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WCAP-18243-NP, Rev. 0, Surry, Units 1 and 2, Heatup and Cooldown Limit Curves for Normal Operation.
	ML18085A164
Person / Time
Site:	Surry
Issue date:	10/31/2017
From:	Mays B, Turicik L; Westinghouse
To:	; Office of Nuclear Reactor Regulation
References
	WCAP-18243-NP, Rev 0
	Download: ML18085A164 (161)
	v • d • e

Serial Number 18-098 Docket Nos. 50-280/281 Attachment 4 WCAP-18243-NP, REV. 0, SURRY UNITS 1 AND 2 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION OCTOBER 2017 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION ENERGY VIRGINIA)

SURRY POWER STATION UNITS 1 AND 2

Westinghouse Non-Proprietary Class 3 WCAP-18243-N P October 2017 Revision 0 Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation

@ Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-18243-NP Revision 0 Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation Benjamin E. Mays*

Structural Design & Analysis III Louis W. Turicik*

Aging Management & License Renewal October 2017 Reviewers: D . Brett Lynch* Approved: Lynn A. Patterson*, Manager Anees Udyawar* Structural Design & Analysis III Alexandria M. Carolan*

Structural Design & Analysis III Arzu Alpan* Laurent P. Houssay*, Manager Radiation Engineering & Analysis Radiation Engineering & Analysis Amy E. Freed*

Aging Management & License Renewal

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

Westinghouse Non-Proprietary Class 3 II RECORD OF REVISION Revision 0: Original Issue WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS LIST OF TABLES ......... ....................... ............ ......... .. .. ... ........... ................. .. ........ ........ ............... .. ... ... ....... V LIST OF FIGURES ............ ................... .. ....... .................................... .. ....... ................... .. ............ ............. viii EXECUTIVE

SUMMARY

............... .. .. ......... ... .. .. .. ............... .. .. .. .................. ..... .. .. ...... ................. .. ............. X 1 INTRODUCTION .. ............. ........... ...... .. ....... ............. ........ ..... ................... .. ................................ 1-1 2 CALCULATED NEUTRON FLUENCE ........... .. ....................... ....................... ... .. ... .. .............. .. 2-1 3 MATERIAL PROPERTY INPUT .. .... .............. .......... ............ ... ... ...... ............................... .......... .. 3-l 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ............... .4-1 4.1 OVERALLAPPROACH ........................................ ............. .. ...... ............... ... ............. ..... 4-l 4.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ......... ........... ............ ..... .. ... .. .. .... ...... .. ........ .. ..... ... ... ................. ...... .. .. .. 4-l 4.3 PRESSURE CORRECTION ........ ... ..... ... .............................. .................................... ..... .4-5 4.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS ...... .. .... ............................... 4-5 4.5 CLOSURE HEADN ESSEL FLANGE REQUIREMENTS ..... ................. ........ .. .. ........ .4-5 4.6 BOLTUP TEMPERATURE REQUIREMENTS .... .. .. ..... .. ........... ..................... ... ... ..... ...4-5 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ......... .................. .. ............. 5-1 6 HEATUPAND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .......... .. ... ........ 6- 1 7 APPLICABILITY OF CURRENT HEATUP AND COOLDOWN LIMITS ....... ........... ............. 7-1 8 REFERENCES ............................................... .......... ....... ...... .... .. ......... ... ......... ............................ 8- l APPENDIXA THERMAL STRESS INTENSITY FACTORS (Ku) ............ ........... ...... ...... ...... A- 1 APPENDIXB REACTOR VESSEL INLET AND OUTLET NOZZLES .......... ... ......... .......... B-1 APPENDIX C OTHER RCPB FERRITIC COMPONENTS ........................... ...... .......... ......... C-1 APPENDIX D LTOP SYSTEM ENABLE TEMPERATURE .................. .. ............................... D-1 APPENDIXE WELD MATERIAL HEAT # 0227 INITIAL RTNoT AND UPPER-SHELF ENERGY DETERMINATION .................................... .. .. .. ................ ................ E- 1 APPENDIXF

SUMMARY

OF THE APPLICABILITY OF P-T LIMIT CURVES FOR SURRY UNITS 1 AND 2 .............. ................................... ....................................... .... .....F-1 WCAP- 18243-NP October 201 7 Revision 0

Westinghouse Non-Proprietary Class 3 IV APPENDIXG CREDIBILITY EVALUATION OF THE SURRY UNITS 1 AND 2 SURVEILLANCE DATA ............. .. ... ........................ .. ............................... .. ..... G-1 APPENDIXH COMPARISON OF AXIAL FLAW AND CIRCUMFERENTIAL FLAW P-T LIMIT CURVES .......................... .. .. .... .. ........................... .. ............................. .. H-1 APPENDIX I SURRY UNITS 1 AND 2 UPPER-SHELF ENERGY EVALUATION AT 68 EFPY .................................................... ........................................ ................... .. .. 1-l APPENDIXJ MATERIAL PROPERTY INPUT COMPARISON ...................................... ...... J-1 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 V LIST OF TABLES Table 2-1 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center for Surry Unit 1 ..... ... .. .. ................................................ .. ........................ ...............................2-2 Table 2-2 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Surveillance Capsule Center for Surry Unit 2 .......................................... ............ .. ............................................................. 2-3 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 ... .................... 2-4 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 ........... ..... ....... 2-5 Table 3-1 Best-Estimate Cu and Ni Weight Percent Values, Initial RTNDT Values, and Initial USE Values for the Surry Unit 1 RPV Beltline and Surveillance Materials ............................ 3-9 Table 3-2 Best-Estimate Cu and Ni Weight Percent Values, Initial RT NDT Values, and Initial USE Values for the Surry Unit l RPV Extended Beltline Materials ...................................... 3-10 Table 3-3 Best-Estimate Cu and Ni Weight Percent Values, Initial RTNDT Values, and Initial USE Values for the Surry Unit 2 RPV Beltline and Surveillance Materials ....... .. ............. .... 3-11 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 .. ................ .. .. .. .... .......... 3-12 Table 3-5 Initial RT NDT Values for the Surry Unit 1 Replacement Reactor Vessel Closure Head and Vessel Flange Materials ..... .................. ................................. ..... .. .. ... .. .. ......................... 3-13 Table 3-6 Initial RT NDT Values for the Surry Unit 2 Replacement Reactor Vessel Closure Head and Vessel Flange Materials ..................................... ... .. ................. .. ... .. .. ....... ........... ........... 3-13 Table 3-7 Surveillance Data for Weld Wire Heat# 299L44 ............................ .... ...................... .... 3-14 Table 3-8 Surveillance Data for Weld Wire Heat # 72445 ............... ....................................... .. ..... 3-15 Table 3-9 Calculation of Position 2.1 CF Values for Surry Unit l ................................................. 3-16 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 ......

...... ...................... ....................................................... .. ........ ........... .. ...... ....................... 3-l 7 Table 3-11 Calculation of Position 2.1 CF Values for Surry Unit 2 ............ ... ... ................. ......... .....3-18 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 .. ....

......... ......................... .......... ...................................................................... ..................... 3-19 Table 5-1 Fluence Values and Fluence Factors for the Vessel Surface, l/4T and 3/4T Locations for the Surry Unit 1 Reactor Vessel Materials at 68 EFPY .......... ............ ......... .................... 5-3 Table 5-2 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Surry Unit 2 Reactor Vessel Materials at 68 EFPY .................................. .................5-4 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 vi Table 5-3 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 1/4T Location ........................ 5-5 Table 5-4 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 3/4T Location ........................ 5-8 Table 5-5 Adjusted Reference Temperature Evaluation for the Surry Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 1/4T Location ................ ...... 5-11 Table 5-6 Adjusted Reference Temperature Evaluation for the Surry Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 3/4T Location ...................... 5-13 Table 5-7 Summary of the Limiting ART Values for Surry Units 1 and 2 at 68 EFPY ................. 5-15 Table 6-1 Surry Units 1 and 2 68 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ K,c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) .. ......... .............................. ...................... .................6-5 Table 6-2 Surry Units 1 and 2 68 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) .. ...................... ........... ..... .. .............................. . 6-8 Table 7-1 Current Surry Power Station P-T Limit Curve Data Points without Pressure Adjustment Plus 10% Margin for Heatup ................................................... .. .................................... ..7-4 Table 7-2 Current Surry Power Station P-T Limit Curve Data Points without Pressure Adjustment Plus 10% Margin for Cooldown ................................ .. ......... ........................................... 7-6 Table 7-3 Data Points for Surry Units 1 and 2 Heatup P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY .......

................................................................................ ......... ............... .. .. ........................... 7-11 Table 7-4 Data Points for Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY .......

..... .................................................................................................................................. 7-13 Table 7-5 Surry Units 1 and 2 Heatup P-T Limit Curve Margin Summary between the Current P-T Limit Curves + 10% Margin and the New P-T Limit Curves to 68 EFPY .................... 7-15 Table 7-6 Surry Units 1 and 2 Cooldown P-T Limit Curve Margin Summary between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY ................. 7-16 Table A-1 K, 1 Values for Surry Units 1 and 2 at 68 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors) ................................................. A-2 TableA-2 K, 1 Values for Surry Units 1 and 2 at 68 EFPY 100°F/hr Cooldown Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors) ................................................. A-3 Table B-1 Calculation of the Surry Unit 1 Nozzle Forging ART Values at 68 EFPY ..................... B-3 Table B-2 Calculation of the Surry Unit 2 Nozzle Forging ART Values at 68 EFPY ..................... B-4 Table B-3 Summary of the Limiting ART Values for the Surry Units 1 and 2 Inlet and Outlet Nozzle Forging Materials ............. ....................... ........................................................................ B-4 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 vii Table E-1 Weld Material Qualification Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat# 0227) ............... .. ... ... .... .. ..... ....... ................ . E-1 Table E-2 Supplemental Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat# 0227) ..... ..... .... ...................... ..... ....... ..... .......................... E-2 Table E-3 Charpy V-Notch Test Data for Surry Unit 2 Surveillance Weld (Heat# 0227) .... .......... E-3 Table F-1 Surry Units 1 and 2 P-T Limit Curve Applicability History ............. ......... .. .. .. .......... .. ... F-1 Table F-2 Data Points for Surry Units 1 and 2 Current Technical Specifications Heatup P-T Limit Curves .. .... ......... ..................... .... .. .. ... ..... .... ........ ..... ....... ... .... ....... .. .. ... ...... ...... .. .... .... ..... . F-5 Table F-3 Data Points for Surry Units l and 2 Current Technical Specifications Cooldown P-T Limit Curves ........ ....... ...... ................ ... .... .......... ......... ......... ............ ...................... ..... .... F-7 Table G-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation for Surry Unit 1

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

.... .. .... ...................... .. .... ................ ....... .......................... ......... .......... ... .......... .... ..... ....... . G-6 Table G-3 Mean Chemical Composition and Temperature for Weld Heat# 299L44 ....... ...... .... ..... G-7 Table G-4 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat # 299L44 ................ ... .. ....... .. .... .... ...... ....... .......... ... .. ... ....... ...... .. ... .. ... .. .. ... ... ... ....... G-8 Table G-5 Mean Chemical Composition and Temperature for Weld Heat # 72445 ........... .. ... .. ...... G-9 Table G-6 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat # 72445 ........ .. ........... .. ...... ....... .......... ... .. ... ...... ............ ... ... ........ ...... .... .... ............. G-1 O Table G-7 Surry Unit 1 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line ... G-11 Table G-8 Surry Unit 2 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line ... G-12 Table G-9 Calculation of Residual vs. Fast Fluence for Surry Units 1 and 2 ... ... ............... ....... .. . G-14 Table I-1 Predicted USE Values at 68 EFPY for Surry Unit 1 .... .......... .. ....... ..... .......... .. ................ I-4 Table I-2 Predicted USE Values at 68 EFPY for Surry Unit 2 ............... .. ... ...... ..... .. .... ........... ........ I-6 Table J-1 Comparison of Previous and Current Initial RT NDT Values for Surry Unit 1 ....... ....... ..... J-1 Table J-2 Comparison of Previous and Current Initial RTNDT Values for Surry Unit 2 ... ............ .... J-2 Table J-3 Comparison of Previous and Current cr1 Values for Surry Unit 1.. ... .. .................. ... .. ....... J-3 Table J-4 Comparison of Previous and Current cr1 Values for Surry Unit 2 .......................... .......... . J-4 Table J-5 Comparison of Previous and Current cr6 Values for Surry Unit 1 ................. .. ............ ... .. J-5 Table J-6 Comparison of Previous and Current cr6 Values for Surry Unit 2 ....... ......... .... .. .............. J-6 Table J-7 Comparison of Previous and Current Unirradiated USE Values for Surry Unit 1 ...... ..... J-7 Table J-8 Comparison of Previous and Current Unirradiated USE Values for Surry Unit 2 .... ....... J-8 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 Vlll LIST OF FIGURES Figure 3-1 RPV Base Metal Material Identifications for Surry Unit l.. .. ... .................. ......... ... .... ..... 3-5 Figure 3-2 RPV Weld Identifications for Surry Unit l ...................................... ................................ 3-6 Figure 3-3 RPV Base Metal Material Identifications for Surry Unit 2 ........... .. .................. ............... 3-7 Figure 3-4 RPV Weld Identifications for Surry Unit 2 ............................ .. ................ ........................ 3-8 Figure 6-1 Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 20, 40, and 60°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Axial Flaw" Methodology (w/ K1c) ............. ................... ...................... ................ .. ........ 6-3 Figure 6-2 Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Axial Flaw" Methodology (w/ Kic) ... ...... .. ... .. ................... ................ . 6-4 Figure 7-1 Surry Units 1 and 2 Heatup P-T Limit Curve Comparison between the Current P-T Limit Curves + l 0% Margin and the New P-T Limit Curves to 68 EFPY ....... ... ..... .. ........... ... .7-8 Figure 7-2 Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between the Current P-T Limit Curves + 10% Margin and the New P-T Limit Curves to 68 EFPY ...................... 7-9 Figure 7-3 Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between the Current P-T Limit Curves + 10% Margin and the New P-T Limit Curves to 68 EFPYMagnified ...7-10 Figure B-1 Comparison of Surry Unit 1 Beltline Cooldown P-T Limits (Including Current P-T Limits without Pressure Adjustment + 10% Margin and New 68 EFPY P-T Limits) to 68 EFPY Nozzle P-T Limits, Without Margins for Instrumentation Errors ........................... ........ B-7 Figure B-2 Comparison of Surry Unit 2 Beltline Cooldown P-T Limits (Including Current P-T Limits without Pressure Adjustment + 10% Margin and New 68 EFPY P-T Limits) to 68 EFPY Nozzle P-T Limits, Without Margins for Instrumentation Errors ................................... B-8 Figure F-1 Surry Units 1 and 2 Heatup P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications ......... ........ .................... ................... ................ ............. ... ... ...... F-3 Figure F-2 Surry Units 1 and 2 Cooldown P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications ................................................................................................. F-4 Figure H-1 Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 20, 40, and 60°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Circumferential Flaw" Methodology (w/ K1c) .......... ...................... .. .................... ... ...... H-2 Figure H-2 Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Circumferential Flaw" Methodology (w/ K1c) ............... ............. .. ..... H-3 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 ix Figure H-3 Surry Units 1 and 2 Heatup P-T Limit Curve Comparison between Limiting "Axial Flaw" Based Curves and "Circumferential Flaw" Based Curves ......................................... ..... H-4 Figure H-4 Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between Limiting "Axial Flaw" Based Curves and "Circumferential Flaw" Based Curves ................................... H-5 Figure I-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 68 EFPY ....... ........ .. ....................... I-8 Figure I-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 68 EFPY ................................... .... . I-9 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 X EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure-temperature (P-T) limit curves for normal operation of the Surry Units 1 and 2 reactor vessels. The heatup and cooldown P-T limit curves were generated using the limiting Adjusted Reference Temperature (ART) values for Surry Units 1 and 2. The limiting ART values which pertain to "axial flaw" materials were those of the Surry Unit 1 Lower Shell Longitudinal Weld L2 (Heat # 299L44, using Position 2.1) at both the 1/4 thickness (l/4T) and 3/4 thickness (3/4T) locations. The limiting ART values which pertain to "circumferential flaw" materials were those of the Surry Unit 1 Intermediate to Lower Shell Circumferential Weld (Heat# 72445, using Position 1.1 or Position 2.2) at the l /4T location and the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat# 0227, using Position 2.1) at the 3/4T location.

The P-T limit curves were generated for 68 effective full-power years (EFPY) using the K1c methodology detailed in the 1998 Edition through 2000 Addenda of the ASME Code,Section XI, Appendix G. The P-T limit curve generation methodology is consistent with the NRC-approved methodology documented in WCAP-14040-A, Revision 4. Heatup rates of 20, 40, and 60°F/hr, and cooldown rates ofO (steady-state),

20, 40, 60, and 100°F/hr were used to generate the P-T limit curves, with the flange requirements and without margins for instrumentation errors. The Surry Units 1 and 2 Subsequent License Renewal (SLR) period of operation, also known as the Subsequent Period of Extended Operation (SPEO), corresponding to 80 years of operation is 68 EFPY. The SLR P-T limit curves can be found in Figures 6-1 and 6-2. As concluded in Section 7, the new 68 EFPY P-T limit curves are bounded by the current Surry Power Station P-T limit curves. Thus, continued use of the current Surry Power Station P-T limit curves is justified through 68 EFPY.

Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 68 EFPY based on the Section 6 P-T limit curves.

Appendix B contains a P-T limit evaluation of the reactor vessel inlet and outlet nozzles based on a l /4T flaw postulated at the inside surface of the reactor vessel nozzle comer, where T is the thickness of the nozzle comer region. As discussed in Appendix B, the P-T limit curves generated based on the limiting cylindrical beltline materials bound the P-T limit curves for the reactor vessel inlet and outlet nozzles for Surry Units 1 and 2 at 68 EFPY.

Appendix C contains discussion of the other ferritic Reactor Coolant Pressure Boundary (RCPB) components relative to P-T limits. As discussed in Appendix C, all of the other ferritic RCPB components meet or are reconciled to the applicable requirements of Section III of the ASME Code.

Appendix D contains the determination of the Low Temperature Overpressure Protection (LTOP) system minimum enable temperature at 68 EFPY.

Appendix E contains an updated evaluation of weld Heat# 0227 initial material properties.

Appendix F contains a brief history of the Surry Units 1 and 2 P-T limit curves.

Appendix G contains an evaluation of the Surry Units 1 and 2 surveillance data credibility.

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 xi Appendix H contains a comparison of the "Axial Flaw" and "Circumferential Flaw" P-T limit curves.

Appendix I contains an evaluation of the Surry Units 1 and 2 Upper-Shelf Energy (USE) at 68 EFPY.

Appendix J contains a comparison of the material property input values used in this evaluation and those used in past evaluations as well as the Updated Final Safety Analysis Report (UFSAR).

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 1- 1 1 INTRODUCTION The purpose of this report is to present the calculations and the development of the Surry Units 1 and 2 heatup and cooldown P-T limit curves for 68 EFPY This report documents the calculated Adjusted Reference Temperature (ART) values, the development of the P-T limit curves for normal operation, and comparison of these new P-T limit curves to the current P-T limit curves in the Surry Power Station Technical Specifications [Ref. 1]. The goal of this report is to demonstrate that the current P-T limit curves in the Surry Power Station Technical Specifications are bounding and remain valid through 80 years of operation. Note that the term "current" is utilized herein regarding P-T limit curves only in reference to the Surry Power Station Technical Specifications [Ref. l] P-T limit curves.

Heatup and cooldown P-T limit curves are calculated using the adjusted RT NDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced

~T NDT, and adding a margin. The unirradiated RT NDT (RT DT(U)) is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F. In instances where insufficient data is available to determine RT NDT(U) using ASME Code methods, alternate estimation methods such as Branch Technical Position (BTP) 5-3 are applied.

RT NDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT NDT at any time period in the reactor's life, ~RTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RT NDT* The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steel. The U.S. Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. 2]. Regulatory Guide 1.99, Revision 2 is used for the calculation of ART values (RTNDT(U) + ~TNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where Tis the thickness of the vessel at the beltline region measured from the clad/base metal interface. The calculated ART values for 68 EFPY are documented in Section 5 of this report. The fluence projections used in calculation of the ART values are provided in Section 2 of this report.

The heatup and cooldown P-T limit curves documented in this report were generated using the most limiting ART values and the NRC-approved methodology documented in WCAP-14040-A, Revision 4

[Ref. 3]. Specifically, the "Axial Flaw" and "Circumferential Flaw" methodologies of the 1998 Edition through 2000 Addenda of ASME Code,Section XI, Appendix G [Ref. 4] were used, which make use of the Kic methodology. The K1c curve is a lower bound static fracture toughness curve obtained from test data gathered from several different heats of pressure vessel steel. The limiting material is indexed to the K1c curve so that allowable stress intensity factors can be obtained for the material as a function of temperature. Allowable operating limits are then determined using the allowable stress intensity factors.

The current P-T limit curves in the Surry Power Station Technical Specifications are based on the more conservative Kr, fracture toughness curve. The methodology utilizing the Kr, fracture toughness curve is equivalent to the Kra methodology, which is discussed further in this report.

The P-T limit curves presented herein were generated without instrumentation errors consistent with the Surry Power Station Technical Specification P-T limit curves. The reactor vessel flange requirements of WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 1-2 10 CFR 50, Appendix G [Ref. 5] have been incorporated in the P-T limit curves. The P-T limit curves generated in Section 6 were compared to the current Surry Units 1 and 2 P-T limit curves, contained in the Technical Specifications [Ref. 1], in Section 7 to determine if adequate margin exists to justify continued use of the Surry Units 1 and 2 current P-T limits through the Subsequent License Renewal (SLR) period of operation.

The P-T limit curves generated in Section 6 bound the P-T limit curves for the reactor vessel inlet and outlet nozzles generated in Appendix B for Surry Units 1 and 2 at 68 EFPY. Additionally, per Section 7, the current maximum allowable Low Temperature Overpressure Protection System (LTOPS) pressurizer Power Operated Relief Valve (PORV) setpoint Technical Specification value of :S 390.0 psig is bounding and will remain valid through the 80-year period of operation. Discussion of the other ferritic RCPB components relative to P-T limits is contained in Appendix C. Appendix D contains a calculation of the Low Temperature Overpressure Protection (LTOP) system enable temperature. Appendix E contains an evaluation of the initial material properties of weld Heat # 0227. Appendix F provides a summary of the Surry Units 1 and 2 P-T limit curves applicability. Appendix G provides a credibility evaluation of the Surry Units 1 and 2 surveillance data. Appendix H provides a comparison of the "Axial Flaw" and "Circumferential Flaw" P-T limit curves. Appendix I contains an evaluation of the Surry Units 1 and 2 Upper-Shelf Energy (USE) values at 68 EFPY. Appendix J contains a comparison of the material property input values used in this evaluation and those used in past evaluations as well as the Updated Final Safety Analysis Report (UFSAR).

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE For the initial 60-year End of License Extension (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. Reactor Pressure Vessel (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" [Ref. 6]. 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 [Ref. 7], materials exceeding a fast neutron fluence (E > 1.0 MeV) of 1.0 x 10 17 n/cm2 at the end of the SLR period are evaluated for changes in fracture toughness. This guidance is consistent with Regulatory Issue Summary (RIS) 2014-11 [Ref. 8].

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, l/4T flaw location in the inlet and outlet nozzle [Refs. 9 and 1OJ, 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 10 17 n/cm2 at SLR. The materials that exceed the 1.0 x 10 17 n/cm2 fast neutron fluence (E > 1.0 MeV) threshold, and were not evaluated in past analyses of record 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 1 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:

(f)(r,fJ, z) = rp(r,fJ) x rpi~;)

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 [Ref. 11] with the BUGLE-96 cross-section library [Ref. 12]. The BUGLE-96 library provides a coupled 47-neutron-,

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-2 20-garnma-ray-group cross-section data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a P 5 Legendre expansion and the angular discretization was modeled with an S 16 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 1 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, Wl, 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 Irradiation Irradiation Fluence from Core ID CycJe(s) Time (E > 1.0 MeV)

Cardinal (EFPY) (n/cm 2)

Axis (0 )

T 15 I 1.1 2.71E+18 w 35 1-4 3.4 3.68E+18 V 15 1-8 8.0 l.80E+19 25 1-12 X 16.1 2.11E+l9 15 13-14 WCAP-18243-NP October 2017 Revision 0

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

Cardinal (EFPY) (n/cm 2)

Axis ( 0 )

X 15 1 1.2 2.97£+18 w 25 1-4 3.8 6.36£+18 V 15 1-8 8.4 1.89£+19 s 45 1-13 15.0 1.07£+19 Wl 15 11-14 5.3 7.80£+18 y 25 1-12 20.3 2.72£+19 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 end of Cycle (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 10 17 n/cm2 at the nozzle forging to vessel shell weld and one inlet nozzle has fast neutron fluence (E > 1.0 MeV) greater than 1.0 x 10 17 n/cm2 at the l/4T nozzle flaw location at 68 EFPY for Surry Unit I . 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 10 17 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 10 17 n/cm 2 at the 1/4T 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 10 17 n/cm 2 through SLR for both Surry Units 1 and 2.

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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 Fast Neutron Fluence (n/cm

Material 32.5 EFPY 54 EFPY 68EFPY 72 EFPY l /4T Flaw in Outlet Nozzle Nozzle I l.53E+l6 2.69E+ l6 3.45E+ l 6 3.67E+l6 Nozzle 2 l.08E+ l6 l.93E+l6 2.49E+l6 2.65E+ l 6 Nozzle 3<*J 4.48E+l6 7.59E+ l6 9.62E+l6 l.02E+ l 7 l/4T Flaw in Inlet Nozzle Nozzle 1101 5.80E+l6 9.82E+ l6 1.24E+l7 l.32E+l7 Nozzle 2 l.40E+l6 2.50E+16 3.22E+l6 3.42E+ l6 Nozzle 3 l.98E+ l6 3.48E+ l6 4.46E+ l6 4.74E+l6 Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle I 3.62E+l6 6.3 5E+ l6 8. 13E+ l6 8.63E+ l6 Nozzle2 2.55E+l6 4.55E+l6 5.86E+l6 6.23E+ l6 Nozzle 3\<I l.06E+ l7 l.79E+ l 7 2.27E+l7 2.40E+ l7 Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1101 l.42E+ l 7 2.40E+ l 7 3.04E+ l7 3.22E+l7 Nozzle 2 3.43E+l6 6.IOE+ l6 7.84E+l6 8.34E+l6 Nozzle 3w 4.85E+l6 8.5 1E+ l 6 l.09E+ l 7 l.16E+ l7 Nozzle Shell 3.64E+ l 8 6.00E+ l8 7.54E+ l 8 7.98E+l 8 Nozzle Shell to Intermediate Shell Circumferential Weld 3.64E+ l 8 6.00E+ l8 7.54E+18 7.98E+ l8 Intermedi ate Shell Plate I 3.17E+ l9 5.06E+ l9 6.29E+l9 6.65E+ l9 Plate 2 3. 17E+l9 5.06E+ l9 6.29E+ l9 6.65E+l9 Intermediate Shell Longitudinal Welds Weld 1 5.75E+l8 9.85E+ l8 l .25E+ l 9 l.33E+l9 Weld 2 5.75E+l8 9.85E+ l 8 l .25E+ l 9 1.33E+l9 Intermediate Shell to Lower Shell Circumferential Weld 3. 18E+ l 9 5.08E+ l 9 6.3 1E+ l9 6.67E+ l9 Lower Shell Plate 1 3.20E+l9 5. IIE+ l 9 6.35E+ l9 6.70E+l9 Plate 2 3.20E+l9 5.I IE+ l 9 6.35E+ l 9 6.70E+ l9 Lower Shell Longitudinal Welds Weld I 5.80E+ l 8 9.94E+l8 l.26E+ l9 1.34E+ l 9 Weld2 5.80E+ l 8 9.94E+ l 8 l .26E+ l9 l.34E+ l9 Lower Shell to Lower Vessel Head Circumferential Weld

< 1E+ l 7 < IE+ l7 < IE+ l7 < 1E+ l7 Notes:

(a) l /4T flaw in Outlet Nozzle 3 is projected to reach 1.0 x 10 17 n/cm2 at approximately 70.7 EFPY.

(b) l/4T flaw in Inlet Nozzle I is projected to reach 1.0 x 10 17 n/cm 2 at approximately 55.0 EFPY.

(c) Outlet Nozzle 3 forging to vessel shell we ld reached 1.0 x 10 17 n/cm 2 at approximately 30.8 EFPY.

(d) Inlet Nozzle I fo rging to vessel shell weld reached 1.0 x 10 17 n/cm2 at approximately 23.2 EFPY.

(e) Inlet Nozzle 3 forging to vessel shell weld is projected to reach 1.0 x 10 17 n/cm2 at approx imately 62.8 EFPY.

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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 Neutron Fluence [n/cm']

Material 31.3 EFPY 54 EFPY 68EFPY 72EFPY 1/4T Flaw in Outlet Nozzle Nozzle I 1.49E+ l6 2.66E+ l6 3.38E+l6 3.58E+l6 Nozzle 2 l.09E+ l6 1.95E+ l6 2.48E+l6 2.63E+ l6 Nozzle 31*i 4.29E+ l6 8.28E+ l6 l.07E+ l7 l.15E+ 17 J/4T Flaw in Inlet Nozzle Nozzle l(bJ 5.55E+16 l .07E+ l7 l.39E+17 l.48E+l7 Nozzle 2 l.41E+16 2.52E+ l 6 3.2 1E+16 3.40E+ l 6 Nozzle 3 1.93E+16 3.44E+ l6 4.37E+16 4.63E+l6 Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle l 3.52E+l6 6.27E+ l6 7.96E+ l6 8.45E+16 Nozzle 2 2.57E+l6 4.60E+l6 5.85E+ l6 6.20E+16 Nozzle 3ici 1.0IE+ l7 l .95E+l7 2.53E+ 17 2.70E+17 Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Nozzle 1<d>

l.36E+l7 2.62E+l7 3.40E+l7 3.62E+ l7 Nozzle 2 3.45E+l6 6.17E+ l6 7.84E+ l6 8.32E+ l6 Nozzle 3l<J 4.73E+ l6 8.41E+l6 l .07E+ l7 l.1 3E+l7 Nozzle Shell 3.52E+l8 6.70E+l8 8.65E+l8 9.21E+ l 8 Nozzle Shell to Intermediate Shell Circumferential Weld 3.52E+ l8 6.70E+ l 8 8.65E+ l 8 9.21E+ l 8 Intermediate Shell Plate l 3.IOE+l9 5.64E+l9 7.20E+ l9 7.65E+l9 Plate 2 3. 10E+l9 5.64E+l9 7.20E+ l9 7.65E+ l9 Intermediate Shell Longitudinal Welds Weld I 5.98E+l8 1.03E+ l9 1.29E+l9 l.36E+ l9 Weld2 5.98E+l8 l.03E+ l9 1.29E+ l9 l.36E+ l9 Intermediate Shell to Lower Shell Circumferential Weld 3. 1 IE+ l9 5.66E+ l9 7.22E+ l9 7.67E+ l9 Lower Shell Plate l 3.12E+ 19 5.68E+ l9 7.26E+l9 7.71E+ l9 Plate 2 3.12E+l9 5.68E+ l9 7.26E+19 7.71E+ l 9 Lower Shell Longitudinal Welds Weld I 6.03E+ I8 1.03E+ l9 l .30E+l9 1.37E+ l9 Weld2 6.03E+l8 1.03E+ l9 l .30E+19 1.37E+ l9 Lower Shell to Lower Vessel Head Circumferential Weld

< IE+ J7 < IE+17 < IE+ l 7 <1E+ l7 Notes:

(a) l /4T flaw in Outlet Nozzle 3 is projected to reach 1.0 x 10 17 n/cm 2 at approximately 63.8 EFPY.

(b) 1/4T flaw in Inlet Nozzle I is projected to reach 1.0 x 10 17 n/cm2 at approximately 50.9 EFPY.

(c) Outlet Nozzle 3 forging to vesse l shell weld reached 1.0 x I 0 17 n/cm 2 at approx imately 31.0 EFPY.

(d) Inlet Nozzle l forgi ng to vessel shell weld reached 1.0 x 10 17 n/cm 2 at approximately 23.5 EFPY.

(e) Inlet Nozzle 3 forging to vessel shell weld is proj ected to reach 1.0 x 10 17 n/cm2 at approximately 63 .9 EFPY.

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-1 3 MATERIAL PROPERTY INPUT The requirements for P-T limit curve development are specified in 10 CFR 50, Appendix G [Ref. 5). The beltline region of the reactor vessel is defined as the following in 10 CFR 50, Appendix G:

"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 exp erience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage. "

Per RlS 2014-11 [Ref. 8] materials which are predicted to experience neutron fluence greater than 1 x 10 17 n/cm 2 (E > 1 MeV) at the end of the licensed operating period must also be evaluated for neutron embrittlement effects. Materials which have not previously been considered in the beltline region, but are predicted to experience neutron fluence greater than 1 x 10 17 n/cm2 are termed "extended beltline" materials.

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 1 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 (Ll) 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 C4339-1 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 #

WCAP-18243-NP October 201 7 Revision 0

Westinghouse Non-Proprietary Class 3 3-2 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% ofIS Longitudinal Weld L4, LS Longitudinal Weld Ll , and 63% of LS Longitudinal Weld L2 were fabricated from weld wire Heat # 8Tl 762, Linde 80 Flux Type, Flux Lot Number 8597. The remaining 37% of LS Longitudinal Weld L2 was fabricated from weld wire Heat# 8Tl 762, Linde 80 Flux Type, Flux Lot Number 8632. Surveillance data does not exist for Heat #

8Tl762.

Based on the results of Section 2 of this report, the materials that exceeded the 1 x 10 17 n/cm 2 (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 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 10 17 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.

Thus, 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 8Tl 762, 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# 8Tl554B, Linde 80 Flux Type, Flux Lot Number 8579. The Surry Unit 2 Inlet 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 1 [Ref. 13). 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 1 and 2 extended beltline. The data supporting this materials summary was gathered primarily from PWROG-16045-NP, Revision O [Ref. 14).

The identification of the RPV 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. The material property inputs used for the subsequent P-T limits 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 [Ref. 15) 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.

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-3 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 0.

Fracture Toughness Properties The fracture toughness properties (initial RT NDT and initial Upper-Shelf Energy [USE]) of most of beltline plate materials were originally determined using NUREG-0800, BTP 5-3 Position 1.1 [Ref. 16]

methodology, with three exceptions. Surry Unit 1 IS Plate C4326-l, Surry Unit 1 LS Plate C4415-l , and Surry Unit 2 LS Plate C4339-l were determined using the ASME Code,Section III [Ref. 17]. Many of the beltline and extended beltline fracture toughness properties were updated per ASME Section III, the General Electric (GE) Method [Ref. 18], and NUREG-0800, BTP 5-3 Position 1.1 methodologies, as described in PWROG-16045-NP, Revision O [Ref. 14]. The initial RT NDT 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 DT =To+ 35°F). The initial RT NDT 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) values. When using these "Master Curve" -generated initial RT NDT values, the CF and margin terms will be adjusted to a minimum of 167°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 RT NDT and initial USE values are documented in PWROG-16045-NP, Revision O 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 1 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 E 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. A comparison of the material property input values utilized herein and those utilized previously is documented in Appendix J.

The initial RT NDT 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 [Ref. 5] 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.

The closure heads at both Surry Units 1 and 2 have been replaced, and the initial RT NDT values of the Surry Units 1 and 2 flange materials were updated in PWROG-16045-NP, Revision O [Ref. 14]. The Surry Unit 1 replacement closure head has an initial RT NUT value of -67°F, determined per ASME Code Section III, NB-2300. The Surry Unit 1 reactor vessel flange has an initial RTNUT of -114.6°F, calculated using the GE methodology. The Surry Unit 2 replacement head has an initial RTNUT value of -60°F, determined per ASME Code Section III, NB-2300. The Surry Unit 2 reactor vessel flange has an initial RTNuT of

-156.3°F, calculated using the GE methodology. See Tables 3-5 and 3-6 for a summary of the initial RTNDT values for these two components at each plant.

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Westinghouse Non-Proprietary Class 3 3-4 Chemistry Factor Values The chemistry factor (CF) values were calculated using Positions 1.1 and 2.1 of Regulatory Guide 1.99, Revision 2 [Ref. 2). 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 G. The calculated capsule fluence values are provided in Tables 2-1 and 2-2 and are used to determine the Position 2.1 CFs as shown in Tables 3-9 and 3-11 for Surry Units 1 and 2, respectively.

Tables 3-10 and 3-12 summarize the Positions 1.1 and 2.1 CF values determined for the Surry Units 1 and 2 RPV beltline and extended beltline materials, respectively.

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-5 vent pipe (HK-23) -~~~~ contTol rod mechanism*

,_..,,___housing (HK-A2 thru A14) shroud support

~cloaure head cap closure* head flange inlet nozzle outlet no~zle/ 3x0iK-S8) 3x(HK-S9) - upper shell Vessel *suppo~t_/

pad inter~:d1te

.... I

. l.

course

lower sh 1

~

core sup4rt (MK-70) ,

instrumentation no~zles (HK-891 thru 8113)

Figure 3-1 RPV Base Metal Material Identifications for Surry Unit 1

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

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-6 V£RTJCAl SC~

t10*

C..326-1

-W06 9.0" 1ao*

t(>R.£

-l.

~

1<<*

90*

CL 9.7"

-6#

C

-wos i10*

_,I 48.3" Figure 3-2 RPV Weld Identifications for Surry Unit 1 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-7 SPARE CROM HEAD AOAP1R CLOSURE HEAD LIFTING LUG CLOSURE Sl\JO(Ml<-62).

NUT(MK-63),SPHE'.RICAL REACTOR VESSEL WASHERS(MK-6~ ond MK-65 O.OSURE HEAD THERMAL SLEEVE F'U 'NEL OUTER 0-RINC CASKET MK- 72 INNER 0-RlNG GASKET MK-71 VESSEL SUPPORT PAO I. VESSEL SUPPORT PAO INTERMEOIA TE ISHELL COURSE I

I LOWER SHELL COURSE CORE SUPPORT GUIDE (MK-70)

LOWER HEAD RING '

INSTRUMENlA TION NOZZLES BOTIOM HEAD CAP MK-991 thru 8113 Figure 3-3 RPV Base Metal Material Identifications for Surry Unit 2

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

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-8 VERTICAL SCAMS.

210*

-W06 C-4331-2 9.0" 1eo*

COR!

-,J.,,..

-i 144* ...., go*

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C4339-1

,so*

I 48.3" Figure 3-4 RPV Weld Identifications for Surry Unit 2 WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-9 Table 3-1 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 Wt.% Wt.% RTNDT(U) a, Initial USE RPV Material (OF)

Cu Ni (OF) (ft-lb)

Reactor Vessel Beltline MaterialiaJ Upper Shell Forging 122Vl09VA1 0.11 0.74 40 0 114 Intermediate Shell Plate C4326-I 0.11 0.55 10 0 11 5 Intermediate Shell Plate C43 26-2 0.11 0.55 11.4 0 94 Lower Shell Plate C44 I 5- 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 ~64(b) 0.33 0.10 0 20.0 (Heat # 2501 7)

Intermediate Shell Longitudinal Welds 64(b) 0.16 0.57 -48.6 18.0 L3 and L4 (Heat # 8Tl 554)

Intermediate to Lower Shell Circwnferential Weld 64(b) 0.22 0.54 -72.5 12.0 (Heat # 72445)

Lower Shell Longitudinal Weld LI 64(b) 0.16 0.57 -48.6 18.0 (Heat # 8Tl554)

Lower Shell Longitudinal Weld L2 64(b) 0.34 0.68 -74.3 12.8 (Heat # 299L44)

Reactor Vessel Surveillance Materiali c)

Lower Shell Plate C44 l 5- l 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 [Ref. 14], unless otherwise noted.

(b) Per Surry Power Stati on UFSAR (Ref. 19], reactor vessel Equi valent Margins Analys is (EMA) report BAW-2494, Revision I [Ref. 20) has been approved for these welds. The EMA is updated for SLR under Pressurized Water Reactor Owners Group (PWROG) PA-MSC-148 1. Linde 80 initial USE values are set to the generic value of64 ft-lbs per BAW-23 13, Revision 7, Supplement I, Revision I [Ref. 13]. Only limited Charpy test information is available for Heat # 25017.

Based on the average Charpy energy value of the weld quali fication tests completed at I 0°F, the USE fo r Heat # 250 17 is at least 64 ft-lbs.

(c) The surveillance plate data was taken to be the same as the vesse l plate data. The surveillance weld data was obtained from BAW-2324, Revision O [Ref. 2 1].

WCAP-1 8243-NP October 201 7 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 a, Initial Wt. % Wt.% RTNDT(U)

RPV Material USE Cu Ni {°F) (OF)

(ft-lb)

Reactor Vessel Extended Beltline Materiali0 J Inlet Nozzle l (Heat # 9-4787) 0.159 0.85 10.3 0 63 Inlet Nozzle 2 (Heat # 9-5078) 0.159 0.87 11.6 0 64 Inlet Nozzle 3 (Heat # 9-48 19) 0.159 0.84 -47.2 0 68 Outlet Nozzle l (Heat # 9-4825- 1) 0.159 0.85 -44.9 0 68 Outlet Nozzle 2 (Heat # 9-4762) 0.159 0.83 -87.5 0 82 Outlet Nozzle 3 (Heat # 9-4788) 0. 159 0.84 -50.2 0 71 Inlet Nozzle to Upper Shell Heat# 299L44 0.34 0.68 -7.0 20.6 64 Welds Heat # 8Tl 762 0.19 0.57 -4.9 19.7 64 Outlet Nozzle to Upper Shell Heat # 8Tl 762 0.19 0.57 -4.9 19.7 64 Welds Heat# 8T l 554B 0.16 0.57 -4.9 19.7 64 Note:

(a) All values were taken from Table 8 of PWROG-16045-NP, Revision O [Ref. 14].

WCAP-18243-NP 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 Wt.% Wt.% RTNDT(U) a, Initial USE RPV Material (OF)

Cu Ni (OF) (ft-lb)

Reactor Vessel Beltline Materiafi0>

Upper Shell Forging l 23V303V Al 0.11 0.72 30 0 104 Intermediate Shell Plate C433 l-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 ~68(b) 0.35 0.10 0 20.0 (Heat# 4275)

Intermediate Shell Longitudinal Welds 64(b) 0.22 0.54 -72 .5 12.0 L3 and L4 (OD 50%) (Heat# 72445)

Intermediate Shell Longitudinal Weld 64(b) 0.19 0.57 -48.6 18.0 L4 (ID 50%) (Heat# 8Tl 762)

Intermediate to Lower Shell Circumferential Weld o<c) o<c) 8ic) 0.187 0.545 (Heat # 0227)

Lower Shell Longitudinal Welds Ll and L2 64(b) 0.19 0.57 -48.6 18.0 (Heat# 8T 1762)

Reactor Vessel Surveillance MaterialidJ 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 [Ref. 14], unl ess otherwise noted.

(b) Per Surry Power Station UFSAR [Ref. 19], reactor vessel EMA report BA W-2494, Revision 1 [Ref. 20] has been approved for these welds. The EMA is 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 1, Revision l [Ref. 13]. 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 E. Since the initial RTNDT is based on measured data, o 1 is equal to 0°F. Per Surry Power Station UFSAR [Ref. 19], reactor vessel EMA report BA W-2494, Revision 1 [Ref. 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 , Revision O[Ref. 22].

WCAP-18243-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 NOT Values, and Initial USE Values for the Surry Unit 2 RPV Extended Beltline Materials a, Initial Wt.% Wt.% RTNDT(U)

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 71 (b)

Rotterdam 0.35 1.0 30 0 Welds Notes:

(a) All values were taken from Table 9 of PWROG-16045-NP, Revision O [Ref. 14).

(b) Per PWROG-16045-NP, Revision O [Ref. 14], this initial USE value is set equal to the USE value of the first tested capsule from WCAP-16001 [Ref. 22). This methodology utilizes BTP 5-3 [Ref. 16], Position 1.2 guidance, as no USE data is available from the supplier.

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-13 Table 3-5 Initial RT NDT Values for the Surry Unit 1 Replacement Reactor Vessel Closure Head and Vessel Flange Materials 0

RPV Material Initial RT DT ( F)

Replacement Closure Head -67'*)

E4381/E4382 Vessel Flange FV-1870 -144.6(b)

Notes:

(a) Value taken from Table 8 of PWROG-16045-NP, Revision O [Ref. 14]. This value is based on ASME Code Section III, NB-2300 criteria. Note that the original Closure Head Flange initial RTNoTwas 10°F per WCAP-14177 [Ref. 23] .

(b) Value taken from Table 8 of PWROG-16045-NP, Revision O [Ref. 14]. This value is based on the GE Methodology. Note that the Vessel Flange Initial RT NDT used in previous reactor vessel integrity calculations was I 0°F as documented in WCAP-14177 [Ref. 23].

Table 3-6 Initial RT NDT Values for the Surry Unit 2 Replacement Reactor Vessel Closure Head and Vessel Flange Materials RPV Material Initial RT NDT ( 0 F)

Replacement Closure Head -60(*)

02Wl-1-l-l Vessel Flange FV-2542 -156.3(b)

Notes:

(a) Value taken from Table 9 of PWROG-16045-NP, Revision O [Ref. 14]. This value is based on ASME Code Section III, NB-2300 criteria. Note that the original Closure Head Flange initial RTNoT was l0°F per WCAP-14177 [Ref. 23].

(b) Value taken from Table 9 of PWROG-16045-NP, Revision O [Ref. 14]. This value is based on the GE Methodology. Note that the Vessel Flange Initial RT NDT used in previous reactor vessel integrity calcul ations was -65°F as documented in WCAP-14177 [Ref. 23].

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-14 Table 3-7 Surveillance Data for Weld Wire Heat # 299L44 Capsule Fluence L\R.TNDT Irradiation Cu Ni CF Capsule Designation<a) Temperature wt.% wt.% (OF) (x 10 19 n/cm2, {°F) {°F)

E > 1.0 MeV)

TMI2-LG1 (CR-3ibl 0.37 0.70 234.0 0.830 216 556 0

Wl(CR-3i > 0.37 0.70 234.0 0.780 262 545 TMI1-E 0.33 0.67 215.2 0.107 74 556 TMil-C 0.33 0.67 215 .2 0.8 82 166 556 TM12-LG1(TMI-2ib> 0.33 0.67 215.2 0.968 226 556 CR3-LGl(ONS-3) 0.36 0.70 230.5 0.779 202 556 A5(ct> 0.23 0.64 175.8 2.75 246.6 556 Suny 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 Suny Unit I : Capsule X 0.23 0.64 175.8 2. 11 234 542 Notes:

(a) Data was obtained from ANP-2650 [Ref. 24), unless otherwise noted. Material source is indicated in parentheses. CR-3 = Crystal River Unit 3, TMI I = Three Mile Island Unit I, ONS = Oconee Nuclear Station Unit 3.

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

(c) Capsule WI was irradiated in Surry Unit 2. The fluence value is updated from ANP-2650 [Ref. 24) per 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.

(d) Data taken from AREVA-17-01417 [Ref. 25).

WCAP-1 8243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-15 Table 3-8 Surveillance Data for Weld Wire Heat # 72445 Capsule Fluence i\R.TNOT Irradiation Cu Ni CF Capsule Designation<*> Temperature wt.% wt.% (OF) (x 10 19 n/cm2, {°F) (OF)

E > 1.0 MeV)

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 (AN0-1t> 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 I: Capsule S 0.23 0.62 172.4 0.829 165 542 Point Beach Unit I: 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 [Ref. 24], unless otherwise noted. Material source is indicated in parentheses.

ANO-I = Arkansas Nuclear One Unit 1 (b) Capsule Wl was irradiated in Surry Unit 2. The fluence value is updated from ANP-2650 [Ref. 24] per Section 2.

The irradiation temperature value is the time-weighted average Tcold considering the cycles th at WI was inside the Surry Unit 2 reactor vessel.

WCAP-18243-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 Fluence FF<c> .dRTNDT (d) 2 RPV Material Capsule .dRTNDT .dRTNDT FF (x 10 19 n/cm2, (OF)

(OF) (OF)

E > 1.0 MeV)

T 0.271 0.644 50 50 32.21 0.415 Lower Shell V 1.80 1.161 113 11 3 131.23 1.349 Plate C44 l 5- l (b) X 2.1 1 1.203 86 86 103.46 1.447 (Longitudinal)

SUM: 266.91 3.2 11 CF r44 1s - 1 = :E(FF

L'i.RTNn1 l + :E(FF2) = (266 .91) + (3.2 1 l ) = 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 TMII-E 0.107 0.431 74 91 39.02 0.185 (Heat# 299L44) TMll-C 0.882 0.965 166 185 178.87 0.931 TMI2-LG1 0.968 0.99 1 226 247 244.95 0.982 CR3-LG I 0.779 0.930 202 216 200.87 0.865 AS 2.75 1.270 246.6 326 413.61 1.612 SUM: 2385.98 9.550 CF '-'*a* H , ao , "" = :E(FF

L'i.RTNDr) + :E(FF2) = (2385.9 8) + (9.550) = 249.8°F CR3-LG I 0.510 0.8 12 139 153 124.24 0.659 CR3-LG2 1.67 1.141 164 178 203 .15 1.303 Surveillance WI 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-I: S 0.829 0.947 165 165 156.32 0.898 PB-1 : R 2.19 1.2 13 155 155 187.48 1.47 1 PB-I : T 2.23 1.2 17 181 172 209.86 1.482 SUM: 1105 .56 7.438 CF Heat # 72445 = l:(FF

L'i.RT NOT) 7 2 l:(FF ) = (1105 .56) 7 (7.438) = 148.6°F Notes:

(a) Fluence and LI.RTNDT data taken from Tables 3-7 and 3-8, unless otherwise noted.

(b) Surry Unit I Lower Shell Plate C44 15- l capsule fluence values obtained from Section 2. Ll.RTNDT values obtained from BAW-2324, R evision O [Ref. 21).

(c) FF = fl uence factor = t<0*28 -0* 1o*iog(t)J.

(d) The surveillance weld LI.RT NDT 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 vesse l we ld chemistry. Pre-adjusted values are listed in the LI.RTNDT co lumn. Temperature adjustment =

I.O*(Tcapsu1e- Tp1an,), where Tplan, = 542°F for Surry Unit I and Tcapsute is the irrad iation temperature in Table 3-7 or 3-8. The temperature adjustment procedure is not uti lized 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 I, a ratio of 1.00 is utilized; this approach is conservative.

WCAP-18243-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 122V109VA1 76.1 ---

Intermediate Shell Plate C4326- l 73.5 ---

Intermediate Shell Plate C4326-2 73.5 ---

Lower Shell Plate C4415- l 66.6 83.1 <*)

Lower Shell Plate C4415-2 73.0 83.1 (a)

Upper to Intermediate Shell Circumferential Weld (Heat# 25017) 152.0 ---

Intermediate Shell Longitudinal Welds 143.9(c)

L3 and L4 (Heat# 8Tl554)

Intermediate to Lower Shell Circumferential 158.0(c) 148.6(c)

Weld (Heat # 72445)

Lower Shell Longitudinal Weld LI 143.9(c) ---

(Heat# 8Tl554)

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

(Heat# 299L44)

Reactor Vessel Extended Beltline Materia/s<b)

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-l shares a heat number with Lower Shell Plate C44 J5-2, 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 SE and Revision 2-A SE [Refs. 26 and 27]. Chemistry Factors must be adjusted to a minimum of 167°F when used in ART calculations. If the Position I .I CF is greater than l 67°F, it is used in calculations.

WCAP-18243-NP October 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-18 Table 3-11 Calculation of Position 2.1 CF Values for Surry Unit z<a>

Capsule Adjusted

{d) FF* Adjusted RPV Fluence FF<c> ARTNDT ARTNDT Capsule ARTNDT FF 2 Material (x 10 19 n/cm2, (OF) (OF)

(OF)

E > 1.0 MeV)

Lower Shell X 0.297 0.668 59.08 59.08 39.45 0.446 Plate C4339- l V 1.89 1.174 79.12 79.12 92.91 1.379 (Longitudinal) y 2.72 1.267 114.22 114.22 144.72 1.605 X 0.297 0.668 48.67 48.67 32.50 0.446 Lower Shell V 1.89 1.174 63 .60 63 .60 74.68 1.379 Plate C4339- l y 2.72 1.267 106.8 1 106.8 1 135.33 1.605 (Transverse) SUM: 519.59 6.860 CF r4no_, = L(FF

t.RTNn -) + L(FF 2) = (5 19.59) + (6.860 = 75.7°F X 0.297 0.668 95.65 95.65 63.86 0.446 Surveillance V 1.89 1.174 140.21 140.21 164.64 1.379 Weld Material y 2.72 1.267 178.32 178.32 225.94 1.605 (Heat # 0227)

SUM: 454.45 3.430 CF Heat # 0227 = L(FF

t.RT'-'"T) + L(FF 2) = (454.45) + (3.430) = 132.5°F CR3-LG1 0.510 0.812 139 152 123.43 0.659 CR3-LG2 1.67 1.141 164 177 202.01 1.303 Surveillance Wl 0.780 0.930 138 140 130.24 0.865 Weld MaterialCb) PB-1: V 0.634 0.872 107 106 92.46 0.761 (Heat# 72445) PB-1: S 0.829 0.947 165 164 155.37 0.898 PB-1: R 2.19 1.2 13 155 154 186.26 1.471 PB-1: T 2.23 1.2 17 181 171 208.64 1.482 SUM: 1098.42 7.438 CF Hear # mA< = L(FF

t.RTNnT) + L(Ff 2) = (1098.42) + (7.438) = 147,7°F Notes:

(a) Fluence and t.RT NOT data are from WCAP- 1600 1, Revis ion O [Ref. 22], unless otherwise noted.

(b) Fluence and t.RTNoT data are from Table 3-8.

(c) FF= fluence factor= t<0 *2s-O.Wlog(())_

(d) The surveillance weld t.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 t.RT NOT column. Temperature adjustment =

l.O*(Tcapsule - 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 utili zed when plant-specific capsules are analyzed alone. The ratio procedure is applicab le only to survei llance welds and the ratio app lied = CF vessel Weld/ CFsurv. Weld* If the ratio procedure yields a ratio less than I, a ratio of 1.00 is utilized; this approach is conservative.

WCAP-18243-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 Beltline Materials Upper Shell Forging 123V303VA1 75 .8 ---

Intermediate Shell Plate C4331-2 83.0 ---

Intermediate Shell Plate C4339-2 73.4 75.ia)

Lower Shell Plate C4208-2 107.3 ---

Lower Shell Plate C4339-l 70.8 75 .ia)

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 l 52.4(c) ---

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

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

Lower Shell Longitudinal Welds L1 and L2 152.4(c) ---

(Heat# 8Tl 762)

Reactor Vessel Extended B eltline Materials

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 I (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 Shell Heat# 8Tl 762 152.4 ---

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 ---

Note:

(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 RTNDT values were established using master curve data. Per BAW-2308 Revi sion 1-A SE and Revision 2-A SE [Refs. 26 and 27) Chemistry Factors must be adjusted to a minimum of 167°F when used in ART calculations. If the Position 1.1 CF is greater than 167°F, it will be used in calculations.

WCAP-18243-NP October 20 17 Revision 0

Westinghouse Non-Proprietary Class 3 4-1 4 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 4.1 OVERALL APPROACH The ASME (American Society of Mechanical Engineers) approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. K 1c is obtained from the reference fracture toughness curve, defined in the 1998 Edition through 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 4]. The K,c curve is given by the following equation:

K le =33 .2+20 .734*e[0.02(T-RT,vor )] (1)

where, K1c (ksi-Vin.) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNOT This K1c curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.
4.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
(2)
where, K 1m stress intensity factor caused by membrane (pressure) stress K1t stress intensity factor caused by the thermal gradients K,c reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNoT C 2.0 for Level A and Level B service limits C 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 4-2 For membrane tension, the corresponding K 1 for the postulated defect is:
Kim = Mm X (pR;/ t) (3) where, Mm for an inside axial surface flaw is given by:
Mm l.85 for Ji < 2, Mm 0.926 Ji for 2 :::; Ji:::; 3.464, Mm 3.21 for Ji > 3.464 and, Mm for an outside axial surface flaw is given by:
Mm l.77 for Ji < 2, Mm 0.893 Ji for 2 :::; .Ji:::; 3.464, Mm 3.09 for Ji > 3.464 Similarly, Mm for an inside or an outside circumferential surface flaw is given by :
Mm 0.89 for Ji < 2, Mm 0.443 Ji for 2 :::; Ji:::; 3.464 ,
Mm 1.53 for Ji > 3.464
where, p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).
For bending stress, the corresponding K 1 for the postulated axial or circumferential defect is:
K 1b = Mb* Maximum Bending Stress, where Mb is two-thirds of Mm (4)
The maximum K 1 produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:
K11 = 0.953 X 10-3 X CR X t2*5 (5) where CR is the cooldown rate in °f/hr., or for a postulated axial or circumferential outside surface defect K1t = 0.753 X 10-3 X HUX t2 5 (6) where HU is the heatup rate in °f/hr.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 4-3 The through-wall temperature difference associated with the maximum thermal K 1 can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-2 for the maximum thermal K 1*
(a) The maximum thermal K 1 relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).
(b) Alternatively, the K 1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a ~ -thickness axial or circumferential inside surface defect using the relationship:
JG,= (1.0359Co + 0.6322C1 + 0.4753C2 + 0.3855C3) * ~ (7) or similarly, K, 1 during heatup for a ~ -thickness outside axial or circumferential surface defect using the relationship:
K1t = (l.043C o + 0.630C1 + 0.481C 2 + 0.40lC3) * ~ (8) where the coefficients C0 , C 1, C2 , and C 3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form :
o-(x) = Co+ C,(x I a)+ C2(x I a) 2 + CJ(x I a)3 (9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).
Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 3] Section 2.6 (equations 2.6.2-4 and 2.6.3-1). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 (equation 2.6.1-1). This equation is solved utilizing values for thermal diffusivity of 0.518 fr/hr at 70°F and 0.379 fr/hr at 550°F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-fr-°F.
At any time during the heatup or cooldown transient, K1c is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code,Section XI, Paragraph G-2120), the appropriate value for RTNoT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated, and then the corresponding (thermal) stress intensity factors, Kit, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained, and from these the allowable pressures are calculated.
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Westinghouse Non-Proprietary Class 3 4-4 For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference l /4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the l/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the i1T (temperature) across the vessel wall developed during cooldown results in a higher value of K1c at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1c exceeds K 11 , the calculated allowable pressure during cooldown will be greater than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the l /4T location, and therefore allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a l /4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1c for the inside l /4T flaw during heatup is lower than the K1c for the flaw during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K 1c values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the l/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The third portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a I/4T flaw located at the l/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 4-5 comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.
4.3 PRESSURE CORRECTION The current Surry Units l and 2 heatup and cooldown limit curves in the Surry Power Station Technical Specifications [Ref. 1] include a pressure correction value of 21.5 psi. This pressure correction later was applied to the curves developed in WCAP-14177 [Ref. 23] to account for the pressure difference between the location of pressure measurement and the reactor vessel. See Appendix F for details. This pressure correction has not been incorporated into the heatup and cooldown limit curves developed in Section 6 of this report. The pressure correction value has been removed from the current Technical Specification heatup and cooldown limit curves in Section 7 to appropriately compare the current Technical Specification P-T limit curves to the P-T limit curves developed in this report.
4.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS Surry Units 1 and 2 are Westinghouse-designed plants; thus, the primary Reactor Coolant System (RCS) piping is stainless steel. Therefore, the lowest service temperature requirements of Paragraph NB-2332 of ASME Code Section Ill [Ref. 17] do not apply to the Surry Units l and 2 reactor vessels. See Appendix C for additional details.
4.5 CLOSURE HEADNESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Ref. 5] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure bead regions must exceed the material unirradiated RT NDT by at least l 20°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is calculated to be 621 psig. The initial RTNDT values of the reactor vessel closure bead and vessel flange are documented in Tables 3-5 and 3-6. The limiting unirradiated RT NDT of -60°F is associated with the Surry Unit 2 replacement reactor vessel closure head, so the minimum allowable temperature of this region is 60°F at pressures greater than 621 psig (without margins for instrument uncertainties). This limit is shown in Tables 6-1 and 6-2.
4.6 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded. It is determined by the highest reference temperature, RTN DT, in the closure flange region. This requirement is established in Appendix G to 10 CFR 50 [Ref. 5]. Per the NRC-approved methodology in WCAP-14040-A, Revision 4 [Ref. 3], the minimum boltup temperature is 60°F or the limiting unirradiated RTNDT of the closure flange region, whichever is higher. Since the limiting unirradiated RT DT of this region is below 60°F per Tables 3-5 and 3-6, the recommended minimum boltup temperature for the Surry Units 1 and 2 reactor vessel is 60°F (without margins for instrument uncertainties). It is noted that the boltup temperature is procedurally controlled at Surry Units 1 and 2 independent from the Technical Specification curves.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-1 5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2 [Ref. 2], the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = Initial RT NDT + ~RTNDT + Margin (10)
Initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section Ill of the ASME Boiler and Pressure Vessel Code [Ref. 17]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided there are sufficient test results to establish a mean and standard deviation for the class.
MTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
MTNDT =CF* f (0.28-0.IOlogf) (11)
To calculate ~RTNDT at any depth (e.g., at l/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth:
£(depth x) -- f su-'acc
"'
e (-O.i 4x) (12) where x inches (reactor vessel cylindrical shell beltline thickness is 8.05 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 11 to calculate the MTNDT at the specific depth.
The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 3].
Tables 5-1 and 5-2 contain the surface fluence values at 68 EFPY, which were used for the development of the P-T limit curves contained in this report. Tables 5-1 and 5-2 also contain the l/4T and 3/4T calculated fluence values and fluence factors (FFs), per Regulatory Guide 1.99, Revision 2. The values in this table will be used to calculate the 68 EFPY ART values for the Surry Units 1 and 2 reactor vessel materials.
Margin is calculated as M = 2 ~ CT i +CT~ . The standard deviation for the initial RT NDT margin term ( cr1) is 0°F when the initial RT NDT is a measured value and l 7°F when a generic value is available, unless a material-specific cr 1 is calculated. The standard deviation for the ~RT NDT margin term, cr6 , is l 7°F for plates or forgings when surveillance data is not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data is used. For welds, cr 6 is equal to 28°F when surveillance capsule data is not used or is non-credible, and is 14°F (half the value) when credible surveillance capsule data is used. The value for cr6 need not exceed 0.5 times the mean value of MT NDT*
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-2 However, for the welds utilizing "Master Curve"-based initial RT NOT values, 0 6 is set equal to 28°F per the safety evaluations (SEs) associated with BAW-2308, Revision 1 and 2 [Refs. 26 and 27).
Contained in Tables 5-3 through 5-6 are the 68 EFPY ART calculations at the l /4T and 3/4T locations for generation of the Surry Units 1 and 2 heatup and cooldown curves.
Surry Unit 1 Inlet Nozzle 1 and Surry Unit 2 Inlet Nozzle 1 and Outlet Nozzle 3 have projected fluence values that exceed the 1 x 10 17 n/cm2 fluence threshold at the l /4T flaw location at 68 EFPY per Tables 2-3 and 2-4. Therefore, per NRC RIS 2014-11 [Ref. 8], neutron radiation embrittlement must be considered herein for these nozzle forging materials. For conservatism, embrittlement is considered for each nozzle forging material. The nozzle forging ART values are calculated using surface fluence values at the 1/4T flaw location for each specific nozzle. Thus, ART calculations for the Surry Units 1 and 2 inlet and outlet nozzle forging materials utilizing the 1/4T and 3/4T fluence values are excluded from Tables 5-3 through 5-6. ART values for the nozzle forging materials are contained in Appendix B.
Finally, the second conclusion ofTLR-RES/DE/CIB-2013-01 [Ref. 28] states that if Ll.RTNDT is calculated to be less than 25°F, then embrittlement need not be considered. This conclusion was applied, as necessary, to the ART calculations documented in Tables 5-3 through 5-6.
The limiting ART values for Surry Units 1 and 2 to be used in the generation of the P-T limit curves are based on multiple materials, since a combination of axial and circumferential flaw materials have the most limiting 1/4T and 3/4T ART values. The limiting ART values for Suny Units 1 and 2 are summarized in Table 5-7. The limiting ART values are less than the ART values utilized to develop the current Surry Units 1 and 2 Technical Specification curves [Ref. 1). Thus, the applicability of the P-T limit curves in the Surry Units 1 and 2 Technical Specifications (based on WCAP-14177 [Ref. 23)) can be extended to 68 EFPY.
Section 7 provides further justification that the applicability of the current Surry Units 1 and 2 Technical Specifications P-T limit curves can be extended to 68 EFPY by directly comparing the 68 EFPY P-T limit curves developed in Section 6 to the 48 EFPY curves in the Surry Units 1 and 2 Technical Specifications.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-3 Table 5-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Surry Unit 1 Reactor Vessel Materials at 68 EFPY Surface Fluence, 1/4T f 3/4T f r<*) (xl0 19 19 1/4T 3/4T Reactor Vessel Material (x 10 n/cm2, (x 10 19 n/cm2, n/cm2, FF FF E > 1.0 MeV) E > 1.0 MeV)
E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 122V I 09V Al 0.754 0.465 0.787 0.177 0.541 Upper to Intermediate Shell Circ.
0.754 0.465 0.787 0.177 0.541 Weld (Heat # 25017)
Intermediate Shell Plates C4326-l 6.29 3.88 1.350 1.48 1.108 and C4326-2 Lower Shell Plates C44 l 5- l and 6.35 3.92 1.352 1.49 1.111 C4415-2 Intermediate Shell Longitudinal Welds L3 and L4 1.25 0.771 0.927 0.294 0.665 (Heat # 8Tl554)
Lower Shell Longitudinal Welds LI (Heat# 8Tl554) and L2 (Heat # 1.26 0.777 0.929 0.296 0.667 299L44)
Intermediate to Lower Shell Circumferential Weld 6.3 1 3.89 1.350 1.48 1.109
<Heat# 72445)
Reactor Vessel Extended Beltline Material/bJ Inlet Nozzle I to Upper Shell Weld 0.0304 0.01 88 0.165 0.00714 0.087 (Heats # 299L44 and# 8T 1762)
Inlet Nozzle 2 to Upper Shell Weld 0.00784 0.00484 0.065 0.00184 0.031
<Heats # 299L44 and # 8T 17 62)
Inlet Nozzle 3 to Upper Shell Weld 0.0109 0.00672 0.083 0.00256 0.040 (Heats# 299L44 and# 8T I 762)
Outlet Nozzle I to Upper Shell Weld 0.00813 0.00502 0.067 0.00191 0.032 (Heats## 8Tl 762 and# 8T 1554B)
Outlet Nozzle 2 to Upper Shell Weld 0.00586 0.00362 0.052 0.00138 0.024 (Heats # 8Tl 762 and# 8Tl554B)
Outlet Nozzle 3 to Upper Shell Weld 0.0227 0.0140 0.137 0.00533 0.070 (Heats# 8T 1762 and# 8T1554B)
Notes:
(a) 68 EFPY fluence values are documented in Table 2-3.
(b) Reactor vessel nozzle forgings are excluded from this table - see Appendix B.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-4 Table 5-2 Fluence Values and Fluence Factors for the Vessel Surface, l/4T and 3/4T Locations for the Surry Unit 2 Reactor Vessel Materials at 68 EFPY Surface Fluence, l/4T f 3/4T f f (a) (x 10 19 l/4T 3/4T Reactor Vessel Material (x 10 19 n/cm2, (x 10 19 n/cm2, n/cm2, FF FF E > 1.0 MeV) E > 1.0 MeV)
E > 1.0 MeV)
Reactor Vessel Be/t/ine Materials Upper Shell Forging 123V303VA1 0.865 0.534 0.825 0.203 0.573 Upper to Intermediate Shell Circ.
0.865 0.534 0.825 0.203 0.573 Weld (Heat# 4275)
Intermediate Shell Plates 7.20 4.44 1.378 1.69 1.145 C4331-2 and C4339-2 Lower Shell Plates C4208-2 and 7.26 4.48 1.380 1.70 1.147 C4339-l Intermediate Shell Longitudinal Welds L3 and L4 1.29 0.796 0.936 0.303 0.673 (Heats# 72445 and 8Tl 762)
Lower Shell Longitudinal Welds 1.30 0.802 0.938 0.305 0.675 LI and L2 (Heat # 8T 1762)
Intermediate to Lower Shell 7.22 4.45 1.379 1.70 1.145 Circumferential Weld (Heat# 0227)
Reactor Vessel Extended Beltline Material/bJ Inlet Nozzle l to Upper Shell Weld 0.0340 0.0210 0.177 0.00798 0.094 (Heat# 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld 0.00784 0.00484 0.065 0.00184 0.031 (Heat# 8T 1762)
Inlet Nozzle 3 to Upper Shell Weld 0.0107 0.00660 0.082 0.00251 0.039 (Heat# 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld 0.00796 0.00491 0.066 0.00187 0.031 (Rotterdam)
Outlet Nozzle 2 to Upper Shell Weld 0.00585 0.00361 0.052 0.00137 0.024 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld 0.0253 0.0156 0.147 0.00594 0.076 (Rotterdam)
Notes:
(a) 68 EFPY fluence values are documented in Table 2-4.
(b) Reactor vessel nozzle forgings are excluded from this table - see Appendix B.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-5 Table 5-3 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the l/4T Location R.G. l/4T l/4T 1.99, Wt. % Wt. % CF<*> FluenceCb> l/4T RT NDT(U) <c> ART DT (d )
(J.cc> <Jt,. (e) Margin RPV Material ART Rev. 2 cu<*> Ni<*> (OF) (x 10 19 n/cm2, FF (OF) (OF) (OF) (OF) (OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel B eltline Materials Upper Shell Forging 122V I 09V A I I.I 0.11 0.74 76. 1 0.465 0.787 40 59.9 0.0 17.0 34.0 133.9 Upper to Intennediate Shell 1.1 0.33 0. 10 152.0 0.465 0.787 0 119.6 20.0 28.0 68.8 188.4 Circumferential Weld (Heat# 250 17)
Intennediate Shell Plate C4326- l 1.1 0.11 0.55 73. 5 3.88 1.350 10 99.2 0. 0 17.0 34.0 143.2 Intermediate Shell Plate C4326-2 I .I 0.11 0.55 73 .5 3.88 1.350 11.4 99.2 0.0 17.0 34.0 144.6 Intermediate Shell Longitudinal Welds I. I 0.16 0.57 167.0 0.771 0.927 -48.6 154.8 18.0 28.0 66.6 172.8 L3 and L4 (Heat # 8T l 554)
Intennediate to Lower Shell I. I 0.22 0.54 167.0 3.89 1.350 -72.5 225.5 12.0 28.0 60.9 213.9
_.. ~~r-~tJ~f~~~~!!~~ '!!. ~~~ .(~!!-~t- ff. ?J.~1?)... -- -- ------- - - - - - - - - - - - - -- ------ --- ---- ---- ------- ------- ----- -------- ---- ----------- ----- ---- -- ---- -- --- --- - --------- ----------- ---- -- -----
Using credible surveillance data 2.1 --- --- 167.0 3.89 1.350 -72.5 225.5 12.0 28.0 60.9 213.9 Lower Shell Plate C44 l 5- l 1.1 0. 102 0.493 66.6 3.92 1.352 20 90.0 0.0 17.0 34.0 144.0 Using credible surveillance data 2.1 --- --- 83. 1 3.92 1.352 20 11 2.3 0.0 8.5 17.0 149.3 Lower Shell Plate C44 J5-2 I.I 0.11 0.50 73 .0 3.92 1.352 4.6 98.7 0.0 17.0 34.0 137.3 Using credible surveillance data 2.1 --- --- 83.1 3.92 1.352 4.6 112.3 0.0 8.5 17.0 133.9 Lower Shell Longitudinal Weld LI I. I 0.16 0.57 167.0 0.777 0.929 -48.6 155.2 18.0 28.0 66.6 173.2 (Heat# 8T l 554)
Lower Shell Longitudinal Weld L2 I. I 0.34 0.68 220.6 0.777 0.929 -74.3 205 .0 12.8 28.0 61.6 192. 3
.............. {t!~~! -~ -~~?-~~).......... .. -. ------ ---- -- -- --------- ------ --- -- ---- ---- --- ----- ----- ------ ------- - -- ------- -- ---- ------ ------ --- ----- ---- --------- ----------- ----- ------
Using credible surveillance data 2.1 --- --- 249.8 0.777 0.929 -74.3 232.1 12.8 28.0 61 .6 219.4 WCAP- 18243-NP October 20 17 Revision 0
Westinghouse Non-Proprietary Class 3 5-6 Table 5-3 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 1/4T Location R.G. 1/4T (c) (d) 1/4T 1.99, Wt.% Wt.% CF<*l Fluence<bl 1/4T RTNDT(U) LiRTNDT 0'1 (c) O'&(e) Margin RPV Material ART Rev. 2 cu<*) Ni<*l (OF) (x 10 19 n/cm2, FF<bl (OF) (OF) (OF) (OF) (OF)
(OF)
Position E > 1.0 MeV)
R eactor Vessel Extended B eltline Materials Inlet Nozzle I to Upper Shell Weld I. I 0.34 0.68 220.6 0.0188 0.165 -7.0 36.5 20.6 18.2 55.0 84.5
-- ----- ---- -- _{!-!~~!-~ -~~?_1:-:l~)_---- ------- -- ------------ ----------- ----------- - - - - - - - - - - - - ------- --------------- --------------- ------ --- --------- ---------- - -- ------ ---
Using credible surveillance data 2.1 --- --- 249.8 0.0188 0.165 -7.0 41 .3 20.6 14.0 49.8 84.1 Inlet Nozzle 2 to Upper Shell Weld I. I 0.34 0.68 220.6 0.00484 0.065 -7.0 0.0 (1 4.4) 20.6 0.0 41.2 34.2

--- -----_{~~~!-~ -~~?_1:-:l~)_ -- --- ---- --- - -------- ---- --------- -- ----------- ---- -- ------- -- ---- ----- --- ------- -------------- - --------- ---- ----- --- ----- --- ---- --- ----

Using credible surveillance data 2.1 --- --- 249.8 0.00484 0.065 -7.0 0.0 (16.3) 20.6 0.0 41.2 34.2 Inlet Nozzle 3 to Upper Shell Weld l.l 0.34 0.68 220.6 0.00672 0.083 -7.0 0.0 (18 .3) 20.6 0.0 41.2 34.2
-- -- ------- -- _@~~! H-~~?~~)_ -- --------- -- ------------ ----------- --- -- --- --- -------- -- --------- -------------- - ------ ----- ---- --- ------ --------- ------ -- -- - ----- ------
Using credible surveillance data 2.1 --- --- 249.8 0.00672 0.083 -7.0 0.0 (20.8) 20.6 0.0 41 .2 34.2 Inlet Nozzle I to Upper Shell Weld l.l 0.19 0.57 152.4 0.0188 0. 165 -4.9 25.2 19.7 12.6 46.8 67.1 (Heat# 8T 1762)
Inlet Nozzle 2 to Upper Shell Weld l.l 0. 19 0.57 152.4 0.00484 0.065 -4.9 0.0 (10.0) 19.7 0.0 39.4 34.5 (Heat# 8T 1762)
Inlet Nozzle 3 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00672 0.083 -4.9 0.0 (12.7) 19.7 0.0 39.4 34.5 (Heat# 8Tl 762)
Outlet Nozzle l to Upper Shell Weld I. I 0.19 0.57 152.4 0.00502 0.067 -4.9 0.0 (10.2) 19.7 0.0 39.4 34.5 (Heat# 8TI 762)
Outlet Nozzle 2 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00362 0.052 -4.9 0.0 (8 .0) 19.7 0.0 39.4 34.5 (Heat# 8Tl 762)
Outlet Nozzle 3 to Upper Shell Weld l.l 0.19 0.57 152.4 0.0140 0. 137 -4.9 0.0 (20.9) 19.7 0.0 39.4 34.5 (Heat# 8Tl 762)
Outlet Nozzle I to Upper Shell Weld I. I 0.16 0.57 143.9 0.00502 0.067 -4.9 0.0 (9 .7) 19.7 0.0 39.4 34.5 (Heat# 8Tl 5548 )
WCAP-1 8243 -NP October 2017 Revision 0
Westinghouse Non-Proprietary C lass 3 5-7 Table 5-3 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 1/4T Location R.G. 1/4T 1/4T 1.99, Wt. % Wt. % CF<*> Fluence<hl 1/4T RT NDT(U) <c> aRTNoT (d)
<I/cl (J/1,. (e) Margin RPV Material ART Rev.2 cu<*) Ni<*> (OF) (x 10 19 n/cm2, FF(bl (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 0.00362 0.052 -4 .9 0.0 (7 .6) 19.7 0 .0 39.4 34.5 (Heat# 8Tl554B)
Outlet Nozzle 3 to Upper Shell Weld I. I 0.16 0 .57 143.9 0.0140 0. 137 -4 .9 0.0 (19.7) 19.7 0 .0 39.4 34.5 (Heat# 8Tl554B )
Notes :
(a) Chemical composition data taken from Tables 3-1 and 3-2 of this repo rt. Chemistry factor values taken from Table 3- 10 of this report.
(b) The I/4T fl uence and l/4T FF values were taken from Table 5-l.
(c) Initial RT NDT values and cr 1 values are from Tables 3-1 and 3-2 of this report.
(d) Calculated t.RTNor values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-20 13-01 [Ref. 28]; actual calculated t.RTNDT values are listed in parentheses for these materials.
(a) As summarized in Appendix G of this report, all surveillance data for Surry Unit l were deemed credible. Per the guidance of Regulatory Guide l.99, Revision 2 [Ref. 2], the base metal cr 6 = l 7°F fo r 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, cr 6 need not exceed O.S*t.RTNDT* For welds utilizing initial RT NDT values based on BA W-2308, cr6 = 28°F per References 26 and 27.
WCAP- 18243-NP October 20 17 Revision 0
Westinghouse Non-Proprietary Class 3 5-8 Table 5-4 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 3/4T Location R.G. 3/4T o,<c) 3/4T 1.99, Wt. % Wt.% CF<*> Fluence 3/4T RT NDT(U)<c> .1.RTNDT (d) Gt;.(e) Margin RPV Material ART Rev. 2 cu<*) Ni<*> (OF) (x 10 19 n/cm2, FF (OF) (OF) (OF) (OF) (OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel B eltline Materials Upper Shell Forging 122VJ09VAI I .I 0.11 0.74 76.J 0.177 0.541 40 41.1 0.0 17.0 34.0 115 .1 Upper to Intermediate Shell I. I 0.33 0.10 152.0 0.177 0.541 0 82.2 20.0 28.0 68.8 151.0 Circumferential Weld (Heat # 25017)
Intermediate Shell Plate C4326- J 1.1 0.11 0.55 73.5 1.48 1.108 JO 81.4 0.0 17.0 34.0 125.4 Intermediate Shell Plate C4326-2 1.1 0.11 0.55 73 .5 1.48 1.108 11.4 81.4 0.0 17.0 34.0 126.8 Intermediate Shell Longitudinal Welds I. I 0.16 0.57 167 .0 0.294 0.665 -48.6 111.0 18.0 28.0 66.6 129.0 L3 and L4 (Heat # 8Tl554)
Intermediate to Lower Shell I.I 0.22 0.54 167.0 1.48 1.109 -72.5 185.2 12.0 28. 0 60.9 173.6
___ 9r_<:l!~f<?~<?t:1~!~~Y!_ <?~~ _(l-:l~_8:t_ ~-??_~1?)___ --- --------- --------- -- ------ ----- -------- - - - - - - - - - - - - - - - - - - - ------- ----- - - - - - - - - - - - - - -------------- - -- - - -- - -- ----- -- -- ------ --- ---- ------ -
Using credible surveillance data 2.1 --- --- 167.0 1.48 1.109 -72. 5 185.2 12.0 28. 0 60.9 173.6 Lower Shell Plate C4415-l I.I 0.102 0.493 66.6 1.49 1.111 20 74.0 0.0 17.0 34.0 128.0 Using credible surveillance data 2.1 --- --- 83.1 1.49 1.111 20 92.3 0.0 8. 5 17.0 129.3 Lower Shell Plate C44 l 5-2 1.1 0.11 0.50 73.0 1.49 1.111 4.6 81.1 0.0 17.0 34.0 119.7 Using credible surveillance data 2.1 --- --- 83.1 1.49 1.111 4.6 92.3 0.0 8. 5 17. 0 113.9 Lower Shell Longitudinal Weld LI I. I 0.16 0.57 167.0 0.296 0.667 -48 .6 111.3 18. 0 28. 0 66.6 129.3 (Heat # 8Tl 554)
Lower Shell Longitudinal Weld L2 I. I 0.34 0.68 220.6 0.296 0.667 -74.3 147.1 12.8 28.0 61.6 134.3

_{!'!<?~!-~ -~~?_I:~:l)_ ------------- ------- --- -- --- -- ------ ---- --- ---- ------ -- - - --- - --------- - - - - ------------ - - - - - - - - - ---- -- -- ---------- - - ------ --------- ---- ------- ---- --- -- --

Using credible surveillance data 2.1 --- --- 249.8 0.296 0.667 -74.3 166. 5 12.8 28. 0 61 .6 153.8 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-9 Table 5-4 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 3/4T Location R.G. 3/4T 3/4T 1.99, Wt.% Wt.% CF<*> Fluence 3/4T RTNDT(U)
(c)
ARTNDT (d) at> (J I!,. (e) Margin ART RPV Material Rev.2 cu<*> Ni<*> (OF) (x 10 19 n/cm2, FF (OF) (OF) (OF) (OF) (OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Extended Beltline Materials Inlet Nozzle 1 to Upper Shell Weld 34.2 I. I 0.34 0.68 220.6 0.00714 0.087 -7.0 0.0 ( 19.1) 20.6 0.0 41.2

rn~~! -~-~??.~~)_------------- -------- -- -- ----------- ----------- - -- - -- - -- - - - - - - - - - - ---- ------ ----- ---- -- ---- ------- ------- ---- -- --- ----------- ---- ------ -

Using credible surveillance data 2.1 --- --- 249.8 0.00714 0.087 -7.0 0.0 (2 1.7) 20.6 0.0 41 .2 34.2 Inlet Nozzle 2 to Upper Shell Weld 34.2 1.1 0.34 0.68 220.6 0.00184 0.031 -7.0 0.0 (6.8) 20.6 0.0 41.2
-- -- ------- -- _{~~~! -~ -~??.~~)_ ---- -- ------ - - --- - - - - - - - - --------- -- ------- -- -- - - - - -- -- ---- - - -- -- - - --- - - - - - - - - - - - - - - - - - - - - ------- ------- --- ------ --- ----- --- ----- ------
Using credible surveillance data 2. 1 --- --- 249.8 0.00184 0.031 -7.0 0.0 (7.7) 20.6 0.0 41 .2 34.2 Inlet Nozzle 3 to Upper Shell Weld 1.1 0.34 0.68 220.6 0.00256 0.040 -7.0 0.0 (8.8) 20.6 0.0 41.2 34.2

---___ {~~~!-~-~??.~~)_ -- ---------- - ------------ --- -------- ---- ---- --- - - - - - - - - - - - - - - - - - - - -------- -- - - - - - - - ------- - ----------- --- --- --- --- --------- -- ------ --- --

Using credible surveillance data 2.1 --- --- 249.8 0.00256 0.040 -7.0 0.0 (10.0) 20.6 0.0 41 .2 34.2 Inlet Nozzle I to Upper Shell Weld I. I 0.19 0.57 152.4 0.00714 0.087 -4.9 0.0 (13.2) 19.7 0.0 39.4 34.5 (Heat# 8T 1762)
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 34.5 (Heat# 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00256 0.040 -4.9 0.0 (6.1) 19.7 0.0 39.4 34.5 (Heat # 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00191 0.032 -4.9 0.0 (4 .8) 19.7 0.0 39.4 34.5 (Heat# 8Tl 762)
Outlet Nozzle 2 to Upper Shell Weld I. I 0.19 0.57 152.4 0.00138 0.024 -4.9 0.0 (3.7) 19.7 0.0 39.4 34.5 (Heat # 8T 1762)
Outlet Nozzle 3 to Upper Shell Weld I.I 0.19 0.57 152.4 0.00533 0.070 -4.9 0.0 (10.7) 19.7 0.0 39.4 34.5 (Heat # 8Tl 762)
Outlet Nozzle 1 to Upper Shell Weld I. I 0.16 0.57 143 .9 0.00191 0.032 -4.9 0.0 (4.5) 19.7 0.0 39.4 34.5 (Heat# 8Tl554B)
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-10 Table 5-4 Adjusted Reference Temperature Evaluation for the Surry Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 3/4T Location R.G. 3/4T 3/4T 1.99, Wt.% Wt.% CF<*> Fluence<bl 3/4T RTNDT(U)
(c)
~TNDT (d) a/c> 0',1 (e) Margin RPV Material cu<*) FF(bl ART Rev. 2 Ni<*> (OF) (x 10 19 n/cm2, (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 0.00138 0.024 -4.9 0.0 (3 .5) 19.7 0.0 39.4 34.5 (Heat# 8TI554B)
Outlet Nozzle 3 to Upper Shell Weld I. I 0.16 0.57 143.9 0.00533 0.070 -4.9 0.0 (IO.I) 19.7 0.0 39.4 34.5 (Heat# 8TI 554B)
Notes:
(a) Chemical composition data taken from Tables 3-1 and 3-2 of this report. Chemi stry factor values taken from Table 3-10 of th is report.
(b) The 3/4T fluence and 3/4T FF values were taken from Table 5-1.
(c) Initial RT NOT values and o 1 values are from Tables 3-1 and 3-2 of this report.
(d) Calculated 6-RTNOT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-20 13-0 1 [Ref. 28] ; actual calculated 6-RTNOT values are listed in parentheses for these materials.
(e) As summarized in Appendix G of this report, all surveillance data for Surry Unit I were deemed credible. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. 2], the base metal o,.. = J7°F fo r Position 1.1, and o,.. = 8.5 °F fo r Position 2.1 with credible surveillance data. Also per Regulatory Gui de 1.99, Revision 2, the weld metal o,.. = 28°F for Position 1.1, and with credible survei ll ance data o,.. = 14°F for Position 2.1. However, o,.. need not exceed 0.5*6.RT NOT* For welds utilizing initial RT NOT values based on BA W-2308, o,.. = 28°F per References 26 and 27.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5- 11 Table 5-5 Adjusted Reference Temperature Evaluation for the Surry Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 1/4T Location R.G. 1/4T RT NDT(U) (cl 1/4T 1.99, Wt. % Wt. % CF<*> Fluence(bl 1/4T ARTNDT(d) at> G t,. (e) Margin RPV Material ART Rev.2 cu<*> Ni<*> (OF) (x 10 19 n/cm2, FF<bl (OF) (OF) (OF) (OF) (OF)
(OF)
Position E > 1.0 MeV)
R eactor Vessel B eltline Materials Upper Shell Forging I23V303VA1 I. l 0.11 0.72 75 .8 0.534 0.825 30 62.5 0.0 17.0 34.0 126. 5 Upper to Intermed iate Shell l.l 0.35 0.10 160.5 0.534 0.825 0 132.3 20.0 28.0 68.8 20 1.2 Circumferential Weld (Heat # 4275) lntennediate Shell Plate C43 3 l-2 I. I 0.12 0.60 83 .0 4.44 1.378 15.0 114.4 0.0 17.0 34.0 163 .4 Intermediate Shell Plate C4339-2 l. l 0.1 1 0.54 73.4 4.44 1.378 7.8 101.2 0.0 17.0 34.0 143 .0 Using non-credible surveillance data 2.1 --- --- 75.7 4.44 1.378 7.8 104.3 0.0 17. 0 34.0 146.1 Intermediate Shell Longitudinal Welds
l. l 0.22 0.54 167.0 0.796 0.936 -72.5 156.3 12.0 28.0 60.9 144.7 L3 and L4 (OD 50%) (Heat # 72445)
Using credible surveillance data 2.1 --- --- 167.0 0.796 0.936 -72.5 156.3 12.0 28.0 60.9 144.7 Intermediate Shell Longitudina l Weld L4 1.1 0.19 0.57 167.0 0.796 0.936 -48 .6 156.3 18.0 28.0 66.6 174.3 (ID 50%) (Heat # 8T I 762)
Intermediate to Lower Shell l.l 0. 187 0. 545 147. 5 4.45 1.379 0 203.4 0.0 28.0 56.0 259.4 Circumferential Weld (Heat # 0227)
Using credible surveillance data 2.1 --- --- 132.5 4.45 1.379 0 182.7 0.0 14.0 28.0 210. 7 Lower Shell Plate C4208-2 l.l 0.15 0.55 107.3 4.48 1.380 -30 148.1 0.0 17.0 34.0 152 .l Lower Shell Plate C4339-I l.l 0.107 0.53 70.8 4.48 1.380 -4.4 97.7 0.0 17.0 34.0 127.3 Using non-credible surveillance data 2.1 --- -- - 75.7 4.48 1.380 -4.4 104.5 0.0 17.0 34.0 134.1 WCAP-1 8243 -NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-1 2 Table 5-5 Adjusted Reference Temperature Evaluation for the Surry Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the l/4T Location R.G. l/4T (c) l/4T 1.99, Wt.% Wt.% CF<*> Fluence<bl l/4T RTNDT(U) ~TNoT (d) 0'1 (c) 0',1 (e) Margin RPV Material cu<*) FF(bl ART Rev. 2 Ni<*> (OF) (x 10 19 n/cm2, (OF) (OF) (OF) (OF) (OF)
(OF)
Position E > 1.0 MeV)
Lower Shell Longitudinal Welds LI and I. I 0.19 0.57 167.0 0.802 0.938 -48.6 156.7 18.0 28.0 66.6 174.6 L2 (Heat# 8Tl 762)
Reactor Vessel Extended Beltline Materials Inlet Nozzle 1 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 69.9 (Heat# 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld I.I 0.19 0.57 152.4 0.00484 0.065 -4.9 0.0 (10.0) 19.7 0.0 39.4 34.5 (Heat# 8Tl 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 34.5 (Heat# 8Tl 762)
Outlet Nozzle 1 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 30.0 (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 30.0 (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 40.0 110.0 (Rotterdam)
Notes:
(a) Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemistry Factor values taken from Table 3-1 2 of this report.
(b) The I/4T tluence and I/4T FF values were taken from Table 5-2.
(c) Initial RTNDT values and cr 1 values are from Tables 3-3 and 3-4 of this report.
(d) Calculated D.RTNDT values less than 25°F have been reduced to zero per TLR-RES/DE/CTB-2013-0 I [Ref. 28]; actual calculated D.RT NDT values are listed in parentheses for these materials.
(e) Per Appendix G 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 [Ref. 2], the base metal cr.,, = 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 cr.,, = 28°F for Position 1.1 , and with credible surveillance data cr.,, = I 4°F for Position 2.1. However, cr.,, need not exceed 0.5* D.RTNDT* For welds utilizing initial RTNDT values based on BA W-2308, cr.,, = 28°F per References 26 and 27.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5- 13 Table 5-6 Adjusted Reference Temperature Evaluation for the Surry Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 3/4T Location R.G. 3/4T (c) (d) o} c) 3/4T 1.99, Wt.% Wt. % CF<*> Fluence(bl 3/4T RTNDT(U) ARTNDT G t,. (e) Margin RPV Material cu<*) FF(bl ART Rev. 2 Ni<*> (OF) (x 10 19 n/cm2, (OF) (OF) (OF) (OF) (OF)
(OF)
Position E > 1.0 MeV)
Reactor Vessel Beltline Materials Upper Shell Forging 123V303VA1 I.I 0.1 1 0.72 75.8 0.203 0.573 30 43.4 0.0 17.0 34.0 107.4 Upper to Intermediate She ll 1.1 0.35 0.10 160.5 0.203 0.573 0 92 .0 20.0 28.0 68.8 160.8 Circumferential Weld (Heat # 4275)
Intennediate Shell Plate C433 l -2 1.1 0.12 0.60 83.0 1.69 1.1 45 15.0 95.0 0.0 17.0 34.0 144.0 Intennediate Shell Plate C4339-2 I. I 0.11 0.54 73.4 1.69 1.145 7.8 84.0 0.0 17.0 34.0 125.8 Using non-credible surveillance data 2.1 --- --- 75.7 1.69 1. 145 7.8 86.6 0.0 17.0 34.0 128.4 Intermediate Shell Longitudinal Welds I. I 0.22 0.54 167.0 0.303 0.673 -72.5 112.4 12.0 28.0 60.9 100.8 L3 and L4 (OD 50%) (Heat # 72445)
Using credible surveillance data 2.1 --- --- 167.0 0.303 0.673 -72.5 112.4 12.0 28.0 60.9 100.8 Intermediate Shell Longitudinal Weld L4 I. I 0. 19 0.57 167.0 0.303 0.673 -48 .6 I 12.4 18.0 28.0 66.6 130.3 (ID 50%) (Heat# 8Tl 762)
Intennediate to Lower Shell 1.1 0.187 0.545 147.5 1.70 1.145 0 168.9 0.0 28.0 56.0 224.9 Circumferential Weld (Heat# 0227)
Using credible surveillance data 2.1 --- --- 132.5 1.70 1.145 0 151.8 0.0 14.0 28.0 179.8 Lower Shell Plate C4208-2 I. I 0.15 0.55 107.3 1.70 1.1 47 -30 123.1 0.0 17.0 34.0 127. 1 Lower Shell Plate C4339- l I. I 0.107 0.53 70.8 1.70 1.147 -4.4 81.2 0.0 17.0 34.0 110.8 Using non-credible surveillance data 2. 1 --- -- - 75.7 1.70 1.147 -4.4 86.8 0.0 17.0 34.0 116.4 Lower Shell Longitudinal Welds LI and I. I 0.19 0.57 167.0 0.305 0.675 -48.6 112 .7 18.0 28.0 66.6 130.7 L2 (Heat# 8Tl 762)
WCAP- 18243 -NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-14 Table 5-6 Adjusted Reference Temperature Evaluation for the Surry Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 68 EFPY at the 3/4T Location R.G. 3/4T 3/4T 1.99, Wt.% Wt.% CF<*> Fluence<b) 3/4T RT NDT(U)<cl ARTNDT (d) <r1 (c) fft,. (e) Margin RPV Material ART Rev. 2 cu<*) Ni<*> (OF) (x 10 19 n/cm2, FF<bl (OF) (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 0.00798 0.094 -4 .9 0.0 ( 14.3) 19.7 0.0 39.4 34.5 (Heat # 8TJ 762)
Inlet Nozzle 2 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00184 0.031 -4.9 0.0(4.7) 19.7 0.0 39.4 34.5 (Heat# 8TJ 762)
Inlet Nozzle 3 to Upper Shell Weld 1.1 0.19 0.57 152.4 0.00251 0.039 -4.9 0.0 (6.0) 19.7 0.0 39.4 34.5 (Heat# 8TJ 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 30.0 (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 30.0 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld I. I 0.35 1.0 272.0 0.00594 0.076 30 0.0 (20.7) 0.0 0.0 0.0 30.0 (Rotterdam)
Notes:
(a) Chemical composition values taken from Tables 3-3 and 3-4 of this report. Chemi stry Factor values taken from Table 3-12 of this report.
(b) The 3/4T fluence and 3/4T FF values we re taken from Table 5-2.
(c) Initial RT NOT values and cr1 values are from Tables 3-3 and 3-4 of this report.
(d) Calculated t.RTNoT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-20 13-01 [Ref. 28); actual calculated t.RTNoT values are listed in parentheses fo r these materials.
(e) Per Appendix G of this report, the surveillance plate data were deemed non -credibl e, whereas the surveill ance data for the weld materials were deemed credib le. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. 2), the base metal cr 6 = J 7°F for Position 1.1 and for Position 2. 1 with non-credible surveillance data. Also per Regulatory Guide 1.99, R evision 2, the weld metal cr6 = 28°F fo r Position I. I , and with credible surveillance data cr 6 = 14°F for Position 2.1. However, cr6 need not exceed O.S*t.RTNOT* For welds utilizi ng initial RT NOT values based on BAW-2308, cr 6 = 28°F per References 26 and 27.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 5-15 Table 5-7 Summary of the Limiting ART Values for Surry Units 1 and 2 at 68 EFPv<*>
l/4T Limiting ART (°F) 3/4T Limiting ART (°F)
Plant Limiting Material Existing 48 EFPY Existing 48 EFPY Curves TLAA Curves TLAA Documented in Evaluation Documented in Evaluation at Technical at 68 EFPY Technical 68 EFPY Specifications<bl Specifications<bl
{Circ Flaw) Circ. Weld:
Intermediate to Lower Shell 213.9 173.6 Circ. Weld, Surry Unit 1 Heat# 72445
{Axial Flaw) Lon~. Weld:
Lower Shell Long. Weld L2 219.4 153.8 Heat # 299L44 (Position 2.1)
{Circ Flaw) Circ. Weld: 228.4 189.5 Intermediate to Lower Shell Circ. Weld, 210.7 179.8 Heat# 0227 Surry Unit 2 (Position 2 .1)
{Axial Flaw) Plate: Not Limiting 144.0 Intermediate Shell Plate C433 l-2 Axial Flaw) Weld: Lower Shell Longitudinal Weld LI and L2 174.6 Not Limiting Heat# 8Tl 762 Notes:
(a) The overall limiting ART values are shown in bold. Since the limiting l/4T ART value is an axial flaw material and the limiting 3/4T ART value is a circumferential flaw material, both the limiting axial flaw and limiting circumferential flaw P-T limits were considered. See Section 6 and Appendix H for details.
(b) The limiting 48 EFPY l/4T and 3/4T ART values in the Techn ical Specifications correspond to the Surry Unit l Intermediate to Lower Shell Circumferential Weld (Heat # 72445). The basis for the P-T limit curves is contained in WCAP-14177, Revision O [Ref. 23]; however, the applicability was extended to 48 EFPY in a later analysis. See Appendi x F for details.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 4 and 5 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4 [Ref. 3]. The curves are generated for the purpose of comparing the current Technical Specifications P-T limit curves to P-T limit curves developed using modem techniques with the goal of extending the applicability of the current Surry Units 1 and 2 Technical Specifications P-T limit curves to 68 EFPY.
Figure 6-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 20, 40, and 60°F/hr applicable for 68 EFPY, with the flange requirements and using the "Axial Flaw" methodology and the limiting "Axial Flaw" ART values summarized in Table 5-7. Figure 6-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of O (steady-state), 20, 40, 60, and 100°F/hr applicable for 68 EFPY, with the flange requirements and using the "Axial Flaw" methodology and the limiting "Axial Flaw" ART values summarized in Table 5-7. The heatup and cooldown curves were generated using the 1998 Edition through the 2000 Addenda ASME Code Section XI, Appendix G. Note that a "Circumferential Flaw" evaluation was also completed to confirm that the "Axial Flaw" methodology and ART values are bounding. See Appendix H for details.
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 6-1 and 6-2. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 6-1 (heatup curve only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig in-service hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 Edition through the 2000 Addenda ASME Code Section XI, Appendix G as follows.
(13)
where, K,m is the stress intensity factor covered by membrane (pressure) stress [see page 4-2, Equation (3)],
2 K1c = 33 .2 + 20.734 e [O.o (T - RTNoTll [see page 4-1 Equation (l)],
T is the minimum permissible metal temperature, and RT NDT is the metal reference nil-ductility temperature.
The criticality limit curve specifies pressure-temperature limits for core operation in order to provide additional margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-2 higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 4 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the inservice hydrostatic leak tests for the Surry Units 1 and 2 reactor vessel at 68 EFPY is 274°F. This temperature is the minimum permissible temperature at which design pressure can be reached during a hydrostatic test per Equation (13). The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 6-1 and 6-2 define all of the above limits for ensuring prevention of non-ductile failure for the Surry Units 1 and 2 reactor vessel for 68 EFPY with the flange requirements and without instrumentation uncertainties. The data points used for developing the heatup and cooldown P-T limit curves shown in Figures 6-1 and 6-2 are presented in Tables 6-1 and 6-2. The P-T limit curves shown in Figures 6-1 and 6-2 were generated based on the limiting "Axial Flaw" ART values for the cylindrical beltline and extended beltline reactor vessel materials. As discussed in Appendix B, the P-T limits developed for the cylindrical beltline region bound the P-T limits for the reactor vessel inlet and outlet nozzles for Surry Units 1 and 2 at 68 EFPY.
The curves developed in this Section are compared to the P-T limit curves in the current Surry Power Station Technical Specifications [Ref. l] in Section 7 with the goal of showing that the current Surry Power Station Technical Specifications P-T limit curves are bounding and appropriate for continued use to 68 EFPY. To allow direct comparison, the curves in the current Technical Specifications have been adjusted by 10% to account for the differences in methodology between the utilization of the Kia and K1c curves.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Surry Unit 1 Lower Shell Longitudinal Weld L2 (Heat# 299L44, Position 2.1)
LIMITING ART VALUES AT 68 EFPY: 1/4T, 219.4°F (Axial Flaw) 3/4T, 153 .8°F (Axial Flaw) 2500 I IOperlimAnalysis Version:5.4 Run :7360 Oper1im.xlsm Version: 5.4.1 ILeak Test um;if-L 2250 j - Critical 2000 Heatup L---":-- " Limits:
~
Rates:
~
60°F/Hr 60°F/Hr 40°F/Hr 40°F/Hr 1750 20°F/Hr
" 20°F/Hr
( !)
en c.. 1500 ,J J
Cl) lo.. Unacceptable
, Operation Ill Ill Cl) lo.. 1250 I ,
I c..
"C Cl) co
, 1000
~
co I IAcceptable I
(.)
750
/ Operation 500 1\ \.
Criticality Limit based on inservice hydrostatic test -
IHeatup Rate:
60°F/Hr temperature (27 4°F) for the service period up to 68 EFPY 250 I 0
0 50 100 150 200 250 Ii300 350 400 450 I
500 550 Moderator Temperature (Deg. F)
Figure 6-1 Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 20, 40, and 60°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Axial Flaw" Methodology (w/ K1c)
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Surry Unit 1 Lower Shell Longitudinal Weld L2 (Heat# 299L44, Position 2. 1)
LIMITING ART VALUES AT 68 EFPY: I/4T, 219.4°F (Axial Flaw) 2500 ,
3/4T, 153.8°F (Axial Flaw)
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-20°F/Hr
-40°F/Hr
-60°F/Hr I
-100°F/Hr 250 I
0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 6-2 Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Axial Flaw" Methodology (w/ K1c)
WCAP-18243-NP October 201 7 Revision 0
Westinghouse Non-Proprietary Class 3 6-5 Table 6-1 Surry Units 1 and 2 68 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App.
G Methodology (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 20°f/hr Heatup 20°F/hr Criticality 40°F/hr Heatup 40°f/hr Criticality 60°F/hr Heatup 60°F/hr Criticality T (0 f) P (psig) T (°F) P (psig) T (°F) P (psig) T (0 f) P (psig) T {°F) P (psig) T (°F) P (psig) 60 621 274 0 60 621 274 0 60 621 274 0 60 665 274 1194 60 665 274 1194 60 664 274 1194 65 666 275 1201 65 666 275 1201 65 664 275 1201 70 668 280 1259 70 668 280 1259 70 664 280 1259 75 670 285 1319 75 670 285 1319 75 664 285 1319 80 673 290 1383 80 673 290 1378 80 664 290 1378 85 675 295 1455 85 675 295 1444 85 664 295 1438 90 678 300 1534 90 678 300 1516 90 664 300 1504 95 681 305 1622 95 681 305 1596 95 665 305 1577 100 685 310 1718 100 685 310 1684 100 668 310 1658 105 689 315 1825 105 689 315 1782 105 673 315 1747 110 693 320 1943 110 693 320 1889 110 679 320 1845 115 698 325 2073 115 698 325 2008 115 687 325 1954 120 703 330 2217 120 703 330 2139 120 696 330 2073 125 709 335 2375 125 709 335 2283 125 707 335 2205 130 716 130 716 340 2443 130 716 340 2351 135 723 135 723 135 723 140 730 140 730 140 730 145 739 145 739 145 739 150 749 150 749 150 749 155 759 155 759 155 759 160 771 160 771 160 771 165 784 165 784 165 784 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-6 Table 6-1 Surry Units 1 and 2 68 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App.
G Methodology (w/ Kie, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 20°F/hr Heatup 20°F/hr Criticality 40°F/hr Heatup 40°F/hr Criticality 60°F/hr Heatup 60°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 170 798 170 798 170 798 175 814 175 814 175 814 180 832 180 832 180 832 185 851 185 851 185 851 190 873 190 873 190 873 195 896 195 896 195 896 200 922 200 922 200 922 205 951 205 951 205 951 210 983 210 983 210 983 215 1018 215 1018 215 1018 220 1057 220 1057 220 1057 225 1100 225 1100 225 1100 230 1148 230 1148 230 1148 235 1201 235 1201 235 1201 240 1259 240 1259 240 1259 245 1319 245 1319 245 1319 250 1383 250 1378 250 1378 255 1455 255 1444 255 1438 260 1534 260 1516 260 1504 265 1622 265 1596 265 1577 270 1718 270 1684 270 1658 275 1825 275 1782 275 1747 280 1943 280 1889 280 1845 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-7 Table 6-1 Surry Units 1 and 2 68 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App.
G Methodology (w/ K 1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 20°F/hr Heatup 20°F/hr Criticality 40°F/hr Heatup 40°F/hr Criticality 60°F/hr Heatup 60°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T {°F) P (psig) T {°F) P (psig) T (°F) P (psig) 285 2073 285 2008 285 1954 290 2217 290 2139 290 2073 295 2375 295 2283 295 2205 300 2443 300 2351 Leak Test Limit T (°F) P (psig) 157 2000 274 2485 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-8 Table 6-2 Surry Units 1 and 2 68 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ Krc, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)
Steady-State 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 (°F) P (psig) T (°F) P (psig) 60 621 60 619 60 573 60 526 60 428 60 665 65 621 65 575 65 527 65 429 65 666 70 623 70 576 70 529 70 431 70 668 75 625 75 579 75 531 75 434 75 670 80 627 80 581 80 534 80 436 80 673 85 630 85 584 85 536 85 439 85 675 90 633 90 587 90 540 90 443 90 678 95 636 95 590 95 543 95 446 95 681 100 640 100 594 100 547 100 451 100 685 105 644 105 598 105 55 1 105 456 105 689 110 648 110 603 110 556 110 462 110 693 115 653 115 608 115 562 115 468 115 698 120 659 120 614 120 568 120 475 120 703 125 665 125 620 125 575 125 483 125 709 130 672 130 628 130 583 130 492 130 71 6 135 679 135 636 135 592 135 502 135 723 140 688 140 645 140 60 1 140 514 140 730 145 697 145 655 145 612 145 527 145 739 150 707 150 666 150 624 150 541 150 749 155 719 155 678 155 637 155 557 155 759 160 731 160 692 160 652 160 574 160 771 165 745 165 707 165 669 165 594 165 784 170 761 170 724 170 687 170 616 170 798 175 778 175 742 175 707 175 640 175 814 180 797 180 763 180 730 180 667 180 832 185 818 185 786 185 755 185 697 185 851 190 841 190 811 190 782 190 730 190 873 195 867 195 839 195 813 195 767 195 896 200 895 200 870 200 847 200 808 200 922 205 927 205 905 205 885 205 854 205 951 210 962 210 943 210 926 210 905 210 983 215 1000 215 985 215 973 215 961 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-9 Table 6-2 Surry Units 1 and 2 68 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)
Steady-State 20°f/hr Cooldown 40°f/hr Cooldown 60°f/hr Cooldown 100°f/hr Cooldown T ( f) 0 P (psig) T (OF) P (psig) T {°f) P (psig) T ( f) 0 P (psig) T {°F) P (psig) 215 1018 220 1043 220 1031 220 1024 220 1023 220 1057 225 1090 225 1083 225 1081 225 1081 225 1100 230 1142 230 1140 230 1140 230 1140 230 1148 235 1200 235 1200 235 1200 235 1200 235 1201 240 1259 240 1259 240 1259 240 1259 240 1259 245 1323 245 1323 245 1323 245 1323 245 1323 250 1394 250 1394 250 1394 250 1394 250 1394 255 1472 255 1472 255 1472 255 1472 255 1472 260 1559 260 1559 260 1559 260 1559 260 1559 265 1655 265 1655 265 1655 265 1655 265 1655 270 1761 270 1761 270 1761 270 1761 270 1761 275 1878 275 1878 275 1878 275 1878 275 1878 280 2007 280 2007 280 2007 280 2007 280 2007 285 2150 285 2150 285 2150 285 2150 285 2150 290 2308 290 2308 290 2308 290 2308 290 2308 295 2483 295 2483 295 2483 295 2483 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-1 7 APPLICABILITY OF CURRENT HEATUP AND COOLDOWN LIMITS The applicability of the current Surry Units 1 and 2 P-T limit curves was determined based on a comparison of the available operating margin between the P-T limits developed in this report at 68 EFPY with those based on WCAP-14177 [Ref. 23], which are contained in the Surry Power Station Technical Specifications (Figures 3.1-1 and 3.1-2) [Ref. l]. A summary of the applicability of the Surry Power Station P-T limit curves is provided in Appendix F. The P-T limit curves presented in Figures 3.1-1 and 3.1-2 of the Surry Power Station Technical Specifications do not contain margins for instrumentation error; however, these curves do contain a pressure adjustment of 21.5 psi. In order to provide a direct comparison between the current Technical Specification P-T limit curves and those developed in this report, the pressure adjustment is removed from the Technical Specification curves for comparison purposes only.
The methodology of the 1998 Edition through 2000 Addenda of ASME B&PV Code,Section XI, Appendix G, along with ASME Code Case N-641 was used in the development of the P-T limit curves contained in this report. Code Case N-641 removes some of the conservatism in P-T limit curves by allowing the use of the K1c reference stress intensity factor, instead of the older, more conservative K 1a reference stress intensity factor, which was used in the development of the Surry Power Station current P-T limit curves. Additionally, the 1998 through the Summer 2000 Addenda Edition of ASME Code Section XI, Appendix G methodology allows use of the less restrictive "Circ-Flaw" methodology, which postulates circumferentially oriented reference defects in circumferential weld materials. Therefore, the P-T limit curves developed in this report took advantage of these updates to the ASME P-T limit methodology and are predicted to contain additional operating margin not present in the curves developed using the older K 1a methodology.
However, when K1c methodology is used, the LTOP system shall limit the maximum pressure in the vessel to 100% of the pressure determined to satisfy Equation (2) of Section 4. Previously, while using K 1., the maximum pressure determined from Equation (2) of Section 4 could be exceeded by 10% by the LTOP system. Therefore, since the current curves utilized the K,a reference stress intensity factor, the P-T limit curve pressure values (without margins for instrumentation error) contained in the Technical Specifications were increased by 10% in order to determine if margin exists between this data and the P-T limit curves developed herein using the K1c reference stress intensity factor. This 10% increase to the pressure values contained in the Technical Specifications is for comparison purposes only. The increased pressure values are not to be used in actual plant operation. Note that before the 10% increase is applied, the current Surry Power Station P-T limit curve data points were increased by 21.5 psi to remove the pressure adjustment so that direct comparison could be made between these pressure values (current curves without pressure adjustment plus 10% margin) and the pressure values for the curves developed in this report. These adjusted values are shown below in Tables 7-1 and 7-2, for heatup and cooldown, respectively.
Additionally, in order for the current Surry Power Station P-T limit curves to be bounded by the curves developed in this report, the criticality temperatures shown in Section 6 must be found to be lower than the minimum criticality temperature in Technical Specifications. In the Surry Power Station Technical Specifications, the minimum criticality temperature was determined to be 538°F. This value of 538°F WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-2 bounds the criticality curves developed in Section 6. Therefore, based on this analysis, significant margin exists between the current Surry Units 1 and 2 criticality temperature and the criticality curves determined in this report.
The pressure and temperature values contained in Tables 7-1 and 7-2 (current curves without pressure adjustment plus 10% margin) were plotted together with the data points from Tables 6-1 and 6-2 of this report, which were developed using the K1c reference stress intensity factor, in Figures 7-1 through 7-3 .
In Figures 7-1 through 7-3, the curves developed in this report (through 68 EFPY; without margins for instrumentation errors) are shown as solid lines, while the curves developed from the data points in Tables 7-1 and 7-2 (current curves without pressure adjustment plus 10% margin) are shown as dashed lines.
The color scheme in the Figures correlates so that the solid and dashed lines have an identical color for each heatup or cooldown rate.
Figure 7-1 shows the comparison of the heatup curves. The corresponding data points, along with the margin between the current Surry Power Station P-T limit curves + 10% margin and the P-T limit curves developed in this report are contained in Table 7-3 .
Figure 7-2 shows the comparison of the cooldown curves. Figure 7-3 shows a magnified version of Figure 7-2 in the lower pressure and temperature region. The corresponding data points, along with the margin between the current Surry Power Station P-T limit curves + 10% margin and the P-T limit curves developed in this report are contained in Table 7-4.
Tables 7-5 and 7-6 contain a summary of the available margin between the P-T limits developed in Section 6 of this report (through 68 EFPY; without margins for instrumentation errors) and the current Surry Power Station P-T limits, contained in the Technical Specifications without pressure adjustment, plus 10% margin.
Per Tables 7-5 and 7-6, the minimum pressure difference (at constant temperature) between the current Technical Specifications [Ref. 1] P-T limit curves (plus a 10% margin) and the new curves developed herein is 109 psi . This 109 psi margin applies to the steady-state curves at 80°F and 85°F, as well as the
-20°F/hr cooldown rate at 80°F. Using visual comparison of the current Technical Specifications P-T limit curves (plus a 10% margin) and the new curves, shown in Figures 7-1 , 7-2, and 7-3 herein, a minimum temperature difference (at constant pressure) of no less than 50°F is identified. These margins of 109 psi and at least 50°F illustrate that adequate margins exist in the current Technical Specifications P-T limit curves to cover typical instrument uncertainties.
P-T Limit Curve Applicability Conclusion Tables 7-5 and 7-6 show that adequate margin exists between the current Surry Power Station P-T limit curves plus 10% margin (to account for the methodology change between K,. to Kie) and the P-T limit curves developed in this report for 68 EFPY. Therefore, the continued use of the current Surry Power Station P-T limit curves as documented in Figures F-1 and F-2 is justified through SLR (68 EFPY).
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Westinghouse Non-Proprietary Class 3 7-3 Low Temperature Overpressure Protection (LTOP) Applicability Conclusion The maximum allowable Low Temperature Overpressure Protection System (LTOPS) pressurizer Power Operated Relief Valve (PORV) setpoint was calculated to be 399.6 psig for the Surry Units 1 and 2 Subsequent License Renewal (SLR) program. The calculation was performed in accordance with the WCAP-14040-A, Revision 4 [Ref. 3) methodology using critical LTOPS input parameters provided by Dominion, and the limiting axial flaw steady state Appendix G limits calculated for the SLR program at 68 Effective Full Power Years (EFPY).
The evaluation showed that the current Technical Specification value of :S 390.0 psig is bounding and will remain valid for the SLR program. Since the maximum allowable PORV setpoint for the SLR program was determined using the methodology in Reference 3, this demonstrates that the current licensing basis PORV setpoint that was developed using K 1* Appendix G limits without applying uncertainties was sufficiently conservative.
Summary of Conclusions
The current P-T limit curves in the Surry Power Station Technical Specifications [Ref. I] remain valid through 68 EFPY.
The 21.5 psi adjustment applied to the Technical Specification P-T limit curves remains applicable per Dominion calculations [Refs. 29 and 30). Note that the LTOP PORV setpoint calculation utilized a conservative 40 psi adjustment.
The margin between the current P-T limit curves in the Surry Power Station Technical Specifications
[Ref. 1] plus 10% and the new K1c curves developed herein is sufficient to cover typical instrument uncertainties.
The nozzle P-T limit curves (documented in Appendix B) are bounded by the current Surry Power Station Technical Specifications [Ref. 1) P-T limit curves through 68 EFPY, and other Reactor Coolant Pressure Boundary ferritic components have been addressed (see Appendix C).
The current Technical Specification PORV setpoint of :S 390.0 psig remains valid through 68 EFPY.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-4 Table 7-1 Current Surry Power Station P-T Limit Curve Data Points without Pressure Adjustment Plus 10% Margin for Heatup<a) 0°F/hr 20°F/hr 40°F/hr 60°F/hr T (°F)
P (psig) P (psig) P (psig) P (psig) 80 564 553 528 503 85 566 553 528 503 90 568 553 528 503 95 571 556 528 503 IOO 573 559 528 503 105 576 563 528 503 110 579 567 530 503 115 583 572 532 503 120 586 578 536 503 125 590 584 540 505 130 594 590 544 507 135 599 597 550 510 140 604 604 555 514 145 609 609 562 518 150 614 614 569 524 155 620 620 576 529 160 627 627 585 536 165 634 634 594 543 170 641 641 603 551 175 649 649 613 560 180 658 658 625 569 185 667 667 637 579 190 677 677 650 590 195 688 688 663 602 200 699 699 678 615 205 711 711 694 628 210 725 725 712 643 215 739 739 730 659 220 754 754 750 676 225 771 771 771 695 230 788 788 788 715 235 807 807 807 736 240 828 828 828 760 245 850 850 850 784 250 873 873 873 811 255 899 899 899 840 260 926 926 926 871 265 955 955 955 904 270 987 987 987 939 275 I020 I020 I020 978 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-5 Table 7-1 Current Surry Power Station P-T Limit Curve Data Points without Pressure Adjustment Plus 10% Margin for Heatup<*>
0°F/hr 20°F/hr 40°F/hr 60°F/hr T (OF)
P (psig) P (psig) P (psig) P (psig) 280 1057 1057 1057 1018 285 1096 1096 1096 1062 290 1138 1138 1138 1110 295 1183 1183 1183 1161 300 1231 1231 1231 1215 305 1283 1283 1283 1273 310 1339 1339 1339 1336 315 1399 1398 1398 1398 320 1463 1456 1456 1456 325 1532 1514 1514 1514 330 1606 1575 1575 1575 335 1686 1639 1639 1639 340 1771 1709 1709 1709 345 1862 1783 1783 1783 350 1960 1863 1863 1863 355 2065 1949 1949 1949 1
360 2178 2040 2040 2040 365 2298 2138 2138 2138 370 2426 2242 2242 2242 375 2564 2355 2355 2355 380 2710 2474 2474 2474 385 2602 2602 2602 Note:
(a) Data is associated with the Surry Power Station current heatup curves contained in the Technical Specifications and based on WCAP-14177
[Ref. 23) evaluations. Ten-percent margin was added to the pressure values after the 21.5 psi pressure adjustment was removed. This ten-percent margin on the pressure values is for comparison purposes only and is not to be used in actual plant operation.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-6 Table 7-2 Current Surry Power Station P-T Limit Curve Data Points without Pressure Adjustment Plus 10% Margin for Cooldown<*>
0°F/hr -20°F/hr -40°F/hr -60°F/hr -100°F/hr T (°F)
P (psig) P (psig) P (psig) P (psig) P (psig) 80 564 518 471 423 324 85 566 520 473 425 325 90 568 522 475 427 328 95 571 525 478 430 330 100 573 527 480 432 333 105 576 530 483 435 335 110 579 533 486 438 339 115 583 537 490 442 342 120 586 540 493 445 346 125 590 544 497 450 350 130 594 549 502 454 355 135 599 553 506 459 360 140 604 558 511 464 366 145 609 563 517 470 372 150 614 569 523 476 379 155 620 575 529 482 386 160 627 582 536 490 394 165 634 589 544 497 403 170 641 597 552 506 412 175 649 605 560 515 422 180 658 614 570 525 433 185 667 624 580 536 445 190 677 634 591 547 458 195 688 645 603 560 472 200 699 657 615 573 487 205 711 670 629 587 504 210 725 684 644 603 521 215 739 699 660 620 541 220 754 716 677 638 561 225 771 733 695 658 584 230 788 752 715 679 608 235 807 772 737 702 634 240 828 793 760 727 662 245 850 817 785 753 693 250 873 842 811 782 726 255 899 869 840 813 761 260 926 898 871 846 800 265 955 929 905 882 841 270 987 963 941 920 885 275 1020 999 979 961 933 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-7 Table 7-2 Current Surry Power Station P-T Limit Curve Data Points without Pressure Adjustment Plus 10% Margin for Cooldown<a) 0°F/hr -20°F/hr -40°F/hr -60°F/hr -100°F/hr T (°F)
P (psig) P (psig) P (psig) P (psig) P (psig) 280 1057 1038 1021 1006 985 285 1096 1080 1065 1054 1040 290 1138 1124 1113 1106 1100 295 11 83 1173 1165 1161 1164 300 1231 1224 1221 1221 1231 305 1283 1280 1281 1283 1283 310 1339 1339 1339 1339 1339 315 1399 1399 1399 1399 1399 320 1463 1463 1463 1463 1463 325 1532 1532 1532 1532 1532 330 1606 1606 1606 1606 1606 335 1686 1686 1686 1686 1686 340 1771 1771 1771 1771 1771 345 1862 1862 1862 1862 1862 350 1960 1960 1960 1960 1960 355 2065 2065 2065 2065 2065 360 2178 2178 2178 2178 2178 365 2298 2298 2298 2298 2298 370 2426 2426 2426 2426 2426 375 2564 2564 2564 2564 2564 380 2710 2710 2710 2710 2710 Note:
(a) Data is associated wi th the Surry Power Station current cooldown curves contained in the Techni cal Specifications and based on WCAP-14177 [Ref. 23] evaluations. Ten-percent margin was added to the pressure values after the 21.5 psi pressure adjustment was removed. This ten-percent margin on the pressure values is for comparison purposes only and is not to be used in actual plant operation.
WCAP-1 8243-NP October 2017 Revision 0
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20°F/Hr orange 40°F/Hr green 60°F/Hr blue 0 ' '
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 7-1 Surry Units 1 and 2 Heatup P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY Note: "New P-T Limits" determined in Section 6. TS = Technical Specifications.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-9 2500 2250 Unacceptable 0 eration 2000 1750 C)
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ca Acceptable u 0 eration 750 Solid Lines: New P-T Limits Dashed Lines: TS + 10% Margin Cooldown Rates:
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-20°F/Hr orange 250 +-----+ -l--r-~1+--== i = - - - - - - - - + - -1
-40°F/Hr green
-60°F/Hr blue
-100°F/Hr red 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature {Deg. F)
Figure 7-2 Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY Note: "New P-T Limits" determined in Section 6. TS = Technical Specifications.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-10 900 +-~ r.-:--~---:-:--:-if--~-l-~~~-+~~-1-- , ~~, i ___::::,,.,,__ ~+-r+,
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-20°F/Hr orange 100 +-----------1------4-------1 -40°F/Hr green
-60°F/Hr blue
-100°F/Hr red 0
0 50 100 150 200 250 300 Moderator Temperature (Deg. F)
Figure 7-3 Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY Magnified Note: "New P-T Limits" determined in Section 6. TS= Technical Specifications.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7- 11 Table 7-3 Data Points for Surry Units 1 and 2 Heatup P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY 0°F/hr 20°F/hr 40°F/hr 60°F/hr
+10% +10% +10% +10%
T New P-T New P-T New P-T New P-T Current Margin<*l Current Margin<*l Current Margin<*l Current Margin<*l (OF) Curves Curves Curves Curves Curves (psi) Curves (psi) Curves (psi) Curves (psi)
(psig) (psig) (psig) (psig)
(psig) (psig) (psig) (psig) 80 564 673 109 553 673 120 528 673 145 503 664 161 85 566 675 109 553 675 122 528 675 148 503 664 161 90 568 678 110 553 678 125 528 678 151 503 664 161 95 571 681 111 556 681 125 528 681 154 503 665 162 100 573 685 112 559 685 126 528 685 157 503 668 165 105 576 689 113 563 689 126 528 689 161 503 673 170 110 579 693 114 567 693 126 530 693 164 503 679 176 115 583 698 115 572 698 126 532 698 166 503 687 184 120 586 703 117 578 703 125 536 703 168 503 696 193 125 590 709 119 584 709 125 540 709 169 505 707 203 130 594 716 121 590 716 126 544 716 171 507 716 209 135 599 723 124 597 723 126 550 723 173 510 723 213 140 604 730 127 604 730 127 555 730 175 514 730 217 145 609 739 130 609 739 130 562 739 177 518 739 221 150 614 749 134 614 749 134 569 749 180 524 749 225 155 620 759 139 620 759 139 576 759 183 529 759 230 160 627 771 144 627 771 144 585 771 187 536 771 235 165 634 784 150 634 784 150 594 784 191 543 784 241 170 641 798 157 641 798 157 603 798 195 551 798 247 175 649 814 165 649 814 165 613 814 201 560 814 255 180 658 832 174 658 832 174 625 832 207 569 832 263 185 667 851 184 667 851 184 637 851 215 579 851 272 190 677 873 196 677 873 196 650 873 223 590 873 283 195 688 896 209 688 896 209 663 896 233 602 896 294 200 699 922 223 699 922 223 678 922 244 615 922 308 205 711 951 240 711 951 240 694 951 257 628 951 323 210 725 983 258 725 983 258 712 983 271 643 983 340 215 739 1018 280 739 1018 280 730 1018 288 659 1018 359 220 754 1057 303 754 1057 303 750 1057 307 676 1057 381 WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-12 Table 7-3 Data Points for Surry Units 1 and 2 Heatup P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10% Margin and the New P-T Limit Curves to 68 EFPY 0°F/hr 20°F/hr 40°F/hr 60°F/hr
+10% +10% +10% +10%
T New P-T New P-T NewP-T New P-T Current Margin<*> Current Margin<*> Current Margin<*> Current Margin<*>
(OF) Curves Curves Curves Curves Curves (psi) Curves (psi) Curves (psi) Curves (psi)
(psig) (psig) (psig) (psig)
(psig) (psig) (psig) (psig) 225 771 1100 330 771 1100 330 771 1100 330 695 1100 405 230 788 1148 360 788 1148 360 788 1148 360 715 1148 433 235 807 1201 393 807 1201 393 807 1201 393 736 1201 464 240 828 1259 431 828 1259 431 828 1259 431 760 1259 499 245 850 1323 473 850 1319 469 850 1319 469 784 1319 534 250 873 1394 521 873 1383 510 873 1378 505 811 1378 566 255 899 1472 574 899 1455 557 899 1444 545 840 1438 598 260 926 1559 633 926 1534 608 926 1516 590 871 1504 633 265 955 1655 700 955 1622 667 955 1596 641 904 1577 673 270 987 1761 774 987 1718 731 987 1684 697 939 1658 718 275 1020 1878 858 1020 1825 804 1020 1782 761 978 1747 769 280 1057 2007 951 1057 1943 886 1057 1889 832 1018 1845 827 285 1096 2150 1055 1096 2073 977 1096 2008 912 1062 1954 891 290 1138 2308 1171 1138 2217 1079 1138 2139 1001 1110 2073 963 295 1183 2483 1300 1183 2375 1192 1183 2283 1100 1161 2205 1045 300 1231 2676 1445 1231 2443 1212 1215 2351 1136 Note:
(a) Margin equals New P-T limit curve data point minus the current (I'echnical Specifications) P-T limit curves + 10% data point for each temperature and rate.
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Westinghouse Non-Proprietary Class 3 7-13 Table 7-4 Data Points fo r Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10%
Margin and the New P-T Limit Curves to 68 EFPY 0°F/hr -20°F/hr -40°F/hr -60°F/hr -100°F/hr
+10% New +10% New +10% New +10% New +10% New T M<*> M<*> M<*> M<*> M<*>
Current P-T Current P-T Current P-T Current P-T Current P-T (OF)
Curves Curves (psi) Curves Curves (psi) Curves Curves (psi) Curves Curves (psi) Curves Curves (psi)
(psig) (psig) (psig) (psig) (psig) (psig) (psig) (osig) (osig) (psig) 80 564 673 109 518 627 109 471 581 110 423 534 111 324 436 113 85 566 675 109 520 630 110 473 584 111 425 536 111 325 439 114 90 568 678 110 522 633 111 475 587 111 427 540 112 328 443 115 95 571 681 111 525 636 111 478 590 112 430 543 113 330 446 117 100 573 685 112 527 640 112 480 594 113 432 547 115 333 451 118 105 576 689 113 530 644 114 483 598 115 435 551 116 335 456 120 110 579 693 114 533 648 115 486 603 116 438 556 118 339 462 123 115 583 698 115 537 653 117 490 608 118 442 562 120 342 468 126 120 586 703 117 540 659 118 493 614 120 445 568 123 346 475 129 125 590 709 119 544 665 121 497 620 123 450 575 126 350 483 133 130 594 716 121 549 672 123 502 628 126 454 583 129 355 492 137 135 599 723 124 553 679 126 506 636 129 459 592 133 360 502 142 140 604 730 127 558 688 130 511 645 133 464 601 137 366 514 148 145 609 739 130 563 697 134 517 655 138 470 612 143 372 527 155 150 614 749 134 569 707 138 523 666 143 476 624 148 379 541 162 155 620 759 139 575 719 144 529 678 149 482 637 155 386 557 171 160 627 771 144 582 731 150 536 692 156 490 652 163 394 574 180 165 634 784 150 589 745 156 544 707 163 497 669 171 403 594 191 170 641 798 157 597 761 164 552 724 172 506 687 181 412 616 204 175 649 814 165 605 778 173 560 742 182 515 707 192 422 640 218 180 658 832 174 614 797 183 570 763 193 525 730 205 433 667 234 185 667 851 184 624 818 194 580 786 206 536 755 219 445 697 252 190 677 873 196 634 841 207 591 811 220 547 782 235 458 730 272 195 688 896 209 645 867 222 603 839 236 560 813 253 472 767 295 200 699 922 223 657 895 238 615 870 255 573 847 274 487 808 321 205 711 951 240 670 927 257 629 905 276 587 885 297 504 854 350 210 725 983 258 684 962 277 644 943 299 603 926 323 521 905 383 215 739 1018 280 699 1000 301 660 985 325 620 973 353 541 961 420 220 754 1057 303 716 1043 327 677 1031 355 638 1024 386 561 1023 462 WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-14 Table 7-4 Data Points for Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between the Current P-T Limit Curves+ 10%
Margin and the New P-T Limit Curves to 68 EFPY 0°F/hr -20°F/hr -40°F/hr -60°F/hr -100°F/hr
+10% New +10% New +10% New +10% New +10% New T MC*> M<*> M<*> M<*> MC*>
Current P-T Current P-T Current P-T Current P-T Current P-T (OF)
Curves Curves (psi) Curves Curves (psi) Curves Curves (psi) Curves Curves (psi) Curves Curves (psi)
(psig) (psig) (psig) (psig) (psig) (psig) (psi2) (psi!?) (psi2) (psi2) 225 771 1100 330 733 1090 357 695 1083 388 658 1081 423 584 1081 497 230 788 1148 360 752 1142 390 715 1140 425 679 1140 461 608 1140 532 235 807 1201 393 772 1200 428 737 1200 463 702 1200 498 634 1200 565 240 828 1259 431 793 1259 465 760 1259 499 727 1259 532 662 1259 596 245 850 1323 473 817 1323 506 785 1323 538 753 1323 570 693 1323 630 250 873 1394 521 842 1394 552 811 1394 583 782 1394 612 726 1394 668 255 899 1472 574 869 1472 603 840 1472 632 813 1472 660 761 1472 711 260 926 1559 633 898 1559 661 871 1559 688 846 1559 714 800 1559 760 265 955 1655 700 929 1655 726 905 1655 750 882 1655 774 841 1655 814 270 987 1761 774 963 1761 798 941 1761 820 920 1761 841 885 1761 876 275 1020 1878 858 999 1878 879 979 1878 899 961 1878 917 933 1878 945 280 1057 2007 951 1038 2007 970 1021 2007 987 1006 2007 1002 985 2007 1023 285 1096 2150 1055 1080 2150 1071 1065 2150 1085 1054 2150 1097 1040 2150 1110 290 1138 2308 1171 1124 2308 1184 1113 2308 1195 1106 2308 1203 1100 2308 1209 295 1183 2483 1300 1173 2483 1311 1165 2483 1318 1161 2483 1322 1164 2483 1319 300 1231 2676 1445 Note:
(a) Margin equals New P-T limit curve data point minus the current (Technical Specifications) P-T limit curves + 10% data point for each temperature and rate.
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Westinghouse Non-Proprietary Class 3 7-15 Table 7-5 Surry Units 1 and 2 Heatup P-T Limit Curve Margin Summary between the Current P-T Limit Curves + 10% Margin and the New P-T Limit Curves to 68 EFPY 0°F/hr 20°F/hr 40°F/hr 60°F/hr T
(OF) Margin Margin Margin Margin (psi) (psi) (psi) (psi) 80 109 120 145 161 85 109 122 148 161 90 110 125 151 161 95 111 125 154 162 100 112 126 157 165 105 113 126 161 170 110 114 126 164 176 115 115 126 166 184 120 117 125 168 193 125 119 125 169 203 130 121 126 171 209 135 124 126 173 213 140 127 127 175 217 145 130 130 177 221 150 134 134 180 225 155 139 139 183 230 160 144 144 187 235 165 150 150 191 241 170 157 157 195 247 175 165 165 201 255 180 174 174 207 263 185 184 184 215 272 190 196 196 223 283 195 209 209 233 294 200 223 223 244 308 205 240 240 257 323 210 258 258 271 340 215 280 280 288 359 220 303 303 307 381 225 330 330 330 405 230 360 360 360 433 235 393 393 393 464 240 431 431 431 499 245 473 469 469 534 250 521 510 505 566 255 574 557 545 598 260 633 608 590 633 265 700 667 641 673 270 774 731 697 718 275 858 804 761 769 280 951 886 832 827 285 1055 977 912 891 290 1171 1079 1001 963 295 1300 1192 1100 1045 300 1445 - 1212 1136 WCAP-18243-NP October2017 Revision 0
Westinghouse Non-Proprietary Class 3 7-16 Table 7-6 Surry Units 1 and 2 Cooldown P-T Limit Curve Margin Summary between the Current P-T Limit Curves + 10%
Mar2in and the New P-T Limit Curves to 68 EFPY
~
0°F/hr -20°F/hr -40°F/hr -60°F/hr -100°F/hr T
(OF) Margin Margin Margin Margin Margin (psi) (psi) (psi) (psi) (psi) 80 109 109 110 111 113 85 109 110 111 111 114 90 110 111 111 112 115 95 111 111 112 113 117 100 112 112 113 115 118 105 113 114 115 116 120 110 114 115 116 118 123 115 115 117 118 120 126 120 117 118 120 123 129 125 119 121 123 126 133 130 121 123 126 129 137 135 124 126 129 133 142 140 127 130 133 137 148 145 130 134 138 143 155 150 134 138 143 148 162 155 139 144 149 155 171 160 144 150 156 163 180 165 150 156 163 171 191 170 157 164 172 181 204 175 165 173 182 192 218 180 174 183 193 205 234 185 184 194 206 219 252 190 196 207 220 235 272 195 209 222 236 253 295 200 223 238 255 274 321 205 240 257 276 297 350 210 258 277 299 323 383 215 280 301 325 353 420 220 303 327 355 386 462 225 330 357 388 423 497 230 360 390 425 461 532 235 393 428 463 498 565 240 431 465 499 532 596 245 473 506 538 570 630 250 521 552 583 612 668 255 574 603 632 660 711 260 633 661 688 714 760 265 700 726 750 774 814 270 774 798 820 841 876 275 858 879 899 917 945 280 951 970 987 1002 1023 285 1055 1071 1085 1097 1110 290 1171 1184 1195 1203 1209 295 1300 1311 1318 1322 1319 300 1445 - - - -
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Westinghouse Non-Proprietary Class 3 8-1 8 REFERENCES
1. Surry Power Station Technical Specifications, Section 3.1.B, Amendments Nos. 285 and 285.
2. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
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. Appendix G to the 1998 Edition through 2000 Addenda of ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."
5. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.
6. 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.
7. 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 MLJ 7J88AJ 58}
8. 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, October 2014. [ADAMS Accession Number MLl4149Al65}
9. 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 ML15041A720}
10. 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}
11. RSICC Computer Code Collection CCC-650, "DOORS 3.2: One, Two- and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998.
12. 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," March 1996.
13. 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 1 and Unit 2,"
AREVA Document No. 77-2313S-007-001 , February 2017.
14. Pressurized Water Reactor Owners Group (PWROG) Report PWROG-16045-NP, Revision 0, "Determination of Unirradiated RT NDT and Upper-Shelf Energy Values of the Surry Units 1 and 2 Reactor Vessel Materials," March 2017.
15. Westinghouse Report WCAP-18028-NP, Revision 0, "Extended Beltline Pressure Vessel Fluence Evaluations Appli~able to Surry Units 1 & 2," September 2015.
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Westinghouse Non-Proprietary Class 3 8-2
16. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements,"
Revision 2, U.S. Nuclear Regulatory Commission, March 2007.
17. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subarticle NB-2300, "Fracture Toughness Requirements for Material."
18. BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data/or BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.
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 1 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. AREVA NP, Inc. Report AREVA-17-01417, "Transmittal of Information for Surry Specific Weld Wire Heat 299L44 (Capsule AS) from BAW-2313 Revision 7, PA-MSC-1200RO Task l," May 2017.
26. NRC Safety Evaluation, "Final Safety Evaluation for Topical Report BAW-2308, Revision 1 'Initial RTNDT of Linde 80 Weld Materials'," August 2005. [ADAMS Accession Number ML052070408]
27. NRC Safety Evaluation, "Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) BAW-2308, Revision 2, 'Initial RTNoT of Linde 80 Weld Materials'," March 2008. [ADAMS Accession Number ML080770349]
28. 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], November 2014. [ADAMS Accession Number MLJ 4318Al 77]
29. Virginia Power Calculation SM-792, Revision 3, "Surry 1 & 2 Composite PIT Limits Curve,"
January 1996.
30. Virginia Power Calculation SM-945, Revision 0, "Surry Unit 1 and 2 Heatup/Cooldown Curves and LTOPS Setpoint," February 1995.
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Westinghouse Non-Proprietary Class 3 A -1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (Kn)
Tables A-1 and A-2 contain the thermal stress intensity factors (Kit) for the maximum heatup and cooldown rates at 68 EFPY for Surry Units 1 and 2 based on the Section 6 P-T limit curves. The reactor vessel cylindrical shell radii to the 1/4T and 3/4T locations are as follows :
1/4T Radius = 80.5 inches
3/4T Radius = 84.5 inches WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-2 TableA-1 K11 Values for Surry Units 1 and 2 at 68 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors)
Water Vessel Temperature l/4T Thermal Stress Vessel Temperature 3/4T Thermal Stress Temp. at l/4T Location for Intensity Factor at 3/4T Location for Intensity Factor (OF) 60°F/hr Heatup (°F) (ksi '>/in.) 60°F/hr Heatup (0F) (ksi '>/in.)
60 56.538 -1.077 55.169 0.604 65 59.879 -2.410 55 .956 1.618 70 63.411 -3.347 57.605 2.400 75 67.274 -4.168 59.939 3.035 80 71.373 -4.775 62.802 3.533 85 75.623 -5.293 66.103 3.936 90 80.052 -5.686 69.739 4.256 95 84.560 -6.022 73 .647 4.515 100 89.192 -6.279 77.760 4.725 105 93.865 -6.504 82.039 4.898 110 98.619 -6.677 86.448 5.039 115 103.396 -6.832 90.960 5.157 120 108.226 -6.952 95.554 5.256 125 113.068 -7.062 100.213 5.339 130 117.946 -7.150 104.924 5.410 135 122.830 -7.232 109.675 5.472 140 127.738 -7.298 114.459 5.526 145 132.648 -7.363 119.269 5.574 150 137.574 -7.416 124.100 5.617 155 142.502 -7.470 128.947 5.657 160 147.439 -7.515 133.806 5.693 165 152 .3 78 -7.561 138.677 5.727 170 157.322 -7.601 143.555 5.759 175 162.268 -7.642 148.440 5.790 180 167.217 -7.679 153.330 5.819 185 172.167 -7.718 158.225 5.847 190 177.119 -7.752 163 .122 5.875 195 182.072 -7.789 168.022 5.902 200 187.025 -7.823 172.925 5.928 205 191.980 -7.858 177.828 5.954 2 10 196.934 -7.891 182.734 5.980 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-3 TableA-2 K11 Values for Surry Units 1 and 2 at 68 EFPY 100°F/hr Cooldown Curves (w/ Flange Requirements, and w/o Margins for Instrument Errors)
Water Vessel Temperature at 1/4T 100°F/hr Cooldown Temp. Location for 100°F/hr 1/4T Thermal Stress (OF) Cooldown (°F) Intensity Factor (ksi --Jin.)
210 233.489 14.320 205 228.412 14 .26 1 200 223 .336 14.202 195 218.259 14.143 190 213.182 14.085 185 208.104 14.025 180 203.027 13.967 175 197.950 13.907 170 192.873 13.849 165 187.796 13.790 160 182.720 13 .73 1 155 177.643 13 .673 150 172.566 13.6 14 145 167.489 13 .556 140 162.413 13.497 135 157.336 13.439 130 152.260 13.381 125 147.183 13 .323 120 142.107 13.265 11 5 137.03 1 13.207 11 0 131.955 13.149 105 126.879 13.092 100 121.803 13.034 95 11 6.728 12.977 90 111.652 12.920 85 106.577 12.863 80 101.502 12.806 75 96.427 12.749 70 9 1.352 12.692 65 86.277 12.636 60 81.203 12.579 WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES As described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. B-1], reactor vessel non-beltline materials may define pressure-temperature (P-T) limit curves that are more limiting than those calculated for the reactor vessel cylindrical shell beltline materials. Reactor vessel nozzles, penetrations, and other discontinuities have complex geometries that can exhibit significantly higher stresses than those for the reactor vessel beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the reference temperatures (RTNDT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.
The methodology contained in WCAP-14040-A, Revision 4 [Ref. B-2] was used in the main body of this report to develop P-T limit curves for the limiting Surry Units 1 and 2 cylindrical shell beltline material; however, WCAP-14040-A, Revision 4 does not consider ferritic materials in the area adjacent to the beltline, specifically the stressed inlet and outlet nozzles. Due to the geometric discontinuity, the inside comer regions of these nozzles are the most highly stressed ferritic component outside the beltline region of the reactor vessel; therefore, these components are analyzed in this Appendix. P-T limit curves are determined for the reactor vessel nozzle comer region for Surry Units 1 and 2 and compared to the P-T limit curves for the reactor vessel traditional beltline region in order to determine if the nozzles can be more limiting than the reactor vessel beltline as the plant ages and the vessel accumulates more neutron fluence. The increase in neutron fluence as the plant ages causes a concern for embrittlement of the reactor vessel above the beltline region. Therefore, the P-T limit curves are developed for the nozzle inside comer region since the geometric discontinuity results in high stresses due to internal pressure and the cooldown transient. The cooldown transient is analyzed as it results in tensile stresses at the inside surface of the nozzle comer.
A l /4T axial flaw is postulated at the inside surface of the reactor vessel nozzle comer, and stress intensity factors are determined based on the rounded curvature of the nozzle geometry. The allowable pressure is then calculated based on the fracture toughness of the nozzle material and the stress intensity factors for the l /4T flaw.
B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (K1c) used for the inlet and outlet nozzle material is defined in Appendix G of the Section XI ASME Code, as discussed in Section 4 of this report. The K1c fracture toughness curve is dependent on the Adjusted Reference Temperature (ART) value for irradiated materials. The ART values for the inlet and outlet nozzle materials are determined using the methodology contained in Regulatory Guide 1.99, Revision 2 [Ref. B-3], which is described in Section 5 of this report, as well as weight percent (wt.%) copper (Cu) and nickel (Ni) values, initial RTNoT values, and projected neutron fluence as inputs. The material properties for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Tables B-1 and B-2 and a summary of the limiting inlet and outlet nozzle ART values for Surry Units 1 and 2 is presented in Table B-3 .
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Westinghouse Non-Proprietary Class 3 B-2 Nozzle Material Properties The Surry Units 1 and 2 nozzle material properties are provided in Tables B-1 and B-2. Copper (Cu) and Nickel (Ni) weight percent (wt. %) values were obtained from PWROG-16045-NP [Ref. B-4] for each of the Surry Units 1 and 2 reactor vessel inlet and outlet nozzles.
Surry Units 1 and 2 nozzle forging initial RT NDT and initial USE values for the inlet and outlet nozzles were also taken from PWROG-16045-NP [Ref. B-4]. The Charpy V-Notch forging specimen orientation for the inlet and outlet nozzles was not reported in the Surry Units 1 and 2 Certified Material Test Reports (CMTRs); thus, it was conservatively assumed that the orientation was the "strong direction" for each nozzle forging. Since each of the nozzle forging materials lacked drop-weight test data, the initial RT NDT values were determined for each of the Surry Units 1 and 2 reactor vessel inlet and outlet nozzle forging materials using the BWRVIP-173-A, Appendix B, Alternative Approach 2 Methodology [Ref. B-5]. The initial RTNDT values for all of the nozzle materials were determined using CV GRAPH, Version 6.02 hyperbolic tangent curve fits through the Charpy data points, in accordance with BWRVIP-173-A, Appendix B, Alternative Approach 2 Methodology [Ref. B-5]. The initial USE values were determined in accordance with the methodology described in ASTM El85-82 [Ref. B-6]. For each of the nozzle forging materials, use of BTP 5-3 Paragraph 1.2 [Ref. B-7] was necessary. The Surry Units 1 and 2 initial RTNDT and initial USE values for the inlet and outlet nozzles materials are summarized in Tables B-1 and B-2.
Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > 1 MeV) exposure of the Surry Units 1 and 2 reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculations were calculated at a location corresponding to the postulated l/4T flaw in nozzle forgings and were chosen at an elevation lower than the actual elevation of the postulated flaw and at the inside surface of the nozzle, for conservatism.
Per NRC RIS 2014-11 [Ref. B-1], embrittlement ofreactor vessel materials, with projected fluence values less than 1 x 10 17 n/cm 2 (E > 1.0 MeV), does not need to be considered. Per Tables 2-3 and 2-4, the only Surry Units 1 and 2 inlet and outlet nozzles determined to receive a projected maximum fluence of greater than 1 x 10 17 n/cm2 (E > 1 MeV) at the l/4T flaw location at 68 EFPY are Surry Unit 1 Inlet Nozzle 1, Surry Unit 2 Inlet Nozzle 1, and Surry Unit 2 Outlet Nozzle 3. For conservatism, the ti.RTNDT for each of the nozzle materials is calculated.
The second conclusion ofTLR-RES/DE/CIB-2013-01 [Ref. B-8] states that if ~RTNDT is calculated to be less than 25°F, then embrittlement need not be considered. This conclusion is applicable to and is applied to each of the Surry Units 1 and 2 inlet and outlet nozzle forging materials. Therefore, the initial RT NDT values documented in Tables B-1 and B-2 are identical to the nozzle ART values.
The neutron fluence values used in the ART calculations for the Surry Units 1 and 2 inlet and outlet nozzle forging materials are summarized in Tables B-1 and B-2. The limiting nozzle ART values used for determination of the nozzle P-T limit curves are summarized in Table B-3 .
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Westinghouse Non-Proprietary Class 3 B-3 Table B-1 Calculation of the Surry Unit 1 Nozzle Forging ART Values at 68 EFPY R.G. Surface 1.99, Wt.% Wt.% CF<*> Fluence<b) Surface RT NDT(U)<c> .1.RTNDT (d) <J/c) Gt,. (e) Margin ART RPV Material 19 Rev. 2 cu<*) Ni<*> (OF) (x 10 n/cm2, FF (OF) (OF) (OF) {°F) (OF) (OF)
Position E > 1.0 MeV)
Inlet Nozzle 1 (Heat # 9-4787) I.I 0.159 0.85 123.5 0.0124 0.127 10.3 0.0 (15 .6) 0.0 0.0 0.0 10.3 Inlet Nozzle 2 (Heat# 9-5078) I.I 0.159 0.87 123.7 0.00322 0.048 11.6 0.0 (5.9) 0.0 0.0 0.0 11.6 Inlet Nozzle 3 (Heat# 9-4819) 1.1 0.159 0.84 123.4 0.00446 0.062 -47.2 0.0 (7.6) 0.0 0.0 0.0 -47.2 Outlet Nozzle 1 (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 -44.9 Outlet Nozzle 2 (Heat# 9-4762) l. l 0.159 0.83 123.3 0.00249 0.039 -87.5 0.0 (4.8) 0.0 0.0 0.0 -87.5 Outlet Nozzle 3 (Heat# 9-4788) I.I 0.159 0.84 123.4 0.00962 0.107 -50.2 0.0 (13.2) 0.0 0.0 0.0 -50.2 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 tluence values taken from Section 2 of this report. FF= tluence factor= t<0*28*0* 10 ' 1og(fl>.
(c) Initial RT NDT values and cr 1 values are from Table 3-2 of this report.
(d) Calculated D.RTNDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-20 13-01 [Ref. B-8]; actual calculated D.RTNDT values are listed in parentheses for these materials.
(e) Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. B-3], the base metal cr 6 = l 7°F for Position 1.1. However, cr 6 need not exceed O.S*D.RTNDT*
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Prop rietary Class 3 8 -4 Table B-2 Calculation of the Surry Unit 2 Nozzle Forging ART Values at 68 EFPY R.G. Surface 1.99, Wt.% Wt.% CF<*> Fluence<hl Surface RTNDT(U)
(c)
ARTNDT (d) o}cl Gt,. (e) Margin ART(f)
RPV Material cu<*) 19 Rev. 2 Ni<*> (OF) (x 10 n/cm2, FF(bl (OF) (OF) (OF) (OF) (OF) (OF)
Position E > 1.0 MeV)
Inlet Nozzle I (Heat# 9-5 104) I.I 0.159 0.84 123.4 0.0 139 0.137 -29.7 0.0 (16.8) 0.0 0.0 0.0 -29.7 Inlet Nozzle 2 (Heat# 9-48 15) I. I 0. 159 0.87 123.7 0.0032 1 0.048 4.5 0.0 (5.9) 0.0 0.0 0.0 4. 5 In let Nozzle 3 (Heat# 9-5205) I. I 0.159 0.86 123.6 0.00437 0.06 1 6.5 0.0 (7 .5) 0.0 0.0 0.0 6.5 Outlet Nozzle I (Heat# 9-4825-2) I. I 0.159 0.85 123 .5 0.00338 0.050 -58 .1 0.0 (6 .2) 0.0 0.0 0.0 -58.1 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 -26.6 Outlet Nozzle 3 (Heat# 9-5086-2) I. I 0.159 0.87 123 .7 0.0 107 0.11 5 -33.8 0.0 (14 .2) 0.0 0.0 0.0 -33.8 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 = flue nce factor= { 0*2s -o.tO' log(f)>.
(c) Initial RT NOT values and cr 1 values are from Table 3-4 of this report.
(d) Calculated D.RTNOT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-20 l 3-0 I [Ref. B-8]; actual calculated D.RTNOT values are listed in parentheses for these materials.
(e) Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. B-3] , the base metal crr, = l 7°F for Position I. I. However, crr, need not exceed O.S*D.RTNOT
Table B-3 Summary of the Limiting ART Values for the Surry Units 1 and 2 Inlet and Outlet Nozzle Forging Materials EFPY Nozzle Material and ID Number Limiting ART Value (°F)
Surry Unit I Inlet Nozzle 2 11.6 (Heat # 9-5078)
Surry Uni t 1 Outlet Nozzle I
-44.9 (Heat# 9-4825-1) 68 Surry Un it 2 Inlet Nozzle 3 6.5 (Heat# 9-5205)
Surry Unit 2 Outlet Nozzle 2
-26.6 (Heat# 9-5086-1)
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Westinghouse Non-Proprietary Class 3 B-5 B.2 NOZZLE COOLDOWN PRESSURE-TEMPERATURE LIMITS Allowable pressures are detennined for a given temperature based on the fracture toughness of the limiting nozzle material along with the appropriate pressure and thermal stress intensity factors. The Surry Units 1 and 2 nozzle fracture toughness used to determine the P-T limits is calculated using the limiting inlet and outlet nozzle ART values from Table B-3. The stress intensity factor correlations used for the nozzle comers are provided in Oak Ridge National Laboratory study, ORNL/TM-2010/246 [Ref.
B-9], and are consistent with ASME PVP2011-57015 [Ref. B-10]. The methodology includes postulating an inside surface 1/4T nozzle comer flaw, and calculating through-wall nozzle comer stresses for a cooldown rate of 100°F/hour.
The through-wall stresses at the nozzle comer location were fitted based on a third-order polynomial of the form:
where, cr = through-wall stress distribution x = through-wall distance from inside surface Ao, A 1, A2, A 3 = coefficients of polynomial fit for the third-order polynomial, used in the stress intensity factor expression discussed below The stress intensity factors generated for a rounded nozzle comer for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010/246. The stress intensity factor expression for a rounded corner is:
K, = ,/na [o.706Ao+ 0.537 (27ta) A1+ 0.448 (~) A2+ 0.393 (::) A3]
where, K1 stress intensity factor for a circular corner crack on a nozzle with a rounded inner radius corner a crack depth at the nozzle corner, for use with 1/4T (25% of the wall thickness)
The reactor vessel nozzle P-T limit curves for Surry Units 1 and 2 are shown in Figures B-1 and B-2, respectively, based on the stress intensity factor expression discussed above. Also shown in these figures are the current Surry Power Station Technical Specification (TS) beltline cooldown P-T limit curves [Ref.
B-11] (without pressure adjustment+ 10% margin) (represented with the dashed lines) and the beltline cool down P-T limit curves developed in this report (represented with the solid lines). These beltline cooldown P-T limit curves are located in Figure 7-2. The current Surry Power Station curves are included in Figures B-1 and B-2 for informational purposes.
Note that the figures show the most limiting P-T limit curves of the inlet and outlet nozzle for each Unit.
The nozzle P-T limits are provided for a cooldown rate of 100 °F/hr, along with a steady-state curve.
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Westinghouse Non-Proprietary Class 3 B-6 An outside surface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve is also not provided since it would be less limiting than the cooldown nozzle P-T limit curves shown in Figures B-1 and B-2 for an inside surface flaw. Additionally, the cooldown transient is more limiting than the heatup transient since it results in tensile stresses at the inside surface of the nozzle comer.
Conclusion Based on the results shown in Figures B-1 and B-2, it is concluded that the nozzle P-T limits are bounded by the traditional cylindrical beltline curves. The minimum pressure difference between the newly developed beltline P-T limit curves and nozzle P-T limit curves is 459 psi for Surry Unit 1 and 545 psi for Surry Unit 2 (based on steady state conditions at 80°F). The minimum pressure difference between the current Surry Power Station Technical Specifications beltline P-T limit curves (plus 10% margin) and nozzle P-T limit curves is 568 psi for Surry Unit 1 and 654 psi for Surry Unit 2 (based on steady state conditions at 80°F). Therefore, the P-T limits provided in Section 6 and in the current Surry Power Station Technical Specifications [Ref. B-11] remain limiting for the beltline and non-beltline reactor vessel components.
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Westinghouse Non-Proprietary Class 3 B-7 2500 IUnacceptable Operation I , ,
I I 2250 l Limiting Nozzle I l'4 ~:...-- Cooldown (-100) 2000 I
..-- Limiting Nozzle IJ 1750 I Steady State I
( !)
en 1500 I I I
I I ll..
Q)
~
I 1/)
1250 J Sri 1/)
I Q)
J
~
ll..
I "O
Q)
I 1 1 ns I 1000
, I
~ Acceptable ns I Operation
~II
(.)
I
-~ ~~/
750 I I
Solid Lines: New P-T Limits
~~ Dashed Lines : TS+ 10% Margin 500
~ -*
250
~ :I\\*~7. -40 Cooldown Rates:
Steady-State black (ss)
-20°F/Hr orange
~
-40°F/Hr green
-60°F/Hr blue
-100°F/Hr red 0
I I
' I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure B-1 Comparison of Surry Unit 1 Beltline Cooldown P-T Limits (Including Current P-T Limits without Pressure Adjustment+ 10% Margin and New 68 EFPY P-T Limits) to 68 EFPY Nozzle P-T Limits, Without Margins for Instrumentation Errors Note: "New P-T Limits" determined in Section 6. TS = Technical Specifications.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-8 2500 I Unacceptable Operation I
2250 I
J Limiting Nozzle .1
... ICooldown (-100) 2000 I
1750
__J Limiting Nozzle 1 Steady State j I C)
Cl) a..
1500 ,I I I I Q)
J J
Sri j 11' 11' Q) 1250 I a..'-
"'O Q) 1 \~
l'O
J 1000 (J ,
l'O Acceptable 0 Operation 750 '~ ~I!
500
~~~
~-----
__ \...7'_ -
~J Solid Lines: New P-T Limits Dashed Lines : TS+ 10% Margin 250 a :1\\ -M l~ -40 Cooldown Rates:
Steady-State black (ss)
-20°F/Hr orange
~
-40°F/Hr green
-60°F/Hr blue
-100°F/Hr red I I I 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure B-2 Comparison of Surry Unit 2 Beltline Cooldown P-T Limits (Including Current P-T Limits without Pressure Adjustment+ 10% Margin and New 68 EFPY P-T Limits) to 68 EFPY Nozzle P-T Limits, Without Margins for Instrumentation Errors Note: "New P-T Limits" determined in Section 6. TS= Technical Specifications.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-9 B.3 REFERENCES B-1 NRC Regulatory Issue Summary 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," U.S.
Nuclear Regulatory Commission, October 2014. [ADAMS Accession Number ML14149A165}
B-2 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.
B-3 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
B-4 Pressurized Water Reactor Owners Group (PWROG) Report PWROG-16045-NP, Revision 0, "Determination ofUnirradiated RTNoT and Upper-Shelf Energy Values of the Surry Units 1 and 2 Reactor Vessel Materials," March 2017.
B-5 BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.
B-6 ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, July 1982.
B-7 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3 , "Fracture Toughness Requirements," Revision 2, U.S . Nuclear Regulatory Commission, March 2007.
B-8 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],
November 2014. [ADAMS Accession Number ML14318A177}
B-9 Oak Ridge National Laboratory Report, ORNL/TM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles -
Revision l," June 2012. [ADAMS Accession Number ML110060164}
B-10 ASME PVP2011-57015, "Additional Improvements to Appendix G of ASME Section XI Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerville, July 2011.
B-11 Surry Power Station Technical Specifications, Section 3.1.B, Amendments Nos. 285 and 285.
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Westinghouse Non-Proprietary Class 3 C- 1 APPENDIX C OTHER RCPB FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [Ref. C-1), requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature requirement (LST) for all RCPB components, which is specified in NB-3211 and NB-2332(b) of the Section III ASME Code, is the relevant requirement that would affect the pressure-temperature (P-T) limits. This requirement is applicable to ferritic materials outside of the reactor vessel with a nominal wall thickness greater than 2 Yi inches, such as piping, pumps and valves [Ref. C-2). The Surry Unit 1 and 2 reactor coolant systems do not contain ferritic materials in the Class 1 piping, pumps and valves per Section 4.4 of this report. Therefore, the LST requirements of NB-2332(b) and NB-3211 are not applicable to the Surry Units 1 and 2 P-T limits.
The other ferritic RCPB components that are not part of the reactor vessel beltline or extended beltline in Surry Unit 1 and 2 consist of the replacement reactor vessel closure heads, replacement steam generators, and pressurizers.
The replacement reactor vessel closure head materials have been considered in the development of the P-T limits, see Section 4.5 of this report for the relevant inputs. Additionally, the Unit 1 replacement reactor vessel closure bead was constructed to the French Construction Code (RCC-M) 1993 Edition with 1st Addenda June 1994, 2nd Addenda June 1995, 3rd Addenda June 1996 and Modification Sheets FM 797, 798, 801, 802, 803, 804, 805, 806, and 807. The sizing calculations, stress and fatigue analysis were performed to ASME Code Section III 1995 Edition through 1996 Addenda. The Design Report and Report of Reconciliation (References 14 and 15 of [C-3)) certify that the closure head meets the design requirements for the ASME Code Section III 1995 Edition through 1996 Addenda. The Unit 2 replacement reactor vessel closure head was constructed to the 1995 Edition through 1996 Addenda Section Ill ASME Code and met all applicable requirements at the time of construction.
The replacement steam generators were constructed to the 1974 Edition through Winter 1976 Addenda .
Section III ASME Code and met all applicable requirements at the time of construction. Therefore, no further consideration is necessary for these components with regards to P-T limits.
The pressurizers were constructed to the 1965 Edition through Winter 1965 Addenda Section III ASME Code and met all applicable requirements at the time of construction. No further consideration is necessary for these components with regards to P-T limits.
C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.
C-2 ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, "Class 1 Components."
C-3 Surry Power Station Updated Final Safety Analysis Report (UFSAR), Revision 48, "Chapter 4:
Reactor Coolant System," September 2016.
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Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D LTOP SYSTEM ENABLE TEMPERATURE D.1 ASME CODE CASE N-641 ASME Code Case N-641 [Ref. D-1] presents alternative procedures for calculating pressure-temperature relationships and low temperature overpressure protection (LTOP) system effective temperatures, Te, and allowable pressures. The procedures provided in Code Case N-641 take into account alternative fracture toughness properties, circumferential and axial reference flaws, and plant-specific LTOP effective temperature calculations.
Per ASME Code Case N-641, the LTOP system shall be effective below the higher temperature determined in accordance with (1) and (2) below. Alternatively, LTOP systems shall be effective below the higher temperature determined in accordance with (1) and (3) below.
(1) a coolant temperature<a> of200°F (2) a coolant temperature<a) corresponding to a reactor vessel metal temperature, for all vessel beltline materials, where Te is defined for inside axial surface flaws as RTNDT + 40°F, and Te is defined for inside circumferential surface flaws as RT DT - 85°F.
(3) a coolant temperature<a> corresponding to a reactor vessel metal temperature, for all vessel beltline materials, where Te is calculated on a plant-specific basis for axial and circumferential reference flaws using the following equation:
Te= RTNDT + 501n [((F
Mm(pRi / t))- 33.2) / 20.734]
Where, F = 1.1, accumulation factor for safety relief valves Mm = the value of Mm determined in accordance with G-2214.1 , ,/in.
p = vessel design pressure, ksig Ri = vessel inner radius, in.
t = vessel wall thickness, in.
Notes:
(a) The coolant temperature is the reactor coolant inlet temperature.
(b) The vessel metal temperature is the temperature at a distance 1/4 of the vessel section thickness from the clad/base metal interface in the vessel beltline region. RT NDT is the highest adjusted reference temperature (for weld or base metal in the beltline region) at a distance 1/4 of the vessel section thickness from the vessel clad/base metal interface as determined by Regulatory Guide 1.99, Revision 2 [Ref. D-2).
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Westinghouse Non-Proprietary Class 3 D-2 Using the ASME Code Case N-641 equations and the following inputs, the Suny Units 1 and 2 LTOP system minimum enable temperature using Cases 2 and 3 was determined.
RTNDT = 219.4°F for 68 EFPY (at l /4T per Table 5-7)
F = 1.1
= 2.627 ...Jin. (See Section 4 for equations used to calculate Mm) p = 2.485 ksig (see Section 4.6, Item 6)
= 78.5 in. (see Section 4.6, Item 4)
= 8.05 in. (see Section 4.6, Item 4)
The LTOP system shall be effective below the higher temperature determined in accordance with (1) and (2) above, which has been determined to be 273°F for 68 EFPY. Alternatively, LTOP systems shall be effective below the higher temperature determined in accordance with (1) and (3) above, which has been determined to be 262°F for 68 EFPY. Therefore, the minimum required enable temperature (without margins for instrument uncertainty) for the Suny Units 1 and 2 reactor vessel is 262°F for 68 EFPY.
D.2 ASME CODE CASE N-514 The LTOP enable temperature can also be calculated based on ASME Code Case N-514 [Ref. D-3]. Per ASME Code Case N-514, the LTOP system shall be effective below the higher temperature determined in accordance with (A) and (B) below.
(A) a coolant temperature<a) of 200°F (B) a coolant temperature(a) corresponding to a reactor vessel metal temperature(b) less than RT NOT +
50°F (a) The coolant temperature is the reactor coolant inlet temperature.
(b) The vessel metal temperature is the temperature at a distance 1/4 of the vessel section thickness from the inside wetted surface in the vessel beltline region. RT NDT is the highest adjusted reference temperature (for weld or base metal in the beltline region) at a distance 1/4 of the vessel section thickness from the vessel wetted inner surface as detennined by Regulatory Guide l.99, Revision 2. For the purpose of this calculation, the inside wetted surface is taken to be the clad/base metal interface.
Using the ASME Code Case N-514 equations and an RTNoT value of219.4°F for 68 EFPY (at l /4T per Table 5-7), the Surry Units 1 and 2 LTOP system enable temperature using Cases (A) and (B) was determined to be 283 °F.
WCAP-18243-NP October 20 l 7 Revision 0
Westinghouse Non-Proprietary Class 3 D-3 D.3 LTOP ENABLE TEMPERATURE CONCLUSION The Surry Power Station Technical Specifications [Ref. D-4) specifies an arming temperature of 350°F, which is conservative and remains valid for the Surry SLR period of operation. The margin to the 350°F value is sufficient to cover uncertainties utilizing either the Code Case N-641 [Ref. D-1) methodology or the more conservative Code Case N-514 [Ref. D-3) methodology.
D.4 REFERENCES D-1 ASME Code Case N-641 , "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 1," ASME International, January 17, 2000.
D-2 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
D-3 ASME Code Case N-514, "Low Temperature Overpressure Protection Section XI, Division l,"
ASME International, dated February 12, 1992.
D-4 Technical Specifications LCO 3.1.G.l.c.(4), Virginia Electric and Power Company, Docket No.
50-280, "Surry Power Station, Unit 1 Renewed Facility Operating License" Amendments 248 and 247.
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Westinghouse Non-Proprietary Class 3 E-1 APPENDIX E WELD MATERIAL HEAT# 0227 INITIAL RTNDT 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 E-1 provides Charpy V-notch test data taken from the Record of Weld Material Qualification for Heat# 0227, Grau Lo flux LW320 per Certified Material Test Reports (CMTRs). Table E-2 provides supplemental Charpy V-notch test data also obtained from CMTRs. Table E-3 provides the Charpy V-notch test data taken from Reference E-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 E-1 Weld Material Qualification Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227)<*>
Temperature Temperature(b> Energy EnergyCb>
(OC) (OF) (kgm/cm 2) (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 [Ref. E-l ].
2 14 223 1 Energy (ft-lbs)= Energy (kgm/cm2) * * .b :cm kg*in
3 *28m ft * (0.315 in.
0.394 in.)
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-2 Table E-2 Supplemental Charpy V-Notch Test Data for Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227)<*>
Temperature Temperature<hl Energy Energy<hl (OC) (OF) (kgm/cm 2) (ft-lb)
- 12 10 8.1 47
- 12 JO 5.5 32
- 12 10 6.6 38
-1 2 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 JO 10.4 60
-12 10 8.5 49
- 12 10 9.5 55
-1 2 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 utili zing the fo rmula below. Note that 0.3 15 inch and 0.394 inch are the nominal dimensions of the Charpy specimen cross section per WCAP-8085 [Ref. E-1].
Energy (ft-lbs)= Energy (kgm/cm2)
14 223 1.b*,cm'
3 *2 0 f t * (0.315 in.
0.394 in.)
kg
Ln m WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-3 Table E-3 Charpy V-Notch Test Data for Surry Unit 2 Surveillance Weld (Heat# 0227)(a)
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 [Ref. E-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 Surry Unit 2 Intermediate to Lower Shell Circumferential Weld initial material properties.
Using the data summarized in Tables E-1 through E-3, the initial RTNDT value for the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat # 0227) must be determined using NUREG-0800, BTP 5-3 guidance [Ref. E-2] and in accordance with the ASME Code Section III, Subarticle NB-2331 requirements [Ref. E-3].
Following NUREG-0800, BTP 5-3 Position 1.1(1) guidance, TNDT "may be assumed to be the temperature at which 41 J (30 ft-lbs) was obtained in Charpy V-notch tests, or -18°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 E-1 through E-
3) at each Charpy V-notch test temperature were used as input to the curve-fitting software, in accordance WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-4 with ASME Code Section III, 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, TNDT for this weld material is set equal to 0°F per BTP 5-3 Position 1.1(1).
This estimate of T NOT and the Charpy V-notch test data in Tables E-1 through E-3 are used to determine RTNOT* Following the requirements of ASME Code Section III, Subarticle NB-2331 , Paragraph (a)(2), the Charpy V-notch test data is first checked at a temperature equal to the drop-weight T NOT 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 T NDT + 60°F. However, multiple Charpy V-notch tests were conducted at T NOT + 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 TNoT limited. For completeness, 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 E-1 through E-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 CVGRAPH figures are contained in the following pages for Charpy V-notch 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 RTNOT*
Tso ft-lb = 32.1°F T35 mils= 5.0°F TCv = Max [T50 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 RTNOT is the higher of TNDT (determined from the drop-weight tests) and Tcv (determined above) minus 60°F.
RTNDT = Max [TNDT, Tcv- 60°F]
RTNOT= Max [0°F, 32.1 - 60°F)
RT NOT= Max [0°F, -27.9°FJ RTNOT= 0°F Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat# 0227) Initial RT:ti!IT = 0°F WCAP-18243-NP October 201 7 Revision 0
Westinghouse Non-Proprietary Class 3 E-5 The current 10 CFR 50, Appendix G [Ref. E-4], requirements specify that USE be calculated based on ASTM El85-82 [Ref. E-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 E185 manual (2016 version). Using the guidelines in ASTM El85-82 and El85-16 [Ref. E-6], the average of all Charpy data 2: 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-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-6 Surry Unit 2 Intermediate to Lower Shell Circumferential Weld CVGraph 6.02: Hyperbolic Tangent Curve Printed on 5/ 18/2017 IA7 PM A = 35.85 B = 33.65 C = 60.43 TO = 5.00 D = 0.00 Correlation Coefficient = 0.%2 Equation is A+ B
JTanh(tT-TO)/(C+D1))]
Upper Shelf Energy = 69.50 (Fixed) Lower Shelf Energy = 2.20 (Fixed)
Tcmp1!l30 ft-lbs= -5.60° F Tcmp(@.15 fl-lbs= 3.50° F Tcmp@5 0 fl- lbs= 32.10° F Plant: Surry 2 Material : WELD Heat: 0227 Orientation: N/A Capsule: Unirradinted Flucncc: O.OOE+ooO n/cm' 100 Ir\
90 80
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CVGraph 6.02 05/JR/2017 Page 1/2 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-7 Plant: Surry 2 Ma terial: WEID Heat: 0227 Orientation: N/A Capsule: Unirradiattd Flue nee: O.OOE+000 n/cm' Surry Unit 2 Intermediate to Lower Shell Circumferential Weld Charpy V-Notch Data Temperature (° F) InputCVN Computed CVN Diffcre.ntial
-1 00 7.0 4.2 2.78
-40 15.5 14.6 0.9 1
-20 27.5 I 22.7 4.83 10 32.0 38.6 -6.63 40 50.0 53.4 -3.41 73 75.0 63. 1 11.91 2 10 69.5 69.4 0.08 300 91.0 69.5 2 1. 50 CVGraph 6.02 05/18/20 17 Page 212 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-8 Surry Unit 2 Intermediate to Lower Shell Circumferential Weld CVGmph 6.02 : Hyperbolic Tangent Curve Printed on 5/ 18/2017 1:57 PM A= 33.50 B= 32.50 C = 59.91 TO= 2.18 D = 0.00 Correla tion Coefficient = 0.980 Eqtmtion is A+ B * [Ta nh((T-TO)/(C+ D1))]
Upper Shelf L.E. = (,6 00 (Fixed) Lower Shelf L.E. = 1.00 (Fixed)
Tcmp@35 mils= 5.00° F Plant: Surry 2 Material: WELD Heat: 0227 Orientation: N/A Capsule : Unirradiated Flucncc: O.OOE+-000 n/cm' 90
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CVGraph 6.02 05/JR/2017 .Page 1/2 WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-9 Plant: Surry 2 Material: WEID Heat: 0227 Orientation: N/A Capsule: Unirradiated Fluence: O.OOE+OOO n/cm' Surry Unit 2 Intermediate to Lower Shell Circumferential Weld Charpy V-Notch Data Temperature (0 F) Input L. E. Computed L. E. Differential
- 100 3.0 3. 1 ..0.08
-40 15.0 13. 8 1.23
-20 25.0 22.0 3.01 10 33.0 I 37.7 -4.72 40 50.0 51.7 . ] .67 73 68.0 60.4 7.59 210 66.0 65 .9 0.06 300 81.0 66.0 15.00 CVGraph 6.02 05/ 18/2017 Page 212 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-10 E.1 REFERENCES E-1 Westinghouse Report WCAP-8085, Revision 0, "Virginia Electric & Power Co. Suny Unit No. 2 Reactor Vessel Radiation Surveillance Program," June 1973 .
E-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.
E-3 ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subarticle NB-2300, "Fracture Toughness Requirements for Material."
E-4 Code of Federal Regulations, 10 CFR 50, Appendix G, "Fracture Toughness Requirements," U.S.
Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.
E-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.
E-6 ASTM El85-16, "Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels," ASTM International, December 2016.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-1 APPENDIX F
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) [Ref. F-1] were first approved in WCAP-14177 [Ref. F-2] for End of License (EOL) and have applicability that was extended to 48 EFPY (per Reference F-12). Table F-1 contains a summary of the applicability of the Surry Units 1 and 2 P-T Limit curves. Figures F-1 and F-2 show the Surry Units 1 and 2 heatup and cooldown curves as currently depicted in the Surry Power Station Technical Specifications [Ref. F-12]. Tables F-2 and F-3 provide the data points corresponding to the heatup and cooldown curves, respectively, as currently depicted in the Surry Power Station Technical Specifications.
Table F-1 Surry Units 1 and 2 P-T Limit Curve Applicability History Subject Content Relevant to Surry Units 1 and 2 P-T Reference Date Document(s) Limit Curves Number(s)
P-T limit curves for 28.8 EFPY for Surry Unit 1 and 29.4 EFPY for Surry Unit 2 were created without WCAP-14177, inclusion of instrumentation errors. Note that this October F-2 Revision 0 evaluation pre-dates the first approval of 1994 Westinghouse's current NRC-approved methodology in WCAP-14040-A, Revision 4 [Ref. F-3].
Per the subject documents, an adjustment of21.5 psi to accommodate for the pressure difference between SM-792, Revision 3 the pressurizer and reactor beltline was applied to the WCAP-14177 curves to create the TS curves.
(Page 18/47)
Additionally, the WCAP-14177 heatup curves are 1995-1996 F-4 and F-5 combined into one bounding heatup curve at SM-945, Revision 0 temperatures of 3 l 5°F and above for the TS. These (Page 26/102) 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 F-6 (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.
NRC approved the P-T limits based on WCAP- December Letter from the NRC F-7 14177 through amendment No. 207. 1995 WCAP-15130, P-T limit curves for End of License Extension April 2001 F-8 Revision 1 (EOLE) were developed.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-2 Table F-1 Surry Units 1 and 2 P-T Limit Curve Applicability History Subject Content Relevant to Surry Units 1 and 2 P-T Reference Date Document(s) Limit Curves Number(s)
The TS P-T Limits were changed to curves based on January Letter from the NRC F-9 WCAP-15130, Revision 1. 2006 The P-T Limits approved under amendment No. 207 Letter from the NRC June 2006 F-10 were reinstated in the TS .
The applicability of the P-T Limits approved under Letter from the NRC May 2011 F-11 amendment No. 207 was extended to 48 EFPY.
This reference represents the most recent TS P-T Letter from the NRC limit curve amendment (No. 285), which June 2015 F-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 1 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-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-3 Surry Units 1 and 2 Reactor Coolant System Heatup Li-mitations Material Property Basis Limiting Material: Surry'Unit 1 Intermediate to Lower Shell Circ Weld Li miting ART Values for Surry 1 at 48 EFPY: 1i 4-T. 228.4°F 3/4* , 18!!.5°F Limiting Boltup Temperature Surry 1 Initial RT"°' Closure Flange Regio n: l06 F 2.500.00 ,..,_,..,....,...,.....,......,.,,.....,;..,.....,..,..,....,..,..,..,..,.._.,."TT,...,....,..,..,.....-,-~.,..,...,..,.....,.,~,...,......,...,-..,...,....,...,....,...,,..,...,...,.........,
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Figure 3. 1- J : Surry Units I and 2 Reactor Coolant System Heatup Limitations (Heamp Rates up lO 60°F/hr) Applicable for 48 EFPY Amendment Nos. 285, 285 Figure F-1 Surry Units 1 and 2 Heatup P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications [Ref. F-12)
WCAP-1 8243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-4 Surry Units 1 and 2 Reactor Coola.n t .System *cooldown limitations Materlal Property Basis . .
limiti ng Material: Surry Unit 1 lntermediate 'to Lower Shell Circ weld Umiting ART Valves for Surry 1 .tt 48 EFl>Y: 1/4-T, 228.4°f
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Figure 3.1-2 : Surry Units l and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 48 EFPY AmendmenL Nos. 285, 285 Figure F-2 Surry Units 1 and 2 Cooldown P-T Limit Curves as Depicted in the Surry Power Station Technical Specifications [Ref. F-12]
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-5 Table F-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) 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 .31 215 650.1 0 215 642.41 215 577.91 220 664.06 220 660.65 220 593.48 225 679.05 225 679.05 225 610.40 230 695.02 230 695 .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 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-6 Table F-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 (OF) 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 11 45 .06 305 11 36.13 310 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 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-7 Table F-3 Data Points for Surry Units 1 and 2 Current Technical Specifications Cooldown P-T Limit Curves Steady-State 20°F/hr Cooldown<a) 40°F/hr Cooldown<a> 60°F/hr Cooldown<a) 100°F/hr Cooldown<a>
T (OF) P (psig) T (OF) P (psig) T (°F) P (psig) T (°F) P (psig) T (OF) P (psig) 80 491.03 80 449.28 80 406.68 80 363.08 80 272.64 85 492.91 85 451.20 85 408.55 85 364.90 85 274.35 90 495.04 90 453.26 90 410.56 90 366.88 90 276.26 95 497.32 95 455 .5 1 95 412.78 95 369.07 95 278.43 100 499.77 100 457.93 100 415.17 100 371.44 100 280.80 105 502.41 105 460.56 105 417.80 105 374.07 105 283.47 110 505 .25 110 463 .38 110 420.62 110 376.91 110 286.38 11 5 508.30 115 466.45 115 423 .72 115 380.04 115 289.62 120 511.58 120 469.75 120 427.05 120 383.42 120 293.15 125 515.10 125 473 .33 125 430.69 125 387.14 125 297.07 130 518.89 130 477.17 130 434.60 130 391.15 130 301.33 135 522.97 135 481.33 135 438 .80 135 395.53 135 306.02 140 527.35 140 485.81 140 443.39 140 400.27 140 311.12 145 532.06 145 490.65 145 448.38 145 405.37 145 316.68 150 537.12 150 495 .77 150 453.76 150 410.94 150 322.74 155 542.46 155 501.40 155 459.60 155 417.02 155 329.39 160 548.3 1 160 507.45 160 465.88 160 423 .56 160 336.58 165 554.60 165 513.99 165 472.69 165 430.68 165 344.44 170 561.37 170 521.01 170 480.02 170 438.28 170 352.94 175 568.64 175 528.61 175 487.96 175 446.61 175 362.16 180 576.47 180 536.76 180 496.41 180 455 .58 180 372.17 185 584.86 185 545.46 185 505 .66 185 465.32 185 383.07 190 593.79 190 554.93 190 515.60 190 475.80 190 394.84 195 603.50 195 565.14 195 526.36 195 487.16 195 407.57 200 613.95 200 576.12 200 537.82 200 499.30 200 421.38 205 625. 19 205 587.83 205 550.33 205 512.54 205 436.28 210 637.24 210 600.55 210 563 .77 2 10 526.80 210 452.44 215 650.10 215 614.27 215 578 .30 215 542.11 215 469.96 220 664.06 220 629.02 220 593 .79 220 558.70 220 488.86 225 679.05 225 644.76 225 610.66 225 576.64 225 509.23 230 695.02 230 661.84 230 628.80 230 595.82 230 531.28 235 712.37 235 680.23 235 648.22 235 616.67 235 555.04 240 730.98 240 699.86 240 669.26 240 638.97 240 580.76 245 750.86 245 721.17 245 691.79 245 663.19 245 608.44 250 772.41 250 743 .88 250 71 6.20 250 689.09 250 638.25 255 795.35 255 768.54 255 742.32 255 717.23 255 670.62 260 820.26 260 794.84 260 770.62 260 747.30 260 705.32 265 846.77 265 823.36 265 800.90 265 779.91 265 742.77 270 875.50 270 853 .82 270 833.62 270 814.86 270 783 .25 275 906.20 275 886.57 275 868.75 275 852.48 275 826.80 WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-8 Table F-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<*l T (°F) P (psig) T {°F) P (psig) T (°F) P (psig) T {°F) P (psig) T (OF) P (psig) 280 939.14 280 922.02 280 906.49 280 892.94 280 873 .65 285 974.78 285 959.97 285 947.11 285 936.54 285 924.18 290 1012.91 290 1000.71 290 990.76 290 983.61 290 978.48 295 1053.86 295 1044.51 295 1037.77 295 1034.14 295 1036.84 300 1097.82 300 1091.61 300 1088.29 300 1088.27 305 1145.06 305 1142.24 305 1142.68 310 1195.82 315 1250.37 320 1308.86 325 1371.62 330 1438.89 335 1511.21 340 1588.69 345 1671.46 350 1760.72 355 1856.03 360 1958.14 365 2067.3 2 370 2184.34 375 2308.98 380 2442.42 Note:
(a) The 20°F/hr and 40°F/hr cooldown curves are identical to the steady-state curve at 3 10°F and above. The 60°F/hr cooldown curve is identical to the steady-state curve at 305°F and above. The 100°F/h r cooldown curve is identi cal to the steady-state curve at 300°F and above.
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Westinghouse Non-Proprietary Class 3 F-9 F.1 REFERENCES F-1 Surry Power Station Technical Specifications, Section 3.1.B, Amendments Nos. 285 and 285.
F-2 Westinghouse Report WCAP-14177, Revision 0, "Surry Units l and 2 Heatup and Cooldown Limit Curves for Normal Operation," October 1994.
F-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.
F-4 Virginia Power Calculation SM-792, Revision 3, "Surry 1 & 2 Composite PIT Limits Curve,"
January 1996.
F-5 Virginia Power Calculation SM-945, Revision 0, "Surry Unit 1 and 2 Heatup/Cooldown Curves and LTOPS Setpoint," February 1995.
F-6 Letter 95-197 from Virginia Electric and Power Company to the Nuclear Regulatory Commission, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Request for Exemption - Code Case N-514, Proposed Technical Specifications Change, Revised Pressureffemperature Limits and LTOPS Setpoint," dated June 8, 1995.
F-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. [ADAMS Accession Number MLOJ 2710054)
F-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.
F-9 Letter from the NRC to Virginia 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]
F-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}
F-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 MLI 111 OAJ 11J F-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 MLJ 51 73AJ02]
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Westinghouse Non-Proprietary Class 3 G-1 APPENDIX G CREDIBILITY EVALUATION OF THE SURRY UNITS 1 AND 2 SURVEILLANCE DATA G.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [Ref. G-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.
G.2 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" [Ref. G-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 G-2 The Suny 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-l and C4415-2
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 L1 (Heat # 8T1554, Linde 80 Flux Type, Lot 8579) and L2 (Heat# 299L44, Linde 80 Flux Type, Lot 8596).
The Suny Unit 2 reactor vessel beltline region traditionally consists of the following materials:
1. Intermediate Shell Plates C433 l-2 and C4339-2
2. Lower Shell Plates C4208-2 and C4339-1
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 L1 (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 [Ref. G-3] and WCAP-8085 Revision O [Ref. G-4], the Suny Units 1 and 2 respective surveillance programs were developed to the requirements of ASTM E185 . 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 Suny Unit 1 reactor vessel beltline materials. Each of the beltline base metal materials has similar chemical properties. Lower Shell Plate C4415-1 has the highest initial RTNoT value (other than the Upper Shell Forging), and is also representative of Lower Shell Plate C44 l 5-2 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 Suny Unit 1 beltline weld materials has similar USE and low initial RT NOT values. Since Heat # 299L44 has the WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 G-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 RTNDT value.
Since Lower Shell Plate C4339-1 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 I beltline weld materials has similar chemical properties and low USE values. Since Heat # 0227 has the highest initial RTNDT 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 I is met for the Surry Units I 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
[Ref. G-5] and WCAP-1600 I [Ref. G-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 I and 2 surveillance programs meet this criterion.
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Westinghouse Non-Proprietary Class 3 G-4 Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of liRTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°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 El 85-82 [Ref. G-7).
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 liRTNoT 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 [Ref. G-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- l
Surry Unit 2 Lower Shell Plate C4339-1
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 G-5 Credibility Assessment Case 1: Lower Shell Plate C44 l 5-l, Lower Shell Plate C4339-l, and Weld Heat # 0227 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 1 and 2 as summarized in Tables G-1 and G-2. Note that when evaluating the credibility of the surveillance weld data, the measured L'.1RTNOT 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 C44 l 5-1 surveillance material data and credibility conclusions pertain to the Lower Shell Plate C4415-l and to Lower Shell Plate C4415-2 (same material heat). The Surry Unit 1, Case 1, chemistry factor is summarized in Table G-1.
Table G-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation for Surry Unit 1 Capsule Fluence<a> (c)
FFb) ARTNDT Material Capsule (x 10 19 FF* ARTNDT(°F) FF2 (Of) n/cm2, E >
1.0 MeV)
Lower Shell Plate T 0.271 0 .644 50 32.21 0.415 C4415-l V 1.80 1.161 11 3 131.23 1.349 (Longitudinal) X 2.11 1.203 86 103.46 1.447 SUM: 266.91 3.211 CF c4415
1= r(FF
LlRTNDT) -c- L(FF )
2
= (266.91) -c- (3 .211) = 83.1°F Notes:
(a) Capsule fluence values taken from Section 2.
(b) FF = fluence factor = t<0*28
0* 10 ' 10gl)_
(c) t,RTNDT values obtained from Table 7-6 of BA W-2324 [Ref. G-5].
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Westinghouse Non-Proprietary Class 3 G-6 Surry 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 Surry Unit 2 Intermediate to Lower Shell Circumferential Weld. Surry Unit 2, Case 1, chemistry factors are summarized in Table G-2.
Table G-2 *Calculation of Interim Chemistry Factors for the Credibility Evaluation for Surry Unit 2 Capsule Fluence<*> (c)
Material Capsule (x 101 9 FFb) ~TNDT FF*~TNDT FF 2
(Of) {°F) n/cm2, E >
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 92.9 1 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 2
CF C4339-I = I(FF
L\.RT NDT) + I(FF ) = (519.59) + (6.860) = 75.7°F Surveillance Weld X 0.297 0.668 95 .65 63 .86 0.446 Material V 1.89 1.174 140.2 1 164.64 1.379 (Heat# 0227) y 2.72 1.267 178.32 225.94 1.605 SUM: 454.45 3.430 2
CF Heat #0227 = L(FF
L\.RTNDT) + I(FF ) = (454.45) + (3.430) = 132.5°F Notes:
(a) Capsule fluence values taken from Section 2.
(b) FF = fluence facto r = t<0*28 - o.IO'log I)_
(c) t.RTNoT values obtained from Table 5- 12 ofWCAP-16001 [Ref. G-6] .
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Westinghouse Non-Proprietary Class 3 G-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 Longihtdinal 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 G-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 LlRTNDT values and the interim CF for weld Heat # 299L44 data from all sources, as shown in Table G-4.
Table G-3 Mean Chemical Composition and Temperature for Weld Heat # 299L44<*>
Cu Ni Inlet Temperature during Temperature Material Capsule Wt.% Wt.% Period of Irradiation (°F) Adjustment (°F)
Weld Metal Heat T 537 -13
299L44 V 0.23 0.64 539 -11 (Surry Unit 1 Data) X 542 -8 TMI2-LG I (CR-3) 556 6 0.37 0.70 Wl(CR-3) 545 -5 Weld Metal Heat TMII-E 556 6
299L44 TMII-C 0.33 0.67 556 6 (Other Plant Data) TMI2-LGl(TMI-2) 556 6 CR3-LG I (ONS-3) 0.36 0.70 556 6 AS 0.23 0.64 556 6 MEAN 0.30 0.67 550 Note:
(a) Data obtained from Table 3-7 or calculated herein.
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Westinghouse Non-Proprietary Class 3 G-8 Table G-4 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat # 299L44 Capsule Chemistry Fluence Adjusted FF*Adj usted ARTNDT Capsule Factor (x 1019 FF<*> ARTNDT(b) ARTNDT FF2 2
(OF)
Position 1.1 n/cm , E > (OF) (OF) 1.0 MeV)
T 175.8 0.27 1 0.644 171 184.9 119.10 0.415 V 175.8 1.80 1.1 61 250 279.6 324.75 1.349 X 175.8 2. 11 1.203 234 264.4 3 18.1 1 1.447 TMI2-LG l(CR-3) 234.0 0.830 0.948 2 16 195.4 185. 15 0.898 WI 234.0 0.780 0.930 262 226.2 210.40 0.865 TMII-E 2 15.2 0.107 0.431 74 76.0 32.72 0.185 TMII-C 215.2 0.882 0.965 166 163.4 157.65 0.93 1 TMI2-LG 1(TMl-2) 2 15.2 0.968 0.991 226 220.4 2 18.39 0.982 CR3-LGI 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: 2 11 3.67 9.550 CF Heat #299L44= :E(FF * ~RTNoT) 7 :E(FF2) = (2113.67) 7 (9.5 50) = 221.3°F Notes:
(a) FF = fl uence factor = t\0. 28
o. io*Jog IJ.
(b) Adjusted D.RTNDT va lues are D.RTNDT values adjusted first to the mean operating temperature using the temperature adjustments in Table G-3, then to the mean chemical composition using the ratio procedure.
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Westinghouse Non-Proprietary Class 3 G-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 1 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 G-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.RT NDT values and the interim CF for Weld Heat # 72445 data from all sources, as shown in Table G-6.
Table G-5 Mean Chemical Composition and Temperature for Weld Heat# 72445<*>
Inlet Temperature Temperature Cu Ni Material Capsule during Period of Adjustment(°F)
Wt.% Wt.%
Irradiation (°F)
CR3-LG1 0.22 0.59 556 11 CR3-LG2 0.22 0.59 556 11 Weld Metal Heat WI 0.22 0.59 545 0
72445 Point Beach Unit I : Capsule V 0.23 0.62 542 -3 (Other Plant Point Beach Unit l : Capsule S 0.23 0.62 542 -3 Data)
Point Beach Unit l : Capsule R 0.23 0.62 541.6 -3.4 Point Beach Unit I: Capsule T 0.23 0.62 533.4 -11.6 MEAN 0.23 0.61 545 Note:
(a) Data obtained from Table 3-8 or calculated herein.
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Westinghouse Non-Proprietary Class 3 G-10 Table G-6 Calculation of Interim Chemistry Factor for the Credibility Evaluation of Weld Material Heat # 72445 Capsule Chemistry Fluence Adjusted FF* Adjusted Factor .aRTNDT (b)
Capsule (x 10 19 FF<*> .aRT DT .aRTNDT FF 2 Position (Of) n/cm2, E > (OF) (Of) 1.1 1.0 MeV)
CR3-LG1 165.5 0.5 10 0.8 12 139 154.5 125.46 0.659 CR3-LG2 165.5 1.67 1.141 164 180.3 205.72 1.303 WI 165.5 0.780 0.930 138 142 .1 132.23 0.865 PB-1 : 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 1.2 13 155 150.1 182.00 1.471 PB- I : Capsule T 172.4 2.23 1.2 17 181 167.7 204.15 1.482 SUM: 1091.3 1 7.438 2
CF Heat #72445= I:(FF
1"1RTNoT) + I:(FF ) = (1091.3 1) + (7.438) = 146.7°F Notes:
(a) FF = fluence factor= t<0 *28 -o.io*Jog ()_
(b) Adjusted LlRTNOT values are LlRTNOT values adjusted first to the mean operating temperature using the temperature adjustments in Table G-5, then to the mean chemical composition using the ratio procedure.
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Westinghouse Non-Proprietary Class 3 G-11 The scatter of ~T NDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. G-1] is presented in Table G-7 for Surry Unit 1 and in G-8 for Surry Unit 2.
Table G-7 Surry Unit 1 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line
<17°F CF Capsule Fluence Measured Adjusted<hl Predicted Scatter (Base (c)
Material Capsule (Slope best-n,) (x 10 19 n/cm 2, FF<*> ART NOT ART NOT ART NOT ART NOT Metal)
(OF) E > 1.0 MeV) (OF) (OF) (OF) (OF) <28°F (Weld)
Lower Shell Plate T 83.1 0.271 0.644 50 50.00 53.54 3.54 Yes C4415-l V 83.1 1.80 1.161 113 113.00 96.51 16.49 Yes (Longitudinal)
X 83.1 2.11 1.203 86 86.00 99.97 13.97 Yes T 221.3 0.271 0.644 171 184.86 142.58 42.28 No V 221.3 1.80 1.161 250 279.63 257.00 22.63 Yes X 221.3 2.11 1.203 234 264.42 266.23 1.81 Yes TMI2-LG1 221.3 0.830 0.948 216 195.36 209.73 14.37 Yes Surveillance WI 221.3 0.780 0.930 262 226.16 205 .87 20.29 Yes We ld Material TMTI-E 221 .3 0.107 0.431 74 76.00 95.28 19.28 Yes (Heat # 299L44)
TMTI-C 221.3 0.882 0.965 166 163.40 213 .51 50.11 No TMI2-LG1 221.3 0.968 0.991 226 220.40 219.28 1.12 Yes CR3-LGI 221.3 0.779 0.930 202 185. 12 205.80 20.68 Yes AS 221.3 2.75 1.270 246.6 295.54 280.99 14.55 Yes CR3-LG1 146.7 0.510 0.812 139 154.50 119.12 35.38 No CR3-LG2 146.7 1.67 1.141 164 180.25 167.43 12.82 Yes Surveillance WI 146.7 0.780 0.930 138 142.14 136.47 5.67 Yes Weld Material PB-1:V 146.7 0.634 0.872 107 102.96 127.97 25.01 Yes (Heat# 72445)
PB-I: S 146.7 0.829 0.947 165 160.38 138.98 2 1.40 Yes PB-1:R 146.7 2.19 1.213 155 150.08 177.90 27.81 Yes PB-I : T 146.7 2.23 1.217 181 167.71 178.58 10.87 Yes Notes:
(a) FF=fluence factor= t<0*28
0* 10 ' 1ogf)_
(b) Adjusted to mean temperature and chemistry, as applicable.
(c) Scatter i'iRT NOT= Absolute Value [Predicted i'iRT NOT - Adjusted i'iRT NOT]-
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 G-12 Table G-8 Surry Unit 2 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line
<17°F CF Capsule Flu ence Measured Adjusted<h) Predicted Scatter (Base 19 F~*> (<)
Material Capsule (Slope best-fiU (x 10 n/cm2, ARTNDT ART ' DT ARTNDT ART 'DT Metal)
{°F) E> l.OMeV) (OF) (OF) {°F) {°F) <2 8°F (Weld)
X 75 .7 0.297 0.668 59.08 59.08 50.54 8.54 Yes Lower Shell Plate C4339-l V 75.7 1.89 1.174 79.12 79. 12 88.89 9.77 Yes (Longitudinal) y 75.7 2.72 1.267 114.22 11 4.22 95 .92 18.30 No X 75.7 0.297 0.668 48.67 48.67 50.54 1.87 Yes Lower Shell Plate C4339-l V 75.7 1.89 1.1 74 63 .60 63 .60 88.89 25.29 No (Transverse) y 75.7 2.72 1.267 106.81 106.8 1 95.92 10.89 Yes X 132.5 0.297 0.668 95.65 95.65 88.47 7. 18 Yes Surveillance Weld Material V 132.5 1.89 1.174 140.21 140.2 1 155.59 15.38 Yes (Heat # 0227) y 132.5 2.72 1.267 178.32 178.32 167.88 10.44 Yes CR3-LGI 146.7 0.510 0.8 12 139 154.50 11 9. 12 35.38 No CR3-LG2 146.7 1.67 1.141 164 180.25 167.43 12.82 Yes WI 146.7 0.780 0.930 138 142. 14 136.47 5.67 Yes Surveillance Weld Material PB-I: V 146.7 0.634 0.872 107 102.96 127.97 25.01 Yes (Heat # 72445)
PB-I: S 146.7 0.829 0.947 165 160.38 138.98 21.40 Yes PB-I : R 146.7 2.1 9 1.213 155 150.08 177.90 27.8 1 Yes PB-I : T 146.7 2.23 1.217 181 167.71 178.58 10.87 Yes Notes:
(a) FF = fluence factor = t< 0*23
0* 10' 10s I)_
(b) Adjusted to mean temperature and chemistry, as applicable.
(c) Scatter 6RTNOT= Absolute Value [Predicted L'.RTNoT - Adjusted 6RTNoTJ.
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 G-7 and G-8 indicate that plate C4415- l , weld Heat# 299144, 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# 299144, weld Heat # 0227 data, and weld Heat # 72445 are deemed "credible",
and C4339-1 is deemed "non-credible" per the third criterion.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 G-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 [Refs. G-3 and G-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 [Ref. G-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 NUREG report. However, the NUREG report does not consider the most up-to-date fluence and ~T NDT values for Surry surveillance capsules. Thus, Table G-9 contains an updated calculation of Residual vs. Fast Fluence, considering the updated capsule fluence and ~RTNDT values for the Surry surveillance capsules.
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Westinghouse Non-Proprietary Class 3 G-14 Table G-9 Calculation of Residual vs. Fast Fluence for Surry Units 1 and 2 Capsule fluence Measured RG 1.99, 19 Residua1<c>
Capsule (x 10 n/cm2, FF Shift<a> Rev. i(bl
(°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 l 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 t.T30 values for the SRM were taken from Table 7-6 ofBAW-2324 [Ref. G-5) for Surry Unit I and Table 5-12 ofWCAP-16001 [Ref. G-6) for Surry Unit 2.
(b) PerNUREG/CR-6413, ORNL/TM-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 121.9°F based on Regulatory Guide 1.99, Revision 2, Position I. I. 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.
G.3 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 # 299L44 data are deemed "credible"
The Weld Heat# 72445 data are deemed "credible" WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 G-15 G.4 REFERENCES G-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.
G-2 Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.
G-3 Westinghouse Report WCAP-7723, Revision 0, "Virginia Electric and Power Co. Surry Unit No.
1 Reactor Vessel Radiation Surveillance Program," July 1971 .
G-4 Westinghouse Report WCAP-8085, Revision 0, "Virginia Electric & Power Co. Surry Unit No. 2 Reactor Vessel Radiation Surveillance Program," June 1973.
G-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.
G-6 Westinghouse Report WCAP-16001 , Revision 0, "Analysis of Capsule Y from Dominion Surry Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003 .
G-7 ASTM El85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.
G-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 MLJ 10070570}
G-9 NUREG/CR-6413; ORNL/TM-13133 , "Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials," April 1996.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 H-1 APPENDIX H COMPARISON OF AXIAL FLAW AND CIRCUMFERENTIAL FLAW P-T LIMIT CURVES Per Table 5-7, the limiting Surry Units 1 and 2 l/4T ART value at 68 EFPY corresponds to an "Axial Flaw" material, while the limiting 3/4T ART value corresponds to a "Circumferential Flaw" material. The following comparison is completed to confirm that the "Axial Flaw" methodology based heatup and cooldown limit curves bound heatup and cooldown limit curves based on the "Circumferential Flaw" methodology.
Figure 6-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 20, 40, and 60°F/hr applicable for 68 EFPY, with the flange requirements and using the "Axial Flaw" methodology and the limiting "Axial Flaw" ART values summarized in Table 5-7. Figure 6-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of O (steady-state), 20, 40, 60, and 100°F/hr applicable for 68 EFPY, with the flange requirements and using the "Axial Flaw" methodology and the limiting "Axial Flaw" ART values summarized in Table 5-7. The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G [Ref. H-1].
Figure H-1 of the Appendix presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 20, 40, and 60°F/hr applicable for 68 EFPY, with the flange requirements and using the "Circumferential Flaw" methodology and the limiting "Circumferential Flaw" ART values summarized in Table 5-7. Figure H-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of O (steady-state), 20, 40, 60, and 100°F/hr applicable for 68 EFPY, with the flange requirements and using the "Circumferential Flaw" methodology and the limiting "Circumferential Flaw" ART values summarized in Table 5-7. The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G.
Note that the "Circumferential Flaw" based heatup and cooldown limitations should not be used in plant operation based the following paragraph.
Figure H-3 shows a comparison of the heatup limit curves developed using the "Axial Flaw" methodology and the "Circumferential Flaw" methodology. Similarly, Figure H-4 shows a comparison of the cooldown limit curves using the "Axial Flaw" methodology and the "Circumferential Flaw" methodology. Figures 6-5 and 6-6 indicate that the curves based on the "Axial Flaw" methodology and the "Axial Flaw" ART values represent the most limiting heatup and cooldown limitations. Therefore, the "Axial Flaw" based heatup and cooldown limit curves, summarized in Figure 6-1 , Figure 6-2, Table 6-1 ,
and Table 6-2 are considered the limiting Surry Units 1 and 2 heatup and cooldown limits generated using the 1998 through the 2000 AddendaASME Code Section XI, Appendix G [Ref. H-1].
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 H-2 MATERIAL PROPERTY BASIS LIMITING MATERIALS: Surry Unit 1 Intermediate to Lower Shell Circumferential Weld (Heat #
72445) and Surry Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat# 0227, Position 2.1)
LIMITING ART VALUES AT 68 EFPY: l /4T, 213. 9°F (Circumferential Flaw) 3/4T, l 79.8°F (Circumferential Flaw) 2500 OperlimAnalysis Version:5.4 Run :16852 2250 LeaL,tLL~I./ 1 1 I/
Operlim .xl sm Version: 5.4.1 I I I Ii Ifl 2000 IUnacceptable Operation I
1750
~~
Heatup ~
6' Rates : L--- ~ I Critical 60°F/Hr ---- ,i Limits:
rn ' ~ I Q.
Cl) 1500 - 40"FIH*
20°F/Hr
_1 I
'j, - -
I D:, 60°F/Hr I 40°F/Hr
"- 20°F/Hr 1/)
1 1/)
Cl)
Q.
1250 I
"C Cl) m I IAcceptable I
, 1000
~ I Operation I m I
(.)
750 Criticality Limit based on 500 inservice hydrostatic test temperature (189°F) for the service period up to 68 EFPY 250 I 0 ~
I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure H-1 Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 20, 40, and 60°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Circumferential Flaw" Methodology (w/ K1c)
Note: Curves generated for informational and comparison purposes only.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 H-3 MATERIAL PROPERTY BASIS LIMITING MATERIALS: Surry Unit 1 Intermediate to Lower Shell Circumferential Weld (Heat #
72445) and Suny Unit 2 Intermediate to Lower Shell Circumferential Weld (Heat# 0227, Position 2.1)
LIMITING ART VALUES AT 68 EFPY: l/4T, 213.9°F (Circumferential Flaw) 3/4T, l 79.8°F (Circumferential Flaw) 2500 Operl imAnalysis Version :5.4 Run:16852 Operlim .xlsm Version: 5.4 .1 I
2250 2000 IUnacceptable Ooeration I
1750 A
~
I 6'
ISteady-State I
(/)
-... \c-b -:
1 Q. 1500 I a,
J 1/)
1/)
II ~ Cool down a,
Q.
1250
-1 ~Rates "C
~
a,
_,, ,. J ~
I I
l
-20°F/Hr
-40°F/Hr
-60°F/Hr
J 1000 0 .,,,. V ' -100°F/Hr n,
(.) IAcceptable Operation I
750 500 250 I
0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure H-2 Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr) Applicable for 68 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G "Circumferential Flaw" Methodology (w/ K 1c)
Note: Curves generated for informational and comparison purposes only.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 H-4 2500 2250 ~
2000 I
l,LEJ--.
,1!-°S 1750 IUnacceptable Operation I
Ii 1/ l
( !)
I l k/
en 1500 a..
Q) lo.
~
/ /
/ / II
/i I'I 1/j 1/j Q) lo.
a..
1250 -
--* /
"'O Q)
C'O
~ 1000 0
I IAcceptable I
/
C'O u Operation 750 c::::~
500 Solid Lines: Axial Flaw P-T Lim its Dashed Lines: Circ. Flaw P-T Limits Heatup Rates:
250 20°F/Hr orange I-40°F/Hr green 60°F/Hr blue 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure H-3 Surry Units 1 and 2 Heatup P-T Limit Curve Comparison between Limiting "Axial Flaw" Based Curves and "Circumferential Flaw" Based Curves WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 H-5 2500 2250 IUnacceptable Operation
~
2000
~
A 1750 ,lt l
il1
( !)
tn 1500
-a. ., /I ~t" I
/~L I 1~~
Q) a..
11' I r;
1250 11' Q) a..
a.
~
I LSSJ/
~ -~-~ .,
~
"'O Q) n,
, 1000 boi 0
L~ Acceptable
--~!J n,
~~~I
(.) Operation 750 Solid Lines: Axial Flaw P-T Limits Dashed Lines : Circ. Flaw P-T Limits 500
-<Ill ~
I Cooldown Rat es :
Steady-State b lack (ss) 250
-20°F/Hr orange
-40°F/Hr green
-60°F/Hr blue
-1 00°F/Hr red 0 I 0 50 100 150 200 250 300 350 400 450 500 550 Moderato r Temperature (Deg. F)
Figure H-4 Surry Units 1 and 2 Cooldown P-T Limit Curve Comparison between Limiting "Axial Flaw" Based Curves and "Circumferential Flaw" Based Curves WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 H-6 H.1 REFERENCES H-1 Appendix G to the 1998 Edition through 2000 Addenda of ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 1-1 APPENDIX I SURRY UNITS 1 AND 2 UPPER-SHELF ENERGY EVALUATION AT 68 EFPY
1.1 INTRODUCTION
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 [Ref. I-1]. 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 [Ref. I-2]. 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 (SLR) Position 1.2 USE values of the vessel materials can be predicted using the corresponding 1/4T 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 l/4T fluence projection. The reduced plant surveillance data was obtained from Table 7-6 ofBAW-2324
[Ref. I-3] for Surry Unit 1. The reduced plant surveillance data was obtained from Table 5-12 of WCAP-16001, Revision O [Ref. I-4] for Surry Unit 2. The surveillance data was plotted in Regulatory Guide 1.99, Revision 2, Figure 2 (see Figures I-1 and I-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 l-1 and I-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. These calculations are summarized in Tables I-1 and I-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-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 1-2
1.2 CONCLUSION
For Surry Unit 1, the limiting USE value at 68 EFPY is 32 ft-lb (see Table I-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 I-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 [Ref. I- 5] 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.
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 l 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 LI, Heat# 8Tl554
Lower Shell Longitudinal Weld L2,Heat # 299L44
Inlet Nozzle to Shell Welds, Heat# 299L44 and# 8Tl 762 (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# 8Tl 762
Intermediate to Lower Shell Circumferential Weld,Heat # 0227
Lower Shell Longitudinal Weld LI and L2, Heat# 8Tl 762
Inlet Nozzle to Shell Welds, Heat# 8Tl 762 (Projected USE > 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 on the following page.
WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 1-3
Surry Unit 1 Outlet Nozzle 1 to Upper Shell Weld
Surry Unit l Inlet Nozzle 1 to Upper Shell Weld
Surry Unit 1 Inlet Nozzle 3 to Upper Shell Weld
Surry Unit 2 Outlet Nozzle 1 to Upper Shell Weld
Surry Unit 2 Inlet Nozzle 1 to Upper Shell Weld
Surry 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 1-1); this value corresponds to the Inlet Nozzle to Upper Shell Welds (Heat# 299L44) using Position 2.2. For Surry Unit 2, the limiting USE value for materials not requiring an EMA at 68 EFPY is also 54 ft-lb (see Table 1-2); this value corresponds to the Outlet Nozzle to Upper Shell Welds using Position 2.1. Except for the materials listed above, all of the beltline and extended beltline materials in the Surry 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-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 I-4 Table 1-1 Predicted USE Values at 68 EFPY for Surry Unit 1 SLR 1/4T SLR Wt.% Initial USE<*> Projected USE RPV Material Fluence(b> USE cu<*) (ft-lb) Decrease<c> (%)
(x 10 19 n/cm 2) (ft-lb)
Position 1.2 Upper Shell Forging 122V l09VA 1 0.11 0.465 11 4 17 95 Upper to Intermediate Shell 0.33 0.465 64 39 39(*)
Circumferential W eld<*> (Heat # 250 17)
Intermediate Shell Plate C4326- l 0.11 3.88 11 5 28 83 Intermediate Shell Plate C4326-2 0.11 3.88 94 28 68 Intermediate Shell Longitudinal Welds 0.16 0.77 1 64 29 45<*>
L3 and u <*l (Heat # 8T l 554)
Intermediate to Lower Shell 0.22 3.89 64 50 3i*>
Circumferential W etd<*>(Heat # 72445)
Lower Shell Plate C44 l 5- l 0.102 3.92 103 27 75 Lower Shell Plate C44 I 5-2 0.1 I 3.92 82 28.5 59 Lower Shell Longitudinal Weld L l(e) 0.1 6 0.777 64 29 45<*>
(Heat # 8TJ554)
Lower Shell Longitudinal Weld Li *>
0.34 0.777 64 41 38(e)
(Heat # 299L44)
Inlet Nozzle 1 to Upper Shell Weld 0.34 0.01 88 64 24 49 (Heat # 299L44)
Inlet Nozzle 2 to Upper Shell Weld 0.34 0.00484 64 24 49 (Heat # 299L44)
Inlet Nozzle 3 to Upper Shell Weld 0.34 0.00672 64 24 49 (Heat # 299L44)
Inlet Nozzle I to Upper Shell Weld 0.19 0 .01 88 64 13 56 (Heat # 8TJ 762)
Inlet Nozzle 2 to Upper Shell Weld 0.19 0.00484 64 13 56 (Heat # 8TJ 762)
Inlet Nozzle 3 to Upper Shell Weld 0.19 0.00672 64 13 56 (Heat # 8TJ 762)
Outlet Nozzle I to Upper Shell Weld 0.19 0.00502 64 13 56 (Heat # 8TJ 762)
Outlet Nozzle 2 to Upper Shell Weld 0.19 0.00362 64 13 56 (Heat # 8TJ 762)
Outlet Nozzle 3 to Upper Shell Weld 0.19 0.0140 64 13 56 (Heat # 8TJ 762)
Outlet Nozzle 1 to Upper Shell Weld 0.1 6 0.00502 64 12 56 (Heat # 8T 1554B)
Outlet Nozzle 2 to Upper Shell W eld 0.1 6 0.00362 64 12 56 (Heat # 8TJ 554B)
WCAP-1 8243-NP October 20 17 Revision 0
Westinghouse Non-Proprietary Class 3 1-5 Table 1-1 Predicted USE Values at 68 EFPY for Surry Unit 1 SLR 1/4T SLR Wt.% Initial USE<*> Projected USE RPV Material Fluence USE cu<*) 19 (ft-lb) Decrease<c>(%)
(x 10 n/cm 2) (ft-lb)
Outlet Nozzle 3 to Upper Shell Weld 0.16 0.01 40 64 12 56 (Heat # 8T l 5548 )
Inlet Nozzle 1 (Heat # 9-4787) 0.1 59 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 0.005 86 82 10 74 Outlet Nozzle 3 (Heat# 9-4788) 0.159 0.0227 71 10.5 64 Position 2.id>
Lower Shell Plate C44 l 5- l 0.102 3.92 103 28 74 Lower Shell Plate C441 5-2 0.11 3.92 82 28 59 Lower Shell Longitudinal Weld Lz<e> 4z(e) 0.34 0.777 64 35 (Heat # 299 L44)
Inlet Nozzle I to Upper Shell Weld 0.34 0.01 88 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 th is report.
(b) The I/4T fluence was calculated using the fl uence data in Table 2-3, the Regulatory Guide 1.99, Revision 2 [Ref. I-2] correlation, and the Surry Units I and 2 reactor vessel wall th ickness of 8.05 inches. The surface fluence at the lowest extent of the nozzle weld was used to represent the inlet and outlet nozzle forgi ngs; th is approach is conservative. Bound ing material fluence values, only, are shown in Figure I- 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 usi ng the materi al-specific Cu wt. % va lues.
(d) Surveillance data (deemed credible per Appendix G) from Table 7-6 of BA W-2324 [Ref. I-3] 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 (i n 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 in BAW-2494, Revision 1 [Ref. I-5], and are included herein to establish a baseline for SLR evaluation. EMAs for these materials are addressed under PA-MSC- 148 1.
WCAP-1 8243-NP October 201 7 Revision 0
Westinghouse Non-Proprietary Class 3 1-6 Table 1-2 Predicted USE Values at 68 EFPY for Surry Unit 2 SLR l/4T SLR Wt.% Initial USE<a> Projected USE RPV Material Fluence(b> USE cu<a) (ft-lb) Decrease<c) (%)
(x 10 19 n/cm 2) (ft-lb)
Position 1.2 Upper Shell Forging 123V303VA1 0.11 0.534 104 18 85 Upper to Intermediate Shell Circumferential 41(*)
0.35 0.534 68 39 Weld(e) (Heat # 4275)
Intermediate Shell Plate C433 l-2 0.12 4 .44 84 30 59 Intermediate Shell Plate C4339-2 0.11 4.44 83 29 59 Intermediate Shell Longitudinal Welds L3 and 0.22 0.796 64 34 4i >
0 L4 (OD 50%/0 > (Heat# 72445)
Intermediate Shell Longitudinal Weld 44(e) 0.19 0.796 64 32 L4 (ID 50%i0 > (Heat # 8T 1762)
Intermediate to Lower Shell Circ. Weld(e) 43(e) 0.187 4.45 82 47 (Heat # 0227)
Lower Shell Plate C4208-2 0.15 4.48 94 35 61 Lower Shell Plate C4339-l 0.107 4.48 101 29 72 0
Lower Shell Longitudinal Weld L1 and Li >
43(e) 0.19 0.802 64 33 (Heat# 8Tl 762)
Inlet Nozzle 1 to Upper Shell Weld 0.19 0.0210 64 14 55 (Heat# 8Tl 762)
Inlet Nozzle 2 to Upper Shell Weld 0.19 0.00484 64 13 .5 55 (Heat# 8Tl 762)
Inlet Nozzle 3 to Upper Shell Weld 0.19 0.00660 64 13 .5 55 (Heat# 8T 1762)
Outlet Nozzle 1 to Upper Shell Weld 0.35 0.00491 71 24 54 (Rotterdam)
Outlet Nozzle 2 to Upper Shell Weld 0.35 0.00361 71 24 54 (Rotterdam)
Outlet Nozzle 3 to Upper Shell Weld 0.35 0.0156 71 24 54 (Rotterdam)
Inlet Nozzle 1 (Heat# 9-5104) 0.159 0.0340 73 12.5 64 Inlet Nozzle 2 (Heat # 9-4815) 0.159 0.00784 66 10 59 Inlet Nozzle 3 (Heat # 9-5205) 0.159 0.0107 67 10 60 Outlet Nozzle 1 (Heat# 9-4825-2) 0.159 0.00796 73 10 66 Outlet Nozzle 2 (Heat # 9-5086-1) 0.159 0.00585 77 10 69 Outlet Nozzle 3 (Heat # 9-5086-2) 0.159 0.0253 71 10.5 64 Position 2_id)
Lower Shell Plate C4339-l 0.107 4.48 101 19 82 Intermediate Shell Plate C4339-2 0.11 4.44 83 19 67 Intermediate to Lower Shell Circ. Weld(e) 43(e) 0.187 4.45 82 42 (Heat# 0227)
Notes on the following page.
WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 1-7 Notes:
(a) Materi al data is from Tables 3-3 and 3-4 of th is report.
(b) The l/4T fluence was calculated using the flu ence data in Table 2-4, the Regulatory Guide 1.99, Revision 2 [Ref. I-2]
correlation, and the Surry Units l and 2 reactor vessel wall th ickness of 8.05 inches. The surface fluence at the lowest extent of the nozzle weld was used to represent the inlet and outlet nozzle fo rgi ngs; this approach is conservative. Bounding material fluence values, only, are shown in Figure I-2 fo r 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-spec ific Cu wt. % values.
(d) Surveillance data (deemed credible and non-credible per Appendix G) from Table 5- 12 ofWCAP- 16001 , Revision O [Ref. I-4] 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 li ne drawn parallel to the existing lines (i n Figure 2 of the Guide) through the surveillance data points should be used in preference to the existing graph lines fo r determining the decrease in USE. Credibility Cri teri on 3 in the Discussion secti on of Regulatory Guide 1.99, Revision 2, indicates that even if the surveillance data are not considered credible for determination oft.RTNOT, "they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, fo llowing 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 in BAW-2494, Revision I [Ref. I-5], and are included herein to establish a baseline for SLR evaluation. EMAs fo r these materials are addressed under PA-MSC- 148 1.
WCAP- 18243 -NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 1-8 100
% Copper Base Metal Weld Weld Line 0.35 Limiting Weld Percent USE
' 0.30 Decrease Is 44% from capsule X ,_
0.30 I 0.25 0.25 0.20 I Upper Limit I 0.20 I 0.15 :-
0.15 0.1 0 0.10 0.05
-- -- ---~ ...
--~
i)
-- :::::.---s--
~
i-
..... _._- I,.-
w ....... "'i.-
1/) i...-- i...- ..... """' ---
~
C:
~
- ~
i--- i.--
i--
..... ... :t --
~
I Plate Line Q.
0
.....- ~~ i,,,, ~ Limiting Plate Percent USE 0
cu Cl 10 ,___ - - - i---i.--
Decrease Is 24% from Capsu le X (longitudinal-orientation)
Longitudinal Shell Jg C:
....--' Inlet Nozzle (Surface) i-1 Flu ence = 3.04 x 1011 n/cm2 I I Welds L1, L2, L3 and f" - I
.... .. I I cu
...cu 0
I I L4 Bounding Fluence
= 7.77 x 101e n/cm 2 1/4T Lower Shell Plates
.... .. Fluence = 3.92 x 1019 n/cm2 .
11..
n Outlet Nozzle (Surface)
I
- / Fluence = 2.27 X 1017 n/cm 2 1/4T Inlet Nozzle to Upper Shell Weld Fluence = 1.88 x 1017 n/cm2 I
1
-- 1/4T Intermediate Shell Plates and Intermediate to Lower Shell Circ. Weld V Bounding Fluence = 3.89
~Ir-" 1/4T Outlet Nozzle to Upper Shell Weld I x10 19 n/cm2 I Fluence = 1.40 x 1011 n/cm2 I V 1/4T Upper Shell Forging and Upper to v ~]..,
~ r-, Intermediate Shell Circ . Weld Fluence = 4.65 x 101e n/cm2 1
1E+17 1E+1B 1E+19 1E+20
Surveillance Material: LS Plate C<<15-1 Neutron Fluence, n/cm 2 (E > 1 MeV) + Surveillance Material: Weld Heat# 299L44 Figure 1-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Surry Unit 1 at 68 EFPY WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 1-9 100 I I
% Copper Plate Line '--'-
I I Base Metal 0.35 I Weld 0.30 Weld Line Limiting Weld Percent USE Limiting Plate Percent USE Decrease is 10% from Capsule X
~ ~
0.30 I I 0.25 (transverse-orientation)
~ ~
~
Decrease Is 22% from Capsule X 0.25 I I 0.20 ~ Upper Limit ,_
~--- --- -
0.20 0.15 \ Lo--
\
-~ ....
0.15 0.10 0.10 0.05
' .... \
- 1,.,,.- ~
~ c!.:..
i..-- ~
~
- i...-1,.- ...- ................ -:=:...--- ~ Lo-- -............... .... ...--~ .... .... ~
i...- ~
w II)
-- - - .,.i.-- -- ...... .--.... ....- .... .... .... ... .--
J i;;......;;
.-.... :::::::::::- ~
~ ~ !!--- L--
C:
a.
,-i.., ~F"
....... t: -_. ..-........ ~
1,.- Lo-- 1,.,,.- ~
.- ._.I- L--- *
- .... .... -- -- ... l-----"-"" -
0 ~ 1,.,,.- ~ i- i.--
i..,l,o 1--"
C - -
~
10 cu
...ca Cl . - -- ' -
,_ .... ~
C:
114T Lower Shell Plates Fluence =4.48 x 1019 n/cm2
~ -.
L--
cu -'-
0 L..- I I
-' I
~ Inlet Nozzle (Surface) Longitudinal Shell Welds 1/4T Intermediate to Lower
,.. Fluence =3.40 x 1017 n/cm2 L1 , L2 , L3 and L4
~
Shell Circumferential Weld
~
~
Bounding Fluence =8.02 f-t Outlet Nozzle (Surface)
.... Heat# 0227 Fluence =4.45 x 1019 n/cm2 L.. x 10 18 n/cm 2
~ Fluence =2.53 x 1011 n/cm2 H1/4T FluenceInlet Nozzle to Upper Shell Weld 1/4T Intermediate Shell Plates Fluence =4.44 x10 19 n/cm2
. ~
~
'-- i--
=2.1 Ox 10 1 n/cm2 1
1/4T Upper Shell Forging and Upper LJ 1/4T Outlet Nozzle to Upper ~r,.
1
.. Fluence Shell Weld
=1.56 x 10 7 n/cm2 1
to Intermediate Shell Circ . Weld Fluence =5.34 x 101s n/cm 2 1E+17 1E+18 1E+19 1E+20
Surveillance Malerlal: LS Plate C4339-1 Neutron Fluence, n/cm2 (E > 1 MeV) + surveillance Weld Data : Heat# 0227 Figure 1-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Surry Unit 2 at 68 EFPY WCAP- 18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 I-10
1.3 REFERENCES
I-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.
I-2 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
I-3 Framatome ANP Report BAW-2324, Revision 0, "Analysis of Capsule X, Virginia Power Surry Unit No. 1, Reactor Vessel Material Surveillance Program," April 1998.
I-4 Westinghouse Report WCAP-16001, Revision 0, "Analysis of Capsule Y from Dominion Surry Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003.
I-5 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.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 J-1 APPENDIX J MATERIAL PROPERTY INPUT COMPARISON This appendix provides tables which compare the material property input values utilized in this report, taken primarily from PWROG-16045-NP [Ref. J-1] , with those utilized in Dominion calculation SM-1008, Addendum OOM [Ref. J-2] and the Surry Power Station Updated Final Safety Analysis Report (UFSAR) [Ref. J-3], as applicable.
Table J-1 Comparison of Previous and Current Initial RT NDT Values for Surry Unit 1 Previous Current Initial Initial Material Identification (a) (b)
RTNDT RTNDT (OF) (OF)
Replacement Reactor Vessel Closure Head Flange
-67 -67 E4381/E4382 Reactor Vessel Flange FV-1870 10 -114.6 Inlet Nozzle 1 (Heat# 9-4787) 60 10.3 Inlet Nozzle 2 (Heat # 9-5078) 60 11.6 Inlet Nozzle 3 (Heat # 9-4819) 60 -47.2 Outlet Nozzle 1 (Heat # 9-4825-1) 60 -44.9 Outlet Nozzle 2 (Heat# 9-4762) 60 -87.5 Outlet Nozzle 3 (Heat # 9-4788) 60 -50.2 Inlet Nozzle to Uooer Shell Welds (Heat# 299L44) --- -7.0 Inlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- -4.9 Outlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- -4.9 Outlet Nozzle to Upper Shell Welds ---
-4.9 (Heat# 8Tl554B)
Upper Shell Forging 122V109VA1 40 40 Upper to Intermediate Shell Circumferential Weld 0 0 (Heat#25017)
Intermediate Shell Plate C4326-l 10 10 Intermediate Shell Plate C4326-2 0 11.4 Intermediate Shell Longitudinal Welds L3 and L4
-48.6 -48.6 (Heat# 8Tl554)
Intermediate to Lower Shell Circumferential Weld
-72.5 -72.5 (Heat # 72445)
Lower Shell Plate C44 l 5- l 20 20 Lower Shell Plate C44 l 5-2 0 4.6 Lower Shell Longitudinal Weld Ll (Heat# 8Tl554) -48.6 -48.6 Lower Shell Longitudinal Weld L2 (Heat# 299L44) -74.3 -74.3 Notes:
(a) The previous initial RT NDT values were taken from the Surry Power Station UFSAR, Table 4.1-14 [Ref. J-3). These values are consistent with those documented in Dominion Calculation SM-1008, Addendum OOM [Ref. J-2]; however, some initial RT NDT values are only listed in the UFSAR.
(b) Current initial RTNor values correspond to the values utilized herein. In some cases, these values have been updated or defined based on evaluations completed in PWROG-16045-NP [Ref. J-1].
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 J-2 Table J-2 Comparison of Previous and Current Initial RTNoT Values for Surry Unit 2 Previous Current Initial Material Identification (a) Initial RT OT (OF)
RT OT(b) (OF)
Replacement Reactor Vessel Closure
-60 -60 Head 02Wl-l-l-l Reactor Vessel Flange FV-2542 -65 -156.3 Inlet Nozzle 1 (Heat# 9-5104) 60 -29.7 Inlet Nozzle 2 (Heat# 9-4815) 60 4.5 Inlet Nozzle 3 (Heat# 9-5205) 60 6.5 Outlet Nozzle 1 (Heat # 9-4825-2) 60 -58 .1 Outlet Nozzle 2 (Heat # 9-5086-1) 60 -26.6 Outlet Nozzle 3 (Heat# 9-5086-2) 60 -33.8 Inlet Nozzle to Upper Shell Welds
--- -4.9 (Heat# 8Tl 762)
Outlet Nozzle to Upper Shell Welds
--- 30 (Rotterdam)
Upper Shell Forging 123V303VA1 30 30 Upper to Intermediate Shell 0 0 Circumferential Weld (Heat# 4275)
Intermediate Shell Plate C4331-2 -10 15.0 Intermediate Shell Plate C4339-2 -20 7.8 Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%) -72.5 -72.5 (Heat# 72445)
Intermediate Shell Longitudinal
-48 .6 -48.6 Weld L4 (ID 50%) (Heat # 8T 1762)
Intermediate to Lower Shell 0 0 Circumferential Weld (Heat# 0227)
Lower Shell Plate C4208-2 -30 -30 Lower Shell Plate C4339-l -10 -4.4 Lower Shell Longitudinal Weld Ll
-48.6 -48.6 and L2 (Heat# 8Tl 762)
Notes:
(a) The previous initial RTNDT values were taken from the Surry Power Station UFSAR, Table 4.1-15 [Ref. J-3). These values are consistent with those documented in Dominion Calculation SM- 1008, Addendum OOM [Ref. J-2); however, some initial RTNDT values are only listed in the UFSAR.
(b) Current initial RTNDT values correspond to the values utilized herein. In some cases, these values have been updated or defined based on evaluations completed in PWROG-16045-NP [Ref. J-1) and Appendix E (for weld Heat # 0227).
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 J-3 Table J-3 Comparison of Previous and Current a 1 Values for Surry Unit 1 Previous a 1 <*> Current a, (b)
Material Identification (OF) (OF)
Inlet Nozzle 1 (Heat# 9-4787) --- 0 Inlet Nozzle 2 (Heat# 9-5078) --- 0 Inlet Nozzle 3 (Heat# 9-4819) --- 0 Outlet Nozzle I (Heat# 9-4825-1) --- 0 Outlet Nozzle 2 (Heat# 9-4762) --- 0 Outlet Nozzle 3 (Heat# 9-4788) --- 0 Inlet Nozzle to Upper Shell Welds (Heat# 299L44) --- 20.6 Inlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- 19.7 Outlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- 19.7 Outlet Nozzle to Upper Shell Welds ---
19.7 (Heat# 8Tl554B)
Upper Shell Forging 122Vl09VA1 0 0 Upper to Intermediate Shell Circumferential Weld 20.0 20.0 (Heat# 25017)
Intermediate Shell Plate C4326- l 0 0 Intermediate Shell Plate C4326-2 0 0 Intermediate Shell Longitudinal Welds L3 and L4 18.0 18.0 (Heat# 8Tl554)
Intermediate to Lower Shell Circumferential Weld 12.0 12.0 (Heat# 72445)
Lower Shell Plate C4415-l 0 0 Lower Shell Plate C4415-2 0 0 Lower Shell Longitudinal Weld Ll 18.0 18.0 (Heat# 8Tl554)
Lower Shell Longitudinal Weld L2 12.8 12.8 (Heat# 299L44)
Notes:
(a) The previous cr 1 values were taken from Dominion Calculation SM-1008, Addendum OOM [Ref. J-2].
(b) Current cr 1 values correspond to the values utilized herein. In some cases, these values have been confirmed or defined based on evaluations completed in PWROG-16045-NP [Ref. J-1]. cr 1 is set equal to O when measured data is used per WCAP-14040-A, Revision 4 [Ref. J-4].
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 J-4 Table J-4 Comparison of Previous and Current a 1 Values for Surry Unit 2 Previous a 1<*> Current a/bl Material Identification (OF) (OF)
Inlet Nozzle 1 (Heat# 9-5104) --- 0 Inlet Nozzle 2 (Heat# 9-4815) --- 0 Inlet Nozzle 3 (Heat# 9-5205) --- 0 Outlet Nozzle 1 (Heat# 9-4825-2) --- 0 Outlet Nozzle 2 (Heat# 9-5086-1) --- 0 Outlet Nozzle 3 (Heat # 9-5086-2) --- 0 Inlet Nozzle to Upper Shell Welds ---
19.7 (Heat# 8Tl762)
Outlet Nozzle to Upper Shell Welds ---
0 (Rotterdam)
Upper Shell Forging 123V303VA1 0 0 Upper to Intermediate Shell 20.0 20.0 Circumferential Weld (Heat # 4275)
Intermediate Shell Plate C4331-2 0 0 Intermediate Shell Plate C4339-2 0 0 Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%) 12.0 12.0 (Heat # 72445)
Intermediate Shell Longitudinal Weld 18.0 18.0 L4 (ID 50%) (Heat# 8Tl 762)
Intermediate to Lower Shell 20.0 0 Circumferential Weld (Heat # 0227)
Lower Shell Plate C4208-2 0 0 Lower Shell Plate C4339-l 0 0 Lower Shell Longitudinal Weld LI 18.0 18.0 and L2 (Heat# 8Tl 762)
Notes:
(a) The previous cr 1 values were taken from Dominion Calculation SM- I 008, Addendum OOM [Ref. J-2).
(b) Current cr1 values correspond to the values utilized herein. In some cases, these values have been confirmed or defined based on evaluations completed in PWROG-16045-NP [Ref. J-1) and Appendix E (for weld Heat # 0227). cr 1 is set equal to O when measured data is used per WCAP-14040-A, Revi sion 4 [Ref. J-4).
WCAP-18243-NP October 201 7 Revision 0
Westinghouse Non-Proprietary Class 3 J-5 Table J-5 Comparison of Previous and Current a.1 Values for Surry Unit 1 Previous <J,1 <*> Current a.1 (b)
Material Identification (OF) (OF)
Inlet Nozzle 1 (Heat# 9-4787) - -- 17.0 Inlet Nozzle 2 (Heat# 9-5078) --- 17.0 Inlet Nozzle 3 (Heat# 9-4819) --- 17.0 Outlet Nozzle 1 (Heat# 9-4825-1) --- 17.0 Outlet Nozzle 2 (Heat# 9-4762) --- 17.0 Outlet Nozzle 3 (Heat# 9-4788) --- 17.0 Inlet Nozzle to Upper Shell Welds (Heat# 299L44) --- 14.0/28.o(c)
Inlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- 28.0 Outlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- 28.0 Outlet Nozzle to Upper Shell Welds (Heat# 8Tl 554B) --- 28 .0 Upper Shell Forging 122Vl09VA1 17.0 17.0 Upper to Intermediate Shell Circumferential Weld 28.0 28.0 (Heat # 25017)
Intermediate Shell Plate C4326-l 17.0 17.0 Intermediate Shell Plate C4326-2 17.0 17.0 Intermediate Shell Longitudinal Welds L3 and L4 (Heat 28.0 28.0
8Tl554)
Intermediate to Lower Shell Circumferential Weld 28.0 28.0 (Heat # 72445)
Lower Shell Plate C4415- l 8.5 8.5/17.0(d)
Lower Shell Plate C4415-2 17.0 8.5/17.0(d)
Lower Shell Longitudinal Weld LI 28 .0 28.0 (Heat# 8Tl554)
Lower Shell Longitudinal Weld L2 28.0 28.0 (Heat# 299L44)
Notes:
(a) The previous crd values were taken from Dominion Calculation SM-1008, Addendum OOM [Ref. J-2].
(b) Current crd values correspond to the values utilized herein; however, values reported in th is table do not consider that crd need not exceed 0.5*L'.RTNDT per Regulatory Guide 1.99, Revision 2 [Ref. J-5]. See Section 5 and Appendix B for the actual crd values utilized in cases where 0.5*L'.RTNDT was limiting.
(c) For Regulatory Guide l.99, Revision 2 [Ref. J-5] Position 2.1 , 14.0°F was uti lized as a resu lt of credible surveillance data.
For Regulatory Guide 1.99, Revision 2 [Ref. J-5] Position 1.1, 28.0°F was utilized (d) For Regu latory Guide 1.99, Revision 2 [Ref. J-5] Position 2.1, 8.5°F was utilized as a resu lt of credible surveillance data.
For Regulatory Guide 1.99, Revision 2 [Ref. J-5] Position 1. 1, 17.0°F was utilized.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 J-6 Table J-6 Comparison of Previous and Current <J,1 Values for Surry Unit 2 Previous a-1 (a) Current <J,1 (b)
Material Identification (OF) (OF)
Inlet Nozzle 1 (Heat# 9-5104) --- 17.0 Inlet Nozzle 2 (Heat# 9-4815) --- 17.0 Inlet Nozzle 3 (Heat# 9-5205) --- 17.0 Outlet Nozzle 1 (Heat# 9-4825-2) --- 17.0 Outlet Nozzle 2 (Heat# 9-5086-1) --- 17.0 Outlet Nozzle 3 (Heat# 9-5086-2) --- 17.0 Inlet Nozzle to Upper Shell Welds ---
28.0 (Heat# 8Tl 762)
Outlet Nozzle to Upper Shell Welds -- -
28.0 (Rotterdam)
Uooer Shell Forging 123V303VA1 17.0 17.0 Upper to Intermediate Shell 28.0 28.0 Circumferential Weld (Heat# 4275)
Intermediate Shell Plate C433 l-2 17.0 17.0 Intermediate Shell Plate C4339-2 17.0 17.0 Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%) 28.0 28.0 (Heat # 72445)
Intermediate Shell Longitudinal Weld 28.0 28.0 L4 (ID 50%) (Heat# 8Tl 762)
Intermediate to Lower Shell 14.0 14.0/28.0(c)
Circumferential Weld (Heat # 0227)
Lower Shell Plate C4208-2 17.0 17.0 Lower Shell Plate C4339-l 17.0 17.0 Lower Shell Longitudinal Weld Ll 28.0 28.0 and L2 (Heat# 8Tl 762)
Notes:
(a) The previous 0 6 values were taken from Dominion Calculation SM-1008, Addendum OOM [Ref. J-2).
(b) Current 0 6 values correspond to the values utilized herein; however, values reported in this tabl e do not consider that 0 6 need not exceed 0.5*~RTNoT per Regulatory Guide 1.99, Revi sion 2 [Ref. J-5). See Section 5 and Appendix B for the actual 0 6 values utilized in cases where 0.5*~RTNDT was limiting.
(c) For Regulatory Guide 1.99, Revision 2 [Ref. J-5] Position 2. 1, I 4.0°F was utilized as a result of credible surveillance data.
For Regulatory Guide 1.99, Revision 2 [Ref. J-5] Position 1.1, 28.0°F was utilized.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 J-7 Table J-7 Comparison of Previous and Current Unirradiated USE Values for Surry Unit 1 Previous Current Material Identification Unirradiated Un irradiated USE<*> (ft-lb) USE(b> (ft-lb)
Inlet Nozzle 1 (Heat# 9-4787) 64 63 Inlet Nozzle 2 (Heat# 9-5078) 64 64 Inlet Nozzle 3 (Heat# 9-4819) 68 68 Outlet Nozzle 1 (Heat# 9-4825-1) 68 68 Outlet Nozzle 2 (Heat# 9-4762) 85 82 Outlet Nozzle 3 (Heat# 9-4788) 72 71 Inlet Nozzle to Upper Shell Welds (Heat# 299L44) --- 64 Inlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- 64 Outlet Nozzle to Upper Shell Welds (Heat# 8Tl 762) --- 64 Outlet Nozzle to Upper Shell Welds (Heat# 8T1554B) --- 64 Upper Shell Forging 122V109VA1 83 114 Upper to Intermediate Shell Circumferential Weld EMA ~64 (Heat# 25017)
Intermediate Shell Plate C4326-l 115 115 Intermediate Shell Plate C4326-2 94 94 Intermediate Shell Longitudinal Welds L3 and L4 (Heat 77/EMA 64
8Tl554)
Intermediate to Lower Shell Circumferential Weld 77/EMA 64 (Heat # 72445)
Lower Shell Plate C4415-1 103 103 Lower Shell Plate C4415-2 83 82 Lower Shell Longitudinal Weld Ll 77/EMA 64 (Heat# 8Tl554)
Lower Shell Longitudinal Weld L2 70/EMA 64 (Heat# 299L44)
Notes:
(a) The previous unirradiated USE values were taken from the Surry Power Station UFSAR, Table 4. 1-14 [Ref. J-3]. These values are consistent with those documented in Dominion Calculation SM-1008, Addendum OOM [Ref. J-2]; however, some initial USE values are only listed in the UFSAR.
(b) Current unirradiated USE values correspond to the values utilized herein. In some cases, these values have been updated or defined based on evaluations completed in PWROG-16045-NP [Ref. J-1]. The current unirradiated USE values for materials previously designated with "EMA" have been updated utilizing a generic value or weld qualification data as described in Table 3- 1.
I WCAP-18243-NP October 2017 I Revision 0
Westinghouse Non-Proprietary Class 3 J-8 Table J-8 Comparison of Previous and Current Unirradiated USE Values for Surry Unit 2 Previous Current Material Identification Unirradiated Unirradiated USEC*> (ft-lb) USE Cb> (ft-lb)
Inlet Nozzle 1 (Heat # 9-5104) 73 73 Inlet Nozzle 2 (Heat # 9-4815) 66 66 Inlet Nozzle 3 (Heat # 9-5205) 66 67 Outlet Nozzle 1 (Heat# 9-4825-2) 74 73 Outlet Nozzle 2 (Heat# 9-5086-1) 79 77 Outlet Nozzle 3 (Heat# 9-5086-2) 73 71 Inlet Nozzle to Upper Shell Welds
--- 64 (Heat# 8Tl 762)
Outlet Nozzle to Upper Shell
--- 71 Welds (Rotterdam)
Upper Shell Forging 104 104 123V303VA1 Upper to Intermediate Shell Circumferential Weld EMA >68 (Heat# 4275)
Intermediate Shell Plate C433 l -2 84 84 Intermediate Shell Plate C4339-2 83 83 Intermediate Shell Longitudinal Welds L3 and L4 (OD 50%) 77/EMA 64 (Heat # 72445)
Intermediate Shell Longitudinal Weld L4 (ID 50%) EMA 64 (Heat # 8Tl 762)
Intermediate to Lower Shell Circumferential Weld 90/EMA 82 (Heat# 0227)
Lower Shell Plate C4208-2 94 94 Lower Shell Plate C4339-l 105 101 Lower Shell Longitudinal Weld EMA 64 Ll and L2 (Heat# 8Tl 762)
Notes:
(a) The previous unirradiated USE values were taken from the Suny Power Station UFSAR, Table 4. 1-15 [Ref. J-3). These values are consistent with those documented in Dominion Calculation SM- 1008, Addendum OOM [Ref. J-2); however, some initial USE values are only listed in the UFSAR or SM- I 008, Addendum OOM.
(b) Current un irradiated USE values correspond to the values util ized herein. In some cases, these values have been updated or defined based on evaluations completed in PWROG-16045-NP [Ref. J-1) and Appendix E (for weld Heat # 0227). The current unirradiated USE values for materials previously designated with "EMA" have been updated utilizing a generic value, weld qualification data, or data in Appendix E as described in Table 3-3.
WCAP-18243-NP October 2017 Revision 0
Westinghouse Non-Proprietary Class 3 J-9 J.1 REFERENCES J-1 Pressurized Water Reactor Owners Group (PWROG) Report PWROG-16045-NP, Revision 0, "Determination of Unirradiated RT NDT and Upper-Shelf Energy Values of the Surry Units 1 and 2 Reactor Vessel Materials," March 2017.
J-2 Dominion Calculation SM-1008, Revision 0, Addendum OOM, "Reactor Vessel Integrity Calculations Supporting a Technical Specifications Change Request (TSCR) to Update the Burnup Applicability Limit for RCS Pressure/Temperature Limits, LTOPS Setpoint, and LTOPS Enabling Temperature at Surry Power Station Units 1 and 2," January 2010.
J-3 Surry Power Station Updated Final Safety Analysis Report, Revision 48, September 2016.
J-4 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.
J-5 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
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ML18085A164: Difference between revisions

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1.1 INTRODUCTION

1.2 CONCLUSION

1.3 REFERENCES

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ML18085A164: Difference between revisions

Revision as of 12:24, 23 February 2020

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SUMMARY

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1.1 INTRODUCTION

1.2 CONCLUSION

1.3 REFERENCES

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