weld (Heat# 0227t>
Temperature Energy Shear Lateral Expansion (OF)
(ft-lb)
(%)
(mils)
-100 7
9 5
-100 7.5 9
5
-100 7
5 3
-40 15.5 17 15
-40 24 37 20
-40 34 33 31
-20 31 53 29
-20 27.5 47 25
-20 29 33 25 10 53 68 47 10 47 58 40 10 35 47 33 40 50 74 50 40 55.5 74 51 40 53.5 68 51 73 75 100 68 73 81 100 72 73 78 100 71 210 69.5 100 66 210 72 100 70 210 86 100 80 300 91 100 82 300 91 100 81 300 91 100 83 Note:
(a) Data obtained from WCAP-8085 (Reference B-1 ). Since the surveillance weld test data provides the most complete record of Charpy V-notch test information, it is appropriate to include this data for determination of the Suny Unit 2 Intermediate to Lower Shell Circumferential Weld initial material properties.
Using the data summarized in Tables B-1 through B-3, the initial RT NOT value for the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227) must be determined using NUREG-0800, BTP 5-3 guidance (Reference B-2) and in accordance with the ASME Code Section III, Subarticle NB-2331 requirements (Reference B-3).
Following NUREG-0800, BTP 5-3 Position 1.1 (1) guidance, T NOT "may be assumed to be the temperature at which 41 J (30 ft-lbs) was obtained in Charpy V-notch tests, or -l 8°C (0°F), whichever was higher." To precisely determine the temperature at which 30 ft-lbs was obtained on the weld specimens, the unirradiated Charpy V-notch data was plotted and fit using a hyperbolic tangent curve-fitting software, CVGRAPH, Version 6.02. Only the minimum data points (from Tables B-1 through B-WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-4
3) at each Charpy V-notch test temperature were used as input to the curve-fitting software, in accordance with ASME Code Section Ill, Subarticle NB-2331, Paragraph (a)(4). The resulting CVGRAPH figures are contained in the following pages for Charpy V-notch absorbed energy and lateral expansion.
Using these figures, the temperature at which 30 ft-lb absorbed energy was achieved was determined to be -5.6°F. Since this value is lower than 0°F, T NDT for this weld material is set equal to 0°F per BTP 5-3 Position 1.1 (1 ).
This estimate of T NDT and the Charpy V-notch test data in Tables B-1 through B-3 are used to determine RTNDT* Following the requirements of ASME Code Section Ill, Subarticle NB-2331, Paragraph (a)(2), the Cbarpy V-notch test data is first checked at a temperature equal to the drop-weight T NDT plus 60°F to determine if the material exhibits at least 50 ft-lb absorbed energy and 35 mils lateral expansion. Charpy V-notch tests were not performed at TNDT + 60°F. However, multiple Charpy V-notch tests were conducted at T NDT + 40°F (0°F + 40°F = 40°F) and did exhibit a minimum of 50 ft-lb absorbed energy and 35 mils lateral expansion. Thus, the test data are T NDT limited. For completeness, the unirradiated Cbarpy V-notch data was plotted and fit using a hyperbolic tangent curve-fitting software, CVGRAPH, Version 6.02. Only the minimum data points (from Tables B-1 through B-3) at each Charpy V-notch test temperature were used as input to the curve-fitting software, in accordance with ASME Code Section III, Subarticle NB-2331, Paragraph (a)( 4). The resulting CV GRAPH figures are contained in the following pages for Charpy V-notcb absorbed energy and lateral expansion.
Using these figures, the temperatures at which 50 ft-lb absorbed energy and 35 mils lateral expansion were achieved may be obtained. In this case, the absorbed energy test data is more conservative than the lateral expansion test data; therefore, it becomes the dominant data set in defining initial RT DT*
T5oft-1b=32.1°F T35 mils= 5.0°f T Cv = Max [T 50 ft-lb, T 35 mils]
Tcv =Max [32.1°F, 5.0°F]
Tcv = 32.1 °F Following the requirements of ASME Code Section III, Subarticle NB-2331, Paragraph (a)(3), the initial RT NDT is the higher of T NDT ( determined from the drop-weight tests) and T cv ( determined above) minus 60°F.
RT NDT = Max [T NDT, T Cv - 60°F]
RT NDT = Max [0°F, 32. l - 60°F]
RT NDT = Max [0°F, -27.9°F]
RTNoT= 0°F Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat# 0227) Initial RT~= 0°F WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-5 The current 10 CFR 50, Appendix G (Reference B-4), requirements specify that USE be calculated based on ASTM El85-82 (Reference B-5). Herein, USE is calculated based on an interpretation of ASTM El85-82 that is best explained by the most recent version of the ASTM El85 manual (2016 version).
Using the guidelines in ASTM El85-82 and El85-16 (Reference B-6), the average of all Charpy data
~ 95% shear is reported as the USE. In some instances, there may be data deemed 'out of family,' which are removed from the determination of the USE based on engineering judgment. However, the use of engineering judgment to remove 'out of family' data was not necessary for this material.
Intermediate to Lower Shell Circumferential Weld (Heat# 0227) USE = Average (75, 81, 78, 69.5, 72, 86, 91, 91, 91 ft-lbs)= 82 ft-lb WCAP-18242-NP October 20 l 7 Revision 0
Westinghouse Non-Proprietary Class 3 B-6 Surry Unit 2 Intermediate to Lower Shell Circumferential Weld CVGmph 6.02: Hypeibolic Tangent Cuive Printed on 5/18/2017 1:47 PM A = 35.85 B = 33.65 C = 60.43 TO = 5.00 D = 0.00 CorrcL1lion Coefficient = 0. 962 Eqmtion is A + B
1Tanh((T-TO)/(C+D1))]
Upper Shelf Energy = 69.50 (Fixed)
Lower Shelf Energy = 2.20 (Fixed)
Tcmp"ti,30 ft-lbs= -5.6(1° F Tcmp(ij/35 ft-lbs= 3.50° F Temp(it50 ft-lbs= 32. !0° F Plant: Surry 2 Orientation: N/A r,:i
.Q -
I
~
bl>
J.,
~
C
~
z
>-u 100 90 80 70
~
60
~
50 40 30 20 10 0
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Westinghouse Non-Proprietary Class 3 B-10 APPENDIX B REFERENCES B-1 Westinghouse Report WCAP-8085, Revision 0, "Virginia Electric & Power Co. Surry Unit No. 2 Reactor Vessel Radiation Surveillance Program," June 1973.
B-2 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 of LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements," Revision 2, U.S. Nuclear Regulatory Commission, March 2007.
B-3 ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subarticle NB-2300, "Fracture Toughness Requirements for Material."
B-4 Code of Federal Regulations, 10 CFR 50, Appendix G, "Fracture Toughness Requirements," U.S.
Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.
B-5 ASTM El85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, July 1982.
B-6 ASTM El85-16, "Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels," ASTM International, December 2016.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C
SUMMARY
OF THE APPLICABILITY OF P-T LIMIT CURVES FOR SURRY UNITS 1 AND 2 The Surry Units 1 and 2 P-T limit curves that are currently in Surry Power Station Technical Specifications (TS) (Reference C-1) were first approved in WCAP-14177 (Reference C-2) for End of License (EOL) and have applicability that was extended to 48 EFPY. Table C-1 contains a summary of the applicability of the Surry Units 1 and 2 P-T Limit curves. Figures C-1 and C-2 show the Surry Units 1 and 2 heatup and cooldown curves as currently depicted in the Surry Power Station Technical Specifications (Reference C-12). Tables C-2 and C-3 provide the data points corresponding to the heatup and cooldown curves, respectively, as currently depicted in the Surry Power Station Technical Specifications.
Table C-1 Surry Units 1 and 2 P-T Limit Curve Applicability History Subject Content Relevant to Surry Units 1 and 2 P-T Date Document(s)
Limit Curves P-T limit curves for 28.8 EFPY for Surry Unit 1 and 29.4 EFPY for Surry Unit 2 were created WCAP-14177, without inclusion of instrumentation errors. Note October Revision 0 that this evaluation pre-dates the first approval of 1994 Westinghouse's current NRC-approved methodology in WCAP-14040-A (Reference C-3).
Per the subject documents, an adjustment of 21.5 psi to accommodate for the pressure difference SM-792, Revision 3 between the pressurizer and reactor beltline was (Page 18/47) applied to the WCAP-14177 curves to create the TS curves. Additionally, the WCAP-14177 heatup 1995-1996 SM-945, Revision 0 curves are combined into one bounding heatup curve at temperatures of 3 l 5°F and above for the (Page 26/ 102)
TS. These calculations also state that no instrumentation uncertainties were added to the P-T limit curves.
The subject document contains the original request to the NRC to incorporate the curves based on WCAP-14177 in the plant TS. It is stated that the NRC Letter Serial curves do include a "correction for the effects of No.95-197 pressure measurement location" and repeats the June 1995 (Page 14/47) statement that instrument uncertainties are not included in the curves. The differences between WCAP-14177 and the TS curves are a result of pressure measurement location adjustments.
Letter from the NRC approved the P-T limits based on WCAP-December NRC 14177 through amendment No. 207.
1995 WCAP-15130, P-T limit curves for End of License Extension April 2001 Revision 1 (EOLE) were developed.
Letter from the The TS P-T Limits were changed to curves based January NRC on WCAP-15130, Revision 1.
2006 WCAP-18242-NP Reference Number(s)
C-2 C-4 and C-5 C-6 C-7 C-8 C-9 October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 C-2 Table C-1 Surry Units l and 2 P-T Limit Curve Applicability History Subject Content Relevant to Surry Units l and 2 P-T Date Reference Document(s)
Limit Curves Number(s)
Letter from the The P-T Limits approved under amendment No.
June 2006 C-10 NRC 207 were reinstated in the TS.
Letter from the The applicability of the P-T Limits approved under May 2011 C-11 NRC amendment No. 207 was extended to 48 EFPY.
Letter from the This reference represents the most recent TS P-T NRC limit curve amendment (No. 285), which June 2015 C-12 administratively alters the P-T limits.
Thus, the limiting ART values used to create the TS curves (based on WCAP-14177) are those used for determination of applicability of the P-T limit curves at 48 and 68 EFPY with updated fluence, material properties, and Position 2.1 chemistry factor values.
In summary, the current Surry Units l and 2 Technical Specifications P-T limit curves are based on WCAP-14177 with minor administrative changes and an applied pressure measurement adjustment of 21.5 psi.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 Surry Units 1 and 2
Reactor Coolant System Heatup Limitations M<1terial Pro~tty 8<1-~li Llmitlr,g Materi<1!: Surry Un't l lntermedL~te to lower Shell Circ Weld Lirniti~ AR Values for SuNV l at 48 £FPY:
1i4* T, 228,-t*F 3/4-T, 18~.s*F limiting Solwp emper.1t.1.1re SUNV 1 Initial RTto01 Chm.ire Flange Re ion: 10°1' D
So I~
1SO 100 lSO SOO ' :J.SO 400
00 SOO SSO
$00 6SCl lftdic:atad C:,ald le! 'Tampe~ture toeg. F)
Figure 3. l -1 : Sorry nits I and 2 Reactor Coolant System Heatup Limitations (Heatup Rate up to 6()1'Flhr) Applicable for 48 EFPY Amendment Nos. 285. 285 Figure C-1 Surry Units 1 and 2 Heatup P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications (Reference C-12)
C-3 WCAP-1 8242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 Surry Units 1 and 2 Reactor Coolant System*Cooldown limitations l\\llatedat Property Basis.
limiting Material: Surry Unit l lntermedi3te*t o Lower Shell Circ We!c:J Umiting ARfVaJues for Surry lat 48 EFPY:
l/4-T, 228.4°F
3/4-T, 189.s°F Umiting Soltup Temperature Surry l lnltl o1J RTl'i01 Ooso~ f~nge Re.Ilion: 10*,:
0 50 100 1SO 200 250 300 l>O 400 4SO S1l0 SSO
~O 6SO lndk1ted Cold L.1 Tampe..-bJre 10.i. F)
Figure 3.1-'2 : Surry Units I and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rat,es up to 100°F/hr) Applicable for 48 EFPY Amendment Nos. 285, 285 C-4 Figure C-2 Surry Units 1 and 2 Cooldown P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications (Reference C-12)
WCAP-1 8242-NP October 2017 Revision 0
WCAP-18242-NP Table C-2 Westinghouse Non-Proprietary Class 3 Data Points for Surry Units 1 and 2 Current Technical Specifications Heatup P-T Limit Curves 20°F/hr Heatup 40°F/hr Heatup 60°F/hr Heatup T (°F)
P (psig)
T (OF)
P (psig)
T (OF)
P (psig) 80 481.37 80 458.18 80 435.67 85 481.37 85 458.18 85 435.67 90 481.37 90 458.18 90 435.67 95 483.93 95 458.18 95 435.67 100 486.72 100 458.18 100 435.67 105 490.47 105 458.56 105 435.67 110 494.36 110 459.95 110 435.67 115 498.90 115 462.32 115 435.67 120 503.72 120 465.33 120 436.02 125 509.01 125 469.06 125 437.35 130 514.66 130 473.29 130 439.40 135 520.80 135 478.13 135 442.24 140 527.35 140 483.42 140 445.69 145 532.06 145 489.29 145 449.82 150 537.12 150 495.54 150 454.52 155 542.46 155 502.48 155 459.85 160 548.31 160 509.95 160 465.76 165 554.60 165 518.06 165 472.31 170 561.37 170 526.78 170 479.45 175 568.64 175 536.22 175 487.28 180 576.47 180 546.25 180 495.68 185 584.86 185 557.20 185 504.91 190 593.79 190 568.96 190 514.88 195 603.50 195 581.64 195 525.68 200 613.95 200 595.13 200 537.32 205 625.19 205 609.81 205 549.78 210 637.24 210 625.58 210 563.31 215 650.10 215 642.41 215 577.91 220 664.06 220 660.65 220 593.48 225 679.05 225 679.05 225 610.40 230 695.02 230 695.02 230 628.60 235 712.37 235 712.37 235 648.03 240 730.98 240 730.98 240 669.08 245 750.86 245 750.86 245 691.56 250 772.41 250 772.41 250 715.90 255 795.35 255 795.35 255 741.88 260 820.26 260 820.26 260 770.01 265 846.77 265 846.77 265 800.02 270 875.50 270 875.50 270 832.44 C-5 October 2017 Revision 0
WCAP-18242-NP Westinghouse Non-Proprietary Class 3 Table C-2 Data Points for Surry Units 1 and 2 Current Technical Specifications Heatup P-T Limit Curves 20°F/hr Heatup 40°F/hr Heatup 60°F/hr Heatup T(°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig) 275 906.20 275 906.20 275 867.18 280 939.14 280 939.14 280 904.40 285 974.78 285 974.78 285 944.39 290 1012.91 290 1012.91 290 987.33 295 1053.86 295 1053.86 295 1033.64 300 1097.82 300 1097.82 300 1083.17 305 1145.06 305 1145.06 305 1136.13 310 1195.82 310 1195.82 310 1193.21 315 1249.10 315 1249.10 315 1249.10 320 1302.07 320 1302.07 320 1302.07 325 1354.80 325 1354.80 325 1354.80 330 1409.89 330 1409.89 330 1409.89 335 1468.87 335 1468.87 335 1468.87 340 1531.93 340 1531.93 340 1531.93 345 1599.71 345 1599.71 345 1599.71 350 1672.05 350 1672.05 350 1672.05 355 1749.91 355 1749.91 355 1749.91 360 1833.09 360 1833.09 360 1833.09 365 1921.95 365 1921.95 365 1921.95 370 2017.08 370 2017.08 370 2017.08 375 2118.96 375 2118.96 375 2118.96 380 2227.79 380 2227.79 380 2227.79 385 2343.89 385 2343.89 385 2343.89 Leak Test Limit T (°F)
P (psig) 333 1978.5 355 2463.5 C-6 October 2017 Revision 0
Table C-3 Steady-State T (°F)
P (psig) 80 491.03 85 492.91 90 495.04 95 497.32 100 499.77 105 502.41 110 505.25 115 508.30 120 511.58 125 515.10 130 518.89 135 522.97 140 527.35 145 532.06 150 537.12 155 542.46 160 548.31 165 554.60 170 561.37 175 568.64 180 576.47 185 584.86 190 593.79 195 603.50 200 613.95 205 625.19 210 637.24 215 650.10 220 664.06 225 679.05 230 695.02 235
712.37 240 730.98 245 750.86 250 772.41 255 795.35 260 820.26 265 846.77 270 875.50 275 906.20 Westinghouse Non-Proprietary Class 3 C-7 Data Points for Surry Units 1 and 2 Current Technical Specifications Cooldown P-T Limit Curves 20°F/hr Cooldown<*>
40°F/hr Cooldown<*>
60°F/hr Cooldown<*>
100°F/hr Cooldown<*>
T(°F)
P (psig)
T (°F)
P (psig)
T(°F)
P (psig)
T (OF)
P (psig) 80 449.28 80 406.68 80 363.08 80 272.64 85 451.20 85 408.55 85 364.90 85 274.35 90 453.26 90 410.56 90 366.88 90 276.26 95 455.51 95 412.78 95 369.07 95 278.43 100 457.93 100 415.17 100 371.44 100 280.80 105 460.56 105 417.80 105 374.07 105 283.47 110 463.38 110 420.62 110 376.91 110 286.38 115 466.45 115 423.72 115 380.04 115 289.62 120 469.75 120 427.05 120 383.42 120 293.15 125 473.33 125 430.69 125 387.14 125 297.07 130 477.17 130 434.60 130 391.15 130 301.33 135 481.33 135 438.80 135 395.53 135 306.02 140 485.81 140 443.39 140 400.27 140 311.12 145 490.65 145 448.38 145 405.37 145 316.68 150 495.77 150 453.76 150 410.94 150 322.74 155 501.40 155 459.60 155 417.02 155 329.39 160 507.45 160 465.88 160 423.56 160 336.58 165 513.99 165 472.69 165 430.68 165 344.44 170 521.01 170 480.02 170 438.28 170 352.94 175 528.61 175 487.96 175 446.61 175 362.16 180 536.76 180 496.41 180 455.58 180 372.17 185 545.46 185 505.66 185 465.32 185 383.07 190 554.93 190 515.60 190 475.80 190 394.84 195 565.14 195 526.36 195 487.16 195 407.57 200 576.12 200 537.82 200 499.30 200 421.38 205 587.83 205 550.33 205 512.54 205 436.28 210 600.55 210 563.77 210 526.80 210 452.44 215 614.27 215 578.30 215 542.11 215 469.96 220 629.02 220 593.79 220 558.70 220 488.86 225 644.76 225 610.66 225 576.64 225 509.23 230 661.84 230 628.80 230 595.82 230 531.28 235 680.23 235 648.22 235 616.67 235 555.04 240 699.86 240 669.26 240 638.97 240 580.76 245 721.17 245 691.79 245 663.19 245 608.44 250 743.88 250 716.20 250 689.09 250 638.25 255 768.54 255 742.32 255 717.23 255 670.62 260 794.84 260 770.62 260 747.30 260 705.32 265 823.36 265 800.90 265 779.91 265 742.77 270 853.82 270 833.62 270 814.86 270 783.25 275 886.57 275 868.75 275 852.48 275 826.80 WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 C-8 Table C-3 Data Points for Surry Units 1 and 2 Current Technical Specifications Cooldown P-T Limit Curves Steady-State 20°F/hr Cooldown<*>
40°F/hr Cooldown<*>
60°F/hr Cooldown<*>
100°F/hr Cooldown<*>
T (OF) 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig)
T (°F)
P (psig) 939.14 280 922.02 280 906.49 280 892.94 280 873.65 974.78 285 959.97 285 947.11 285 936.54 285 924.18 1012.91 290 1000.71 290 990.76 290 983.61 290 978.48 1053.86 295 1044.51 295 1037.77 295 1034.14 295 1036.84 1097.82 300 1091.61 300 1088.29 300 1088.27 1145.06 305 1142.24 305 1142.68 1195.82 1250.37 1308.86 1371.62 1438.89 1511.21 1588.69 1671.46 1760.72 1856.03 1958.14 2067.32 2184.34 2308.98 2442.42 Note:
(a) The 20°F/hr and 40°F/hr cooldown curves are identical to the steady-state curve at 310°F and above. The 60°F/hr cooldown curve is identical to the steady-state curve at 305°F and above. The 100°F/hr cooldown curve is identical to the steady-state curve at 300°F and above.
WCAP-18242-NP October 201 7 Revision 0
Westinghouse Non-Proprietary Class 3 C-9 APPENDIX C REFERENCES C-1 Surry Power Station Technical Specifications, Section 3.1.B, Amendments Nos. 285 and 285.
C-2 Westinghouse Report WCAP-14177, Revision 0, "Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," October 1994.
C-3 Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.
C-4 Virginia Power Calculation SM-792, Revision 3, "Surry 1 & 2 Composite PfT Limits Curve,"
January 1996.
C-5 Virginia Power Calculation SM-945, Revision 0, "Surry Unit 1 and 2 Heatup/Cooldown Curves and LTOPS Setpoint," February 1995.
C-6 Letter 95-197 from Virginia Electric and Power Company to the Nuclear Regulatory Commission, "Virginia Electric and Power Company, Surry Power Station Units l and 2, Request for Exemption -
Code Case N-514, Proposed Technical Specifications Change, Revised Pressurefremperature Limits and LTOPS Setpoint," dated June 8, 1995.
C-7 Letter from the NRC to Virginia Electric and Power Company, "Surry Units 1 and 2 - Issuance of Amendments RE: Surry, Units 1 and 2 Reactor Vessel Heatup and Cooldown Curves," dated December 28, 1995. [Agencywide Documents Access and Management System (ADAMS)
Accession Number ML012710054].
C-8 Westinghouse Report WCAP-15130, Revision 1, "Surry Units 1 and 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation," April 2001.
C-9 Letter from the NRC to Vuginia Electric and Power Company, "Surry Power Station, Unit Nos. 1 and 2 - Issuance of Amendments on Reactor Coolant System Pressure and Temperature Limits,"
dated January 3, 2006. [ADAMS Accession Number ML053550091].
C-10 Letter from the NRC to Virginia Electric and Power Company, "Surry Power Station, Unit Nos. 1 and 2 - Issuance of Amendments to Reinstate Previous Reactor Coolant System Pressure and Temperature Limits," dated June 29, 2006. [ADAMS Accession Number ML061710242].
C-11 Letter from the NRC to Virginia Electric and Power Company, "Surry Power Station, Unit Nos. 1 and 2 - Issuance of Amendments regarding Reactor Vessel Heatup and Cooldown Curves for 48 Effective Full-Power Years," dated May 31, 2011. [ADAMS Accession Number MLllllOAlll].
C-12 Letter from the NRC to Virginia Electric and Power Company, "Surry Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Clarification of Reactor Coolant System Heatup and Cooldown Limitation Technical Specification Figures," dated June 26, 2015. [ADAMS Accession Number ML15173Al02].
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D EMERGENCY RESPONSE GUIDELINES The Emergency Response Guideline (ERG) limits were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event (Reference D-1 ). Generic categories of limits were developed for the guidelines based on the limiting inside surface RT NOT* These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants.
The highest value of RT NOT for which the generic category ERG limits were developed is 250°F for a longitudinal flaw and 300°F for a circumferential flaw. Therefore, if the limiting vessel material has an RT NOT that exceeds 250°F for a longitudinal flaw or 300°F for a circumferential flaw, plant-specific ERG P-T limits must be developed:
The ERG category is determined by the magnitude of the limiting RT NOT value, which is calculated the same way as the RT PTS values are calculated in Section 4 of this report. The material with the highest RT NOT defines the limiting material, which for Surry Unit 1 is the Lower Shell Longitudinal Weld L2 Position 2.1) and for Surry Unit 2 is the Upper to Intermediate Shell Circumferential Weld Heat# 4275 (Position 1.1 ). Table D-1 identifies ERG category limits and the limiting material RT NOT values at 68 EFPY for Surry Units 1 and 2.
Table D-1 Evaluation of Surry Units 1 and 2 ERG Limit Category ERG Pressure-Temperature Limits (Reference D-1)
Applicable RT NDT Value<a>
ERG P-T Limit Category RT NOT < 200°F Category I 200°F < RT NOT < 250°F Category II 250°F < RT NOT < 300°F Category Ill b Limiting RT NOT Value(b>
Reactor Vessel Material RT NOT Value @ 68 EFPY Surry Unit 1 Lower Shell Longitudinal 253.2 Weld L2 Heat # 299L44 (Position 2.1)
Surry Unit 2 Upper to Intermediate Shell Circumferential Weld Heat#
222.8 4275 (Position 1.1)
Notes:
(a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially oriented flaws are applicable up to 300°F.
(b) Values taken from Tables 4-1 and 4-2.
WCAP-18242-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 D-2 Per the ERG limit guidance document (Reference D-1), some vessels do not change categories for operation through the end of license. However, when a vessel does change ERG categories between the beginning and end of operation, a plant-specific assessment must be performed to determine at what operating time the category changes. Thus, the ERG classification need not be changed until the operating cycle during which the maximum vessel value of actual or estimated real-time RT OT exceeds the limit on its current ERG category. Note that Surry Units l and 2 are currently in ERG Category II.
Conclusion of ERG P-T Limit Categorization Per Table D-1, the limiting material for Surry Unit 2 (Upper to Intermediate Shell Circumferential Weld Heat# 4275 (Position 1.1 )) has an RT NOT value less than 250°F through 68 EFPY. Therefore, Surry Unit 2 remains in ERG Category II through SLR (68 EFPY). Per Table D-1, the limiting material for Surry Unit 1 (Lower Shell Longitudinal Weld L2 Heat# 299L44 (Position 2.1)) has an RT N OT above 250°F and corresponds to a longitudinally oriented flaw. Therefore, Surry Unit 1 will switch from ERG Category II to ERG Category Illa during the SLR extended period of operation. ERG Category Illa guidance does not currently exist for longitudinally oriented flaws; however, the PWROG is currently addressing ERG Category Illa through PA-MSC-1524, Revision 0. The EFPY at which Surry Unit 1 will switch from ERG Category II to ERG Category Illa is projected to be 64.9 EFPY. Since the limiting RT NOT is very close to the limit (253.2°F vs. 250°F limit) and the limiting material is in the Surry Unit 1 and other plant surveillance capsules, the date at which Surry Unit 1 changes ERG categories may change in the future as a result of surveillance capsule test results and/or fluence projection updates. Based on the current evaluation, the Surry Unit 1 ERG classification should be changed prior to the beginning of the operating cycle in which 64.9 EFPY will be reached.
APPENDIX D REFERENCES D-1 Westinghouse Owners Group Document HF04BG, "Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 2," April 2005.
WCAP-18242-NP October 2017 Revision 0

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