L-18-030, WCAP-18102-NP, Rev 1, Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation.

WCAP-18102-NP, Rev 1, Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation.

ML18099A125
Person / Time
Site:	Beaver Valley
Issue date:	02/28/2018
From:	Markivich A, Mcnutt D Westinghouse
To:	Office of Nuclear Reactor Regulation
Shared Package
ML18099A122	List: L-18-030, WCAP-18102-NP, Rev 1, Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation. ML18099A123
References
L-18-030 WCAP-18102-NP, Rev 1
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Westinghouse Non-Proprietary Class 3 WCAP-18102-NP February 2018 Revision 1 Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation

@Westinghouse

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

WESTIN GHOUS E NON-PR OPRIET ARY CLASS 3 WCAP -18102 -NP Revision 1 Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Donald M. McNut t Ill*

Structural Design and Analys is Ill Alex J. Markiv ich*

Radiation Engine ering and Analys is Februa ry 2018 Reviewers: Benjam in E. Mays* Approve d: Lynn A. Patterson*, Manage r Structur al Design and Analysi s III Structural Design and Analysi s Ill Jesse J. Klingensmith* Laurent P. Houssay *, Manage r Radiatio n Enginee ring and Analysi s Radiation Enginee ring and Analysi s

• E lectronically approved records are authenticated in the electroni c documen t managem ent system.

Westingh ouse Electric Compan y LLC 1000 Westing house Dr.

Cranberry Townshi p, PA 16066

© 2018 Westingh ouse Electric Compan y LLC All Rights Reserved

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westin ghouse Non-P ropriet ary Class 3 II RECORD OF REVISION Rev. Description 0 Origin al Issue Update d Appen dix Eby remov ing referen ce to TLR-R ES/DE /CIB-2 013 -01 and associa ted text. Also, Tab le E-1 was update d to utilize I only calcu lated values of 11RTNDT in the RT PTS calcula tions instead of setting calcula ted values less than 25 °F equal to 0°F. Chang es are indicat ed using change bars WCAP -18102 -NP February 20 I 8 Revisi on I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Ill TABLE OF CONTENTS LIST OF TABLE S ... ... ............. .. ... .. ....................................... ..........................

....................... ............. ......... V LIST OF FIGURE S ............................. .................................................................

.......................... ........... viii EXECU TIVE SUMMA RY ..............................................................................

.......................... .................. ix INTRO DUCTI ON ................... ....... ........ ... .... ................ ... ...... .................. ..

.. .. .. ....... .. ............ ...... . 1- 1 2 CALCU LATED NEUTR ON FLUEN CE .......... .. ..... ....... .. .. ...... ... ....... .. ....

... .......... .... ..... .... .... .. ... 2-1 2.1 INTROD UCTION ...........................................................................................

................ 2-I 2.2 DISCRE TE ORDIN ATES ANALY SIS .................................................................

.......... 2-I 2.3 CALCU LATION AL UNCER TAINTI ES ................................................. ....

.. .. .............. . 2-3 3 FRACT URE TOUGH NESS PROPE RTlES .......... ... ........ ... ... ..... ..........................

.................... ... 3-I 4 SURVE ILLANC E DATA .... ........... ... ............ ......................... ....... ........... .......

......... ......... .. .... ..... 4-1 5 CHEMI STRY FACTO RS ...... ........... ........ ........ .. .............. ..... .. .... ........ .. ........

.......... ....... ....... .... ... 5-1 6 CRITER IA FOR ALLOW ABLE PRESSU RE-TEM PERAT URE RELAT IONSHI PS ........ ........ 6-1 6.1 OVERA LLAPP ROACH ..............................................................................

................... 6-I 6.2 METHO DOLOG Y FOR PRESSU RE-TEM PERAT URE LIMIT CURVE DEVEL OPMEN T ....................... ........ .. .. .. ..... ... ....... ................. ... ... .... ..... ...

..................... 6-I 6.3 CLOSU RE HEAD/ VESSE L FLANG E REQUI REMEN TS ........ ..... ............

.. ....... .... ..... 6-5 7 CALCU LATION OF ADJUS TED REFER ENCE TEMPE RATUR E .............

............................. 7-1 8 HEATU P AND COOLD OWN PRESSU RE-TEM PERAT URE LIMIT CURVE S .. ....... .............. 8-1 9 REFER ENCES .... .... .... ....... ... ... .... .... .. ....... .. ....... ................... .. ....... .. ... ..... ............

......... ........ ....... 9-1 APPEN DIX A THERM AL STRESS INTENS ITY FACTO RS (K 11) ************* ****** *********.*** ***

.* *** . A- I APPEN DIX B REACT OR VESSEL INLET AND OUTLE T NOZZL ES ......... ... ...... .......

... .... B-1 APPEN DIX C OTHER REACT OR COOLA NT PRESSU RE BOUND ARY FERRIT IC COMPO NENTS ........ .......... ................. ... .... .. ..... .......... ... ...... .... .. ...... .. .............

. ....................... ... C-1 APPEN DIX D BEAVE R VALLE Y UNIT I SURVE ILLANC E PROGR AM CREDIB ILITY EVALU ATION .......... ...... ........ .... .... .. .... .. .... .. ... ....... ... ... .. .... .. ..... ........... .............

. ....................... D-1 APPEN DIX E PRESSU RIZED THERM AL SHOCK EVALUATION ...... .. ...... ..... ........ .........

E-1 WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westing house Non-Pro prietary Class 3 iv APPENDIX F VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS .... ....... .. ......................

........ .... .................... F-1 APPENDIX G SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE .......... ...........

.... G-1 WCAP- 18102-N P Februar y 2018 Revisio n I

• • • This record was final approved on 2/27/2018 8:36:20 AM . (

This statement was added by the PRIME system upon its validation

)

Westinghouse Non-Propri etary Class 3 V

LIST OF TABLES Table 2-1 Pressu re Vessel Materi al Weld Axial Locati ons ........... ......

... ....... .. ....... .. .............. ..... ...... 2-5 Table 2-2 Reacto r Core Therm al Power Level fo r Beave r Valley Unit 1 ................ .. ..... ...... ... .. ..... . 2-5 Table 2-3 Calcul ated Fast N eutron Fluenc e Rate and Fluenc e (E >

1.0 MeV) at the Pressu re Vessel Clad/ Base Metal lnterfa ce .... .................. ................... ... .. ......

.......... .................................. 2-6 Table 2-3 Calcul ated Fast N eutron Fluenc e Rate and Fluenc e (E >

1.0 MeV) at the Press ure Vessel Clad/ Base Metal lnterfa ce .. ........... ..... ..... .... ......... .......... ..........

.......... ........ ...................... 2-7 Table 2-4 Calcul ated Iron Displa cemen ts per Atom at the Press ure Vesse l Clad/ Base Metal Interfa ce

...... ... ... .... ... .. ... ........ ....... ...... 2-8 Table 2-5 Calcul ated Fast Neutro n Fluenc e (E > 1.0 MeV) at the Pressu re Vessel Plates ............... 2-9 Table 2-6 Calcul ated Iron Displa cemen ts per Atom at the Pressu re Vessel Pl ates .. ..... .. ..... .......... 2- 10 Table 2-7 Calcul ated Fast Neutro n Fluenc e (E > 1.0 MeV) at the Pressu re Vessel Outlet Nozzle ,

Inlet Nozzle , and Circum ferenti al Welds ... ....... .. .... ..... ..........

...... ....... ......................... .. 2-11 Table 2-8 Calcul ated Iron Di splace ments per Atom at the Pressu re Vessel Outlet Nozzle, Inlet Nozzle , and Circum ferenti al Welds ...... ..... .......... .......... ..........

. ......... ......... ................... 2- l 2 Table 2-9 Calcul ated Fast Neutro n Fluenc e (E > 1.0 MeV) at the Pressu re Vessel Longit udinal Welds .. .. ............... ... ........ ......................... ... .......... .......... ..........

.......... .......... .......... ....... 2-13 Table 2-10 Calcul ated Iron Displa cemen ts per Atom at the Pressu re Vessel Longit udinal Welds .. 2-14 Table 2- 11 Calcul ated Fast Neutro n Fluenc e (E > 1.0 MeV) at the Center of the Survei llance Capsu les ..... .... ...... .... .. .. ... .......... ... ... ... ............................. ......

..... ........................... ..... .... 2-15 Table 2-1 2 Summ ary of Calc ulated Survei llance Capsu le Lead Factor s ...... .......... .. ...... ..... ............ 2-16 Table 2-13 Calcul ational Uncert ainties ...... ........ .. .... .. ..... .......... ..........

....... .. .. ...... .... ....... ....... .. .. .... .. 2-17 Table 3-1 Summ ary of Beave r Valley Unit I Reacto r Vessel Base Metal Materi al Initial RT NDT Determ ination Metho dologies .. .. ................. ... .............. ......

.............. ..................... ... ..... .. 3-2 Table 3-2 Summ ary of the Best-E stimat e Cu and Ni Weigh t Percen t and Initial RT NDT Values for the Beave r Valley Unit I Reacto r Vessel Materi als .......... ..........

.......... .......... .......... .......... ... 3-3 Table 3-3 Summ ary of Beave r Valley Unit I Replac ement Reacto r Vessel Closur e Head and Vessel Flange Initial RTNoT Values ... .. .... .... ... ....... .. ................... .......

...... .. ...... ... ... ..... .... .. ... ... .... . 3-4 Table 4-1 Beave r Valley Unit I Survei llance Capsu le Data ..............

.. ................ ... .............. ... ... ..... .4-2 Table 4-2 St. Luci e Unit 1 and Millsto ne Unit 2 S urveill ance Capsu le Data for Weld Heat # 90136

... ........ ........... ..... .. ..... ........ 4-3 Table 5-1 Calcul ation of Beave r Valley Unit I C he mistry Factor Value for Lower Shell Plate B6903 - I Using Survei llance Capsu le Data ...... ... ...... ..........

....... ... .. .... ... ............ ............. . 5-2 Table 5-2 Calcul ation of Beave r Valley Unit 1 C he mistry Factor Value for Weld Heat # 30542 4 Using Survei llance Capsu le Data ....... ... ... ... ... .... .. .... .. ..........

. ...... ... ........ ... ... .. ............... ... 5-2 WCAP -18102 -NP Februa ry 2018 Revisio n I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation )

Westinghouse Non-Proprietary Class 3 VI Table 5-3 Calculatio n of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat # 90136 Using Surveillan ce Capsule Data ..... .. .................. .. ......... ....... ..................... .. ......... ........ .. 5-3 Table 5-4 Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors .................. 5-4 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY .. .. ................. .. ............. .. 7-3 Table 7-2 Adjusted Reference Temperatu re Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 1/4T Location .. ...... ..... .... ........................ ..... 7-4 Table 7-3 Adjusted Reference Temperatu re Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the 3/4T Location .............................................. 7-6 Table 7-4 Summary of the Limiting ART Values Used in the Generation of the Beaver Valley Unit I Heatup and Cooldown Curves at 50 EFPY ............... .. .................... ................................. 7-8 Table 8-1 Beaver Valley Unit 1 50 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodolo gy (w/ K1c, w/ Flange Notch, and w/o Margins for Instrument ation Errors) ............ .. ....................................... .......... .... .. ..... ........................ .. 8-5 Table 8-2 Beaver Valley Unit 1 50 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodolo gy (w/ K1c, w/ Flange Notch, and w/o Margins for Instrument ation Errors) .. .................. ......... .... ........................................... .... 8-7 TableA-1 K1t Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Heatup Curves (w/ Flange Requireme nts and w/o Margins for Instrument Errors) .................................................. A-2 Table A-2 K 11 Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Cooldown Curves (w/ Flange Requireme nts and w/o Margins for Instrument Errors) .................................................. A-3 Table 8-1 ART Calculatio ns for the Beaver Valley Unit 1 Reactor Vessel Nozzle Materials at 50 EFPY .............................. ... .................... .................................. ... ......... .... ... ..... .. ... ........... 8-3 Table 8-2 Summary of the Limiting ART Values for the Beaver Valley Unit 1 Inlet and Outlet Nozzle Materials .............. .. ..... .......... ... ....... ....... ......... ...... ... ..... ....... .. .............. .... ........... 8-4 Table D-1 Mean Chemical Compositi on and Operating Temperatu re for St. Lucie Unit 1 and Millstone Unit 2 ....... ......... .................. ... ............................. .... .... .......... ..................... ..... D-4 Table D-2 Operating Temperatu re Adjustmen ts for the St. Lucie Unit 1 and Millstone Unit 2 Surveillan ce Capsule Data .................................................... ..... ...... ... ................ ............ D-5 Table D-3 Calculatio n of Weld Heat # 90136 Interim Chemistry Factor for the Credibility Evaluation Using St. Lucie Unit 1 and Millstone Unit 2 Surveillan ce Capsule Data .... . D-5 Table D-4 Best-Fit Evaluation for Surveillan ce Weld Metal Heat# 90136 Using St. Lucie Unit 1 and Millstone Unit 2 Data ..................... ........ .............. .. .......... ....... ....................................... 0-6 Table D-5 Calculatio n of Interim Chemistry Factors for the Credibility Evaluation Using Beaver Valley Unit I Surveillan ce Capsule Data ........................................................................ D-7 Table D-6 Beaver Valley Unit 1 Surveillan ce Capsule Data Scatter about the Best-Fit Line .......... D-8 WCAP-181 02-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Pr oprieta ry Class 3 Vil Table E-1 RT rTs Calcul ations for the Beave r Valley Unit I Reacto r Vesse l Materi als at 50 EFPY .... ..

.. ....... .. ...... ... .. ... ... ... .... .. ..... E-3 Table F-1 Nucle ar Param eters Used in the Evalua tion ofNeu tron Sensor s ...... .. ...... ...... .............. F-10 Table F-2 Monthly Therm al Genera tion during the First 24 Fuel Cycles

...... ....... ........ .. ...... .... .. ... F-11 Table F-3 Calcul ated Fast Neutro n (E > 1.0 MeV) Fluenc e Rate and Ci Factor s at the Survei llance Capsu le Center, Core Midpla ne Elevat ion ..............................

...................... .. ....... ..... .. F-16 Table F-4a Measured Sensor Activities and Reacti on Rates of Survei llance Capsu le V ... ........ ... .. F-18 Table F-4b Measured Senso r Activities and Reacti on Rates of Survei llance Capsu le U .. ........ ..... . F-19 Table F-4c Meas ured Senso r Activities and Reacti on Rates of Survei llance Capsu le W ... ....... .... . F-20 Table F-4d Measured Senso r Activities and Reaction Rates of Survei llance Capsu le Y ..... .. ...... ... F-21 Table F-4e Measu red Sensor Activities and Reacti on Rates of Survei llance Capsu le X ..... .......... . F-22 Table F-5 Compa rison of Measured, Calcul ated, and Best- Estima te Reacti on Rates at the Survei llance Capsule Center ......... .. .. ... ...... .. ... .......... ... ..... ...

........ .. ... ..... ... ...... ...... .... .... F-23 Table F-6 Compa rison of Calcul ated and Best-E stimat e Expos ure Rates at the Survei llance Capsu le Center ... ..... .............. .. ........ .. ........ .. ..... .. .... .... ... ........ ..... .. ... ...

................. ..... .. ...... ... ....... . F-26 TableF -7 Compa rison of Measured/Ca lculate d (M/C) Sensor Reacti on Rate Ratios Includ ing all Fast Neutro n Thresh old Reacti ons ...... .... ..............................

.... ....... ... .... .. ..... .... ..... .. .... F-27 Table F-8 Compa rison of Best-E stimate /Calcu lated (BE/C ) Expos ure Rate Ratios ..... .............. .. F-27 Table G-1 Survei llance Capsu le Withdrawal Sched ule .... ... ... ....................

........................... .. ........ G-1 WCAP -18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM

. ( This statement was added by the PRIME system upon its validation)

Westing house Non-Pro prietary Class 3 viii LIST OF FIGURES Figure 2-1 Beaver Valley Unit 1 r,0 Reacto r Geome try at the Core Midpla ne; Octant with No Surveil lance Capsul es .. .... ......... ..... .......... ...... ... ..... ......... .. .............

....... .... ..... .. .............. 2-18 Figure 2-2 Beaver Valley Unit I r,0 Reacto r Geome try at the Core Midpla ne; Octant with Surveil lance Capsul es .. ............... .... .. .... ....... .... ....... ....... .. .. .......... .....

...... .. ... ...... ... ..... ... . 2-19 Figure 2-3 Beaver Valley Unit 1 r,z Reacto r Geome try ... ...... .... ..... ..... .... ..... ....

.................. ..... ... .. ... 2-20 Figure 8-1 Beaver Valley Unit 1 Reacto r Coolan t System Heatup Limitat ions (Heatu p Rates of 60 and 100°F/hr) Applica ble for 50 EFPY (with Flange Requir ements and withou t Margin s for Instrum entatio n Errors) using the 1998 Edition through the 2000 Addend a App. G Method ology (w/ K1c) ....... ..... ... ......... .. ..... ........... ....... ........... ...........

...... ..... ..... ..... .......... 8-3 Figure 8-2 Beaver Valley Unit I Reacto r Coolan t System Cooldo wn Limitat ions (Coold own Rates of 0, -20, -40, -60, and -100°F/hr) Applic able for 50 EFPY (with Flange Requir ements and withou t Marg ins for Instrum entatio n Errors) using the 1998 Edition through the 2000 Adden da App. G Method ology (w/ K 1c) ........... .. .............. ..............

................. ... .... ..... ... . 8-4 Figure 8-1 Compa rison of Beaver Valley Unit 1 Beltlin e P-T Limits to Inlet Nozzle Limits .......... 8-6 Figure B-2 Compa rison of Beaver Valley Unit I Beltlin e P-T Limits to Outlet Nozzle Limits ....... 8-7

• WC AP-181 02-NP Februar y 2018 Revisio n I

• • • This record was final approved on 2/27/2018 8:36:20 AM .

( This statemen t was added by the PRIME system upon its validation

)

Westinghouse Non-Proprietary Class 3 IX EXEC UTIV E

SUMMARY

This report provid es the metho dolog y and results of the generation of heatup and coold own pressure-tempe rature (P-T) limit curves for normal opera tion of the Beave r Valley Unit I reactor vessel. The heatup and coold own P-T limit curves were gener ated using the limiting Adjus ted Reference Temp eratur e (ART) values for Beaver Valley Unit I increased by a small margin for conservatism. The limiting ART values were those of Lower Shell Plate 86903 -1 (Posit ion I. I) at both 1/4 thickness ( I /4 T) and 3/4 thickn ess (3/4T ) locations. The P-T limit curves were generated using the Ki e metho dolog y detail ed in the 1998 Editio n through the 2000 Adden da of the ASME (Ame rican Society of Mechanical Engin eers)

Code, Sectio n XI , Appen dix G. The P-T limit curve genera tion methodology is consis tent with the NRC-appro ved metho dolog y documented in WCA P-140 40-A

, Revision 4.

The P-T limit curve s were generated for 50 effect ive full-power years (EFPY ) using heatup rates of 60 and I 00°F/hr, and cool down rates of 0, -20, -40, -60, and -100°F /hr. The curve s were develo ped with the flange requir ement s and without margins for instrumenta tion errors. They can be found in Figures 8-1 and 8-2.

Appen dix A contai ns the thermal stress intensity factor s for the maxim um heatup and coold own rates at 50 EFPY .

Appen dix B contai ns a P-T limit evalua tion of the reacto r vessel inlet and outlet nozzles based on a flaw postulated at the inside surface of the reacto r vessel nozzle corner. As discus sed in Appen dix B, the P-T limit curves genera ted based on the limiting cylindrical beltline material (Lowe r Shell Plate 86903 -1) bound the P-T limit curves for the reacto r vessel inlet and outlet nozzles for Beave r Valley Unit 1 at 50 EFPY.

Appen dix C contai ns discussion of the other ferriti c Reactor Coola nt Pressure Boundary (RCP B) compo nents relative to P-T limits . As discus sed in Appen dix C, all of the other ferritic RCPB compo nents meet the applicable requir ement s of Sectio n II] of the ASME Code.

Appen dix D contai ns an updated credibility evalua tion for Beaver Valley Unit 1 consid ering all applicable sister plant surveillance program data and updated Beave r Valley Unit I survei llance capsu le fluence values.

Appen dix E contai ns a Pressurized Therm al Shock (PTS) evaluation for the reacto r vessel materials at 50 EFPY for Beave r Valley Unit I. In the previo us Beave r Valley Unit 1 analysis of record , the limiting reactor vessel plate material, Lowe r Shell Plate 86903

-1 , was predicted to excee d the RT PTs screen ing criteria of 270°F for plates at 39.6 EFPY of plant operat ion. Howe ver, as discussed in Appen dix E, this material, while still the limiting material, is now predic ted to remain under the RTPTs screen ing limit through 50 EFPY (end oflice nse extens ion [EOL E]).

Appen dix F contai ns an evaluation of the neutro n dosimetry contained in the Beave r Valley Unit 1 survei llance capsu les withdrawn to date.

Appen dix G contai ns an updated survei llance capsu le withdrawal sched ule for Beave r Valley Unit 1.

WCAP -18102 -NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION Heatup and cool down P-T limit curves are calculate d using the adjusted RT NDT (referenc e nil-ductility temperat ure) correspon ding to the limiting beltline region material of the reactor vessel. The adjusted RT NDT of the limiting material in the core region of the reactor vessel is determin ed by using the unirradia ted reactor vessel material fracture toughnes s propertie s, estimatin g the radiation -induced

~RT NDT, and adding a margin. The unirradia ted RT NDT is designate d as the higher of either the drop weight nil-ductility transition temperat ure (NOTT) or the temperat ure at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansio n (normal to the major working direction )

minus 60°F.

RT NDT increases as the material is exposed to fast-neutron radiation. Therefor e, to find the most limiting RT NDT at any time period in the reactor's life, ~ RT NDT due to the radiation exposure associate d with that time period must be added to the unirradia ted RT NDT (RT NDT( U)), The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The U.S . Nuclear Regulatory Commiss ion (NRC) has published a method for predictin g radiation embrittle ment in Regulatory Guide 1.99, Revision 2 [Ref. I]. Regulato ry Guide 1.99, Revision 2 is used for the calculatio n of Adjusted Referenc e Tempera ture (ART) values (RT NDT(U) + ~RT DT + margins for uncertain ties) at the l /4T and 3/4T locations, where Tis the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown P-T limit curves documen ted in this report were generated using the most limiting ART values (increased by a small margin for conserva tism) and the NRC-app roved methodo logy documen ted in WCAP-1 4040-A , Revision 4 [Ref. 2]. Specifically, the K c methodol 1 ogy of the 1998 Edition through the 2000 Addenda of ASME Code,Section XI, Appendi x G [Ref.

3] was used. The K1c curve is a lower bound static fracture toughnes s 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 temperat ure.

Allowabl e operating limits are then determin ed using the allowabl e stress intensity factors.

The following statemen t excludes the fluence method . For the purpose of this plant-spe cific evaluatio n, the P-T limit curve generatio n method of WCAP-1 4040-A , Revision 4 is identical to the P-T limit curve generatio n method of WCAP-1 4040-NP -A , Revision 2 [Ref. 21] with the addition of the allowanc e for Beaver Valley Unit 1 to also utilize ASME Code Case N-640 and ASME Code Section XI , Appendi x G (1995 Edition through 1996 Addenda ). The fluence method utilized is detailed in Section 2.

The purpose of this report is to present the calculatio ns and the developm ent of the Beaver Valley Unit 1 heatup and cooldown P-T limit curves for 50 EFPY. This report documen ts the calculate d ART values and the developm ent of the P-T limit curves for normal operation. The calculate d ART values for 50 EFPY are documen ted in Section 7 of this report. The fluence projectio ns used in calculatio n of the ART values are provided in Section 2 of this report.

The P-T limit curves herein were generated without instrumen tation errors. The reactor vessel flange requirem ents of 10 CFR 50, Appendix G [Ref. 4] have been incorpora ted in the P-T limit curves. As discussed in Appendix B, the P-T limit curves generated in Section 8 bound the P-T limit curves for the reactor vessel inlet and outlet nozzles for Beaver Valley Unit 1 at 50 EFPY. Discussio n of the other ferritic RCPB compone nts relative to P-T limits is contained in Appendix C.

WCAP-18 102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westingh ouse Non-Pro prietary Class 3 1-2 Append ix D contains an updated credibil ity evaluation for surveill ance data applicab le to Beaver Valley Unit I , and Append ix E contains a Pressuri zed Thermal Shock evaluati on for the Beaver Valley Unit I reactor vessel materials at 50 EFPY.

Append ix F contains an evaluati on of the neutron dosimet ry containe d in the Beaver Valley Unit I surveill ance capsules withdraw n to date.

Append ix G contains an updated surveill ance capsule withdrawal schedul e for Beaver Valley Unit I.

WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM. ( This statement was added by the PRIME system upon its validation)

Westingh ouse Non-Pro prietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

A discrete ordinate s (SN) transpor t analysis was perfonn ed for the Beaver Valley Unit I reactor to determin e the neutron radiatio n environ ment within the reactor pressure vessel. In this analysis , radiation exposur e paramet ers were establish ed on a plant- and fuel-cyc le-speci fic basis. An evaluati on of the dosimet ry sensor sets from the first five surveilla nce capsules is provided in Append ix F. The dosimet ry analysis shows that the +/-20% ( 1cr) acceptan ce criteria specifie d in Regulato ry Guide 1.190 [Ref. 6) is met. The validate d calculat ions form the basis for providin g projectio ns of the neutron exposur e of the reactor pressure vessel for operatin g periods extendin g to 60 EFPY.

All of the calculat ions describe d in this section and in Append ix F were based on nuclear cross-se ction data derived from the Evaluate d Nuclear Data File (ENDF) database (Specifi cally, ENDF /B-Vl).

Furtherm ore, the neutron transpor t evaluati on methodo logies follow the guidanc e of Regulat ory Guide 1.190 [Ref. 6). Additio nally, the tluence calculat ions herein were perform ed in accorda nce with WCAP-14040-A, Revision 4 [Ref. 2), which is a topical report having an NRC approve d method that complie s with NRC Regulat ory Guide 1.190. The method used for the tluence calculat ions is the same as that employe d during the analysis of Beaver Valley Unit 1 Capsule Y docume nted in WCAP- 15571, Revisio n 0 [Ref. 22].

2.2 DISCRETE ORDINATES ANALYSIS In perform ing the fast neutron exposur e evaluati ons for the Beaver Valley Unit 1 reactor vessel , a series of fuel-cyc le-speci fic forward transpor t calculat ions were complet ed using the followin g three-di mension al tluence rate synthesi s techniqu e:

cp(r, 0, z) = cp(r,0) x cp(r, z) cp(r) where cp(r, 0, z) is the synthesi zed three-di mension al neutron tluence rate distribu tion, cp(r, 0) is the transpor t solution in r,0 geometr y, cp(r, z) is the two-dim ensiona l solution for a cylindri cal reactor model using the actual axial core power distribut ion, and cp(r) is the one-dim ensional solution for a cylindri cal reactor model using the same source per unit height as that used in the r,0 two-dim ensiona l calculat ion.

This synthesi s procedu re was complet ed for each operatin g cycle at Beaver Valley Unit I.

Plan views of the r,0 geometr y of the Beaver Valley Unit 1 reactor at the core midplan e are shown in Figures 2-1 and 2-2. In each of these figures , a single octant is depicted showing the arrangem ent of surveilla nce capsules , where Figure 2-1 shows an octant with no surveilla nce capsules , and Figure 2-2 shows an octant with surveilla nce capsules. The maximu m exposur e of the pressure vessel occurs in octants with no surveill ance capsules . In develop ing these analytic al models, nominal design dimensi ons were employe d for the various structural compon ents. Likewis e, water tempera tures, and hence, coolant densitie s in the reactor core and downco mer regions of the reactor were taken to be represen tative of full-power operatin g conditio ns. The coolant densitie s were treated on a fuel-cyc le-speci fic basis. The reactor core itself was treated as a homoge neous mixture of fuel , cladding, water, and miscella neous core WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-2 structures such as fuel assembly grids and guide tubes. The geometric mesh description of the r,0 reactor model consiste d of 189 radial by 75 azimuth al intervals. Mesh sizes were chosen to ensure that proper converg ence of the inner iterations was achieve d on a pointwise basis.

The pointwise inner iteration fluence rate converg ence criterion utilized in the r,0 calculations was set at a value of 0.001.

The r,z model used for the Beaver Valley Unit 1 calculat ions is shown in Figure 2-3 . The model extends radially from the centerlin e of the reactor core out to a location interior to the neutron shie ld tank and over an axial span from an elevation approxim ately five feet below to five feet above the active fuel. As in the case of the r,0 models, nominal design dimensi ons and full-power coolant densities were employe d in the calculations. In this case, the homoge nous core region was treated as an equivale nt cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitl y included in the model.

The r,z geometr ic mesh description of this reactor model consiste d of 189 radial by 223 axial intervals

. As in the case of the r,0 calculations, mesh sizes were chosen to ensure that proper converg ence of the inner iterations was achieved on a pointwi se basis. The pointwise inner iteration fluence rate converg ence criterion utilized in the r,z calculat ions was also set at a value of 0.00 I.

The one-dim ensiona l radial model used in the synthesi s procedure consiste d of the same 189 radial mesh intervals included in the r,z model. Thus, radial synthesi s factors could be determined on a meshwi se basis through out the entire geometry.

The data utilized for the core power distribu tions in plant-specific transpor t analyses included cycle-dependent fuel assembl y initial enrichm ents, bumups , and axial power distribu tions. This information was used to develop spatial- and energy- depende nt core source distributions average d over each individual fuel cycle. Therefo re, the results from the neutron transpor t calculations provided data in terms of fuel cycle-av eraged neutron fluence rate, which when multiplied by the appropr iate fuel cycle length, generated the incremental fast neutron exposur e for each fuel cycle. In construc ting these core source distributions, the energy distribution of the source was based on an appropr iate fission split for uranium and plutonium isotopes based on the initial enrichm ent and burnup history of individual fuel assemblies.

From these assembl y-depen dent fission splits, compos ite values of energy release per fission , neutron yield per fission , and fission spectrum were determi ned.

All of the transpor t calculations supporti ng this analysis were carried out using the DORT discrete ordinates code [Ref. 7] and the BUGLE -96 cross-se ction library [Ref.

8]. The BUGLE -96 library provides a 67-grou p coupled neutron -gamma ray cross-section data set produced specifically for light-water reactor (LWR) applications. In these analyses , anisotropic scatterin g was treated with a P5 Legendre expansi on and angular discretiz ation was modeled with an S1 6 order of angular quadratu re.

Energy- and space-d ependen t core power distribu tions, as well as system operatin g temperatures, were treated on a fuel-cycle-specific basis.

In Table 2-1 , axial locations of the Beaver Valley Unit I pressure vessel material welds in terms of the transport models are provided. The axial position of each material is indexed to z = 0.0 cm, which correspo nds to the midplane of the active fuel stack.

Cycle-specific calculat ions were performed for Cycles I through 24, with core thermal powers given in Table 2-2.

WCAP-1 8102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westingho use Non-Prop rietary Class 3 2-3 Neutron exposure data pertinent to the pressure vessel clad/base metal interface are given in Tables 2-3 and 2-4 for fast neutron tluence rate and tluence (E > 1.0 MeV) and iron displacem ents per atom (dpa),

respectively. In each case the data are provided for each operating cycle of the Beaver Valley Unit 1 reactor. The vessel exposure data are presented in tenns of the maximum exposure experienc ed by the pressure vessel at azimuthal angles of 0°, 15°, 30°, and 45° relative to the core cardinal axes as well as the maximum exposure anywhere on the reactor pressure vessel.

Calculate d fast neutron tluence (E > 1.0 MeV) and dpa for the pressure vessel plates are provided in Tables 2-5 and 2-6, respectively. Calculate d fast neutron tluence (E > 1.0 MeV) and dpa for the pressure vessel circumfe rential welds are provided in Tables 2-7 and 2-8, while the equivalen t data for the longitudinal welds are provided in Tables 2-9 and 2-10, respectively.

ln Tables 2-3 through 2-10, calculate d exposure values are projected to 32, 36, 40, 48, 50, and 60 EFPY.

Projections were based on the burnup weighted average of Cycles 22 through 24 power distributions and reactor operating conditions with the a rated core power of 2900 MWt. The projected results will remain valid as long as future plant operation is consisten t with these assumptions.

In Table 2-11 , calculated fast neutron tluence (E > 1.0 MeV) for the surveillan ce capsule for the Beaver Valley Unit I reactor is provided. In Table 2-12, a summary of the lead factors for each capsule at the time of removal from the reactor ( or at end of Cycle 24 if still inserted) is provided.

2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculate d neutron exposure of the Beaver Valley Unit I reactor pressure vessel materials is based on the recomme nded approach provided in Regulato ry Guide 1.190. In particula r, the qualification of the methodo logy was carried out in the following four stages:

I. Compari son of calculations with benchma rk measurements from the Pool Critical Assembl y (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2. Compari sons of calculations with surveilla nce capsule and reactor cavity measurem ents from the H.B. Robinson power reactor benchma rk experiment.

3. An analytical sensitivity study addressin g the uncertainty compone nts resulting from importan t input paramete rs applicable to the plant-specific transport calculatio ns used in the neutron exposure assessme nts.

4. Compari sons of the plant-specific calculatio ns with all available dosimetry results from the Beaver Valley Unit 1 surveillance program.

The first phase of the methods qualification (PCA comparis ons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniqu es and associated cross sections. This phase, however, did not test the accuracy of commercial core neutron source calculations, nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (H. B. Robinson comparis ons) addressed uncertainties in these additiona l areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluatio ns.

The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculatio nal methods' approxim ations as well as to a lack of WCAP-18 102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghou se Non-Propri etary Class 3 2-4 knowledge relative to various plant-spec ific input parameters . The overall calculational uncertainty applicable to the Beaver Valley Unit I analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessmen t (compariso ns with Beaver Valley Unit I measurem ents) was used solely to demonstrate the validity of the transport calculation s and to confirm the uncertainty estimates associated with the analytical results. The compariso n was used only as a check and was not used in any way to modify the calculated surveillanc e capsule or pressure vessel neutron exposures.

Table 2-13 summarize s the uncertainties developed from the first three phases of the methodolo gy qualification. Additional information pertinent to these evaluations is provided in Reference 2. The net calculation al uncertainty was determined by combining the individual componen ts in quadrature Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurem ent compariso ns given in Appendix F support these uncertainty assessmen ts for Beaver Valley Unit I.

WCAP-181 02-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westingho use Non-Prop rietary Class 3 2-5 Table 2-1 Pressure Vessel Material Weld Axial Locations Axial Location<*>

Material Inches cm Lower Shell to Lower Closure Head Weld -121.10 -307.59 Lower Shell to Intermed iate Shell Weld -20.50 -52.07 Intermed iate Shell to Upper Shell Weld 80.20 203.71 Inlet Nozzle to Upper Shell Weld - Lowest Extent 100.46 255.17 Outlet Nozzle to Upper Shell Weld - Lowest Extent 102.71 260.88 Note:

(a) Axial locations are with respect to the core midplane at O cm .

Table 2-2 Reactor Core Thermal Power Level for Beaver Valley Unit 1 Cycle Core Power (MWt)

I 2652 2 2652 3 2652 4 2652 5 2652 6 2652 7 2652 8 2652 9 2652 10 2652 11 2652 12 2652 13 2652 14 2660(a) 15 2689 16 2689 17 2689 18 2799(b) 19 2900 20 2900 21 2900 22 2900 23 2900 24 2900 Notes:

(a) There was a mid-cycle uprate during Cycle 14.

(b) There were two mid-cycle uprates during Cycle 18.

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Pro prietary Class 3 2-6 Table 2-3 Calcula ted Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressur e

Vessel Clad/Ba se Metal Interfac e Cumula tive Maxim um Fast Neutron Fluence Rate (E > 1.0 MeV)

Operati ng (n/cm 2-s)

Cycle Elevation<*>

Time (EFPY) oo 15° 30° 45° Maxim um (cm)

I 1.2 5.51E+ I0 2.53E+I 0 1.42E+I 0 9.20E+0 9 5.51E+ I0 -54.00 2 1.9 5.79E+I O 2.69E+I 0 1.54E+I 0 l.0IE+I 0 5.79E+ I0 104.00 3 2.7 6.28E+ I0 2.87E+I 0 1.58E+I 0 1.00E+I 0 6.28E+I O -54.00 4 3.6 4.70E+I O 2.21E+I 0 1.21E+I O 7.72E+0 9 4.70E+ I0 -54.00 5 4.8 4.56E+ I0 2.15E+I 0 1.18E+I 0 7.79E+0 9 4.56E+ I0 -54.00 6 5.9 3.52E+ I0 l.91E+I 0 l.19E+I 0 7.65E+0 9 3.52E+ I0 -54.00 7 7.1 4.30E+ I 0 2.14E+I 0 l.16E+I 0 7.67E+0 9 4.30E+ I 0 -60.00 8 8.2 4.23E+ I0 2.17E+I 0 l.19E+I O 7.52E+0 9 4.23E+ I0 -54.00 9 9.6 3.79E+ I0 l.99E+I 0 l.19E+I O 8.25E+0 9 3.79E+1 0 -54.00 10 10.8 2.96E+ I0 l.56E+I 0 l.07E+I 0 7.49E+0 9 2.96E+ I0 -60.00 11 11.8 2.97E+I O l.50E+I 0 l.llE+I 0 8.I0E+0 9 2.97E+ I0 -54.00 12 12.9 3.08E+ 10 1.63E+I 0 1.16E+I 0 7.23E+0 9 3.08E+ I0 -54.00 13 14.3 3.21E+ I0 l.65E+I 0 I.IIE+I O 7.49E+0 9 3.21E+ I0 -58.00 14 15.6 3.25E+ I0 l.44E+I 0 8.45E+0 9 5.86E+0 9 3.25E+ I0 -60.00 15 16.9 2.77E+ I0 l.38E+I 0 9.34E+0 9 6.49E+0 9 2.77E+ I0 -118.00 16 18.4 3.24E+ I0 l.63E+I 0 9.60E+0 9 6.72E+0 9 3.24E+ I0 -54.00 17 19.6 3.18E+ I0 l.57E+I 0 8.99E+0 9 5.99E+0 9 3.18E+ I0 -60.00 18 21.0 3.71E+ I0 l.78E+I 0 9.87E+0 9 6.51E+0 9 3.71E+ I0 -54.00 19 22.5 3.22E+ I0 l.68E+I 0 1.0IE+I 0 7.29E+0 9 3.22E+ I0 -54.00 20 23.8 3.78E+ I0 l.80E+I 0 l.02E+I 0 7.19E+0 9 3.78E+I O 50.00 21 25.2 4.02E+ I0 l.81E+I 0 9.69E+0 9 6.61 E+09 4.02E+ I0 50.00 22 26.6 3.74E+ I0 l.73E+I 0 I.00E+I 0 7.20E+0 9 3.74E+ I0 50.00 23 28.0 3.93E+ I0 1.81E+I 0 1.0IE+I 0 7.05E+0 9 3.93E+ I0 -54.00 24 29.3 3.44E+ I 0 l.69E+I 0 9.71 E+09 6.84E+0 9 3.44E+ I 0 50.00 Note:

(a) The elevation represen ts the distance between the maximum exposure location and the core midplane at O cm.

WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 2-7 Table 2-3 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Cumulative Maximum Fast Neutron Fluence (E > 1.0 MeV)

Operating (n/cm 2) Elevation<a>

Cycle Time (EFPY) oo 15° 30° 45° Maximum (cm)

I 1.2 2.02E+18 9.27E+17 5.18E+17 3.37E+17 2.02E+18 -54.00 2 1.9 3.23E+18 1.49E+ 18 8.39E+17 5.47E+17 3.23E+l 8 -54.00 3 2.7 4.79E+18 2.21 E+ 18 l.23 E+ 18 7.96E+17 4.79E+18 -54.00 4 3.6 6.16E+18 2.85E+ 18 l.58E+18 l.02E+18 6.16E+18 -54.00 5 4.8 7.87E+l8 3.66E+18 2.03E+18 l.31E+18 7.87E+l8 -54.00 6 5.9 9.10E+ 18 4.33E+18 2.44E+ 18 l.58E+18 9.10E+18 -54.00 7 7.1 l.08E+19 5.17E+l8 2.90E+18 l.88E+l8 l.08E+19 -54.00 8 8.2 l.23E+ l9 5.92E+18 3.31E+18 2.15E+18 l.23E+ 19 -54.00 9 9.6 1.39E+19 6.79E+l8 3.83E+l8 2.50E+18 1.39£+19 -54.00 10 10.8 1.50E+l 9 7.37E+l8 4.24E+l8 2.79E+l8 l.50E+ l 9 -54.00 11 11.8 l.59E+19 7.83E+18 4.58E+18 3.03E+18 l.59E+l 9 -54.00 12 12.9 1.70E+ 19 8.41E+ 18 4.99E+l8 3.29E+18 l.70E+19 -54.00 13 14.3 l.84E+ 19 9.12E+18 5.47E+ 18 3.62E+18 l.84E+l 9 -54.00 14 15.6 l.98E+l9 9.72E+18 5.82E+18 3.86E+l8 l.98E+19 -54.00 15 16.9 2.09E+ 19 l.03E+ 19 6.21E+18 4.13E+18 2.09E+19 -54.00 16 18.4 2.24E+19 l.10E+ 19 6.65E+18 4.43E+ 18 2.24E+19 -54.00 17 19.6 2.36E+l9 l.16E+l9 6.99E+ 18 4.67E+l8 2.36E+19 -54.00 18 21.0 2.53E+19 l.24E+l9 7.43E+18 4.95E+18 2.53E+19 -54.00 19 22.5 2.68E+19 l.32E+ l9 7.90E+18 5.29E+18 2.68E+19 -54.00 20 23.8 2.84E+19 1.40E+19 8.33E+l8 5.60E+l8 2.84E+19 -54.00 21 25.2 3.01E+19 l.48E+ 19 8.75E+18 5.89E+18 3.01E+19 -54.00 22 26.6 3.17E+1 9 l.55E+ 19 9.19E+18 6.20E+18 3.17E+l9 -54.00 23 28.0 3.34E+1 9 l.63 E+ l9 9.62E+18 6.50E+18 3.34E+19 -54.00 24 29.3 3.48E+ 19 1.70E+19 I.00E+19 6.78E+18 3.48E+ 19 -54.00 32.0 3.80E+l9 l.85 E+ 19 1.09E+19 7.38E+18 3.80E+l9 -54.00 36.0 4.27E+1 9 2.07E+l9 1.21E+19 8.27E+18 4.27E+ 19 -54.00 40.0 4.73E+l9 2.29E+19 l.34E+l 9 9.15E+18 4.73E+l9 -54.00 Future 48.0 5.66E+l9 2.72E+19 1.59E+19 l.09E+19 5.66E+19 -54.00 50.0 5.89E+19 2.83E+19 l.65E+19 1.14E+19 5.89E+l9 -54.00 60.0 7.06E+19 3.38E+l 9 l.96E+19 l.36E+ 19 7.06E+19 -54.00 Note:

(a) The elevation represents the distance between the maximum exposure location and the core midplane at O cm.

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse No n-Propriet ary Class 3 2-8 Table 2-4 Calculated Iron Displacements per Atom at the Pressure Vessel Clad/Base Metal Interface Cumulat ive Maximum Iron Displacements per Atom Operatin g (dpa) Elevation (a)

Cycle Time (EFPY) oo 15° 30° 45° Maximum (cm)

I 1.2 3.23E-03 I .50E-03 8.29E-04 5.40E-04 3.23E-03 -54.00 2 1.9 5.18E-03 2.42E-03 I .34E-03 8.77E-04 5. I 8E-03 -54.00 3 2.7 7.69E-03 3.58E-03 I .97E-03 1.28E-03 7.69E-03 -54.00 4 3.6 9.88E-03 4.62E-03 2.53E-03 I .64E-03 9.88E-03 -54.00 5 4.8 I .26E-02 5.93E-03 3.24E-03 2. 11 E-03 1.26E-02 -54.00 6 5.9 1.46E-02 7.01 E-03 3.9 IE-03 2.54E-03 1.46E-02 -54.00 7 7.1 1.73E-02 8.37E-03 4.64E-03 3.02E-03 1.73 E-02 -54.00 8 8.2 1.97E-02 9.59E-03 5.3 I E-03 3.44E-03 I .97E-02 -54.00 9 9.6 2.23E-02 l . I0E-02 6.14E-03 4.02E-03 2.23E-02 -54.00 10 10.8 2.41 E-02 l.19 E-02 6.78E-03 4.47E-03 2.41 E-02 -54.00 11 11.8 2.56E-02 1.27E-02 7.32E-03 4.86E-03 2.56E-02 -54.00 12 12.9 2.73E-02 1.36 E-02 7.98E-03 5.28E-03 2.73E-02 -54.00 13 14.3 2.95E-02 I .48 E-02 8.75E-03 5.80E-03 2.95E-02 -54.00 14 15.6 3.17E-02 I .57E-02 9.3 1E-03 6.19E-03 3.17E-02 -54.00 15 16.9 3.36E-02 I .67E-02 9.93E-03 6.62E-03 3.36E-02 -54.00 16 18.4 3.59E-02 l.79E-02 I .06E-02 7.IIE-03 3.59E-02 -54.00 17 19.6 3.79E-02 I .88E-02 I . 12E-02 7.48E-03 3.79E-02 -54.00 18 21.0 4.05E-02 2.01 E-02 I .19E-02 7.94E-03 4.05E-02 -54.00 19 22 .5 4.29E-02 2. 14E-02 I .26E-02 8.48E-03 4.29E-02 -54.00 20 23.8 4.55E-02 2.26E-02 1.33E-02 8.97E-03 4.55E-02 -54.00 21 25.2 4.83E-02 2.39E-02 1.40E-02 9.44E-03 4.83E-02 -54.00 22 26.6 5.09 E-02 2.5 1E-02 1.47 E-02 9.94E-03 5.09E-02 -54.00 23 28.0 5.36E-02 2.64E-02 l .54E-02 1.04 E-02 5.36E-02 -54.00 24 29.3 5.59E-02 2.75E-02 l .60E-02 I .09E-02 5.59E-02 -54.00 32.0 6.09E-02 2.99E-02 1.74E-02 I .18E-02 6.09E-02 -54.00 36.0 6.84E-02 3.35E-02 I .94E-02 1.33E-02 6.84E-02 -54.00 40.0 7.59E-02 3.70E-02 2.14E-02 1.47 E-02 7.59E-02 Future -54.00 48.0 9.08 E-02 4.41 E-02 2.54E-02 l.75E-02 9.08E-02 -54.00 50.0 9.45E-02 4.59E-02 2.64E-02 1.82E-02 9.45E-02 -54.00 60.0 1.1 3E-01 5.47E-02 3.14E-02 2.l 7E-02 l.13E-0 1 -54.00 Note:

(a) The elevation represents the di stance between the maxim um exposure location and the core midplane at O cm .

WCAP-18 102-NP February 2018 Revisio n 1

• • • This record was final approved on 2/27/20 18 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Pro prietary Class 3 2-9 Table 2-5 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Plates Maximum Fast Neutron Fluence Cumulative Operating (E > 1.0 MeV)

Cycle (n/cm 2)

Time (EFPY) Upper Intermediate Lower Shell Shell Shell I 1.2 I.79E+ l 7 2.01E+18 2.02E+ l8 2 1.9 3.87E+ l7 3.22E+ l8 3.23E+18 3 2.7 5.48E+ 17 4.78E+18 4.79E+ l8 4 3.6 6.83E+ l7 6.14E+ l8 6.16E+18 5 4.8 8.81E+17 7.85E+18 7.87E+ l8 6 5.9 9.89E+17 9.09E+ l8 9.I0E+1 8 7 7.1 l.l6E+l 8 l.08 E+ 19 1.08E+ l9 8 8.2 1.28E+ l8 l.22E+1 9 l.23E+l 9 9 9.6 I .44E+I 8 l.39E+l 9 l.39E+ l9 10 10.8 l.54 E+ l8 l.50E+l 9 1.50E+19 11 11.8 l.63 E+ 18 l.59E+ l9 l.59E+l 9 12 12.9 I.73E+ l8 I.70E+ 19 1.70E+ I 9 13 14.3 1.86E+ l 8 1.84E+l 9 1.84E+ 19 14 15 .6 2.06E+18 I.97E+l 9 l.98E+ 19 15 16.9 2.24E+18 2.09E+ l9 2.09E+19 16 18.4 2.45E+ 18 2.24E+ l9 2.24E+ l9 17 19.6 2.64E+ l8 2.36E+ I 9 2.36E+ 19 18 21.0 2.84E+ 18 2.52E+ l9 2.53E+ l9 19 22.5 3.04E+l 8 2.67E+ l9 2.68E+ l9 20 23.8 3.24E+18 2.83E+ 19 2.84E+ l9 21 25.2 3.47E+18 3.01E+19 3.01 E+19 22 26.6 3.67E+ l8 3.17E+ l9 3.17E+1 9 23 28.0 3.88E+ I 8 3.34E+ l9 3.34E+ l9 24 29.3 4.07E+ l8 3.48E+ l9 3.48E+19 32.0 4.48E+ 18 3.79E+ l9 3.80E+19 36.0 5.08£+1 8 4.26E+ 19 4.27E+ l9 40.0 5.68E+ l8 4.72E+ l9 4.73E+19 Future 48.0 6.88E+ l8 5.65E+ l9 5.66E+19 50.0 7.18E+18 5.88E+ l9 5.89E+19 60.0 8.68E+l 8 7.04E+ l9 7.06E+ l9 WCAP-1 8102-NP February 2018 Revision I

• • • Th is record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westingho use Non-Prop rietary Class 3 2-10 Table 2-6 Calculated Iron Displacements per Atom at the Pressure Vessel Plates Cumulative Maximum Iron Displacements per Atom Operating (dpa)

Cycle Time Upper Intermediate Lower (EFPY) Shell Shell Shell 1 1.2 2.98E-04 3.23E-03 3.23E-03 2 1.9 6.43E-04 5. l 7E-03 5.18E-03

.) 2.7 9.l lE-04 7.68E-03 7.69E-03 4 3.6 1.14E-03 9.86E-03 9.88E-03 5 4.8 1.47E-03 1.26E-02 l .26E-02 6 5.9 l .65E-03 1.46E-02 1.46E-02 7 7.1 l.92E-03 1.73E-02 l .73E-02 8 8.2 2.14E-03 l .96E-02 1.97E-02 9 9.6 2.41 E-03 2.23E-02 2.23E-02 10 10.8 2.57E-03 2.41 E-02 2.41 E-02 11 11.8 2.72E-03 2.55E-02 2.56E-02 12 12.9 2.89E-03 2.73E-02 2.73E-02 13 14.3 3.1 I E-03 2.95E-02 2.95E-02 14 15.6 3.43E-03 3.16E-02 3.17E-02 15 16.9 3.73E-03 3.35E-02 3.36E-02 16 18.4 4.08E-03 3.58E-02 3.59E-02 17 19.6 4.40E-03 3.78E-02 3.79E-02 18 21.0 4.73E-03 4.04E-02 4.05E-02 19 22.5 5.05E-03 4.28E-02 4.29E-02 20 23.8 5.39E-03 4.54E-02 4.55E-02 21 25.2 5.77E-03 4.82E-02 4.83E-02 22 26.6 6. l0E-03 5.08E-02 5.09E-02 23 28.0 6.46E-03 5.35E-02 5.36E-02 24 29.3 6.77E-03 5.58E-02 5.59E-02 32.0 7.45£-03 6.08E-02 6.09E-02 36.0 8.45E-03 6.83E-02 6.84E-02 40.0 9.44E-03 7.57E-02 7.59E-02 Future 48.0 1. I 4E-02 9.06E-02 9.08E-02 50.0 1.19E-02 9.43E-02 9.45E-02 60.0 1.44E-02 l.13E-01 1.13£-01 WCAP-18 102-NP February 20 I 8 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 2-11 Table 2-7 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds Maximu m Fast Neutron Fluence (E > 1.0 MeV)

Cumulative (n/cm 2)

Operating Cycle Lower Shell Lower Shell Time Outlet Intermed iate Inlet to to Lower (EFPY) Nozzte<a> Nozzte<a> Shell to Intermed iate Closure Upper Shell Shell Head I 1.2 3.29E+15 4.36E+15 1.79E+l7 2.01E+l8 4.53E+14 2 1.9 6.99E+15 9.29E+ 15 3.87E+17 3.22E+18

,., 7.35E+14

.) 2.7 9.78E+l5 l.30E+l6 5.48E+l7 4.78E+18 l.13E+ I 5 4 3.6 l.24E+ 16 1.65E+16 6.83E+17 6.14E+18 1.47E+l5 5 4.8 l.64E+l6 2.17E+16 8.81E+l7 7.85E+18 l.86E+15 6 5.9 l.85E+16 2.45E+16 9.89E+17 9.09E+l8 2.16E+15 7 7.1 2.18E+l6 2.89E+l6 l.16E+l8 l.08E+19 2.60E+l5 8 8.2 2.44E+16 3.23E+16 1.28E+l 8 l.22E+ I9 2.95E+15 9 9.6 2.76E+16 3.64E+16 l.44E+18 1.39E+l9 3.35E+15 10 10.8 2.98E+l6 3.93E+16 l.54E+18 l.50E+l 9 3.63E+15 11 11.8 3.17E+16 4.18E+16 l.63E+ 18 l.59E+l 9 3.85E+15 12 12.9 3.39E+l6 4.47E+ I 6 1.73E+18 l.70E+ 19 4.1 IE+15 13 14.3 3.68E+16 4.84E+l6 l.86E+l 8 l.84E+19 4.45E+15 14 15.6 4.15E+16 5.46E+16 2.06E+18 l.97E+19 4.84E+15 15 16.9 4.60E+16 6.05E+16 2.24E+l8 2.09E+19 5.19E+l5 16 18.4 5.12E+16 6.73E+l6 2.45E+l 8 2.24E+l9 5.61E+15 17 19.6 5.59E+16 7.33E+l6 2.64E+18 2.36E+ 19 5.98E+l5 18 21.0 6.04E+ 16 7.92E+l6 2.84E+18 2.52E+l9 6.41E+15 19 22.5 6.5 1E+16 8.53E+ l6 3.04E+18 2.67E+19 6.80E+15 20 23.8 6.97E+16 9.14E+l6 3.24E+18 2.83E+l9 7.21E+l5 21 25.2 7.49E+ 16 9.80E+l6 3.47E+l 8 3.01E+19 7.67E+l5 22 26.6 7.96E+16 l.04E+17 3.67E+ 18 3.17E+19 8.08E+ 15 23 28.0 8.44E+16 l.10E+17 3.88E+ 18 3.34E+19 8.52E+ 15 24 29.3 8.89E+16 l.16E+ 17 4.07E+l8 3.48E+ 19 8.91E+15 32.0 9.83E+16 1.29E+ l7 4.48E+ 18 3.79E+19 9.74E+15 36.0 1.12E+17 l.47E+17 5.08E+18 4.26E+l9 1.10E+16 40.0 l.26E+ l 7 l.65E+17 5.68E+l 8 4.72E+l 9 Future 1.22E+16 48.0 1.54E+l 7 2.0IE+l7 6.88E+18 5.65E+19 1.46E+16 50.0 1.61 E+ 17 2.10E+17 7.18E+18 5.88E+19 l.53E+16 60.0 1.95E+l 7 2.55E+17 8.68E+18 7.04E+ 19 l.83E+16 Note:

(a) Exposure for outlet and inlet nozzles is at the lowest extent of the weld.

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Pro prietary Class 3 2-12 Table 2-8 Calculated Iron Displac ements per Atom at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds Maximum Iron Displac ements per Atom Cumulative (dpa)

Operating Cycle Lower Shell Lower Shell Time Outlet Interme diate Inlet to to Lower (EFPY) Nozzte<al Nozzte<al Shell to Interme diate Closure Upper Shell Shell Head 1 1.2 l.OIE-05 1.19E-05 2.98E-04 3.23E-03 3.12E-06 2 1.9 2.03E-05 2.38E-05 6.43E-04 5. 17E-03 5.06E-06 3 2.7 2.94E-05 3.43E-05 9.1 I E-04 7.68E-03 7.76E-06 4 3.6 3.70E-05 4.33E-05 l.14E-03 9.86E-03 1.01 E-05 5 4.8 4.76E-05 5.57E-05 1.47E-03 I .26E-02 l.27E-05 6 5.9 5.45E-05 6.36E-05 1.65E-03 1.46E-02 l .48E-05 7 7.1 6.39E-05 7.47E-05 1.92E-03 I. 73E-02 1.78E-05 8 8.2 7.18E-05 8.39E-05 2.14E-03 l.96E-0 2 2.02E-05 9 9.6 8.16E-05 9.53E-05 2.41 E-03 2.23E-02 2.30E-05 IO 10.8 8.76E-05 I .02E-04 2.57E-03 2.4IE-0 2 2.49E-05 11 11.8 9.28E-05 l .08E-04 2.72E-03 2.55E-02 2.64E-05 12 12.9 9.89E-05 I. l 5E-04 2.89E-03 2.73E-02 2.82E-05 13 14.3 l .07E-04 1.25E-04 3.I lE-03 2.95E-02 3.05E-05 14 15.6 l.16E-0 4 1.35E-04 3.43E-03 3.16E-02 3.30E-05 15 16.9 1.24E-04 I .45E-04 3.73E-03 3.35E-02 3.53E-05 16 18.4 l .34E-04 I .57E-04 4.08E-03 3.58E-02 3.82E-05 17 19.6 l .43E-04 l .68E-04 4.40E-03 3.78E-02 4.07E-05 18 21.0 1.54E-04 I.80E-04 4.73E-03 4.04E-02 4.35E-05 19 22.5 1.64E-04 l.92E-04 5.05E-03 4.28E-02 4.62E-05 20 23.8 l.75E-0 4 2.05E-04 5.39E-03 4.54E-02 4.90E-05 21 25 .2 l .87E-04 2.19E-04 5.77E-03 4.82E-02 5.20E-05 22 26.6 1.98E-04 2.3 IE-04 6.IOE-03 5.08E-02 5.48E-05 23 28.0 2.09E-0 4 2.44E-04 6.46E-03 5.35E-02 5.78E-05 24 29.3 2.19E-0 4 2.56E-04 6.77E-03 5.58E-02 6.04E-05 32.0 2.41 E-04 2.81 E-04 7.45E-03 6.08E-02 6.60E-05 36.0 2.72E-0 4 3.19E-04 8.45E-03 6.83E-02 7.43E-05 40.0 3.04E-04 3.56E-04 9.44E-03 7.57E-02 Future 8.25E-05 48.0 3.68E-0 4 4.30E-04 l .14E-02 9.06E-02 9.90E-05 50.0 3.83E-0 4 4.48E-04 1.19E-02 9.43E-02 l.03E-0 4 60.0 4.63E-04 5.41 E-04 I .44E-02 1. I 3E-01 1.24E-04 Note:

(a) Exposure for outlet and inlet nozzles is at the lowest extent of the weld WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Pro prietary Class 3 2-13 Table 2-9 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Longitudinal Welds Maximum Fast Neutron Fluence Cumulative (E > 1.0 MeV)

Operating Cycle (n/cm 2)

Time (EFPY) 45° Intermediate 45° Lower Shell Shell 1 1.2 3.36E+ l7 3.37E+ l7 2 1.9 5.46E+ 17 5.47E+ l7 3 2.7 7.95E+17 7.96E+17 4 3.6 l.02E+I 8 l.02E+l 8 5 4.8 l.31E+l 8 l.31E+1 8 6 5.9 I.SSE+ 18 I.SSE+ 18 7 7.1 l.88E+l 8 l.88E+l 8 8 8.2 2.14E+ l8 2.15E+ l8 9 9.6 2.50E+18 2.50E+18 10 10.8 2.78E+18 2.79E+ l8 11 11.8 3.03E+18 3.03E+ l8 12 12.9 3.29E+18 3.29E+18 13 14.3 3.61E+18 3.62E+18 14 15 .6 3.85E+18 3.86E+ 18 15 16.9 4.12E+18 4.13E+ l8 16 18.4 4.42E+ 18 4.43E+ l8 17 19.6 4.66E+ l8 4.67E+ l8 18 21.0 4.94E+ l8 4.95E+ 18 19 22.5 5.28E+ l8 5.29E+18 20 23.8 5.58E+18 5.60E+ l8 21 25.2 5.87E+ l8 5.89E+ l8 22 26.6 6.18E+ l8 6.20E+18 23 28.0 6.48E+18 6.50E+ l8 24 29.3 6.77E+ l8 6.78E+ l8 32.0 7.37E+18 7.38E+l 8 36.0 8.25E+ l8 8.27E+ l8 40.0 9.13E+ l8 9.15E+18 Future 48.0 l.09E+l 9 l.09E+1 9 50.0 1.13E+19 l.14E+l 9 60.0 l.35E+ 19 l.36E+ 19 WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Pro prietary Class 3 2-14 Table 2-10 Calculated Iron Displac ements per Atom at the Pressur e Vessel Longitu dinal Welds Maxim um Iron Displac ements per Cumula tive Atom Operati ng Cycle (dpa)

Time (EFPY) 45° Interme diate 45° Lower Shell Shell 1 1.2 5.39E-04 5.40E-04 2 1.9 8.76E-04 8.77£-0 4

.) 2.7 l .28E-03 I .28E-03 4 3.6 1.63£-03 I .64E-03 5 4.8 2.1 0E-03 2.11 E-03 6 5.9 2.53£-0 3 2.54£-0 3 7 7.1 3.02E-03 3.02E-03 8 8.2 3.44E-03 3.44E-03 9 9.6 4.0IE-0 3 4.02E-03 10 10.8 4.46£-0 3 4.47E-03 11 11.8 4.85£-0 3 4.86E-03 12 12.9 5.27E-03 5.28E-03 13 14.3 5.79E-03 5.80E-03 14 15.6 6.18E-03 6.19E-03 15 16.9 6.61 E-03 6.62£-0 3 16 18.4 7.I0E-0 3 7.11£-03 17 19.6 7.47E-03 7.48E-03 18 21.0 7.93E-03 7.94E-03 19 22.5 8.47E-03 8.48E-03 20 23.8 8.95E-03 8.97E-03 21 25.2 9.42E-03 9.44£-03 22 26.6 9.92E-03 9.94E-03 23 28.0 1.04£-02 1.04E-02 24 29.3 l .09E-02 I .09E-02 32.0 l .18E-02 l.18E-02 36.0 I .32E-02 1.33£-02 40.0 I .46E-02 I .47E-02 Future 48.0 l.75E-0 2 l.75E-0 2 50.0 I .82E-02 l.82E-0 2 60.0 2.17E-02 2.17E-02 WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westing house Non-Pro prietary Class 3 2-15 Table 2-11 Calcul ated Fast Neutro n Fluenc e (E > 1.0 MeV) at the Center of the Survei llance Capsul es Cumula tive 1 eutron (E > 1.0 MeV) Fluence (n/cm 2)

Operati ng Time V u w y X s<a) T(b) z<c)

Cycle (EFPY) (150) (25°) (25°) (25°) (]50) (45°/25°/15°) (35°/25°) (35°/15° )

I 1.2 2.97E+ 18 1.99E+l8 l.99E+ l8 l.99E+ l8 2.97E+l 8 l.03E+ l8 I.32 E+ l8 l.32E+I 8 2 1.9 -- 3.26E+ I8 3.26E+ l8 3.26E+ 18 4.84E+ I 8 I.69 E+ 18 2.17E+ l8 2.17E+l 8 3 2.7 -- 4.81E+ l8 4.81E+ l8 4.81E+ l8 7.17E+ I8 2.46E+ I8 3.19E+ 18 3.19E+l 8 4 3.6 -- 6.18E+ I8 6. 18E+ l8 6.18E+ l8 9.25E+ I8 3.15E+I 8 4.08E+l 8 4.08E+l 8 5 4.8 -- -- 7.90E+ 18 7.90E+ l8 l.19 E+ 19 4.04E+ l8 5.22E+ l8 5.22E+ l8 6 5.9 -- -- 9.52E+l 8 9.52E+ l8 l.40 E+ l9 4.86E+ l8 6.29E+ l8 6.29E+ I8 7 7.1 -- -- -- 1.13E+ l9 l.67E+ I9 5.77E+ l8 7.44E+ l8 7.44E+ l8 8 8.2 -- -- -- 1.29E+ 19 l.91 E+ l9 6.56E+ l8 8.47E+ l8 8.47E+ 18 9 9.6 -- -- -- l.49E+ l9 2.19E+ 19 7.68E+ l8 9.85E+ I8 9.85E+ l8 10 10.8 -- -- -- l.64E+ l9 2.38 E+ l9 8.52E+ l8 l.09 E+ I9 l.09 E+ l9 11 11.8 -- -- -- l. 76E+ l9 2.52E+ l9 9.28E+ I8 I.21E+ I9 I.24E+l 9 12 12.9 -- -- -- 1.92E+ l9 2.7IE+ 19 I.0I E+ l 9 I.37E+ 19 I .42E+ 19 13 14.3 -- -- -- 2.I0E+ l9 2.93E+ l9 I.I0E+I9 l.55E+ l9 l.65 E+ l9 14 15.6 -- -- -- -- 3. 12 E+ I9 l.18 E+ l9 I.68 E+ I9 1.84 E+ 19 15 16.9 -- -- -- -- 3.30E+ I9 I.26E+ 19 l.83 E+ l9 2.02E+ l9 16 18.4 -- -- -- -- 3.54E+ l9 I.35E+ I9 2.00E+ l9 2.26E+ 19 17 19.6 -- -- -- -- 3.74E+ l9 I.42 E+ l9 2.l3E+l 9 2.45E+ l9 18 21.0 -- -- -- -- 3.99E+ l9 l.5IE+ l9 2.29E+ l9 2.70E+ l9 19 22.5 -- -- -- -- 4.24E+ l9 l.61 E+ l9 2.47E+ l9 2.95E+ l9 20 23.8 -- -- -- -- 4.49E+ l9 l.78 E+ 19 2.63 E+ l9 3.20E+l 9 21 25 .2 -- -- -- -- 4.74E+ l9 l.94 E+ 19 2.80E+ l9 3.46E+ 19 22 26.6 -- -- -- -- 4.99E+ l9 2.1 I E+ l9 2.96E+ l9 3.70E+ l9 23 28.0 -- -- -- -- -- 2.36E+l 9 3. l3 E+ l9 3.95E+ l9 24 29.3 -- -- -- -- -- 2.58E+ l9 3.28E+ l9 4.l8E+ 19 32.0 -- -- -- -- -- 3.06E+ l9 3.60E+ l9 4.65E+ l9 36.0 -- -- -- -- -- 3.77E+ l9 4.07E+ l9 5.36E+ l9 40.0 -- -- -- -- -- 4.48E+ l9 4.54E+ I9 6.07E+ 19 Future 48.0 -- -- -- -- -- 5.90E+ l9 5.49E+l 9 7.49E+l 9 50.0 -- -- -- -- -- 6.25E+ l9 5.72E+l 9 7.84E+ l9 60.0 -- -- -- -- -- 8.02E+ l9 6.91E+ I9 9.61E+ l9 Notes:

(a) Capsule S was moved to a 25° location after Cycle 19 and to a 15° location after Cycle 22.

(b) Capsule Twas mo ved to a 25 ° location after Cycle IO.

(c) Capsule Z was moved to a 15° location after Cycle I 0.

WCAP- 18102-N P Februar y 2018 Revisio n 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statemen t was added by the PRIME system upon its validation)

Westinghou se No n-Proprietary Class 3 2-16 Table 2-12 Summary of Calculated Surveillance Capsule Lead Factors Capsule ID And Location Status Lead Factor V(165 °) Withdrawn EOC I 1.47 u (65°) Withdrawn EOC 4 1.00 w (245°) Withdrawn EOC 6 1.05 y (295 °) Withdrawn EOC 13 1.14 X (285 °) Withdrawn EOC 22 1.57 s (45 °/295 °/285 °i *) In Reactor Q,74(d)

T (5 5°/65 °i b) In Reactor 0.94(d)

Z (305 °/ ]65 °ic) In Reactor J.2Q(d)

Notes:

(a) Ca psule S was moved to the 295° location after Cycle 19 and to the 285° location after Cycle 22.

(b) Caps ule Twas moved to the 65° location after Cycle I 0.

(c) Capsule Z was moved to the 165° location after Cycle I 0.

(d) The lead factors fo r the capsules remain ing in the reactor are calcul ated based on End of Cycle (EOC) 24, the last co mpleted operating cyc le.

WCA P- 18102-N P February 2018 Revision I

• • • This record wa s final a pproved on 2/27/2018 8:36:20 AM . ( This s tatement was added by the PRIME system upon its validation)

Westinghou se Non-Propri etary Class 3 2-17 Table 2-13 Calculational Uncertainties Uncertainty Description Capsule Vessel Inner Radius PCA Compariso ns 3% 3%

H. B. Robinson Compariso ns 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertaint y for Factors not Explicitly Evaluated 5% 5%

Net Calculatio nal Uncertaint y 12% 13%

WCAP-181 02-NP February 2018 Revision 1

... This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse No n-Proprietary Class 3 2-18 c.,

- --_,l.......lgko - - -Sh,I

...0 -

o -

".r*

a-

\!!

"'t;-

33.8 67.5 10t2 135.0 168.8 202-5 (cm)

Figure 2-1 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Mid plane; Octant with No Surveillance Capsules WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/20 18 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Propri etary C lass 3 2-19

. - --llsk- - - -Sh,j g-

"'~-

I-33.8 67.5 1012 135.0 168.8 202-5 236.2 270.0

[cm]

Figure 2-2 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Mid plane; Octant with Surveillance Capsules WCAP-181 02-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This sta tement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-20 z

..,~ ;-,,a::,--:-==, :::,a=

~ l====== ====iia

,_.110 co

~*

~in E,,;

co I

N co

~

I

,_.co 0

7

~J---- -----

N I

'i'),O 59.6 119.2 178,8 238.-4 298.o R

[cml Figure 2-3 Beaver Valley Unit 1 r,z Reactor Geometry WCAP-18102 -NP February 2018 Revision I

• • • Thi s record was final approved on 2/27/2018 8:36:20 AM . ( Thi s statement was added by the PRIME system upon its va lidation )

Westinghouse Non-Propri etary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The requiremen ts for P-T limit curve developme nt are specified in 10 CFR 50, Appendix G [Ref.

4]. The beltline region of the reactor vessel is defined as the following in IO CFR 50, Appendix G:

"the region of the reactor vessel (sh ell 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. "

The Beaver Valley Unit 1 beltline materials traditional ly included the intennedia te and lower shell plate and weld materials; however, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref.

9],

any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 10 17 n/cm 2 (E > 1.0 MeV) at the end of the licensed operating period should be considered in the developme nt of P-T limit curves. The materials that exceed this fluence threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met. As seen from Tables 2-5 and 2-7 of this report, the extended beltline materials include the upper shell forging, upper to intermediate shell girth weld, the nozzle to upper shell welds, and the nozzle forging materials. The inlet and outlet nozzles are considered a part of the extended beltline, as the exposure at the nozzle welds is conservatively used to represent the exposure at the nozzles. Per NRC RIS 2014-11 the nozzle materials must be evaluated for their potential effect on P-T limit curves regardless of exposure

- See Appendix B for more details.

As part of this P-T limit curve developme nt effort, the methodolo gy and evaluation s used to detennine the initial RT NDT values for the Beaver Valley Unit 1 reactor vessel beltline and extended beltline base metal materials were reviewed and updated, as appropriat e. Table 3-1 contains a summary of these methodolo gies. Summary of the best-estim ate copper (Cu) and nickel (Ni) contents, in units of weight percent (wt. %), as well as initial RT DT values for the reactor vessel beltline and extended beltline materials are provided in Table 3-2 for Beaver Valley Unit I. Table 3-3 contains a summary of the initial RT NDT values of the reactor vessel flange and replaceme nt reactor vessel closure head. These values serve as input to the P-T limit curves "flange-no tch" per Appendix G of IO CFR 50 - See Section 6.3 for details.

WCAP-181 02-NP February 2018 Revision l

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of Beaver Valley Unit 1 Reactor Vessel Base Metal Material Initial RTNoT Determination Methodolog ies Reactor Vessel Material Methodology Upper Shell Forging BTP 5-3, Paragraph Bl.I (3t>[Ref. 10]

Intermediate and Lower Shell Plates ASME Code, Section 111 , Subsection NB-2300(bl [Ref. 11]

Inlet and Outlet Nozzle Forgings BWRVIP-173 -A, Alternate Approach icl [Ref. 12]

Notes:

(a) The Beaver Valley Unit I Certified Material Test Report (CMTR) does not list the orientation of the Charpy V-Notch tests results fo r the upper shell forging materi al. However, eve n though BTP 5-3 , Paragraph B 1. 1(3) methodology fo r SA-508, Class 2 material must be used due to this lack of Charpy V-Notch orientation, the initial RT N OT for this material rema ins drop-weight limited and is confirmed (See Table 3-2) due to exce ll ent Charpy V-notch test results (i n the assumed strong-orientation).

(b) The reacto r vesse l beltline plate materi al initial RT NOT va lues were determined in accordance with the methodo logy of ASME Code, Section lll, Subsection NB-2300 [Ref. 11] utilizing CVGraph, Version 6.02 as documented in Westinghouse Lener MCOE-L TR-15-15-NP, Rev ision I [Ref. 12].

(c) T he initial RT NOT va lues of the Beaver Valley Unit I inlet and outlet nozzles were determined in accordance with BWRVIP-173-A [Ref. 13) - See Append ix B for more detail s.

WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-3 Table 3-2 Summary of the Best-Estimate Cu and Ni Weight Percent and Initial RT NDT Values for the Beaver Valley Unit 1 Reactor Vessel Materiats< 11 >

Fracture Chemical Reactor Vessel Material Toughness Composition and Identification Number Heat Number Property Wt.% Wt.% Initial RT NDT(c)

Cu Ni {°F)

Reactor Vessel Beltline Materials Intermediate Shell Plate 86607-1 C4381-l (b) 0. 14 0.62 26 .8 Intermediate Shell Plate 86607-2 C4381-2(b) 0.14 0.62 53.6 Lower Shell Plate 86903-1 C63 I 7-I (b) 0.21 0.54 13 .1 Lower Shell Plate 87203-2 C6293-ib) 0.14 0.57 0.4 Intermediate to Lower Shell Girth Weld 11-714 90136 0.27 0.07 -56 Intermediate Shell Longitudinal Welds 305424 0.28 0.63 -56 19-714 A&8 Lower Shell Longitudinal Welds 305414 0.34 0.61 -56 20-714 A&8 Reactor Vessel Extended Beltline Materials Upper Shell Forging 86604 123V339VAI 0. 12 0.68 40 305414(d) 0.34 0.61 -56 (3951 & 3958)

AOFJ 0.03 0.93 IO Upper Shell to Intermediate Shell Girth Weld 10-714 FOIJ 0.03 0.94 10 EODJ 0.02 1.04 IO HOCJ 0.02 0.93 10 Inlet Nozzle 86608-1 95443-1 0. 10 0.82 48.5 Inlet Nozzle 86608-2 95460-1 0.10 0.82 -15 .2 Inlet Nozzle 86608-3 95712-1 0.08 0.79 11.4 EODJ 0.02 1.04 10 FOIJ 0.03 0.94 10 HOCJ 0.02 0.93 IO Inlet Nozzle Welds D8IJ 0.02 0.97 10 1-7178, 1-717D, l-717F EOEJ 0.01 1.03 10 ICJJ 0.03 0.99 10 JACJ 0.04 0.97 10 Outlet Nozzle 86605-1 95415-1 0.13 0.77 -26.2 Outlet Nozzle 86605-2 95415-2 0.13 0.77 3.3 Outlet Nozzle 86605-3 95444-1 0.09 0.79 IO. I ICJJ 0.03 0.99 10 Outlet Nozzle Welds 1O8J 0.02 0 .97 10 l-717A , l-717C, l-717E JACJ 0.04 0.97 10 HOCJ 0.02 0.93 10 WCAP-18102-N P February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry C lass 3 3-4 Table 3-2 Summary of the Best-Estima te Cu and Ni Weight Percent and Initial RTNoT Values for the Beaver Valley Unit 1 Reactor Vessel Materials(*)

Chemical Fracture Reactor Vessel Material Composition Toughness Heat Number Property and Identification Number Wt.% Wt.% Initial RT NOT (c)

Cu Ni (OF)

Outlet Nozzle Welds EODJ 0.02 1.04 10 l-717A, l-717C, l-717E (continued) FOIJ 0.03 0.94 10 Surveillance Weld Data(*>

Beaver Valley Unit I 305424 0.26 0.61 ---

St. Luc ie Unit I 0.23 0.07 ---

90136 Millstone Unit 2 0.30 0.06 ---

Fort Calhoun 305414 0.35 0.60 ---

Notes:

(a) All va lues origina lly doc um ented in WCA P-15571 , Suppleme nt I, Revision 2 [Re f. 14], unless otherwise noted.

(b) The reactor vesse l beltline plate mate ri al heat numbers were taken fro m the Beaver Valley U nit I CMTRs.

(c) The initia l RTNDT va lues for a ll the reactor vesse l we lds are gene ric. The initia l RTNDT va lues for the base metal materia ls were updated or co nfirmed as discussed in Table 3- I .

(d) The chem istry values for we ld Heat # 3054 14, as reported in WCAP-15571 , Suppleme nt I, Revis ion 2 [Ref. 14],

were rounded up for cons istency w ith the reactor vesse l be ltline we ld material that shares thi s same heat number.

(e) Surveillance data ex ists fo r weld Heat # 90 I 36, # 305424, and# 305414 from multiple so urces; see Section 4 for more detail s. The data for Beaver Valley Unit I we ld metal Heat # 90136 was taken from WCAP-17896- NP [Ref. 5]. The data fo r St. Lucie Un it I we ld metal Heat# 90136 was take n from the St. Lucie Unit I License Amendment Req uest for Exte nded Powe r Uprates, Attachment 5, Tab le 2. 1.2-4 [Ref. 15]. The data for Millstone Un it 2 weld meta l Heat 90136 was taken from Table 4-1 of WCAP-16012 [Ref. 16]. The data for Fort Calhoun weld meta l Heat# 305414 was take n from Table 5.2-4b of the 20 12 Beaver Va lley Unit I P-T Limits revision report [Ref. I 7].

Table 3-3 Summary of Beaver Valley Unit I Replacement Reactor Vessel Closure Head and Vessel Flange Initial RT NOT Values Initial RT NOT Reactor Vessel Material Methodology (OF)

_4(a) ASME Code, Section Ill , Subsection NB-2300 Replacement Closure Head

[Ref. 11]

I0(b) BWRVIP-173 -A, Alternate Approach 2 Vessel Flange

[Ref. 12]

Notes:

(a) The initial RTNDT va lu e of the rep lace me nt reactor vesse l (RV) closure head was taken from WCAP-16799- NP, Rev ision I [Ref. I 8] and was dete rmined in accordance with the methodo logy of ASME Code, Section Ill , Subsection NB-23 00 [Ref. 11 ].

(b) The initial RT NDT va lue of the vesse l fl ange was updated, utilizi ng the methodo logy of BWRVIP-173-A, Altern ate A pproach 2 [Ref. 13], from the va lue documented in WCA P-1 6799-NP, Rev ision I [Ref. 18].

WCA P-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghou se Non-Propri etary Class 3 4-1 4 SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2, calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillanc e program. In addition to the plant-specific surveillanc e data, data from surveillanc e programs at other plants which include a reactor vessel beltline or extended beltline material should also be considered when calculating Position 2.1 chemistry factors. Data from a surveillanc e

program at another plant is often called ' sister plant' data.

The surveillance capsule plate material for Beaver Valley Unit 1 is from Lower Shell Plate B6903-1.

The surveillanc e capsule weld material for Beaver Valley Unit I is Heat # 305424, which is applicable to the intermedia te shell longitudin al welds. Table 4-1 summarize s the Beaver Valley Unit I surveillanc e data for the plate material and weld material (Heat # 305424) that will be used in the calculation of the Position 2.1 chemistry factor values for these materials. The results of the last withdrawn and tested surveillanc e capsule, Capsule X, were documente d in WCAP-17 896-NP [Ref. 5]. Appendix D concludes that the Beaver Valley Unit I surveillanc e plate and weld (Heat # 305424) material are non-credib le; therefore, a full margin term will be utilized in the ART calculation s contained in Section 7.

The Beaver Valley Unit I reactor vessel intermediate to lower shell girth weld seam was fabricated using weld Heat # 90136. Weld Heat # 90136 is contained in the St. Lucie Unit I and Millstone Unit 2 surveillanc e programs. Thus, the St. Lucie Unit I and Millstone Unit 2 data will be used in the calculation of the Position 2.1 chemistry factor value for Beaver Valley Unit 1 weld Heat # 90136. Table 4-2 summarize s the applicable surveillanc e capsule data pertaining to weld Heat # 90136. The combined surveillanc e data is deemed credible per Appendix D; however, as a result of the Millstone Unit 2 surveillanc e data including both weld Heat # 90136 and 10137, the Position 2.1 chemistry factor calculation s for weld Heat # 90136 will utilize a full margin term for conservati sm. See Appendix D for details.

The Beaver Valley Unit I reactor vessel upper shell to intermedia te shell girth weld seam and lower shell longitudinal weld seams were fabricated using weld Heat # 305414. Weld Heat # 305414 is contained in the Fort Calhoun surveillanc e program . Thus, in WCAP-15 571 , Supplemen t I, Revision 2, the Fort Calhoun data was used to calculate the Position 2.1 chemistry factor value for Beaver Valley Unit 1 weld Heat# 305414, which is used herein. Furthermore, Appendix D of WCAP-17 896-NP [Ref. 5] concluded that the weld Heat# 305414 data is non-credible; therefore, a full margin term will be utilized in the ART calculation s contained in Section 7.

WCAP-181 02-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-2

-I Table 4-1 Beaver Valley Unit 1 Surveillance Capsule Data w*

iil Capsule Fluence<*>

0 Material Capsule Measured 30 ft-lb Transitio n a.

0 (x 10 19 n/cm2, E > 1.0 MeV) Tempera ture Shift(h) (°F)

~

Q)

Vi V 0.297 127.9 g:

~ u 0.6 18 118.3 Q)

Lower Shell Plate 86903-1 (Longitudinal) w

""0< y 0.952 147.7 (1) c..

2.10 141.7 0

, X 4.99 175.8

, V 0.297

-.J 138.0 0

~

u 0.618 132.1 Lower Shell Plate 86903-1 (Transverse) w (X) ex, 0.952 180.2 w

0) y 2.10 i0 0

166.9

• s: X 4.99 179.0

-I V 0.297 159.8 u

w* 0.618 164.9 Vi Surveillance Weld Material (Heat # 305424) oi ro w 0.952 186.3 3 y (1) 2.10 178.5

?.

~

Q)

X 4.99 237 .8 Vi Q)

Notes:

c..

c..

(1) (a) Data was taken from Appendix F, Section F.1.1 .

c..

r::r (b) Data was taken from Table 5-10 of WCAP-178 96-NP [Ref. 5].

5' (1)

"U

0

~

m Vi ro Vi 3

C:

"0 ui

~

C:

!!?.

o*

2, WCAP-18102-NP February 2018 Revision I

Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data for Weld

-I Heat# 90136 cii" al Capsule Fluence<*l 8 Measured 30 ft-lb Transition Inlet Temperature

a. Material Capsute<*l (x 10 19 n/cm2, E > 1.0 Temperature<bl

Temperature Shift<*l (°F) Adj ustment<<l 0,

MeV)

(/)

(OF) (OF)

~ 97° 0.5174 72.34 541

~

0, St. Lucie Unit I -1.7 "O 104° 0.7885 67.4 "O Data 544.6 1.9 0 284°

< 1.243 68.0 546 .3 (1)

C. 3.6 0 97° 0.324 65.93 Millstone Unit 2 544.3 1.6 N

, 104° 0.949 52. 12

...... Data 547.6 4.9

, 83 0 1.74 0

~

56.09 548 .0 5.3 ex, ex, Notes:

w (a) For surveillan ce weld heat# 90136, data pertaining to the St. Lucie Unit 0)

N I were taken from the St. Lucie Unit I License Amendme nt Request for Power Uprates, Attachme nt 5, Table 2.1. 1-3 [Ref. 15). Data pertaining lo Extended 0 Millstone Unit 2 were taken from Table 5-10 of WCA P-16012 [Ref. 16).

)>

(b) Inlet temperatu res were calculated as the average in let temperatu re from all the previously completed cycles at the time of capsule withdraw (c) Temperat ure adjustmen t = l.0*(Tcapsu le - Tpla, ), where Tplant = 542.7° a l.

11 F fo r Beaver Valley Unit I. 542.7° F is the cycle-by- cycle average

-I

,- temperature for Beaver Valley Unit I for Cycle I through Cyc le 22. The downcom er cii" temperatu re adjustmen t procedure is app lied to the weld 6 RTNDT data for St. Lucie Unit I and Millstone Un it 2 capsules in the Position 2. 1 chemistry each of the factor calculatio n - See Section 5 for more details.

s

(/)

ro 3

(1)

l.

0,

(/)

0, C.

C.

(1)

C.

r::r g.

(1) il

o

~

m

(/)

(/)

ro 3

C:

"O 0

~

~

C:

~

6" 2-WCAP-1 8102-NP February 2018 Revision I

Westingho use Non-Prop rietary Class 3 5-1 S CHEMISTRY FACTORS The chemistr y factors (CFs) were calculate d using Regulato ry Guide 1.99, Revision 2, Positions 1.1 and

2. 1. Position 1.1 chemistry factors for each reactor vessel material are calculate d using the best-estim ate copper and nickel weight percent of the material and Tables I and 2 of Regulato ry Guide 1.99, Revision

2. The best-estim ate copper and nickel weight percent values for the Beaver Valley Unit I reactor vessel materials are provided in Table 3-2 of this report.

The Position 2.1 chemistry factors are calculate d for the materials that have available surveilla nce program results. The calculatio n is performe d using the method described in Regulato ry Guide 1.99, Revision 2. The Beaver Valley Unit I surveilla nce data as well as the applicabl e sister plant data was summari zed in Section 4 of this report, and will be utilized in the Position 2.1 chemistr y factor calculatio ns in this Section.

The Position 2.1 chemistry factor calculatio ns are presented in Tables 5-1 through 5-3 for Beaver Valley Unit I reactor vessel materials that have associate d surveillan ce data. These values were calculate d using the surveilla nce data summariz ed in Section 4 of this report. Note that the Position 2.1 chemistr y factor for weld Heat # 305414 was previousl y reported in WCAP-1 5571 , Supplem ent I, Revision 2 [Ref. 14]

and is therefore not recalcula ted . All of the surveilla nce data is adjusted for irradiatio n temperat ure and chemical composit ion differenc es in accordan ce with the guidance presented at an industry meeting held by the NRC on February 12 and 13 , 1998 [Ref. 19]. Margin will be applied to the ART calculatio ns in Section 7 accordin g to the conclusio ns of the credibilit y evaluatio n for each of the surveilla nce materials ,

as documen ted in Section 4.

The Position 1. 1 chemistry factors are summari zed along with the Position 2. 1 chemistry factors in Table 5-4 for Beaver Valley Unit I .

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-2 Table 5-1 Calculation of Beaver Valley Unit 1 Chemist ry Factor Value for Lower Shell Plate B6903-1 Using Surveillance Capsule Data LS Plate B6903-1 Capsule r*> (<)

Capsule (x 10 19 n/cm2, E > 1.0 FF(b) ART NOT FF*ARTNDT 2 Data (O F) FF (OF)

MeV)

V 0.297 0.6677 127.9 85.40 0.446 u 0.618 0.8652 118 .3 102.35 0.749 Longitudinal Orientatio n w 0.952 0.9862 147.7 145.66 0.973 y 2.10 1.2018 141.7 170.30 1.444 X 4.99 1.4020 175 .8 246.46 1.965 V 0.297 0.6677 138.0 92. 14 0.446 u 0.618 0.8652 132.1 114.29 0.749 Transvers e Orientatio n w 0.952 0.9862 180.2 177 .72 0 .973 y 2. 10 1.2018 166.9 200.58 1.444 X 4.99 1.4020 179.0 250.95 1.965 SUM: 1585.86 11.154 CF Ls Pl ateB6903 -I = L(FF

• L'i RTNDT)-;.. L(FF 2) = (1585 .86) -;.. ( 11.154) = 142.2°F Notes:

(a) f = fluence.

(b) FF = fluence factor = ti 0*28

• 0* 10* 10gl)_

(c) L'i RTNoT values are the measured 30 ft-lb shi ft values. All values are taken from Table 4-1 of this report.

Table 5-2 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat# 305424 Using Surveillance Capsule Data Capsule J<*l Weld Metal ARTNDT

(<)

Capsule (x 10 19 n/cm2, E > 1.0 FF(b) FF*ARTNDT 2 Heat# 305424 (OF) FF MeV) (OF)

V 0.297 0.6677 169.4 (159.8) 11 3. 10 0.446 u 0.618 0.8652 174.8 (164.9) 151.23 0.749 Beaver Valley Unit I Data w 0.952 0.9862 197.5 ( 186.3 ) 194.76 0.973 y 2. 10 1.2018 189.2 (178 .5) 227.40 1.444 X 4.99 1.4020 252. l (237 .8) 353.39 1.965 SUM: I 039.87 5.577 CF we ld Heat # 305424 = L(FF

• L'i RT NDT) -;.. L(FF 2 ) = (1039.87) -;.. (5.577) = 186.5°F Notes:

(a) f= fluence .

(b) FF= fluence factor = ti 028

• 0 io*log f)_

(c) L'iRT NDT values are the measured 30 ft-lb shift values. The 6.RTNDT values are adj usted using the ratio procedure to acco unt for differences in the surveillance weld chemistry and the reactor vesse l weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 4-1 of this report). Ratio applied to the Beaver Valley Unit I surve illance data = CF vessel Weld/ CFsurv. Weld = 191.7°F / I8 l .6°F = 1.06.

WCAP-18 102-NP February 2018 Revision l

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westingho use Non-Prop rietary Class 3 5-3 Table 5-3 Calculation of Beaver Valley Un it 1 Chemistry Factor Value for Weld Heat# 90136 Using Surveillance Capsule Data Weld Metal Capsule t<">

Capsule FF(b) ARTNDT (r) FF*ARTNDT Heat# 90136 (x 10 19 n/cm2, E > 1.0 MeV) (OF) FF2 (OF) 97° 0.5174 0.8160 82.6 (72.34)

St. Lucie Unit I 67.44 0.666 104° 0.7885 0 .9333 81.1 (67.4)

Data 75.68 0.871 284° 1.243 1.0606 83.8 (68.0) 88.85 1.125 97° 0.324 0.6902 67.5 (65.93)

Millstone Unit 2 46.61 0.476 Data<d) 104° 0.949 0.9853 57.0(52.1 2) 56 .18 0.971 83 0 1.74 1.1523 61.4 (56.09) 70 .74 1.328 SUM: 405.50 5.437 CF weld Heat 1190 136 = L(FF

• L'.RTNDT) -;- L(FF 2 ) = (405 .50) -;- (5.437) = 74.6°F Notes:

(a) f = fluence.

r (b) FF = fluence factor = 0 28

• O I O* log f)_

(c) t-.RTNoT va lues are the measured 30 ft- lb shift values. The t-.RTNDT va lues are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chem istry and the reactor vessel we ld chemistry (pre-adjusted va lues are listed in parentheses and were taken from Tab le 4-2 of this report). The temperature adjustment s are listed in Table 4-2. Ratio app lied to the St.

Lucie Unit I surveillance data

= CF vessel weld / CFsurv. weld = I 24.3 °F I I 06.6°F = 1.17. A ratio of 1.00 was conservativ ely applied to the Millstone Unit 2 surveillance data, since CFvessel Weld < CFsu,v Weld*

(d) Millstone Unit 2 surveillance data contains specimens from both weld Heat # 90136 and we ld Heat # IO 137. However, this inclusion of an additional heat is not expected to negatively impact the subsequent reactor vesse l integrity calculation results, as additional conservati sms are in place. See Appendix D for more details.

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/20 18 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 5-4 Table 5-4 Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (°F) and Identification Number Heat Number Position 1.1 <*> Position 2.1 Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-I 100.5 ---

Intermediate Shell Plate B6607-2 C438 l-2 100.5 ---

Lower Shell Plate B6903-1 C6317-I 147.2 l42.2(b)

Lower Shell Plate B7203-2 C6293-2 98 .7 -

Intermediate to Lower Shell Girth Weld 11-714 90136 124.3 74.6(c)

Intermediate Shell Longitudinal Welds 305424 191.7 186.5(d)19-714 A&B Lower Shell Longitudinal Welds20-714 A&B 305414 210 .5 2]6.9(e)

Reactor Vessel Extended Beltline Materials Upper Shell Forging B6604 l23V339VAI 84.2 ---

305414 210 .5 216 .9(e)

AOFJ 41.0 ---

Upper Shell to Intermediate Shell Girth FOIJ 41.0 ---

Weld 10-714 EODJ 27.0 ---

HOCJ 27.0 ---

Inlet Nozzle B6608-1 95443-1 67.0 ---

Inlet Nozzle B6608-2 95460-1 67.0 ---

Inlet Nozzle B6608-3 95712-1 51.0 ---

EODJ 27.0 ---

FOIJ 41.0 ---

HOCJ 27.0 ---

Inlet Nozzle Welds 1-717B, 1-717D, l-717F DBIJ 27.0 ---

EOEJ 20.0 ---

ICJJ 41.0 ---

JACJ 54.0 ---

Outlet Nozzle B6605- l 95415-l 95.3 ---

Outlet Nozzle B6605-2 95415-2 95.3 ---

Outlet Nozzle B6605-3 95444-1 58.0 ---

ICJJ 41.0 ---

IOBJ 27.0 ---

Outlet Nozzle Welds 1-717 A, l-7 I 7C, l-717E JACJ 54.0 ---

HOCJ 27.0 ---

EODJ 27.0 ---

WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westingho use Non-Prop rietary Class 3 5-5 Table 5-4 Summar y of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemist ry Factors Reactor Vessel Material Chemistr y Factor {°F) and Identification Number Heat Number Position 1.1 <*) Position 2.1 Outlet Nozzle Welds l-717A, l-717C, 1-7l7E FOIJ 41.0 ---

(continued )

Surveillance Weld Data Beaver Valley Unit I 305424 181.6 ---

St. Lucie Unit 1 90136 106.6 ---

Millstone Unit 2 135 .5 ---

Fort Calhoun 305414 212.0 ---

Notes:

(a) Position 1.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-2 of this report and Tables I and 2 of Regul atory Guide 1.99, Revision 2.

(b) Position 2.1 chemistry factor was taken from Table 5-1 of this report. As discussed in Section 4, the surveillance plate data is not credible.

(c) Position 2. 1 chemistry factor was taken from Table 5-3 of thi s report. As discussed in Section 4, the surveillanc e weld data for Heat # 90136 is credible; however, no reduction in the margin term will be taken.

(d) Position 2.1 chemistry factor was taken from Table 5-2 of this repo rt. As discussed in Section 4, the surveillanc e weld data for Heat # 305424 is not credible.

(e) Position 2.1 chemistry factor was taken from WCAP-15571 , Supplemen t I, Revision 2 [Ref. 14]. As discussed in Section 4, the surve il lance we ld data for Heat # 305414 is not credible.

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 6.1 OVERALL APPROACH The ASME 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, Kie, for the metal temperature at that time. Kie is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 3). The K1c curve is given by the following equation:

K =33 2+ 20 734

• e [ o.oz(T- tffNor >l (1) le *

• where, K1e (ksi in.) = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NOT This K1e 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 I , SA-508-1 , SA-508-2, and SA-508-3 steel.

6.2 METHODOL OGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPM ENT The governing equation for the heatup-cooldow n analysis is defined in Appendix G of the ASME Code as follows:

(2)

where, Kim stress intensity factor caused by membrane (pressure) stress K1t stress intensity factor caused by the thermal gradients K1c reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NOT 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-18102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Propri etary Class 3 6-2 For membrane tension , the correspond ing K 1 for the postulated defect is:

Kim = M,,, X (pJ?i / f) (3) where, Mm for an inside axial surface flaw is given by:

Mm = 1.85 for t < 2, Mm = 0.926 t for 2 ~ t ~ 3.464 ,

Mm 3.21 for f > 3.464 and, Mm for an outside axial surface flaw is given by:

Mm 1.77 for f < 2, Mm 0.893 f for 2 :<S; -Ji :<S; 3.464, M 01 3.09 for f > 3.464 Similarly, Mm for an inside or an outside circumfere ntial surface flaw is given by:

Mm 0.89 for f < 2, Mm 0.443 f for 2 :<S; -Ji :<S; 3.464 ,

Mm =: 1.53 for t > 3.464 where:

p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).

For bending stress, the correspond ing K 1 for the postulated axial or circumfere ntial defect is:

K,b = 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 circumfere ntial inside surface defect of G-2120 is:

Kit = 0 .953x l0- 3 X CR X t 25 (5) where CR is the cooldown rate in °F/hr. , or for a postulated axial or circumfere ntial outside surface defect K1t = 0.753xl0- 3 x HU x t 25 (6) where HU is the heatup rate in °F/ hr.

WCAP-181 02-NP February 2018 Revision l

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-3 The through-wal l temperature difference associated with the maximum thermal K can be determined 1

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, relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-22 l 4.3(a)( 1) and (2).

(b) Alternatively , the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness axial or circumferent ial inside surface defect using the relationship :

K11 = (1.0359 Co + 0.6322 C 1+ 0.4753 C2 + 0.3855C3) * ,J;;;, (7) or s imilarly, K 11 during heatup for a 1/4-thickness outside axial or circumferential surface defect using the relationship:

K i, = (l .043 C o + 0 .630 C1 + 0.481 C 2 + 0.401 C J) * ~ (8) where the coefficients C 0 , C 1, C 2 and C 3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form :

a-(x) =Co+ C,(x I a)+ C2(x I a )2 + C3(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-tem perature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-1404 0-A , Revision 4, " Methodology Used to Develop Cold Overpressur e Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2] 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-1404 0-A ,

Revision 4 ( equation 2.6.1-1 ). This equation is solved utilizing values for thermal diffusivity of 0.518 ft 2/hr at 70° F and 0.379 ft 2/hr at 550°F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft 2-° F.

At any time during the heatup or cooldown transient, K,c is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of l /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 RT NDT, 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 correspondin g (thermal) stress WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-4 intensity factors , K 11 , for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained, and from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference I /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-tempe rature 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 L'1 T (temperature) across the vessel wall developed during cooldown results in a higher value of Kie at the l/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K,e exceeds K11 , 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 I /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-temper ature 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 K 1e for the inside 1/4T flaw during heatup is lower than the K 1c 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 Kie 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 I /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-temper ature 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 WCAP-18102-N P February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-5 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-tem perature 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-poi nt 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.

6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS IO CFR Part 50, Appendix G [Ref. 4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RT NDT by at least 120°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 replacement reactor vessel closure head and vessel flange are documented in Table 3-3 . The limiting unirradiated RT NDT of 10°F is associated with the vessel flange of the Beaver Valley Unit 1 vessel , so the minimum allowable temperature of this region is I 30°F at pressures greater than 621 psig (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.

WCAP-18102 -NP February 2018 Revision l

• • • This record was final approved on 2/27/2018 8:36 :20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 7-1 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulato ry Guide 1.99, Revision 2, the adjusted reference temperat ure (ART) for each material in the beltline region is given by the following expressio n:

ART= Initial RT NDT + ~ RT NDT + Margin (l 0)

Initial RT NDT is the reference temperatu re for the unirradia ted material as defined in paragraph NB-2331 of Section Ill of the ASME Boiler and Pressure Vessel Code [Ref. 11]. If measured values of the initial RT NDT for the material in question are not available , generic mean values for that class of material may be used, provided if there are sufficien t test results to establi sh a mean and standard deviation for the class.

~RT NDT is the mean value of the adjustme nt in reference temperat ure caused by irradiatio n and should be calculate d as follows :

~RT DT =CF* f (028-0.101ogf)

(11)

To calculate ~RT NDT at any depth (e.g. , at I/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth:

f.(de pth x) -- f.s,,rf'ace

• e (-0.24x) (12) where x inches (reactor vessel cylindrical shell beltline thickness is 7.875 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 ~RT NDT at the specific depth.

The projected reactor vessel neutron fluence was updated for this analysis and documen ted in Section 2 of this report. The evaluatio n methods used in Section 2 are consisten t with the methods presented in WCAP-1 4040-A , Revision 4, " Methodo logy Used to Develop Cold Overpres sure Mitigatin g System Setpoints and RCS Heatup and Cooldow n Limit Curves" [Ref. 2].

Table 7-1 contains the surface fluence values at 50 EFPY, which were used for the developm ent of the P-T limit curves contained in this report. Table 7-1 also contains the l /4T and 3/4T calculate d fluence values and fluence factors, per Regulato ry Guide 1.99, Revision 2 . The values in this table will be used to calculate the 50 EFPY ART values for the Beaver Valley Unit I reactor vessel material s.

Margin is calculate d as M = 2 -Jc, } + c, ~ . The standard deviation for the initial RT NDT margin term (cr 1) is 0°F when the initial RT NDT is a measured value, and I 7°F when a generic value is available

. The standard deviation for the ~RTNDT margin term , cr 6 , is I 7°F for plates or forgings when surveillan ce data is not used or is non-cred ible, and 8.5°F (half the value) for plates or forgings when credible surveillan ce data is used. For welds, cr 6 is equal to 28°F when surveilla nce capsule data is not used or is non-cred ible, and is 14°F (half the value) when credible surveillan ce capsule data is used . The value for cr 6 need not exceed 0.5 times the mean value of ~RT NDT

• Containe d in Tables 7-2 and 7-3 are the 50 EFPY ART calculatio ns at the I/4T and 3/4T locations for generatio n of the Beaver Valley Unit I heatup and cooldown curves.

WCAP-18 102-NP February 2018 Revision l

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 7-2 The inlet and outlet nozzle forging materials for Beaver Valley Unit I have projected fluence values that exceed the I x 10 17 n/cm 2 fluence threshold at 50 EFPY per Table 2-7; therefore , per NRC RJS 2014-11

[Ref. 9], neutron radiation embrittle ment must be considere d herein for these materials

. The nozzle ART calculatio ns conserva tively utilize the maximum fluence value for each nozzle material, as documen ted in Appendix 8. Thus, ART calculatio ns for the inlet and outlet nozzle forging materials utilizing the l/4T and 3/4T fluence values are excluded from Tables 7-2 and 7-3 , respectively.

Finally, the second conclusio n of TLR-RE S/ DE/CIB-2013-01 [Ref. 20] states that if ~RT NOT is calculate d to be less than 25°F, then embrittle ment need not be considere d. This conclusio n was applied, as necessary, to the ART calculatio ns documen ted in Tables 7-2 and 7-3.

The limiting ART values for Beaver Valley Unit I to be used in the generatio n of the P-T limit curves are based on Lower Shell Plate 86903-1 (Position 1.1 ). For conserva tism, limiting ART values were rounded to the nearest whole number, then increased by 0.5°F. The increased limiting ART values, using the

" Axial Flaw" methodo logy, for Lower Shell Plate 86903-1 are summari zed in Table 7-4.

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westingho use Non-Prop rietary Class 3 7-3 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, l/4T and 3/4T Location s

for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY Surface l/4T f 3/4T f Fluence, f*l 1/4T Reactor Vessel Region (n/cm 2 , 3/4T (n/cm2, (n/cm2, FF FF E > 1.0 MeV) E > 1.0 MeV)

E > 1.0 MeV)

Reactor Vessel Beltline Materials Intermedi ate Shell Plates 5.88x 10 19 3.666 X 10 19 1.3370 1.425 X 10 19 1.0982 19 Lower Shell Plates 5.89 }0 3.672 10 19 X X 1.3374 1.427 X 10 19 1.0987 Intermedi ate to Lower Shell 19 5.88 X (0 3.666 10 19 1.3370 1.425 10 19 Girth Weld X X 1.0982 Intermedi ate Shell J.13 X 10 19 7.04x 10 18 0.9018 2.74 10 18 0.6469 Longitudi nal Welds X Lower Shell Longitudinal l.14x 10 19 7.11 X 10 18 0.9042 2.76 10 18 Welds X 0.6492 Reactor Vessel Extended Beltline Materials Upper Shell Forging 7.18xl0 18 4.48 X 10 18 0.7764 1.74 10 18 X 0.5366 Upper to Intermedi ate Shell 7.18x 10 18 4.48x 10 18 0.7764 1.74 10 18 Girth Weld X 0.5366 Inlet Nozzle to Upper Shell 2.I0x I0 17<bl 1.31 10 17 0.1313 Weld - Lowest Extent X 5.09 X 10 16 0.0679 Outlet Nozzle to Upper 1.61 X 10 17(b) 1.00 10 17 0.1099 Shell W eld - Lowest Extent X 3.90 X 10 16 0.0556 Notes:

(a) 50 EF PY flue nce va lues we re taken from Tables 2-5 , 2-7, and 2-9.

(b) The fluence for the inlet and outlet nozzle to upper shell welds was also used as the fluence fo r the inlet and outlet nozzle materials. The actual nozz le forging fluence va lues, at the location of a postulated flaw along the nozzle corne r region , are ex pected to be lower since they are further away from the active core.

WCAP-18102-NP February 20 I 8 Revision I

• • • This record was fina l approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 7-4 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at

-I the l/4T Location cii" ro Reactor Vessel Material and ID CF I/4T Fluence Heat Number I/4T RT NDT(U)

(a)

ARTNDT (b)

CJ/*> fl t;.(c) Margin ART(d)

Number

(")

a.

0

{°F) (n/cm2, E > 1.0 MeV) FF (OF) {°F) (OF) (OF) (OF) (OF)

~

Ill

(/)

Reactor Vessel Beltline Materials g Intermediate Shell Plate 86607-1 C4381-I 100.5 3.666 10 19 X 1.3370 26.8 134.4 0 17 34.0 195.2

~

Intermediate Shell Plate 86607-2 C4381-2 100 .5 3.666 10 19 Ill

""O X 1.3370 53 .6 134.4 0 17 34.0 222.0

""O Lower Shell Plate 86903-1 C6317-1 147.2 3.672 10 19 1.3374

~ X 13.1 196 .9 0 17 34.0 244.0 Using Beaver Valley Unit I

(!)

C.

0

, C6317-I 142.2 3.672 10 19 1.3374 13.1 190.2 surveillance data X 0 17 34.0 237.3

-.J Lower Shell Plate 87203-2 C6293-2 98 .7 3.672 10 19 1.3374 X 0.4 132.0 0 17 34.0 166.4 0

~

ex, Intermediate to Lower Shell Girth 90136 124.3 3.666x 10 19 1.3370 -56 166 .2 ex, w Weld 11-714 17 28 65 .5 175.7 0) j(:, Using St. Lucie Unit I and Millstone 0

90136 74.6 3.666 )0 19 1.3370 -56 99.7

)> Unit 2 surveillance data X 17 28 65.5 109.3

~

Intermediate Shell Longitudinal

-I

,- 305424 191.7 0.704 10 19 0.9018 -56 172.9 cii" Welds19-714 A&8 X 17 28 65.5 182.4

(/)

al" Using Beaver Valley Unit I m 305424 186.5 0.704 10 19 3 surveillance data X 0.9018 -56 168.2 17 28 65.5 177.7

(!)

~

~

Lower Shell Longitudinal Welds Ill 305414 210.5 0.711 )0 19 0.9042 -56 190.3

(/)

Ill 20-714 A&8 X 17 28 65 .5 199.9 C.

C. Using Fort Calhoun surveillance data 305414 216.9 0.7]) )0 19

(!)

C.

X 0.9042 -56 196.1 17 28 65.5 205 .6 CT Reactor Vessel Extended Beltline Materials<*>

9- Upper Shell Forging 86604 123V339VA1 84.2 ro 0.448x 10 19 0.7764 40 65.4 0 17 34.0 cl 139.4

o Upper Shell to Intermediate Shell 305414

~

Girth Weld 10-714 210.5 0.448x 10 19 0.7764 -56 163.4 17 28 m (3951 & 3958) 65 .5 172.9

(/)

'<(/)

Using Fort Calhoun surveillance data 305414 m 216.9 0.448x 10 19 0.7764 -56 168.4 3 (3951 & 3958) 17 28 65.5 177.9 C

""O 0 AOFJ 41.0 0.448 10 19 0.7764 10 31.8

, X 17 15 .9 46.6 88.4 ui Upper Shell to Intermediate Shell FOIJ 41.0 0.448 10 19 0.7764 IO X

31.8 17 15.9 46.6 88.4

~ Girth Weld I 0-714 ( continued) EODJ 27.0 0.448x 10 19 0.7764 a: IO 0.0 (21 .0) 17 0 34.0 44 .0

!!?. HOCJ 5* 27.0 0.448 10 19 0.7764 IO 0.0 (21.0) 2-X 17 0 34.0 44.0 WCAP-18102-NP February 2018 Revision I

Westinghouse Non-Proprietary Class 3 7~5 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at

-I

r the 1/4T Location en*

m Reactor Vessel Material and ID CF l/4T Fluence

()

Heat Number l/4T RTNDT( U) (a) ARTNDT(b ) cs,<*> CSf!.(c)

Margin ART(d)

Number (OF) (n/cm2, E > 1.0 MeV) a.

0 FF (OF) (OF) (OF) (OF) (OF) (OF)

~

Q) EODJ 27.0 0.01 3 1 X 10 19 0 .13 13 10 en 0.0 (3 .5) 17 0 34.0 44.0

~ FOIJ 41.0 0.0131 X 10 19 0.13 13 10

~ 0.0 (5 .4) 17 0 34.0 44 .0 Q) 1'.J HOCJ 27.0 0 .Q] 3 J X 10 19 0.13 13 10 Inlet Nozzle Welds 1-717B, 1-7 17D, 0.0 (3.5) 17 0 34.0 44.0 a<

1'.J DBIJ 27.0 0.0] 3 1 X 10 19 0.13 13 10 (1)

Cl.

l-7 17F 0.0 (3. 5) 17 0 34.0 44.0 0 EOEJ 20.0 0.01 3 ] 10 19 0.13 13 10

, X 0.0 (2.6) 17 0 34.0 44.0

ICJJ 41.0 0.0] 3 ) X 10 19 0.13 13 10 0.0 (5 .4)

-.J 17 0 34.0 44 .0

JACJ 54.0 0.01 3 ]

0 X 10 19 0.13 13 10 0.0 (7.1) 17

~

0:,

0 34.0 44.0 ICJJ 41.0 0.0100 ]0 19 0.1099 10 0.0 (4.5) 0:,

w X

17 0 34.0 44.0 O>

._

, IOBJ 27.0 0.0100 X 10 19 0 .1099 10 0.0 (3.0) 0 17 0 34.0 44.0 Outlet Nozzle Welds l -717A, l-717C, JA CJ 54.0 0.Q]QQ 10 19 0.1099

)>

s

:

X 10 0.0 (5 .9) 17 0 34.0 44 .0 l-717E HOCJ 27.0 0.0] 00 ]0 19 X 0.1099 10 0.0 (3.0) 17 0 34.0

-I

r 44 .0 en* EODJ 27.0 0.0]00 ]0 19 0.1099 IO en X

0.0 (3.0) 17 0 34.0 44.0 or FOIJ 41.0 0.0100 ]0 19 0.1099 ro X 10 0.0 (4.5 ) 17 0 34.0 44 .0 3 Notes:

(1) a (a) T he plate and fo rg ing materi al initi al RT NDT va lues are measured va lues.

~ Th e initia l RT NDT va lues fo r all of the reacto r vessel we ld s are generic; Q) en reactor vesse l we lds. hence cr 1 = l 7°F fo r all Q)

Cl. (b) As di scussed in Section 7, ca lcul ated L\RTNoT values less than 25° F have Cl. bee n reduced to zero per TLR- RES/DE/C IB-20 13-0 I [Ref. 20]. Actual (1)

Cl. are li sted in parentheses. ca lcul ated L\ RT NDT va lues 0-

'< (c) As disc ussed in Section 4, the surve ill ance pl ate and we ld Heat # 3054 14 5' and # 305424 data were dee med non-credible. T he surve ill ance we ld data (1) deemed credible; however, pe r Section 4 and Appendi x D, a full margin fo r Heat # 901 36 was term will be used. Per the gui dance o f Regul atory G uide 1.99, Rev ision "tJ er,:,= l 7°F fo r Pos ition 1. 1 and Pos ition 2. 1 w ith non-credi ble surve ill ance 2 [Ref. I], the base meta l

~ data, and the we ld meta l crf!. = 28°F fo r Position 1.1 and 2. 1 with non-credi

s

: ble surve illance data.

m Since a fu ll margin term w ill be used fo r Heat # 901 36, er,:, = 28°F w ith credible surve illance data fo r Position 2. 1 fo r thi s weld heat. However, en crf!. need not exceed

'< 0.5*L\ RT/',,'1)T*

en ro (d) The Regul atory Gui de 1.99, Rev ision 2 methodology was used to ca lcul ate ART va lues. A RT = RT NDT(UJ + L\RT NDT + Marg in.

3 C (e) As di scussed in Section 7, the inl et and outlet nozzle fo rg in g material 1'.J A RT ca lcul ations utilizing a l/4T tluence va lue are exc luded from thi 0

, materia l A RT va lues necessary to perform the fracture mechanic s evaluatio s table because the nozz le ns on these materi als are th ose calculated util iz ing a max imum tluence

~ These maximum ART values are documented in Table 8 -1. value fo r each materi a l.

~

C:

a-WCAP-18102-NP February 2018 Revision I

Westinghouse Non-Proprietary Class 3 7-6 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the

-I 3/4T Location

<ii" ro Reactor Vessel Material and ID CF 3/4T Fluence 0

Number Heat Number 3/4T FF RTNDT( U)

(a)

ARTNDT (b) CJ/*> Gt.(c)

Margin ART(d) a.

0 (Of) (n/cm2, E > 1.0 MeV) (Of) (Of) (Of) (Of) (Of) (Of)

1, Q) Reactor Vessel Beltline Materials en Intermediate Shell Plate 86607- I C4381-1 100.5 1.425 I 0 19

~ X 1.0982 26.8 110.4 0 17 34.0 171.2

~

Intermediate Shell Plate 86607-2 C438 l-2 100.5 1.425 I 0 19 1.0982 Q)

"O "O

X 53 .6 110.4 0 17 34.0 198.0 Lower Shell Plate 86903-1 2 C6317- l 147.2 l.427x 10 19 1.0987 13 .1 161 .7 0 17 34.0 208.8 Using Beaver Valley Unit 1 (I)

C.

0

, C6317-l 142.2 l.427x 10 19 1.0987 13.1 156.2 surveillance data 0 17 34.0 203 .3

~

Lower Shell Plate 87203-2 C6293-2 98.7 l.427x 10 19 1.0987 0.4 108.4 0 17 34.0 142.8 0

~ Intermediate to Lower Shell Girth 00 90136 124.3 l.425x 10 19 1.0982 -56 136.5 00 w Weld 11-714 17 28 65 .5 146 .0 a,

0

• s:

Using St. Lucie Unit 1 and Millstone Unit 2 surveillan ce data 901 36 74.6 l.425x 10 19 1.0982 -56 81.9 17 28 65.5 91.4 Intermediate Shell Longitudinal

-I

,- 305424 191.7 0.274 10 19 0.6469 -56 124.0

<ii" Weld s19-714 A&8 X 17 28 65 .5 133 .5 en or Using Beaver Valley Unit 1 ro 305424 186.5 0.274 (0 19 3

(I) surveillan ce data X 0.6469 -56 120.6 17 28 65.5 130.2

~

1, Lower Shell Longitudinal Weld Q) en 305414 210.5 0.276x 10 19 0.6492 -56 136.6 Q)20-714 A&B 17 28 65 .5 146.2 C.

C. Using Fort Calhoun surveillan ce data 305414 216.9 0.276 X 10 19 (I)

C. 0.6492 -56 140.8 17 28 65.5 150.3 O"

Reactor Vessel Extended Beltline Materiats<e>

T Upper Shell Forging 86604 123V339V AI 84.2 0.174 X 10 19 (I) 0.5366 40 45.2 0 17 34.0 119.2

-0

o Upper Shell to Intermediate Shell 305414

~ Girth Weld 10-714 210 .5 0.174 X 10 19 0.5366 -56 113 .0 17 28 rn (3951 & 3958) 65.5 122.5 en en Using Fort Calhoun surveillance data 305414 ro 216.9 0.174x 10 19 0.5366 -56 116.4 3 (3951 & 3958) 17 28 65.5 125.9 C

"O 0 AOFJ 41.0 0. ]74 X 10 19 0.5366 10 0.0 (22 .0) 17 0 34.0 44.0

~ Upper Shell to Intermediate Shell FOIJ 41.0 19

< 0.174xl0 0.5366 10 0.0 (22.0) 17 0 34.0 Girth Weld 10-714 (continued) 44.0

~

a: EODJ 27.0 0.174 10 19 0.5366

~

X 10 0.0 (14.5) 17 0 34.0 44.0 5* HOCJ 27.0 0.174 10 19 0.5366

, X 10 0.0 (14.5) 17 0 34.0 44.0 WCAP-18 102-NP February 2018 Revision I

Westinghouse Non-Proprietary Class 3 7-7 Table 7-3 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the

-I 3/4T Location w*

ro Reactor Vessel Materia l and ID CF 3/4T Fluence (a) 8 Heat Number 3/4T FF RT NDT(U) iiRTNDT (b) u?l u l!.(c) Margin ART(d)

Number

a. (°F) (n/cm2, E > 1.0 MeV) (OF) (OF) (OF) (°F) (°F) (OF)

Q) EODJ 27.0 0.00509 x 10 19 0.0679 10 0.0 ( 1.8)

(/)

17 0 34.0 44.0 CJ FOIJ 41.0 0.00509 x 10 19 0.0679 e1. 10 0.0 (2 .8) 17 0 34.0 44.0 Q)

HOCJ 27.0 0.00509 ]0 19 0.0679

-0

-0 Inlet Nozzle Welds 1-7178 , I-717D, X 10 0.0 ( 1.8) 17 0 34.0 44.0 0 DBIJ 27.0 19 l-7l7F 0.00509 X 10 0.0679 10 0.0 ( 1.8) 17 0 Cl)

a. 34.0 44.0 EOEJ 20.0 0.00509 x 10 19 0.0679 0

CJ 10 0.0(1.4) 17 0 34.0 44.0 IV t3 ICJJ 41.0 0.00509 x 10 19 0.0679 10 0.0 (2 .8) 17 0 34.0 44.0

~ JACJ 54.0 0.00509 10 19

~ X 0.0679 10 0.0 (3 .7) 17 0 34.0 44.0 00 ICJJ 41.0 0.00390 J0 19 0.0556 10 0.0 (2.3) 00 w

X 17 0 34.0 44.0 cr, !OBJ 27 .0 0.00390 )0 19 0.0556 r-.:,

0 X 10 0.0 (1.5) 17 0 34.0 44.0 Outlet Nozzle Welds I- 717 A, JACJ 54.0

• s:: 0.00390 X )0 19 0.0556 10 0.0 (3.0) 17 0 34.0 44.0 l-717C, l-717E HOCJ 27.0 0.00390 X 10 19 0.0556 10 0.0(1 .5) 17 0 34.0

-I

-::,- 44.0 w* EODJ 27.0 0.00390 ]0 19 0.0556

(/)

X 10 0.0(1 .5) 17 0 34.0 44.0 iii FOJJ 41.0 0.00390 CD X ]0 19 0.0556 10 0.0 (2.3) 17 0 34.0 3

44.0 Cl)

Notes:

~

(a) The plate and forging material initial RT NDT values are measured values.

The initial RT NDT va lues for all of the reactor vesse l welds are generic; Q)

(/) reactor vesse l welds. hence o 1 = l 7° F for all Q)

a. (b) As di scussed in Section 7, ca lcul ated t.RTr-.'DT va lues less th an 25 ° F

a. have been reduced to zero per TLR-RES / DE/CJB-2 0 13-0 I [Ref. 20]. Actual Cl)

a. are li sted in parenthes es. calcu lated t.RT NDT va lues 0-

'< (c) As discussed in Sectio n 4, the survei ll ance plate and weld Heat # 3054 I CT 4 and# 305424 data were deemed non-credi ble. The surveillan ce weld data for Heat# 90 136 was Cl) deemed credible; however, per Section 4 and Append ix D, a full margin

-u term will be used. Per the guidance of Regulator y Guide 1.99, Revision

o of!.= I 7° F for Position 1. 1 and Position 2.1 with non-credi ble surveillan 2 [Ref. I], the base metal ce data, and the weld metal 0 6 = 28°F for Position 1. 1 and 2. 1 with non-credi

~ Since a full margin term wi ll be used for Heat # 90136, 0 = 28° F w ble survei ll ance data.

m 6 ith credib le survei llance data for Position 2. 1 fo r this weld heat. However,

(/) 0 6 need not exceed

(/)

0.5*L'.RTNDT*

CD (d) The Regulator y Guide 1.99, Revision 2 methodol ogy was used to calculate 3 ART values. ART = RT NDT(U) + L'.RT NDT + Margin.

C (e) As discussed in Section 7, the inlet and outlet nozzle forging material

-0 ART calculatio ns utilizing a 3/4T fluence va lue are excluded from this 0

material ART va lues necessary to perform the fracture mechanic s eva luations table because the nozzle CJ on these materials are those ca lcul ated utilizing a maximum fluence value

~ These maximum ART values are document ed in Table B-1. for each material.

e1.

1i

~

o*

2.

WCAP-18102-NP February 2018 Revision 1

Westinghouse Non-Propri etary Class 3 7-8 Table 7-4 Summary of the Limiting ART Values Used in the Generatio n of the Beaver Valley Unit 1 Heatup and Cooldown Curves at 50 EFPY 1/4T Limiting ART<a> 3/4T Limiting ART<">

244.5 °F 209.5 °F Lower Shell Plate B6903-1 (Position 1.1)

Note:

(a) The A RT values used for P-T limit curve development in this repo1t are the limiting 1/4T and 3/4T ART values calculated in Tables 7-2 and 7-3 rounded to the neare st whole number, then increased by 0.5°F to add additional conservatism .

WCAP-181 02-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary C lass 3 8-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure -temper ature limit curves for normal heatup and cooldow n of the primary reactor coolant system have been calculat ed for the pressure and tempera ture in the reactor vessel cylindri cal beltline region using the methods discusse d in Sections 6 and 7 of this report. This approve d methodo logy is also presente d in WCAP- 14040-A , Revisio n 4.

Figure 8-1 presents the limiting heatup curves without margins for possible instrume ntation errors using heatup rates of 60 and I 00°F/hr applicab le for 50 EFPY, with the flange requirem ents and using the "Axial Flaw" methodo logy. Figure 8-2 presents the limiting cooldow n curves without margins for possible instrum entation errors using cool down rates of -0, -20, -40, -60, and - I 00°F /hr applicab le for 50

• EFPY, with the flange requirem ents and using the "Axial Flaw" methodo logy . The heatup and cooldow n curves were generate d using the 1998 Edition through the 2000 Addend a ASME Code Section XI ,

Append ix G.

Allowab le combina tions of tempera ture and pressure for specific tempera ture change rates are below and to the right of the limit lines shown in Figures 8-1 and 8-2. This is in addition to other criteria, which must be met before the reactor is made critical, as discusse d in the followin g paragrap hs.

The reactor must not be made critical until pressure -temper ature combina tions are to the right of the criticality limit line shown in Figure 8-1 (heatup curve only). The straight-line portion of the criticali ty limit is at the minimu m permiss ible tempera ture for the 2485 psig inservic e hydrosta tic test as required by Append ix G to IO CFR Part 50. The governi ng equation for the hydrosta tic test is defined in the 1998 Edition through the 2000 Addend aASME Code Section XI , Append ix Gas follows:

(13) where, Kim is the stress intensity factor covered by membra ne (pressur e) stress, Ki e = 33.2 + 20.734 e [0 02(T-RTNoTll, T is the minimu m permiss ible metal tempera ture, and RT oT is the metal referenc e nil-duct ility tempera ture.

The criticali ty limit curve specifie s pressure -temper ature limits for core operatio n in order to provide addition al margin during actual power product ion. The pressure -temper ature limits for core operatio n (except for low power physics tests) are that: I) the reactor vessel must be at a tempera ture equal to or higher than the minimu m tempera ture required for the inservice hydrosta tic test, and 2) the reactor vessel must be at least 40°F higher than the minimu m permiss ible tempera ture in the correspo nding pressure -

tempera ture curve for heatup and cooldow n calculat ed as describe d in Section 6 of this report. For the heatup and cooldow n curves without margins for instrume ntation errors, the minimu m tempera ture for the inservice hydrosta tic leak tests for the Beaver Valley Unit I reactor vessel at 50 EFPY is 30 I °F; this tempera ture value is calculat ed based on Equatio n (13). The ve1tical line drawn from these points on the WCAP-1 8102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 8-2 pressure- temperat ure curve, intersecti ng a curve 40°F higher than the pressure- temperat ure limit curve, constitut es the limit for core operation for the reactor vessel.

Figures 8-1 and 8-2 define all of the above limits for ensuring preventio n of non-duct ile failure for the Beaver Valley Unit I reactor vessel for 50 EFPY with the flange requirem ents and without instrume ntation uncertainties. The data points used for developi ng the heatup and cooldown P-T limit curves shown in Figures 8-1 and 8-2 are presented in Tables 8-1 and 8-2 . The P-T limit curves shown in Figures 8-1 and 8-2 were generated based on the limiting ART values for the cylindric al beltline and extended beltline reactor vessel materials. These ART values were slightly increased to add additiona l margin; this approach is conserva tive. As discussed in Appendi x B, the P-T limits develope d for the cylindric al beltline region bound the P-T limits for the reactor vessel inlet and outlet nozzles for Beaver Valley Unit 1 at 50 EFPY.

WCAP-18 102-NP February 20 I 8 Revision l

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 8-3 MATERI AL PROPER TY BASIS LIMITIN G MATERIAL : Lower Shell Plate B6903-1 using Reg ulatory Guide 1.99 Position 1.1 data LIMITIN G ART VALUES AT 50 EF PY: 1/4T, 244.5° F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 i;=============:;:I .--.- -.---- --- ~

j operlim Vers,on:S 4 Run :19454 Opertim .xlsm Version : 5.4 2250

!Leak Test Limit~

2000 Unacceptable Acceptable 0 eration 0 eration 1750 1500 tn

-~

iii Q.

1250

(/)

(/)

~

0.

-0 S 1000

I

~

n:J u

750 Criticality Limit based on 500 inservice hydrostatic test temperature (301°F) for the service period up to 50 EFPY 250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-1 Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F/hr) Applicable for 50 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G Methodo logy (w/ K1c)

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/20 18 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 8-4 MATERJ AL PROPER TY BASIS LIMITIN G MATERIAL: Lower Shell Plate B6903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITIN G ART VALUES AT 50 EF PY: l/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 ,;::::===========::::::;------- --- -7 loperlim Version :5.4 Run:19454 Operlim .xlsm Version : 5.41 2250 2000 Unacceptable Acceptabl e 0 eration 0 eration 1750 1500

.2>

t/1 C.

a, 1250 t/1 t/1 a,

a.

'C a, 1000 (J

ni

(.)

750 Cooldown Rates

°F/Hr 500 Steady-St ate

-20

-40

-60

-100 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderat or Tempera ture (Deg. F)

Figure 8-2 Beaver Valley Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, and -100°F/hr) Applicab le for 50 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ K1c)

WCAP-18 102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-5 Table 8-1 Beaver Valley Unit 1 50 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodo logy (w/

K 1c, w/ Flange Notch, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality I00°F/hr Heatup I00°F/hr Criticality T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 0 301 0 60 0 301 0 60 602 301 1190 60 552 301 947 65 602 305 1241 65 552 305 990 70 602 310 1303 70 552 310 1042 75 602 315 1358 75 552 315 1099 80 602 320 1417 80 552 320 1162 85 602 325 1483 85 552 325 1232 90 602 330 1555 90 552 330 1310 95 602 335 1636 95 552 335 1395 100 602 340 1724 100 552 340 1488 105 602 345 1821 105 552 345 1592 II 0 603 350 1929 110 552 350 1706 115 604 355 2048 115 552 355 1832 120 606 360 2179 120 552 360 1971 125 609 365 2324 125 552 365 2124 130 612 370 2483 130 552 370 2292 135 616 135 552 375 2464 140 621 140 553 145 627 145 555 150 633 150 557 155 640 155 561 160 648 160 565 165 657 165 570 170 667 170 575 175 678 175 582 180 691 180 590 185 704 185 598 190 719 190 608 195 736 195 619 200 755 200 631 205 775 205 645 210 798 210 660 215 823 215 677 220 851 220 696 225 882 225 717 WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-6 Table 8-1 Beaver Valley Unit 1 50 EFPY Heatup Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodo logy (w/

K1c, w/ Flange Notch, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality I00°F/hr Heatup 100°F/hr Criticality T {°F) P (psig) T (°F) P (psig) T {°F) P (psig) T (°F) P (psig) 230 915 230 741 235 953 235 766 240 994 240 795 245 1040 245 827 250 1085 250 861 255 1132 255 900 260 1184 260 943 265 1241 265 990 270 1303 270 1042 275 1358 275 1099 280 1417 280 1162 285 1483 285 1232 290 1555 290 1310 295 1636 295 1395 300 1724 300 1488 305 1821 305 1592 310 1929 310 1706 315 2048 315 1832 320 2179 320 1971 325 2324 325 2124 330 2483 330 2292 335 2464 Leak Test Limit T (°F) P (psig) 283 2000 301 2485 WCAP-18102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghou se Non-Propri etary Class 3 8-7 Table 8-2 Beaver Valley Unit 1 50 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodol ogy (w/ K1c, w/ Flange Notch, and w/o Margins for Instrumentation Errors)

Steady State -20°F/hr. -40°F/hr. -60°F/hr. -100°F/hr.

T {°F) P (psig) T {°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 607 60 563 60 518 60 426 65 621 65 608 65 564 65 519 65 426 70 621 70 609 70 565 70 520 70 427 75 621 75 610 75 566 75 521 75 428 80 621 80 611 80 567 80 522 80 429 85 621 85 613 85 569 85 523 85 431 90 621 90 614 90 570 90 525 90 432 95 621 95 616 95 572 95 527 95 434 100 621 100 618 100 574 100 529 100 436 105 621 105 621 105 576 105 531 105 439 110 621 110 621 110 579 110 534 110 442 115 621 115 621 115 582 115 537 115 445 120 621 120 621 120 585 120 541 120 449 125 621 125 621 125 589 125 545 125 453 130 621 130 621 130 593 130 549 130 458 130 680 130 637 135 598 135 554 135 464 135 684 135 641 140 603 140 559 140 470 140 689 140 646 145 609 145 566 145 477 145 694 145 652 150 615 150 572 150 485 150 700 150 658 155 623 155 580 155 494 155 706 155 665 160 630 160 588 160 504 160 713 160 672 165 639 165 598 165 515 165 721 165 680 170 649 170 609 170 527 170 729 170 689 175 660 175 620 175 541 175 739 175 700 180 672 180 633 180 556 180 749 180 711 185 685 185 648 185 573 185 761 185 723 190 700 190 664 190 593 190 774 190 737 195 717 195 682 195 614 195 788 195 752 200 735 200 702 200 637 200 803 200 769 205 755 205 724 205 664 205 821 205 788 210 778 210 748 210 693 WCAP-181 02-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-8 Table 8-2 Beaver Valley Unit 1 50 EFPY Cooldown Curve Data Points using the 1998 Edition through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange Notch, and w/o Margins fo r Instrumentation Errors)

Steady State -20°F/hr. -40°F/hr. -60°F/hr. -100°F/hr.

T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 210 840 210 808 215 802 215 775 215 725 215 861 215 831 220 830 220 805 220 761 220 884 220 856 225 860 225 838 225 801 225 910 225 884 23 0 894 230 875 230 846 230 938 230 915 235 93 1 235 916 235 895 235 970 235 949 240 973 240 961 240 949 240 1004 240 987 245 1018 245 1011 245 1010 245 1043 245 1029 250 1069 250 1067 250 1067 250 1085 250 1075 255 I 125 255 11 25 255 1125 255 11 32 255 11 27 260 11 83 260 11 83 260 I 183 260 I 184 260 1183 265 1241 265 1241 265 1241 265 1241 265 1241 270 1305 270 1305 270 1305 270 1305 270 1305 275 1375 275 1375 275 1375 275 1375 275 1375 280 1452 280 1452 280 1452 280 1452 280 1452 285 1537 285 1537 285 1537 285 1537 285 1537 290 1632 290 1632 290 1632 290 1632 290 1632 295 1736 295 1736 295 1736 295 1736 295 1736 300 1851 300 1851 300 1851 300 1851 300 185 1 305 1979 305 1979 305 1979 305 1979 305 1979 3 10 2120 310 2120 3 10 21 20 310 2120 310 2120 315 2275 315 2275 315 2275 315 2275 315 2275 320 2448 320 2448 320 2448 320 2448 320 2448 WCA P-18102-NP February 2018 Rev ision 1

• • • This record was final approved on 2/2 7/201 8 8:36 :20 AM . ( Thi s statement was add ed by the PRIME system upon its validation )

Westinghouse Non-Proprietary Class 3 9-l 9 REFERENCES I. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U. S.

Nuclear Regulatory Commission, May 1988.

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.

3. Appendix G to the 1998 Edition through the 2000 Addenda of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI , Division I, " Fracture Toughness Criteria for Protection Against Failure."

4. Code of Federal Regulations, IO CFR Part 50, Appendix G, " Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243 , dated December 19, 1995.

5. Westinghouse Report WCAP-17896- NP, Revision 0, "Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program,"

September 2014 .

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

7. RSICC Computer Code Collection CCC-650, " DOORS3 .2: One-, Two- and Three-Dimensio nal Discrete Ordinates Neutron/Photon Transport Code System," April 1998.

8. RSICC Data Library Collection DLC-185 , " BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/ B-Vl for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

9. 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 14, 2014. [Agencywide Document Management System (A DAMS)

Accession Number ML14149AJ65]

I 0. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 L WR Edition, Branch Technical Position 5-3 , " Fracture Toughness Requirements,"

Revision 2, U.S . Nuclear Regulatory Commission, March 2007.

11. ASME Boiler and Pressure Vessel (B&PV) Code, Section Ill, Division I , Subsection NB , Section NB-2300, "Fracture Toughness Requirements for Material."

12. Westinghouse Letter MCOE-L TR-15-15-NP, Revision I, " Determination of Unirradiated RT NOT Values of the Four Beaver Valley Unit I Reactor Vessel Beltline Plate Materials Using a Hyperbolic Tangent Curve Fit," dated July 6, 2015 .

13. BWRVJP-173-A : BWR Vessel and Internals Project: Evaluation of Chemistry Data f or BWR Vessel Nozzle Forging Materials. EPRI , Palo Alto, CA : 2011. I 022835 .

14. Westinghouse Report WCAP-15571 Supplement I, Revision 2, "Analysis of Capsule Y from the Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program," September 2011 .

15. Florida Power & Light Letter L-20 I 0-078, Attachment 5, " License Amendment Request Extended Power Uprates Licensing Report Florida Power & Light St. Lucie Nuclear Plant, Unit I," April 20 I 0.

{ADAMS Accession Number MLJOJ 160193]

16. Westinghouse Report WCAP-16012 , Revision 0, "Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003.

WCAP-18102-N P February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 9-2

17. FirstEnergy Nuclear Operating Company Letter L-12-077, " Pressure and Temperature Limits Report Revision," dated April 5, 2012 .

18. Westinghouse Report WCAP-16799-N P, Revision I, "Beaver Valley Power Station Unit I Heatup and Cooldown Limit Curves for Normal Operation," June 2007.

19. K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/lndustry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number MLI 100705 70}

20. U.S. NRC Technical Letter Report TLR-RES/DE/CIB-2013-01 , " Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels," Office of Nuclear Regulatory Research [RES] , dated November 14, 2014. [ADAMS Accession Number MLl431 8Al 77}

21. Westinghouse Report WCAP-14040-N P-A, Revision 2, " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996.

22. Westinghouse Report WCAP-15571 , Revision 0, " Analysis of Capsule Y from Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program," November 2000.

WCAP-18102-N P February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 A-1 APPEN DIX A THER MAL STRES S INTEN SITY FACTO RS (K1t)

Tables A- I and A-2 contain the thermal stress intensity factors (K1t) for the maximum heatup and cooldown rates at 50 EF PY for Beaver Valley Unit I. The reactor vessel cylindric al shell radii to the 1/4T and 3/4T locations are as follows:

• 1/4T Radius= 80.625 inches

• 3/4T Radius= 84.562 inches WCAP-18 102-NP February 20 I 8 Revision 1

• • • This record was final approved on 2/27/20 18 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghou se Non-Propri etary Class 3 A-2 Table A-1 K 11 Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements and w/o Margins for Instrumen t Errors)

Water Vessel Temperature l /4T Thermal Stress Vessel Temperatu re 3/4T Thermal Stress Temp. at l/4T Location for Intensity Factor at 3/4T Location for Intensity Factor (OF) 100°F/hr Heatup (°F) (ksi in.) 100°F/hr Heatup (°F) (ksi in.)

60 56.130 -0 .987 55 .065 0.493 65 58 .927 -2.377 55.425 1.455 70 62.129 -3 .521 56.315 2.377 75 65.562 -4.586 57.748 3.208 80 69.262 -5.475 59.641 3.929 85 73.079 -6 .273 61.944 4.558 90 77 .089 -6 .948 64.601 5. 101 95 81.193 -7 .553 67.562 5.578 100 85.435 -8 .069 70.788 5.991 105 89.755 -8.5 3 1 74.238 6.353 110 94. 171 -8.928 77.881 6.671 115 98.650 -9.285 81.690 6.951 120 103.196 -9 .594 85.642 7.198 125 107.790 -9.875 89.717 7.418 130 11 2.433 -I 0. 118 93.898 7.612 135 117. 114 -10.341 98.171 7.785 140 121.829 - I 0.535 102.523 7.940 145 126.574 -10 .715 106.944 8.080 150 131.343 -10 .873 111.424 8.206 155 136.136 -11.020 115 .955 8.320 160 140 .945 -11.151 120.5 29 8.423 165 145 .773 -11.275 125 .142 8.519 170 150 .6 13 -11.385 129.788 8.606 175 155.467 -11 .49 1 134.462 8.687 180 160 .33 0 -11.586 139. 161 8.762 185 165 .204 -11.678 143.881 8.833 190 170.083 -11.763 148.620 8.899 195 174.972 -11.845 153.374 8.961 200 179.864 -11.920 158 .143 9.020 205 184.764 -11.995 162.923 9.077 2 10 189.666 -12 .064 167.713 9. 131 WCAP-181 02-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-3 Table A-2 K 11 Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Cooldown Curves (w/ Flange Requirements and w/o Margins for Instrument Errors)

Water Vessel Temperature at l/4T -100°F/hr Cooldown Temp. Location for -100°F/hr l/4T Thermal Stress (OF) Cooldown (°F) Intensity Factor (ksi in.)

210 232.426 13.510 205 227 .3 52 13.454 200 222 .278 13.398 195 217.204 13.342 190 2 12.131 13 .286 185 207.057 13 .230 180 201 .983 13 . 175 175 196.909 13 . 119 170 191.836 13.063 165 186 .762 13 .008 160 181.688 12.952 155 176.615 12.897 150 171.541 12.842 145 166.468 12 .786 140 161.395 12.731 135 156 .322 12.676 130 151.249 12.622 125 146 .176 12.567 120 141.103 12.512 115 136.03 1 12.457 II 0 130.958 12.403 105 125.8 86 12 .349 100 120.813 12 .295 95 115 .741 12.240 90 110.669 12. 187 85 I 05 .597 12. 133 80 I 00.526 12.079 75 95.454 12.025 70 90.382 11 .972 65 85. 3 11 11.919 60 80.241 11 .865 WCAP-181 02-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 B-1 APPENDIX B REACTO R VESSEL INLET AND OUTLET NOZZLE S As described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. B-1 ], reactor vessel non-beltline materials may define pressure-tem perature (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 discontinuiti es 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 (RT Nor) 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-1404 0-A , Revision 4 [Ref. B-2] was used in the main body of this report to develop P-T limit curves for the limiting Beaver Valley Unit I cylindrical shell beltline material; however, WCAP-1404 0-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 corner 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 corner region for Beaver Valley Unit I 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 embrittlemen t of the reactor vessel above the beltline region . Therefore, the P-T limit curves are developed for the nozzle inside corner 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 corner.

An axial flaw is postulated at the inside surface of the reactor vessel nozzle corner 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 flaw. A discussion of the flaw depth is located in Appendix 8.2.

B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (Kie) used for the inlet and outlet nozzle material is defined in Appendix G of the Section XI ASME Code, as discussed in Section 6 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. 8-3], which is described in Section 7 of this report, and weight percent (wt.

%) copper (Cu) and nickel (Ni), initial RT NDT value, and projected neutron fluence as inputs. The ART values for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Table 8-1 and a summary of the limiting inlet and outlet nozzle ART values for Beaver Valley Unit I is presented in Table 8-2.

Nozzle Material Properties Copper and nickel weight percent values and the subsequent Position 1.1 CF values, were previously documented in WCAP-15571 , Supplement I, Revision 2 [Ref. 8-4] and are taken directly from this source for this analysis. The initial RT NDT values were determined for each of the Beaver Valley Unit I WCAP-18102 -NP February 20 I 8 Revision l

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 8-2 reactor vessel inlet and outlet nozzle forging materials using the BWRVIP- 173-A [Ref. 8-5] Alternativ e

Approach 2 methodolo gy, contained in Appendix B of that report. For all six of the Beaver Valley Unit I inlet and outlet nozzle materials, CVGraph Version 6.02 was utilized to plot the material-sp ecific Charpy V-Notch impact energy data from the Certified Material Test Reports (CMTRs) to determine the transition temperatur es at 35 ft-lb and 50 ft-lb, as specified in the Alternative Approach 2 methodolo gy.

The 35 ft-lb and 50 ft-lb temperatur es were then evaluated per the Alternative Approach 2 methodolo gy presented in B WRVIP- I 73-A, to determine the initial RT NOT values for the inlet and outlet nozzle materials for Beaver Valley Unit 1. The Charpy V-Notch forging specimen orientation for the inlet and outlet nozzles was not reported in the CMTRs; thus, it was conservati vely assumed that the orientation was the "strong direction" for each nozzle forging. Therefore, the 50 ft-lb transition temperatur es for the inlet nozzles were increased by 30° F to provide conservati ve estimates for specimens oriented in the weak direction per the Alternative Approach 2 methodolo gy in BWRVIP- 173-A .

The material properties of the Beaver Valley Unit I inlet and outlet nozzle forging materials are documente d in Table B-1.

Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > I MeV) exposure of the Beaver Valley Unit I reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculation s were calculated at the lowest extent of the nozzles (i.e., the nozzle to nozzle shell weld locations) and were chosen at an elevation lower than the actual elevation of the postulated flaw, which is at the inside corner of the nozzle, for conservati sm.

Per Table 2-7, the inlet nozzles are determined to receive a projected maximum fluence of 2.10 x 10 17 n/cm 2 (E > I MeV) at the lowest extent of the nozzles at 50 EFPY. Similarly, the outlet nozzles are projected to achieve a maximum fluence value of 1.61 x 10 17 n/cm 2 (E > 1 MeV) at the lowest extent of the nozzles at 50 EFPY . Per NRC RIS 2014-11 [Ref. 8- I] , embrittlem ent of reactor vessel materials, with projected fluence values greater than I x 10 17 n/cm2, must be considered. However, the second conclusion of TLR-RES/ DE/CIB-2 0 I 3-0 I [Ref. B-6] states that if !iRTNOT is calculated to be less than 25° F, then embrittlem ent may be ignored. This conclusion was applicable to each of the Beaver Valley Unit 1 nozzle materials.

The neutron fluence values used in the ART calculation s for the Beaver Valley Unit I inlet and outlet nozzle forging materials are summarize d in Table B-1.

The use of the embrittlem ent conclusion ofTLR-RE S/DE/CIB -2013-01 [Ref. B-6], and thus the limiting ART values summarize d in Table B-2, will remain unchanged as long as the fluence values assigned to the inlet and outlet nozzles remain below 7.98 x I 0 17 n/cm 2 (E > 1.0 MeV) and 4. IO x I 0 17 n/cm 2 (E > 1.0 MeV), respectively. lfthese fluence values are reached, embrittlem ent must be considered and the nozzle ART values reported herein will increase.

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Westinghouse Non-Proprietary Class 3 B-3 Table B-1 ART Calculations for the Beaver Valley Unit 1 Reactor Vessel Nozzle Materia

~

T ls at 50 EFPY al Wt.% Wt.% CF<*> Fluence at Lowest 8 Reactor Vessel Material Extent of Nozzle FF(b) RTNDT(U)

(c)

ARTNDT (d) Gu a l!.(e) Margin ART

a. cu<*> Ni<*> (OF) (OF) (OF) (OF) (OF) (OF) (OF)

(n/cm2, E > 1.0 MeV)

!ll

(/)

Inlet Nozzle B6608-1 0.10 0.82 67.0 0.0210

~ X 10 19 0. 1773 48.5 0 (11.9) 0 0 0.0 48 .5

~ Inl et Nozzle B6608-2 0.10 0.82 67.0 0.0210

!ll X 10 19 0. 1773 -15 .2 0(11.9) 0 0 0.0

-0 -15.2

-0 Inlet Nozzle B6608-3 0.08 0.79 51.0 0.0210

~ X 10 19 0. 1773 11.4 0 (9.0) 0 0 0.0 11.4 (I)

Outlet Nozzle B6605-1 0.1 3 0.77 95.3 0.0161 C.

0 X 10 19 0.1501 -26. 2 0 ( 14.3) 0 0 0.0 -26. 2 Outlet Nozzle B6605-2 0. 13 0.77 95 .3 0.0]61 10 19 N

X 0.1501 3.3 0 (14 .3 ) 0 0 0.0 3.3

--.J Outlet Nozzle B6605-3 0.09 0.79 58.0 0.016) 10 19

X 0.1501 IO. I 0 (8 .7) 0 0 0.0 10 .1

~

0, Notes:

0, w (a) Cu and N i wt. % va lues, as we ll as CF va lues were obtained from WCA P-155 7 1, Suppleme nt I, Revision 2 [Ref. 8 -4].

(j)

(b) Fluence va lues co nse rvati ve ly correspond to 50 EF PY flu ence va lues i--J at the lowest ex tent o f the nozzle we ld. FF va lues we re ca lculated using 0

Rev ision 2. Regul atory G ui de 1.99,

}>

$'. (c) RT NDT(Ul values were determine d usin g the Alternati ve Approach 2 methodo logy as described in A ppendix B o f BWRVIP -1 73-A .

(d) Calculate d L'. RTNoT values less than 25°F have bee n reduced to zero

~

per TLR-RES /DE/C IB-201 3-0 1 [Re f. B-6]. Actua l ca lculated L'. RT NDT

T

• parenthes es fo r these mate ri als. va lues are listed in

(/)

iii (e) Per Regulatory Guide 1.99, Rev isio n 2, the base meta l nozzle fo rging mate rial s crl!. = I 7°F fo r Pos iti on 1. 1 witho ut surve ill ance data. However, CD crl!. need not exceed 3 0. 5*1'.RT NDT* Thi s conclusio n is consistent with footnote (d) sin ce embrittlem ent is not considere d for any o f the six Beaver Va ll ey Unit I nozzles; hence (I) crl!. = 0.

?.

!ll

(/)

!ll C.

C.

(I)

C.

cr g.

(I)

"U

0

~

m

(/)

'<(/)

CD 3

C

-0 0

ui

~

5:

~

cs-2-

WCAP-18102-NP February 2018 Revision I

Westinghouse Non-Proprietary Class 3 B-4 Table B-2 Summary of the Limiting ART Values for the Beaver Valley Unit 1 Inlet and Outlet Nozzle Materials Nozzle Material and ID Limiting ART Value EFPY Number (OF)

Inlet Nozzle B6608-1 48.5 50 Outlet Nozzle B6605-3 10.1 B.2 NOZZLE COOLDOWN PRESSURE-TEMPERATURE LIMITS Allowable pressures are determined 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 Beaver Valley Unit I nozzle fracture toughness used to determine the P-T limits is calculated using the limiting inlet and outlet nozzle ART values from Table B-2. The stress intensity factor correlations used for the nozzle comers are provided in ORNL study, ORNL/TM-201 0/246 [Ref. B-7], and are consistent with ASME PVP201 l-57015 [Ref. B-8]. The methodology includes postu lating an inside surface 1/4T nozzle comer flaw, and calculating through-wall nozzle comer stresses for a cooldown rate of 100°F/hour.

For one of the inlet nozzles, the ART is 48.5°F for material B6608-1 as shown in Table B-1. Using a l/4T circular corner flaw creates a situation where the reactor vessel inlet nozzle curve for the 48.5 °F ART value becomes very slightly more limiting than the traditional beltline curves at the lowest temperature region (i.e., 60°F) of the P-T limit curves. At this temperature region, the typical pressure vessel is depressurized and at atmospheric pressure.

Therefore, in lieu of using a I /4T circular corner flaw depth for the limiting inlet nozzle, Article G-2120 of the ASME Section XI Code [Ref. B-9] states that for sections greater than 12" thick, the postulated defect for the 12" section may be used. The thickness of the reactor vessel inlet nozzle is approximately 15"; therefore, a 3" postulated flaw depth (1 /4

• 12" = 3") may be used for the limiting inlet nozzle P-T curve calculations. A 3" postulated flaw depth is approximately 20% or 1/5T of the through wall thickness for the limiting inlet nozzle. Thus, a l /5T flaw is only used for the Inlet Nozzle B6608-1 , and all other nozzles use the l /4T flaw depth for the P-T limits evaluation. The Appendix G evaluation also contains additional conservatism, such as factor of 2 on pressure and a lower bound fracture K1c curve, which can account for the use of a I/5T flaw depth.

The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the fonn:

where, cr = through-wall stress distribution x = through-wall distance from inside surface A 0, A 1, A 2, A 3 = coefficients of polynomial fit for the third-order polynomial, used in the stress intensity factor expression di sc ussed below WCAP-18102-N P February 2018 Revision 1

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Westinghou se Non-Propri etary Class 3 B-5 The stress intensity factors generated for a rounded nozzle corner for the pressure and thermal gradient were calculated based on the methodolo gy provided in ORNL/TM -20 I 0/246. The stress intensity factor expression for a rounded corner is:

K, = ../rra [0. 706Ao+ 0.537 (:a) A 1+ 0.448 ( ~) A2+ 0.393 (::)A 3]

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 stress intensity factors for the inlet nozzle l/5T postulated flaw depth are calculated using the same equations as the 1/4T flaw depth. Although the K 1 equation above only calculates the stress intensity factor at the deepest point on a postulated flaw and the use of a large l /4T and l/5T flaw bounds the potential non-conse rvatisms of evaluating K 1 at only the deepest point.

The Beaver Valley Unit I reactor vessel inlet and outlet nozzle P-T limit curves are shown in Figures 8-1 and B-2, respectivel y, based on the stress intensity factor expression discussed above; also shown in these figures are the traditional beltline cooldown P-T limit curves from Figure 8-2. The nozzle P-T limit curves are provided for a cool down rate of I 00°F/hr, along with a steady-stat e curve.

An outside surface flaw in the nozzle was not considered because the pressure stress is significant ly 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 curve 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 corner.

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 beltline curves. Therefore, the P-T limits provided in Section 8 for 50 EFPY remain limiting for the beltline and non-beltline reactor vessel componen ts.

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Westinghouse Non-Proprietary Class 3 B-6 2500 Inlet Nozzle 2250 Cooldown

-100 °F/hr 2000 Inlet Nozzle Steady State 1750

--~ 1500 1/)

C.

Cl)

s 1/)

1250 1/)

Cl) ll.

- E Cl) 1000 1/)

en C:

0 750 0

0

...0 Cooldown Rates

( .)

°F/Hr Cl) Steady-State

~ 500 -20

-40

-60

-100 Boltup 250 Temperature 0

0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature {°F)

Figure B-1 Comparison of Beaver Valley Unit 1 Beltline P-T Limits to Inlet Nozzle Limits WCAP-18102-NP February 2018 Revision I

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Westing house Non-Proprietary Class 3 B-7 2500 2250 - Outlet Nozzle Cooldown

-100 °F/hr 2000 -

Outlet Nozzle Steady State 1750

- 1500 C) t/1 0.

a,

~

I t/1 1250 t/1 a,

a.

~

- E a,

1000 -

t/1 Cl)

C:

ta 0 750 0

u Cooldown

~

0 Rates

( .) °F/Hr ta a, Steady-State 0:: 500 -20

-40

-60

-100 250 Boltup Temperature 0

0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (°F)

Figure B-2 Comparison of Beaver Valley Unit 1 Beltline P-T Limits to Outlet Nozzle Limits WCAP-18102-N P February 2018 Revision I

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Westinghouse Non-Proprietary Class 3 B-8 B.3 REFERENCES 8-1 NRC Regulatory Issue Summary 2014-11 , " Informat ion on Licensing Applications for Fracture Toughne ss Requirem ents for Ferritic Reactor Coolant Pressure Boundary Compone nts," U.S.

Nuclear Regulatory Commiss ion, October 14, 2014. [ADAMS Accession Number ML14149Al65}

8-2 Westingh ouse Report WCAP-1 4040-A , Revision 4, " Methodo logy Used to Develop Cold Overpres sure Mitigatin g System Setpoints and RCS Heatup and Cooldow n Limit Curves,"

May 2004.

8-3 Regulatory Guide 1.99, Revision 2, " Radiation Embrittle ment of Reactor Vessel Materials

," U.S.

Nuclear Regulatory Commiss ion, May 1988.

8-4 Westinghouse Report WCAP-15571 Supplem ent I, Revision 2, "Analysis of Capsule Y from the Beaver Valley Unit I Reactor Vessel Radiation Surveilla nce Program, " Septembe r 2011.

8-5 BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI , Palo A Ito, CA : 2011. 1022835.

8-6 U.S. NRC Technical Letter Report TLR-RES /DE/CIB -2013-0 I, " Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels," Office of Nuclear Regulatory Research

[RES],

dated Novembe r 14, 2014. [ADAMS Accession Number ML14318AJ 77}

8-7 Oak Ridge National Laborato ry Report, ORNL/T M-2010/ 246, "Stress and Fracture Mechani cs Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles -

Revision I," June 2012.

8-8 ASME PVP201 l-57015, "Additio nal Improvements to Appendi x G of ASME Section XI Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerv ille, July 2011.

8-9 Appendix G to the 200 I Edition through the 2003 Addenda of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI , Division 1, " Fracture Toughne ss Criteria for Protection Against Failure."

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Westinghouse Non-Propri etary Class 3 C-1 APPENDIX C OTHER REACTOR COOLANT PRESSURE BOUNDARY FERRITIC COMPONENTS IO CFR Part 50, Appendix G [Ref. C-1 ], requires that all Reactor Coolant Pressure Boundary (RCPB) componen ts meet the requiremen ts of Section Ill of the ASME Code. The lowest service temperatur e

requiremen t for all RCPB componen ts, which is specified in NB-2332(b ) of the Section III ASME Code, is the relevant requiremen t that would affect the pressure-te mperature (P-T) limits. The lowest service temperatur e (LST) requireme nt ofNB-233 2(b) of the Section III ASME Code is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2 1/2 inches [Ref. C-2].

Note that the Beaver Valley Unit 1 reactor coolant system does not have ferritic materials in the Class 1 piping, pumps or valves. Therefore, the LST requiremen ts ofNB-233 2(b) are not applicable to the Beaver Valley Unit 1 P-T limits and the only ferritic RCPB componen ts that are not part of the reactor vessel beltline or extended beltline consist of the replaceme nt steam generators , the replaceme nt reactor vessel closure head and the pressurizer.

The replaceme nt steam generators (RSG) were designed and evaluated to the 1989 Edition Section Ill ASME Code and met all applicable requiremen ts at the time of construction. Furthermo re, the RSGs have not undergone neutron embrittlem ent that would affect P-T limits. Therefore, no further considerat ion is necessary for these componen ts with regards to P-T limits.

The replaceme nt reactor vessel closure head materials have been considered in the developme nt of the P-T limits, see Section 6.3 of this report for further detail. The replaceme nt reactor vessel closure head was constructe d to the 1989 Ed ition Section Ill ASME Code and met all applicable requiremen ts at the time of construction. Furthermo re, the replaceme nt reactor vessel closure head has not undergone neutron embrittlem ent that would affect P-T limits.

T he pressurize r was constructe d to the 1965 Edition through 1966 Winter Addenda Section Ill ASME Code and ASME Code Case-1401 and met all applicable requiremen ts at the time of constructio n and is original to the plant. Furthermo re, the pressuri zer has not undergone neutron embrittlem ent that would affect P-T limits. Therefore, no further considerat ion is necessary for this componen t with regards to P-T limits .

C.1 REFERENCES C-1 Code of Federal Regulation s, IO CFR Part 50, Appendix G, "Fracture Toughness Requireme nts,"

U.S. Nuclear Regulatory Commissio n, Federal Register, Volume 60, No. 243, December 19, 1995.

C-2 ASME B&PV Code Section Ill , Division I, NB-2332, "Material for Piping Pumps, and Valves, Excluding Bolting Material."

WCAP-181 02-NP February 2018 Revision I

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Westinghouse Non-Prop rietary Class 3 D-1 APPENDIX D BEAVER VALLEY UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION D.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [Ref. D-1] describes general procedur es acceptab le to the NRC staff for calculatin g the effects of neutron radiation embrittle ment of the low-alloy steels currently used for light-wat er-cooled reactor vessels. Position 2.1 of Regulatory Guide 1.99, Revision 2, describes the method for calculatin g the adjusted reference temperat ure of reactor vessel beltline materials using surveillan ce capsule data. The methods of Position 2.1 can only be applied when two or more credible surveillan ce data sets become available from the reactor in question.

The credibility of all surveillan ce program data previousl y applicable to the Beaver Valley Unit I reactor vessel was assessed in WCAP-1 7896-NP [Ref. D-2]. However, since this evaluatio n, additional weld Heat# 90136 surveillan ce capsule data from the Millstone Unit 2 surveilla nce program has been deemed applicabl e, and the Beaver Valley Unit I surveilla nce capsule tluence values have been updated. Thus, this Appendix documen ts the necessary updates to the credibility evaluatio n of surveilla nce program data applicabl e to Beaver Valley Unit I.

The Millstone Unit 2 surveillan ce program includes two distinct welds, Heat# 90136 and Heat# 10137.

In previous analyses, this weld surveilla nce data was treated as one combined weld and subsequently analyzed together. However, these two weld metal heats were not melted together into a tandem weld ;

they were individua lly deposited . It cannot be determin ed with full confiden ce how much of the overall surveillan ce weld is which weld metal heat and, furthenno re, exactly which weld heat specimen s are contained in which surveillan ce capsules in the Millstone Unit 2 program .

The Millstone Unit 2 (combine d) surveilla nce weld data met the second and third credibility criteria of Regulato ry Guide 1.99, Revision 2 [Ref. D-1]. Addition ally, Table D-2 of WCAP-1 6012 [Ref. D-3]

indicates that all of the measured weld t.RT NDT values were within the I-sigma scatter band; therefore ,

suggestin g that there is good agreemen t between the measured capsule data and the embrittle ment correlatio ns. If the two heats of weld material were evaluated individually, one would expect that the scatter in the data would decrease since the irradiated material would embrittle differentl y for the two separate welds with different, as-measu red, copper and nickel contents. However ,

since the (combine d) weld material already passes the Regulatory Guide 1.99, Revision 2 credibilit y analysis, a re-evaluation of the material (as two separate heats) is not expected to significantly change the overall results of the subseque nt reactor vessel integrity analyses. Thus, the surveillan ce weld metal will be considere d to be only Heat # 90136 for the evaluatio ns contained herein. All currently determin ed input data for Position 2.1 chemistry factor determin ation (See Section 5) and surveillan ce data credibilit y assessme nt documen ted in this Appendix will be used "as-is," as documen ted in the Millstone Unit 2 surveillance capsule analyses of record.

For conserva tism, no reduction in the margin term of Regulatory Guide 1.99, Revision 2 [Ref. D-1] and 10 CFR 50.61 [Ref. D-4] was taken to account for the additional uncertainties, despite the data remainin g credible (see Section D.2). Additionally, the Beaver Valley Unit I intermed iate to lower shell girth weld seam 11-714 (Heat# 90136) was assigned the most limiting calculated ART and RTPTS values during the evaluatio ns contained in Section 7 and Appendi x E, respectively. Thus, since the values determin ed using Position 2.1 are less conserva tive than the values determin ed using Position 1.1 , the more conserva tive WCAP-18 102-NP February 2018 Revision 1

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Westinghouse Non-Proprieta ry Class 3 D-2 Position 1. 1 values were used . However, despite these additional conservatism s, the Beaver Valley Unit 1 intermediate to lower shell girth weld seam 11-714 (Heat# 90136) was not the limiting material in any evaluation.

D.2 EVALUATION The only required updates to the previously determined credibility conclusions are to update Criterion 3 of Regulatory Guide 1.99, Revision 2 [Ref. D-1] to include the combined surveillance capsule data set for weld Heat # 90136 from both the St. Lucie Unit I and Millstone Unit 2 surveillance programs and to update Criterion 3 of Regulatory Guide 1.99, Revision 2 [Ref. D-1] utilizing the Beaver Valley Unit l surveillance capsule fluence values documented in Appendix F. These evaluations are documented herein. Criterion # 1, 2, 4, and 5 conclusions remain unchanged from those documented in Appendix D of WCAP-1789 6-NP [Ref. D-3]. Note also that the credibility assessment of Heat# 305414 data remains valid as documented in Appendix D of WCAP-1789 6-NP [Ref. D-2].

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of

~RTN OT values about a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2 [Ref. D-1] normally should be less than 28°F for welds and l 7°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82 [Ref. D-5].

The functional form of the least-squares method as described in Regulatory Guide 1.99, Revision 2 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ~RTNOT values about this line is less than 28°F for the weld.

Following is the calculation of the best-fit line as described in Reference D-1 . In addition, the recommende d 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. D-6]. At this meeting, the NRC presented five cases. Of the five cases, Case 5 ("Surveillan ce Data from Other Sources Only") most closely represents the situation for the Beaver Valley Unit I reactor vessel intermediate to lower shell girth weld seam 11-714 (Heat # 90136) as described below. Of the five cases, Case I

(" Surveillance data available from plant but no other source") most closely represents the situation for the Beaver Valley Unit I surveillance materials as described below.

Heat # 90 I 36 (Case 5) - This weld heat pertains to the intermediate to lower shell girth weld seam 11-714 in the Beaver Valley Unit I reactor vessel. This weld heat is not contained in the Beaver Valley Unit I surveillance program. However, it is contained in the St. Lucie Unit I and Millstone Unit 2 surveillance programs . NRC Case 5 per Reference D-6 is entitled "Surveillanc e Data from Other Sources Only" and most closely represents the situation for Beaver Valley Unit 1 weld Heat# 90136.

Lower Shell Plate 86903-1 (Case 1) - This plate material will be evaluated using the NRC Case l guidelines as described above.

Weld Heat # 305424 (Case I) - This weld heat pertains to the intermediate shell longitudinal welds in the Beaver Valley Unit I reactor vessel. NRC Case 1 per Reference D-6 regarding " Surveillance data available from the plant but no other source" most closely represents the situation for Beaver Valley Unit I weld Heat# 305424.

WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry C lass 3 D-3 Credibility Assessment Case 5: Weld Heat# 90136 (St. Lucie Unit I Data Only)

Following the NRC Case 5 guidelines, the St. Lucie Unit I and Millstone Unit 2 surveillance weld metal (Heat# 90136) will be evaluated for credibility. Weld Heat # 90136 pertains to Beaver Valley Unit I reactor vessel intermediate to lower shell girth weld seam 11-714, but is not contained in the Beaver Valley Unit I surveillance program.

In accordance with the NRC Case 5 guidelines, the data from only St. Lucie Unit I will be analyzed first, since the irradiation environment for St. Lucie Unit I is judged closer to that of Beaver Valley Unit I as evidenced by the temperature adjustments documented in Table 4-2. This assessment was performed in Appendix D of WCAP-1789 6-NP [Ref. 0-2] and concluded that the surveillance data for Heat# 90136 from St. Lucie Unit I only was credible. Therefore, in accordance with Case 5, the combined data from both St. Lucie Unit I and Millstone Unit 2 will now be assessed to determine the credibility conclusion for all applicable data for weld Heat# 90136.

WCAP-18102 -NP February 2018 Revision I

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Westinghouse Non-Proprietary Class 3 D-4 Credibility Assessmen t Case 5: Weld Heat# 90136 (All data)

In accordanc e with the NRC Case 5 guidelines , the data from St. Lucie Unit 1 and Millstone Unit 2 will now be analyzed together. Data is adjusted to the mean chemical composition and operating temperatu re of the surveillance capsules. This is performed in Table D-1 .

Table D-1 Mean Chemical Composition and Operating Temperature for St. Lucie Unit 1 and Millstone Unit 2 Cu Ni Inlet Temperatu re during Material Capsule Wt.%<"l Wt. %<al Period of Irradiation (°F/bJ 97° 541 Weld Metal Heat # 90 136 104° 0.23 0.07 544.6 (St. Lucie Unit I Data) 284° 546.3 97° 544.3 Weld Metal Heat # 90136 104° 0.30 0.06 547.6 (Millstone Unit 2 Data) 83 0 548.0 MEAN 0.265 0.065 545.3 Note:

(a) Chemistry data obtained from Table 3-2 .

(b) Temperature data obtained from Table 4-2.

Therefore, the St. Lucie Unit I and Millstone Unit 2 surveillance capsule data will be adjusted to the mean chemical compositio n and operating temperatur e calculated in Table D-1.

St. Lucie Unit I data I 2 l.2°F (calculated per Table I of Regulatory Guide 1.99, Revision 2 [Ref. D-1] using Cu Wt. % =

0.265 and Ni Wt. % = 0.065 per Table D-1)

CF Surv. We ld (St. L ucie Unit I) 106.6°F (from Table 5-4)

Ratio= 121.2 + 106.6 = 1.14 (applied to St. Lucie Unit I surveillance data for weld Heat# 90136 in the credibi lity evaluation )

Millstone Unit 2 data CF Mean 121.2°F CF Surv. We ld (Mill stone Unit 2) 135.5°F (from Table 5-4)

Ratio= 121.2 + 135.5 = 0.89 (applied to Millstone Unit 2 surveillance data for weld Heat# 90136 in the credibility evaluation )

The capsule-specific temperatur e adjustmen ts are as shown in Table D-2.

WCAP-181 02-NP February 20 18 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse No n-Proprietary Class 3 D-5 Table D-2 Operating Temperature Adjustments for the St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data Inlet Temperatu re during Mean Operating Temperatu re Material Capsule Period of Irradiation (°F) Temperatu re (°F) Adjustmen t (°F) 97° 541 -4.3 Weld Metal Heat # 90 I 36 104° 544.6 -0.7 (St. Lucie Unit I Data) 284 ° 546.3 1.0 545.3 97° 544.3 -1.0 Weld Metal Heat # 90136 104° 547.6 2.3 (Millstone Unit 2 Data) 83 ° 548.0 2.7 Using the chemical compositio n and operating temperatur e adjustments described and calculated above, an interim chemistry factor is calculated for weld Heat # 90136 using the St. Lucie Unit I and Millstone Unit 2 data. This calculation is shown in Table 0-3 below.

Table D-3 Calculation of Weld Heat# 90136 Interim Chemistry Factor for the Credibility Evaluation Using St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data Material Capsule Capsule r*> FF(b) ART NOT (c)

FF*ARTNDT (x 10 19 n/cm2, E > 1.0 MeV) FF2

{°F) (OF)

Weld Metal Heat # 97° 0.5 174 0.8160 77.6 (72.34) 63.29 0.666 90136 (St. Lucie 104° 0.7885 0.9333 76.0 (67.4) 70.97 0.871 Unit I Data) 284° 1.243 1.0606 78 .7 (68 .0) 83.43 1. 125 Weld Metal Heat # 97° 0.324 0.6902 57.8 (65.93) 39.89 0.476 90136 (Millstone 104° 0.949 0.9853 48.4 (52. 12) 47.72 0.971 Unit 2 Data) 83 ° 1.74 1.1523 52 .3 (56.09) 60.29 1.328 SUM: 365 .59 5.437 CFHea, #90 136= L(FF

• L'. R T NoT) + L(FF 2) = (365.59) + (5.437) = 67.2°F Notes:

(a) f = fluence .

r (b) FF = fluence factor = 0 28

• 0 IO'log f)_

(c) .tlRTNDT va lues are the measured 30 ft-lb shift va lues. Each .tlRT DT va lu e has first been adjusted according to the temperature adj ustments summarized in Tab le D-2. Then, the .tlRTNDT values for each surve illance weld data point are adjusted by the ratios determ ined previously for weld Heat # 90 136 (pre-adjusted va lues are listed in parentheses and were take n from Tab le 4-2).

The scatter of LlRTN OT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2, Position 2. 1 [Ref. 0-1] is presented in Table 0-4.

WCAP-18102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/20 18 8:36:20 AM . ( This statement was added by the PRIME system upon its validation )

Westinghouse Non-Proprietary Class 3 0-6 Table D-4 Best-Fit Evaluation for Surveillance Weld Metal Heat# 90136 Using St. Lucie Unit 1 and Millstone Unit 2 Data CF Capsule f Measured Predicted Residual Material Capsule (Slopebes1-fi1) (x 10 19 n/cm2, FF ARTNDT ARTNDT ARTNDT <28°F (Weld)

(OF) E > 1.0 MeV) (OF) (OF) (OF)

Weld Metal Heat # 97° 67.2 0.5174 0.8160 77.6 54.8 22.7 Yes 90136 (St. Lucie 104° 67.2 0.7885 0.9333 76.0 62.7 13.3 Yes Unit I Data) 284° 67.2 1.243 1.0606 78 .7 71.3 7.4 Yes Weld Metal Heat # 97° 67.2 0.324 0.6902 57.8 46.4 11.4 Yes 90136 (Millstone 104° 67.2 0.949 0.9853 48.4 66.2 17.8 Yes Unit 2 Data) 83 0 67.2 1.74 1.1523 52.3 77.4 25 . 1 Yes The scatter of ~RT NDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. D-1 ], should be less than 28°F for weld metal. Table D-4 indicates that I 00% (six out of six) of the surveillance data points fall within the +/- I cr of 28°F scatter band for surveillanc e weld materials. Therefore, the surveillanc e weld material (Heat # 90136) is deemed "credible" per the third criterion when all available data is considered .

WCAP-18102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Prop rietary Class 3 D-7 Credibili ty Assessm ent Case I: Lower Shell Plate B6903- l and Weld Heat# 305424 Followin g the NRC Case I guideline s, the Beaver Valley Unit I surveillance plate and weld metal (Heat#

305424) will be evaluated for credibility. Note that when evaluatin g the credibilit y of the surveillance weld data, the measured liRT NOT values for the surveilla nce weld metal do not include the adjustme nt ratio procedur e of Regulato ry Guide 1.99, Revision 2, Position 2.1 , since this calculatio n is based on the actual surveillan ce weld metal measured shift values. The chemistry factors for the Beaver Valley Unit 1 surveillan ce plate and weld material contained in the surveillan ce program were calculate d in accordan ce with Regulato ry Guide 1.99, Revision 2, Position 2.1 and are presented in Table D-5. The scatter of t;RT N OT values about the functional form of a best-fit line drawn as described in Regulato ry Position 2.1 is presented in Table D-6.

Table D-5 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Beaver Valley Unit I Surveillance Capsule Data Capsule f 11 >

(c)

Material Capsule (x 10 19 n/cm2, E FF(b) ~RTNDT FF*~RTNDT (OF) FF2 (OF)

> 1.0 MeV)

V 0.297 0.6677 127.9 85.40 0.446 Lower Shell u 0.618 0.8652 118.3 102.35 0.749 Plate B6903-1 w 0.952 0 .9862 147.7 145.66 0.973 (Longitud inal) y 2.10 1.2018 141.7 170.30 1.444 X 4.99 1.4020 175.8 246.46 1.965 V 0.297 0.6677 138.0 92 . 14 0.446 Lower Shell u 0.618 0.8652 132. 1 114.29 0.749 Plate B6903-1 w 0.952 0.9862 180.2 177 .72 0.973 (Transver se) y 2.10 1.20 I 8 166.9 200 .58 1.444 X 4.99 1.4020 179.0 250.95 1.965 SUM : 1585.86 11.154 CF 6 6903 .1= I(FF * ~RTNoT) + I(FF 2) = (1585.86) +( 11.154) = 142.2°F V 0.297 0.6677 159.8 106.70 0.446 Surveillance u 0.618 0.8652 164.9 142.67 0.749 Weld Metal w 0.952 0.9862 186.3 183.73 0.973 (Heat #305424) y 2.10 1.2018 178.5 214.52 1.444 X 4 .99 1.4020 237.8 333 .38 1.965 SUM : 981.0 I 5.577 CF Surv . Weld = I(FF * ~RT Nor) + I(FF 2) = (981.01) + (5.577) = 175.9°F Notes:

(a) f = tluence

• r (b) FF = fluen ~e factor = 0*28 " 0 IO' log O (c) L'>RT NDT values are the measured 30 ft-lb shift values.

WCAP-18 102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 D-8 Table D-6 Beaver Valley Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line Capsule CF Measured Predicted Residual <17°F Fluence Material Capsule (Slopebest-fit) FF ARTNDT ARTNDT (x 10 19 n/cm2, ARTNDT (Base Metal)

(OF) (OF) (OF) (OF) <28°F (Weld)

E > 1.0 MeV)

V 142.2 0.297 0.6677 127.9 94 .9 33.0 No Lower Shell Plate u 142.2 0.618 0.8652 118.3 123 .0 4.7 Yes 86903-1 w 142.2 0.952 0.9862 147.7 140.2 7.5 Yes (Longitudinal )

y 142.2 2.10 1.2018 141.7 170.9 29.2 No X 142.2 4.99 1.4020 175.8 199.4 23.6 No V 142.2 0.297 0.6677 138.0 94.9 43.1 No Lower Shell Plate u 142.2 0.618 0.8652 132.1 123.0 9.1 Yes 86903-1 w 142.2 0.952 0.9862 180.2 140.2 40.0 No (Transverse) y 142.2 2.10 1.2018 166.9 170.9 4.0 Yes X 142.2 4.99 1.4020 179.0 199.4 20.4 No V 175.9 0.297 0.6677 159.8 117.4 42.4 No Surveillance Weld u 175 .9 0.618 0.8652 164.9 152.2 12.7 Yes Material w 175 .9 0.952 0.9862 186 .3 173.5 12.8 Yes (Heat # 305424) y 175 .9 2.10 1.2018 178.5 211.4 32 .9 No X 175.9 4.99 1.4020 237 .8 246.6 8.8 Yes From a statistical point of view, +/- I cr would be expected to encompass 68% of the data. Table D-6 indicates that only four of the ten surveillance data points fall inside the +/- I cr of I 7°F scatter band for surveillance base metals; therefore, the plate data is deemed " non-credible " per the third criterion.

Table D-6 indicates that only three of the five surveillance data points fall inside the +/- 1cr of 28°F scatter band for surveillance weld materials ; therefore, the surveillance weld data is deemed " non-credible " per the third criterion.

D.3 CONCLUSION In conclusion, the combined surveillance data from St. Lucie Unit 1 and Millstone Unit 2 for weld Heat#

90136 may be applied to the Beaver Valley Unit I reactor vessel weld. The Position 2.1 chemistry factor calculation, as applicable to the Beaver Valley Unit I reactor vessel weld, is contained in Section 5. This Position 2.1 CF value could be used with a reduced margin term in the ART calculations contained in Section 7 and the RT PTs calculations contained in Appendix E. However, consistent with the discussion in Section D. I of this Appendix, the ART and RT PTs values calculated with the Position 2.1 CF value for weld Heat# 90136 will utilize a full margin term for conservatism . Additionally, the Beaver Valley Unit 1 surveillance plate and weld data remain non-credible , as concluded in WCAP-1789 6-NP [Ref. D-2].

WCAP-18102 -NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 D-9 D.4 REFERENCES D-1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U. S.

Nuclear Regulatory Commission , May 1988.

D-2 Westinghous e Report WCAP-1789 6-NP, Revision 0, " Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," September 2014.

D-3 Westinghouse Report WCAP-1601 2, Revision 0, "Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003.

D-4 Code of Federal Regulations, 10 CFR 50.61 , "Fracture Toughness Requirement s for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243 , dated December 19, 1995, effective January 18, 1996.

D-5 ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," ASTM , July 1982.

D-6 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/lndust ry Workshop on RPV Integrity Issues, February 12, 1998.

[ADAMS Accession Number MLJ 10070570}

WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Propri etary Class 3 E-1 APPENDIX E PRESSURIZED THERM AL SHOCK EVALUATION E.1 PRESSURIZED THERMAL SHOCK CALCULATIONS Pressurize d Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coo lant accident (LOCA) or steam line break. Such transients may challenge the integrity of the reactor pressure vessel (RPV) under the following conditions : severe overcoolin g of the inside surface of the vessel wall followed by high repressuriz ation; significant degradatio n of vessel material toughness caused by radiation embrittlem ent; and the presence of a critical-size defect anywhere within the vessel wall.

In 1985, the U.S. NRC issued a formal ruling on PTS (10 CFR 50.61 [Ref. E-1]) that established screening criteria on PWR vessel embrittlem ent, as measured by the maximum reference nil-ductilit y

transition temperatur e in the limiting beltline componen t at the end of license, termed RT PTS*

RT PTS screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumfere ntial weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlem ent in accordance with the criteria through the end of license. The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlem ent. These revisions make the procedure for calculating the reference temperatur e for pressurized thermal shock (RT PTs) values consistent with the methods given in Regulatory Guide 1.99, Revision 2 [Ref E-2].

These accepted methods were used with the clad/base metal interface fluence values of Section 2 to calculate the following RTPTS values for the Beaver Valley Unit 1 RPV materials at 50 EFPY (EOLE).

The EOLE RT PTS calculation s are summarize d in Table E-1. The following changes and updates to the analysis of record for PTS at Beaver Valley Unit I, WCAP-15571 Supplemen t 1, Revision 2 [Ref.

E-3] ,

have been incorporat ed into the calculation s contained in Table E-1 of this letter report.

1. Incorporat ion of the Capsule X results as documente d in WCAP-17 896-NP, Revision O [Ref. E-4], updated reactor vessel fluence values (See Section 2), surveillanc e capsule irradiated material testing results for Lower Shell Plate 86903-1 and Intermedia te Shell Longitudin al Welds I 9-714 A&B (Heat# 305424) (See Sections 4 and 5), and revised credibility conclusion s (See Appendix D).

2. Incorporation of sister plant surveillanc e capsule test results for weld Heat # 90136 from the Millstone Unit 2 reactor vessel surveillanc e capsule program (See Sections 4, 5 and Appendix D).

Due to the uncertainty in the incorporat ion of the surveillanc e data from Millstone Unit 2 (two wire heats were used in the Millstone 2 surveillanc e weld, with some specimens being Heat #

90136 and others from another weld wire [Heat # 10137]), a full-margi n term was used for this material in the RT rTs calculation s contained in Table I, even though the revised credibility analysis confirmed that Heat# 90136 remained credible.

3. Incorporation of revised initial reference nil-ductility transition temperatur e (RT NDT(U)) values for the four Beaver Valley Unit I reactor vessel plate materials as documente d in Westingho use Letter MCOE-L TR-15 NP, Revision I [Ref. E-5] (See Section 3). It was also concluded that the upper shell forging material for the Beaver Valley Unit I reactor vessel has an appropriat e RT NDT(U) value even though Branch Technical Position (BTP) 5-3 , Paragraph B 1.1 (3) [Ref. E-6]

WCAP-181 02-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 E-2 methodology for SA-508, Class 2 forging material must be used , due to lack of clear definition of Charpy V-notch orientation. The initial RT NDT value of this material remains drop-weight limited due to the excellent Charpy V-notch test results, as documented in its Certified Material Test Report (CMTR) (See Section 3).

4. Utilization of BWRVlP-17 3-A [Ref. E-7] to redefine the initial RTNoTvalues of the six Beaver Valley Unit I nozzle forging materials (See Section 3 and Appendix B).

WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 E-3

-I

,- Table E-1 RT PTs Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY iii" ro Reactor Vessel Material and ID CFC*> EOLE FluenceC*>

8 I Heat Number RTNDT(U) (a) ARTNDT G,1(b) Margin RTPTS (c)

a. Number (OF) (n/cm2, E > 1.0 MeV) FF (OF) {°F)

Gu (OF) (OF) (OF) (OF) i

(/) Reactor Vessel Beltline Materials

~ Intermediate Shell Plate 86607-1 C4381-l 100 .5 5.88 X 10 19

~ 1.4330 26.8 144.0 0 17 34.0 204.8 Ql Intermediate Shell Plate B6607-2 C438 1-2 100.5 5.88 X 10 19 "O

"O 1.4330 53.6 144.0 0 17 34.0 231.6 0 Lower Shell Plate B6903-1 C6317-1 147.2 5.89 X 10 19 (l) 1.4333 13. 1 211.0 0 17 34.0 258 .1 0.. Using Beaver Valley Unit I 0

, C63 17-l 142.2 5.89 10 19 1.4333 13 . 1 203 .8 l'v surveillance data X 0 17 34.0 250.9

i

-.J Lower Shell Plate 87203-2 C6293-2 98.7 5.89 10 19 1.4333

i X 0.4 141.5 0 17 34.0 175.9 0

lntennedi ate to Lower Shell Girth co

~

co 90136 124.3 5.88 10 19 1.4330 -56 178 . 1 17 28 w

Weld I 1-714 X 65.5 187 .6 a,

i-.:> Using St. Lucie Unit I and 0

90136 74.6 5.88 10 19 1.4330 -56 106.9

)> Millstone Unit 2 surveillance data X 17 28 65 .5 116 .4 s::

Intermediate Shell Longitudinal

-I

,- 305424 191.7 l.13xl0 19 1.0341 -56 198.2 17 iii" Welds19-714 A&B 28 65 .5 207.8

(/)

iii Using Beaver Valley Unit I ro 305424 186.5 1.13 10 19 1.0341 3 surveillance data X -56 192.9 17 28 65 .5 202.4 (l)

?.

Lower Shell Longitudinal Welds Ql 305414 210.5 1.14 10 19 1.0366 -56 218.2

(/)20-714 A&8 X 17 28 65.5 227.7 Ql 0..

0..

(l)

Using Fort Calhoun surveillance 0.. 305414 216.9 1.14 10 19 1.0366 -56 224.8 17 rr data X 28 65.5 234.4 SC (l) Reactor Vessel Extended Beltline Materials

"~

o Upper Shell Forging B6604 Upper Shell to Intermediate Shell 123V339V A I 84.2 0.718 X 10 19 0.9071 40 76 .4 0 17 34.0 150.4 m 305414 Girth Weld 10-714 210.5 0.718 X 10 19 0.9071 -56 190.9 17 28 65.5 (3951 & 3958) 200.5

(/)

ro

(/)

3 Using Fort Calhoun surveillance 305414 C: 216.9 0.718 10 19 0.9071 -56 196.7 17 28 "O

0 data (3951 & 3958)

X 65.5 206.3 vi AOFJ 41.0 0.718 10 19 0.9071 10 37.2 X

17 18.6 50.4 97.6

< Upper Shell to Intermediate Shell FOIJ 41.0 0.7 18 10 19 0.9071 10 37.2 I

~

a: X 17 18.6 50.4 97.6 Girth Weld 10-714 (continued) EODJ 27.0

~ 0.718 10 19 0.9071 10 24.5 I

5*

X 17 12.2 41.9 76.4 HOCJ 27.0 0.718 10 19 0.9071 10 X 24.5 17 12.2 41.9 76.4 WCAP-18 102-NP February 2018 Revision I

Westinghouse Non-Proprietary Class 3 E-4

-1 Table E-1 RT rTs Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials

<ii" at 50 EFPY

'°

{") I Reactor Vessel Material and ID Heat Number CF<*> EOLE Fluence<*> RTNDT(U)

(a)

ARTNDT <f,1 (b) Margin (c)

a. RTrTS 0

Number (OF) (n/cm2, E > 1.0 MeV) FF <fu

(OF) (OF) (OF) (OF) (OF) (OF)

Q)

Inlet Nozzle B6608-1 95443-1 67.0 0.02)0 10 19

(/)

X 0.1773 48 .5 11.9 0 5.9 11.9 72.3

~ Inlet Nozzle B6608-2 95460-1 67.0 0.02(0 )0 19 0.1773 -15.2

~ X 11.9 0 5.9 11.9 8.6 Q)

Inlet Nozzle B6608-3 95712-1 51.0 0 .02)0 10 19 "O

"O X 0.1773 11.4 9.0 0 4.5 9.0 29.5

~ EODJ 27.0 0.0210x 10 19 0.1773 10 4.8 (1)

Q.

17 2.4 34.3 49.1 FOIJ 41.0 0.0210 X 10 19 0 .1773 0

, 10 7.3 17 3.6 34.8 52.0 N HOCJ 27.0 0.02)0 10 19 0.1773 iZl

-.J Inlet Nozzle Welds 1-717B, X 10 4.8 17 2.4 34.3 49. 1 iZl DBIJ 27.0 0.0210 (0 19 0.1773 0 1-717D, 1-7l7F X 10 4.8 17 2.4 34.3 49.1 EOEJ 20.0

~

0) 0.0210 X 10 19 0.1773 10 3.5 17 1.8 34.2

0) 47.7 w ICJJ 41.0 0.02)0 10 19 0.1773 a,

j\:,

X 10 7.3 17 3.6 34.8 52.0 0 JACJ 54.0 0.02)0 )0 19 0.1773

)>

X 10 9.6 17 4.8 35.3 54.9 s: Outlet Nozzle B6605-1 95415-1 95 .3 0.0)6( (0 19 0. 1501 -26 .2 X

14.3 0 7.2 14.3 2.4

-1 Outlet Nozzle B6605-2 95415-2 95 .3 0.0(6] )0 19

,- X 0. 1501 3.3 14.3 0 7.2 14.3 31.9

<ii" Outlet Nozzle B6605-3 95444-1 58.0 0.0161 (0 19

(/)

lii X 0 .1501 JO . I 8.7 0 4.4 8.7 27.5 ro ICJJ 41.0 0.0)6( X )0 19 0.1501 10 6.2 17 3.1 3 34.6 50.7 (1)

!OBJ 27.0 0.016( 10 19 0.1501

l.

Outlet Nozzle Welds 1-717 A, X 10 4.1 17 2.0 34.2 48 .3

JACJ 54.0 0.0(6) )0 19 0.1501 10 Q)

(/)

l-717C, 1-717£ X 8.1 17 4.1 35.0 53.1 Q)

HOCJ 27.0 0.0)61 10 19 0.1501 Q.

Q.

X 10 4.1 17 2.0 34.2 48.3 (1)

EODJ 27.0 0.016) )0 19 0.1501 Q.

cr X 10 4.1 17 2.0 34.2 48.3

'< FOIJ 41.0 0.016] 10 19 0. 1501

. X 10 6.2 17 3.1 34.6 50.7 Notes

(1)

-0

o (a) CF values were taken from Table 5-4, fluence values were taken from Tables

~

m

(/)

I (b) As discussed in Section 4, the surveillance plate and weld Heat # 305414 2-5 , 2-7, and 2-9, and RT NDT(U) values were taken from Table 3-2.

and # 305424 data were deemed non-credible. The surveillan ce weld data deemed credible; however, per Section 4 and Appendix D, a full marg in for Heat # 90136 was ro

(/)

term will be used. Per the guidance of 10 CFR 50.61 [Ref. E-1], the base 3 Position 1.1 and Position 2.1 with non-credible surveillan ce data, and the metal cr,1 = I 7°F for weld metal cr,1 = 28°F for Position 1.1 and 2.1 with non-credible surveillan C:

"O margin term will be used for Heat # 90136, cr,1 = 28°F with credible surveillan ce data. Since a full ce data for Position 2.1 for this weld heat. However, cr need not exceed 0.5*li.RT I

0

, 6 NDT*

(c) The 10 CFR 50.61 [Ref. E- 1) methodology was utilized in the calculation

~ of the PTS values.

~

C:

~

ci" 2.,

WCAP-1 8102-NP February 20 I 8 Revision 1

Westinghouse Non-Propri etary Class 3 E-5 E.2 PRESSURIZED THERMAL SHOCK CONCLUSIONS The Beaver Valley Unit I limiting RTPTs value for base metal or longitudina l weld materials at 50 EFPY is 258.1 °F (see Table E-1 ), which correspond s to Lower Shell Plate B6903-1 (using Position I. 1). The Beaver Valley Unit I limiting RTPTS value for circumfere ntially oriented welds at 50 EFPY is 206.3°F (see Table E-1 ), which correspond s to the Upper Shell to Intermedia te Shell Girth Weld I 0-714 (Heat# 305414, using Position 2.1 ).

Therefore, all of the beltline and extended beltline materials in the Beaver Valley Unit I reactor vessel are below the RT rTs screening criteria of 270°F for base metal and/or longitudina l welds, and 300°F for circumfere ntially oriented welds through EOLE (50 EFPY).

In the PTS analysis of record for Beaver Valley Unit I , WCAP-155 71 Supplemen t I , Revision 2 [Ref. E-3], the limiting reactor vessel plate material , Lower Shell Plate B6903- I, was predicted to exceed the RTrTs screening criteria of 270°F for plates at 39.6 EFPY of plant operation. However, with the reevaluatio n of the Beaver Valley Unit I reactor vessel beltline plate material initial RT NDT values

[Ref.

E-5], along with incorporat ion of the Capsule X results [Ref. E-4], this material, while still the limiting material , is now predicted to remain under the RTrTs screening limit through EOLE. With considerat ion of the revised initial RT NDT value for Lower Shell Plate B6903-1 , along with incorporat ion of the Capsule X results, this material is now predicted to remain under the RTrTs screening limit through a minimum of 80 EFPY of plant operation.

WCAP-181 02-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 E-6 E.3 REFERENCES E-1 Code of Federal Regulations, IO CFR 50.61 , " Fracture Toughness Requiremen ts for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243 , dated December 19, 1995, effective January 18, 1996.

E-2 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission , May 1988.

E-3 Westinghous e Report WCAP-15571 Supplement I, Revision 2, "Analysis of Capsule Y from the Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program," September 2011.

E-4 Westinghous e Report WCAP-1789 6-NP, Revision 0, "Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit I Reactor Vessel Radiation Surveillance Program," September 2014.

E-5 Westinghous e Letter MCOE-L TR-15-15-N P, Revision I , " Determinatio n of Unirradiated RT NDT Values of the Four Beaver Valley Unit I Reactor Vessel Beltline Plate Materials Using a Hyperbolic Tangent Curve Fit," dated July 6, 2015.

E-6 NUREG-080 0, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 L WR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirement s," Revision 2, U.S. Nuclear Regulatory Commission , March 2007.

E-7 BWRVIP-1 73-A: BWR Vessel and Int ernals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. I 022835.

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Westinghouse Non-Proprietary Class 3 F-1 APPENDIX F VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIME TRY MEASUREMENTS F.1 NEUTRON DOSIMETRY Comparison s of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn from service to date at Beaver Valley Unit 1 are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculation al and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [Ref. F-1 ]. One of the main purposes for presenting this material is to demonstrate that the overall measuremen ts agree with the calculated and least-squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 2.2 of this report.

F.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the five neutron sensor sets analyzed to date as part of the Beaver Valley Unit I Reactor Vessel Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal , and calculated neutron exposure of each of these dosimetry sets were as follows :

Irradiation Iron Atom Azimuthal Withdrawal Time Fluence (E>I.O MeV) Displacemen ts Capsule ID Location Time IEFPYI ln/cm2I ldpal V 15° End of Cycle I 1.2 2.97E+ 18 4.93E-03 u 25° End of Cycle 4 3.6 6. 18E+ l8 I .00E-02 w 25 ° End of Cycle 6 5.9 9.52E+ l8 l .54E-02 y 25 ° End of Cycle 13 14.3 2.I0E+ l9 3.40E-02 X 15° End of Cycle 22 26.6 4.99E+ l9 8.25E-02 The azimuthal locations included in the above tabulation represent the first octant equivalent azimuthal angle of the geometric center of the respective surveillance capsules.

The passive neutron sensors included in the evaluations of Surveillance Capsules V, U, W, Y, and X are summarized as follows :

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Westinghouse Non-Proprieta ry Class 3 F-2 Reaction of Sensor Material Interest Capsule V Capsule U Capsule W Capsule Y CapsuleX Copper-63 63 Cu(n,a 6 X

) °Co X X X X 54 4 lron-54 Fe(n,p)' Mn X X X X X 58 58 Nickel-58 Ni(n,p) Co X X X X X Uranium-238 238 U(n,f)1 31Cs X X LOST X X Neptunium-2 37 z31Np(n,f) m es X X X X X Cobalt-Alumi num

• s9Co(n ,y)6oCo X X X X X

• The cobalt-alumin um measurement s for this plant include both bare and cadmium-cov ered wire sensors.

Pertinent physical and nuclear characteristi cs of the passive neutron sensors are listed in Table F-1 .

The use of passive monitors such as those listed above does not yield a direct measure of the energy-dependent neutron tluence rate at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-depe ndent neutron tluence rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron tluence rate level incident on the various monitors may be derived from the activation measuremen ts only if the irradiation parameters are well known. In particular, the following variables are of interest:

• the measured specific activity of each monitor,

• the physical characteristi cs of each monitor,

• the operating history of the reactor,

• the energy response of each monitor, and

• the neutron energy spectrum at the monitor location.

Results from the radiometric counting of the neutron sensors from Capsules V, U, W, Y, and X are documented in References F-2 through F-6, respectively . In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-alumi num sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.

The irradiation history of the reactor over the irradiation periods experienced by Capsules V, U, W, Y, and X was based on the monthly power generation of Beaver Valley Unit I from initial reactor criticality through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representatio n for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The irradiation history applicable to Capsules V, U, W, Y, and X is given .in Table F-2.

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Westinghouse Non-Proprietary Class 3 F-3 Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation:

A R=- - - -n -p- - - - - - - - - -

NoFYL - 1 C 1 [1-e- A11] [e-Ai,1 _,]

1;J P ref where:

R Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P,ef (rps/nucleus).

A = Measured specific activity (dps/g).

No Number of target element atoms per gram of sensor material.

F Atom fraction of the target isotope in the target element.

y Number of product atoms produced per reaction.

p Average core power level during irradiation period j (MW).

J P,ef Maximum or reference power level of the reactor (MW).

CJ Calculated ratio of ~(E > 1.0 MeV) during irradiation period j to the time weighted average ~(E > 1.0 Me V) over the entire irradiation period.

Decay constant of the product isotope (I /sec).

tJ Length of irradiation period j (sec).

Decay time following irradiation period j (sec).

n Total number of irradiation periods.

and the summation is carried out over the total number of monthly intervals comprising the irradiation period.

In the equation describing the reaction rate calculation, the ratio [Pj]/[P,ed accounts for month-by-mont h variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology discussed in Section 2.2, accounts for the change in sensor reaction rates caused by variations in fluence rate level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Ci is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those WCAP-18102-N P February 20 I 8 Revision I

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Westinghouse Non-Proprietary Class 3 F-4 employing low-leakage fuel management, the additional Ci term should be employed. The impact of changing fluence rate levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low-leakag e to low-leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another. The fuel-cycle-speci fic neutron fluence rate values along with the computed values for Ci are listed in Table F-3 . These fluence rate values represent the cycle-dependen t results at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.

In performing the dosimetry evaluations for the surveillance capsules, the sensor reaction rates measured at the locations in the capsule holder were indexed to the geometric center of the capsules. This indexing procedure required correcting the measured reaction rates by the application of analytically determined spatial gradients. For the Beaver Valley Unit I surveillance capsules, the gradient correction factors for each sensor reaction were obtained from the reference forward transport calculations and were used in a multiplicative fashion to relate individual measured reaction rates to the corresponding value at the geometric center of the surveillance capsule. The correction factors applied to the Beaver Valley Unit I sensor reaction rates are summarized as follows:

Correction Factor Capsule V Capsule U Capsule W Capsule Y Capsule X 63 Cu(n,a) Radial Gradient 0.956 0.956 0.956 0.956 0.956 54 Fe(n,p)

Radial Gradient 1.050 1.051 1.05) 1.051 1.050 58 Ni(n

,p) Radial Gradient 1.158 1.163 1. 163 1.163 1.158 238 U(n ,f) Radial Gradient 1.000 1.000 N/A 1.000 1.000 237 Np(n,f) Radial Gradient 1.000 1.000 1.000 1.000 1.000 59Co(n,y) Radial Gradient 0.957 0.956 0.956 0.956 0.957 59Co(n,y) Cd Radial Gradient 1. 157 1.155 1.155 1.155 1.157 Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238 U measurements to account for the presence of 235 U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.

Corrections were also made to the 238 U and 237Np sensor reaction rates to account for gamma-ray-ind uced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Beaver Valley Unit I fission sensor reaction rates are summarized as follows:

Correction Factor Capsule V Capsule U Capsule W Capsule Y Capsule X 235 U Impurity/ Pu Build-in 0.873 0.860 N/A 0.806 0.712 23s U(y,f) 0.955 0.960 N/A 0.960 0.956 238 Net U Correction Factor 0.834 0.826 N/A 0.774 0.681 231Np(y,f) 0.982 0.983 0.984 0.984 0.983 WCAP-18102-N P February 2018 Revision I

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Westinghouse Non-Proprietary Class 3 F-5 These factors were applied in a multiplicative fashion to the decay corrected uranium and neptunium fission sensor reaction rates.

Results of the sensor reaction rate determinations for Capsules V, U, W, Y, and X are given in Table F-4a through Table F-4e. In Table F-4a through Table F-4e, the measured specific activities, decay corrected saturated specific activities, and computed reaction rates for each sensor indexed to the radial center of the capsule are listed. The fission sensor reaction rates are listed both with and without the applied corrections for 238 U impurities, plutonium build-in, and gamma-ray-ind uced fission effects.

F.1.2 Least-Squares Evaluation of Sensor Sets Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best estimates for key exposure rate parameters such as q>(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties. For example:

relates a set of measured reaction rates, Ri, to a single neutron spectrum, <pg, through the multigroup dosimeter reaction cross section, O'i g, each with an uncertainty o. The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

For the least-squares evaluation of the Beaver Valley Unit I surveillance capsule dosimetry, the FERRET code [Ref. F- 7] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (q>(E > 1.0 MeV) and dpa) along with associated uncertainties for the five in-vessel capsules analyzed to date.

The application of the least-squares methodology requires the following input:

I. The calculated neutron energy spectrum and associated uncertainties at the measurement location.

2. The measured reaction rates and associated uncertainty for each sensor contained in the multipl.e foil set.

3. The energy-depende nt dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.

For the Beaver Valley Unit I application, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 2.2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section F.1.1.

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Westinghouse Non-Proprietary Class 3 F-6 The dosimetry reaction cross sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [Ref. F-8].

The uncertainties associated with the measured reaction rates, dosimetry cross sections, and calculated neutron spectrum were input to the least-squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, "Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" [Ref. F-9].

The following provides a summary of the uncertainties associated with the least-squares evaluation of the Beaver Valley Unit I surveillance capsule sensor sets.

Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is ensured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.

After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least-squares evaluation:

Reaction Uncertainty 63 Cu(n,cx) 6°Co 5%

54 Fe(n,p)5 4Mn 5%

58 N i( n,p)58Co 5%

238 U(n,f) 137 Cs 10%

231Np(n,f)1 37Cs 10%

59 Co(n,y)6°Co 5%

These uncertainties are given at the I cr level.

Dosimetry Cross-Section Uncertainties The reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library. This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multigroup structure for use in least-squares adjustment applications. These cross sections have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 Me V neutron sources.

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Westinghouse Non-Proprietary C lass 3 F-7 For sensors included in the Beaver Valley Unit I surveillance program, the following uncertainties in the fission spectrum averaged cross sections are provided in the SNLRML documentation package.

Reaction Uncertainty 63 Cu(n,a 6

) °Co 4.08-4.16%

54Fe(n,p)s4Mn 3.05-3 . 11 %

ssNi(n,p )ssco 4.49-4.56%

23s U(n ,t)1 31Cs 0.54-0.64%

231Np(n,f)1 31 Cs I 0.32-10.97%

59 Co( n, y)6°Co 0.79-3 .59%

These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.

Calculated Neutron Spectrum The neutron spectra input to the least-sq uares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:

M gg ' =R 2 + R g

• R g ' *Pgg '

11 where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg* specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:

where, (g-g')2 H = - -~-

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Westinghouse Non-Propri etary Class 3 F-8 The first term in the correlation matrix equation specifies purely random uncertainti es, while the second term describes the short-rang e correlation s over a group range y (0 specifies the strength of the latter term). The value of 8 is 1.0 when g = g', and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the Beaver Valley Unit I calculated spectra was as follows:

Fluence Rate Normaliza tion Uncertaint y (R11 ) 15%

Fluuence Rate Group Uncertaint ies (Rg, Rg,)

(E > 0.0055 MeV) 15%

(0.68 eV < E < 0.0055 MeV) 25%

(E < 0.68 eV) 50%

Short Range Correlatio n (0)

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Fluence Rate Group Correlatio n Range (y)

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 F.1.3 Comparisons of Measurements and Calculations Results of the least-squar es evaluation s of the dosimetry from the Beaver Valley Unit I surveillanc e

capsules withdrawn to date are provided in Tables F-5 and F-6. In Table F-5, measured, calculated

, and best-estim ate values for sensor reaction rates are given for each capsule. Also provided in this tabulation are ratios of the measured reaction rates to both the calculated and least-squar es adjusted reaction rates.

These ratios of MIC and measured- to-best-est imate (M/BE) illustrate the consistenc y of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustmen

t. In Table F-6, compariso n of the calculated and best-estim ate values of neutron fluence rate (E >

1.0 MeV) and iron atom displacem ent rate are tabulated along with the best-estim ate-to-calc ulated (BE/C) ratios observed for each of the capsules.

The data compariso ns provided in Tables F-5 and F-6 show that the adjustmen ts to the calculated spectra are relatively small and well within the assigned uncertainti es for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross sections. Further, these results indicate that the use of the least-squar es evaluation results in a reduction in the uncertainti es associated with the exposure of the surveillanc e capsules. From Section 2.3 of this report, it may be noted that the uncertainty associated with the unadjusted calculation of neutron fluence (E > 1.0 MeV) and iron atom displaceme nts at the surveillanc e capsule locations is specified as 12% at the 1cr level. From Table F-6, it is noted that the correspond ing uncertainti es associated with the least-squar es adjusted exposure parameters have been reduced to 6% for neutron fluence rate (E > 1.0 Me V) and 6-7% for iron atom displaceme nt rate.

Again, the uncertainti es from the least-squar es evaluation are at the I cr level.

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Westinghouse Non-Proprietary Class 3 F-9 Fu1ther comparisons of the measurement results (from Tables F-5 and F-6) with calculations are given in Tables F-7 and F-8. These comparisons are given on two levels. In Table F-7, calculations of individual threshold sensor reaction rates are compared directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table F-8, calculations of fast neutron exposure rates in terms of

~(E > 1.0 MeV) and dpa/s are compared with the best-estimate results obtained from the least-squares eval uation of the capsule dosimetry results. These two levels of comparison yield consistent and similar results with all measurement-to -calculation comparisons falling well within the +/-20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of measured and calculated sensor reaction rates, the MIC comparisons for fast neutron threshold reactions range from 0.95 to 1.12. The overall average MIC ratio for the entire set of Beaver Valley Unit I data is 1.0 I with an associated standard deviation of 8.2%.

In the comparisons of best-estimate and calculated fast neutron exposure parameters, the corresponding BEIC comparisons for the capsule data sets range from 0.91 to 1.03 for neutron fluence rate (E > 1.0 Me V) and from 0.93 to 1.05 for iron atom displacement rate. The overall average BEIC ratios for neutron fluence rate (E > 1.0 MeV) and for the iron atom displacement rate are 0.98 with a standard deviation of 4.5% and 0.99 with a standard deviation of 4.4%, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 2.2 of this report are validated for use in the assessment of the condition of the materials comprising the beltline and extended beltline region of the Beaver Valley Unit I reactor pressure vessel.

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Westinghouse Non- Proprietary C lass 3 F-10 Table F-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors Reaction of Target Atom 90% Response Product Fission Yield Monitor Material Interest Fraction Range (Mevti> Half-life (%)

Copper-63 63 C u (n ,a) 0.6917 5.0 - 11.9 5.272 y n/a 54 lron-54 Fe (n,p) 0.0585 2.2 - 8.5 3 12. 11 d n/a 58 Nickel-58 Ni (n,p) 0.6808 1.7 - 8.4 70.82 d n/a 238 Uranium-238 U (n ,f) 1.0000 1.4 - 7.2 30.07 y 6.02 Neptunium-23 7 231Np (n ,f) 1.0000 0.4 - 4 .8 30.07 y 6 .17 Coba lt-Aluminum 59 Co (n,y) 0.0015 non-threshold 5.272 y n/a Note:

(a) The 90% response range is defined such that, in th e neutro n spectrum characteri stic of the Beaver Valley Unit 1 surve illance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximate ly 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons wi th energies above the upper limi t.

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Westinghouse Non-Proprietary Class 3 F-11 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

May-76 250 Oct-78 0 Mar-81 0 Aug-83 0 Jun-76 101578 Nov-78 0 Apr-81 915873 Sep-83 233976 Jul-76 109048 Dec-78 504540 May-81 1453681 Oct-83 1779388 Aug-76 112413 Jan-79 787965 Jun-81 1874754 Nov-83 1695803 Sep-76 431584 Feb-79 1433040 Jul-81 1080807 Dec-83 1893541 Oct-76 608570 Mar-79 336146 Aug-81 1850579 Jan-84 1552319 Nov-76 199669 Apr-79 0 Sep-81 1765793 Feb-84 1811471 Dec-76 410132 May-79 0 Oct-81 1651329 Mar-84 1263132 Jan-77 189115 Jun-79 0 Nov-81 1782396 Apr-84 1815222 Feb-77 0 Jul-79 0 Dec-81 1148933 May-84 1812753 Mar-77 708513 Aug-79 650167 Jan-82 0 Jun-84 1533814 Apr-77 965179 Sep-79 1419642 Feb-82 0 Jul-84 1737076 May-77 1688148 Oct-79 786544 Mar-82 0 Aug-84 1949986 Jun-77 1049724 Nov-79 692354 Apr-82 0 Sep-84 1674388 Jul-77 1489440 Dec-79 0 May-82 0 Oct-84 658805 Aug-77 1116291 Jan-80 0 Jun-82 0 Nov-84 0 Sep-77 0 Feb-80 0 Jul-82 975423 Dec-84 0 Oct-77 40194 Mar-80 0 Aug-82 1597914 Jan-85 1230666 Nov-77 1030814 Apr-80 0 Sep-82 994760 Feb-85 1495792 Dec-77 1828830 May-80 0 Oct-82 1633910 Mar-85 1567714 Jan-78 1570520 Jun-80 0 Nov-82 1868403 Apr-85 1519174 Feb-78 1672227 Jul-80 0 Dec-82 1810831 May-85 1568263 Mar-78 1903683 Aug-80 0 Jan-83 1734339 Jun-85 1888526 Apr-78 1385543 Sep-80 0 Feb-83 1598708 Jul-85 1815511 May-78 0 Oct-80 0 Mar-83 1939771 Aug-85 1799541 Jun-78 141161 Nov-80 216989 Apr-83 1885670 Sep-85 1627814 Jul-78 1621228 Dec-80 916651 May-83 1732947 Oct-85 1491565 Aug-78 0 Jan-81 1118512 Jun-83 585214 Nov-85 1668503 Sep-78 0 Feb-81 878386 Jul-83 0 Dec-85 1951848 WCAP-18102-NP February 2018 Revision l

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Westinghouse Non-Proprietary Class 3 F-12 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Jan-86 1949567 Jun-88 1577195 Nov-90 1899189 Apr-93 0 Feb-86 1543257 Jul-88 1941 3 12 Dec-90 1437202 May-93 0 Mar-86 1955574 Aug-88 1816437 Jan-91 962970 Jun-93 522652 Apr-86 1776190 Sep-88 1615227 Feb-91 1698024 Jul-93 1967310 May-86 825172 Oct-88 1541992 Mar-91 1920600 Aug-93 1899044 Jun-86 0 Nov-88 1202942 Apr-91 734753 Sep-93 1903751 Jul-86 0 Dec-88 1242361 May-91 0 Oct-93 735881 Aug-86 170868 Jan-89 1392655 Jun-91 0 Nov-93 779517 Sep-86 168936 1 Feb-89 1379326 Jul-91 223412 Dec-93 1935108 Oct-86 1845418 Mar-89 1720567 Aug-91 1900996 Jan-94 1306074 Nov-86 1790031 Apr-89 1457829 Sep-91 1406382 Feb-94 1769531 Dec-86 1955537 May-89 1348415 Oct-91 1348372 Mar-94 1920450 Jan-87 1901768 Jun-89 1542116 Nov-91 60240 Apr-94 1892679 Feb-87 1685908 Jul-89 1946984 Dec-91 1967980 May-94 1064786 Mar-87 1952434 Aug-89 1819779 Jan-92 1968906 Jun-94 39682 Apr-87 1506172 Sep-89 43522 Feb-92 1839523 Jul-94 670158 May-87 20911 Oct-89 0 Mar-92 1966962 Aug-94 1759891 Jun-87 1667777 Nov-89 0 Apr-92 1821532 Sep-94 1902497 Jul-87 1886816 Dec-89 82504 May-92 1878892 Oct-94 1968574 Aug-87 1841589 Jan-90 1717462 Jun-92 1902940 Nov-94 1902781 Sep-87 1752375 Feb-90 1766224 Jul-92 1965639 Dec-94 1724467 Oct-87 1945202 Mar-90 1862208 Aug-92 1604887 Jan-95 72153 Nov-87 1762196 Apr-90 154601 3 Sep-92 1629699 Feb-95 0 Dec-87 469765 May-90 1751901 Oct-92 494631 Mar-95 1225380 Jan-88 0 Jun-90 1871935 Nov-92 1620623 Apr-95 190302 1 Feb-88 0 Jul-90 1053499 Dec-92 1768591 May-95 1967916 Mar-88 1677626 Aug-90 1779924 Jan-93 1769550 Jun-95 1900816 Apr-88 1884007 Sep-90 1851557 Feb-93 1599541 Jul-95 1961257 May-88 1929940 Oct-90 1671278 Mar-93 1318121 Aug-95 1381737 WCAP-18102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-13 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Sep-95 1902415 Feb-98 0 Jul-00 1726698 Dec-02 1997926 Oct-95 1968103 Mar-98 0 Aug-00 1970932 Jan-03 1998176 Nov-95 1854945 Apr-98 0 Sep-00 1907941 Feb-03 1573244 Dec-95 1457253 May-98 0 Oct-00 1974214 Mar-03 242589 Jan-96 1965199 Jun-98 0 Nov-00 1900553 Apr-03 22364 Feb-96 1810977 Jul-98 0 Dec-00 1970715 May-03 1827829 Mar-96 1164783 Aug-98 973181 Jan-0 I 1950143 Jun-03 1867655 Apr-96 0 Sep-98 1909045 Feb-01 1705263 Jul-03 1998416 May-96 1103853 Oct-98 1974419 Mar-01 1971346 Aug-03 1997525 Jun-96 1767857 Nov-98 1906308 Apr-01 1287692 Sep-03 1934204 Jul-96 1965530 Dec-98 1968795 May-01 1970566 Oct-03 2001020 Aug-96 847659 Jan-99 1709974 Jun-0 I 1684701 Nov-03 1817339 Sep-96 1901466 Feb-99 1019933 Jul-0 I 1958104 Dec-03 1998755 Oct-96 1965965 Mar-99 1940265 Aug-01 1929939 Jan-04 1985616 Nov-96 1898890 Apr-99 716664 Sep-01 0 Feb-04 1869377 Dec-96 1959481 May-99 1410478 Oct-01 1328004 Mar-04 1899108 Jan-97 1957638 Jun-99 1906371 Nov-01 1718236 Apr-04 1920338 Feb-97 1744174 Jul-99 1967597 Dec-01 1823296 May-04 1998373 Mar-97 1153954 Aug-99 1958611 Jan-02 1990211 Jun-04 1931522 Apr-97 1023580 Sep-99 1494468 Feb-02 1801795 Jul-04 1998792 May-97 1963402 Oct-99 1951486 Mar-02 1989618 Aug-04 1998284 Jun-97 1705565 Nov-99 1823131 Apr-02 1837779 Sep-04 1933040 Jul-97 11589 Dec-99 1970481 May-02 1998037 Oct-04 1055791 Aug-97 1752251 Jan-00 1968186 Jun-02 1932939 Nov-04 855204 Sep-97 1683909 Feb-00 746198 Jul-02 1997832 Dec-04 1999043 Oct-97 0 Mar-00 0 Aug-02 1997954 Jan-05 1994898 Nov-97 0 Apr-00 1013809 Sep-02 1933446 Feb-05 1805552 Dec-97 0 May-00 1881060 Oct-02 2001352 Mar-05 1998423 Jan-98 468225 Jun-00 1907972 Nov-02 1157615 Apr-05 1931487 WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-14 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

May-05 1911916 Oct-07 418056 Mar-IO 2151255 Aug-12 2154521 Jun-05 1933922 Nov-07 2085894 Apr-IO 1924250 Sep-12 2084237 Jul-05 1998525 Dec-07 2155422 May-IO 2153565 Oct-12 2153783 Aug-05 1998289 Jan-08 2154241 Jun-I 0 2084317 Nov-12 2087475 Sep-05 1932792 Feb-08 1925350 Jul-I 0 2152557 Dec-12 2154578 Oct-05 2000893 Mar-08 2106637 Aug-IO 2153379 Jan-13 2154191 Nov-05 1933689 Apr-08 2084532 Sep-10 2083825 Feb-13 1758987 Dec-05 1998375 May-08 2154156 Oct-10 36743 Mar-13 2151090 Jan-06 1998236 Jun-08 2083935 Nov-IO 1800831 Apr-13 2083989 Feb-06 734160 Jul-08 2150483 Dec-IO 2127093 May-13 2154100 Mar-06 0 Aug-08 2154107 Jan-I I 2154831 Jun-13 2084869 Apr-06 643633 Sep-08 2084264 Feb-I I 1944444 Jul-13 2154058 May-06 1833886 Oct-08 2154327 Mar-I I 2151515 Aug-13 2153987 Jun-06 1934547 Nov-08 2087649 Apr-I I 1752956 Sep-13 1935163 Jul-06 1998137 Dec-08 2154107 May-11 2154209 Oct-13 0 Aug-06 1666197 Jan-09 2154018 Jun-I I 2084557 Nov-13 1397606 Sep-06 1864587 Feb-09 1945788 Jul-I I 2147118 Dec-13 2155010 Oct-06 2062087 Mar-09 2150315 Aug-I I 2154308 Jan-14 528851 Nov-06 1991907 Apr-09 1236557 Sep-I I 2084682 Feb-14 1891737 Dec-06 2058622 May-09 661421 Oct-I I 2152575 Mar-14 2151295 Jan-07 2058939 Jun-09 2085427 Nov-I I 2087006 Apr-14 2084711 Feb-07 1828585 Jul-09 2155082 Dec-I I 2154406 May-14 2153565 Mar-07 2022146 Aug-09 2139233 Jan-12 2154094 Jun-14 2084535 Apr-07 2085903 Sep-09 2084445 Feb-12 2015414 Jul-14 2153613 May-07 21 39605 Oct-09 2154064 Mar-12 2134478 Aug-14 2154036 Jun-07 2085743 Nov-09 2087074 Apr-12 499226 Sep-14 2084732 Jul-07 2155428 Dec-09 2153726 May-12 1389815 Oct-14 1848055 Aug-07 2152836 Jan-10 2153807 Jun-12 2085729 Nov-14 2087509 Sep-07 1514542 Feb-10 1944710 Jul-12 2151237 Dec-14 2153981 WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F- 15 Ta ble F-2 Monthly Thermal Generation during the First 24 Fuel Cycles Thermal Thermal Thermal Thermal Month- Generation Month- Generation Month- Generation Month- Generation Year (MWt-hr) Year (MWt-hr) Year (MWt-hr) Year (MWt-hr)

Jan-15 2 154 143 Feb-15 1945558 Mar-15 2 15 11 45 Apr-15 1245394 May-15 473968 Jun-15 1922472 Jul-1 5 2 155674 Aug- 15 2 155704 Sep-15 2085592 Oct-1 5 215480 1 Nov- 15 2088157 Dec-15 2 155053 Jan-16 2154801 Feb- 16 2015642 Mar-16 2 151 9 16 Apr-16 2084943 May-16 2 154636 Jun-16 2084938 Jul-16 2 153963 Aug-1 6 2149558 Sep-16 1365500 WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/20 18 8:36:20 AM . ( Thi s statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-16 Table F-3 Calculated Fast Neutron (E > 1.0 MeV) Fluence Rate and Ci Factors at the Surveillance Capsule Center, Core Midplane Elevation Cycle q>(E > 1.0 MeV) ln/cm 2-sl Length Fuel Cycle IEFPSI Capsule V Capsule U CapsuleW Capsule Y Capsule X I 3.66E+07 8.IOE+ I0 5.43E+ I0 5.43E+ I0 5.43E+ I0 8.l0E+ I0 2 2.26E+07 - 5.62£+ 10 5.62E+ I0 5.62E+ I0 8.27£+10 3 2.49E+07 - 6.21E+ I0 6.21£+ 10 6.21E+ I0 9.38E+ l0 4 2.91£+07 - 4.73£+ 10 4.73E+ 10 4.73E+ I0 7.14£+ 10 5 3.76E+07 - - 4.56E+ I0 4.56£+ 10 6.94£+ 10 6 3.5IE+07 - - 4.63E+ I0 4.63E+ I0 6.15E+ I0 7 3.95£+07 - - - 4.47£+ 10 6.80E+ I0 8 3.48E+07 - - - 4.70£+ 10 6.98£+ 10 9 4.35E+07 - - - 4.60£+ 10 6.46£+ 10 10 3.77E+07 - - - 3.86E+ I0 4.86E+ I0 11 3.05E+07 - - - 4.06E+ I0 4.78£+ 10 12 3.58£+07 - - - 4.45E+ I 0 5.18£+ 10 13 4.3 IE+07 - - - 4.17E+ I0 5.23£+ 10 14 4. l6E+07 - - - - 4.54£+ 10 15 4. 19E+07 - - - - 4.30E+ I0 16 4.56E+07 - - - - 5.25E+ I0 17 3.89E+07 - - - - 5.02E+ 10 18 4.39E+07 - - - - 5.69E+ I0 19 4.65£+07 - - - - 5.42£+ 10 20 4.27E+07 - - - - 5.82E+ l0 21 4.44E+07 - - - - 5.82E+ I0 22 4.33E+07 - - - - 5.55£+ 10 Time Weighted Average Fluence Rate 8. I0E+ I0 5.46£+ 10 5.12E+ I0 4.66E+ I0 5.94E+ I0 WCAP-18102-N P February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-17 Table F-3 Calculated Fast Neutron (E > 1.0 MeV) Fluence Rate and Ci Factors at the Surveillance Capsule Center, Core Midplane Elevation Cycle cj Length Fuel Cycle IEFPSI Capsule V Capsule U Capsule W Capsule Y Capsule X I 3.66E+07 1.000 0.994 1.060 1.165 1.364 2 2.26E+07 - 1.029 1.097 1.206 1.393 3 2.49E+07 - 1.137 1.212 1.332 1.579 4 2.91E+07 - 0.867 0.924 1.016 1.202 5 3.76E+07 - - 0.891 0.979 1.168 6 3.51E+07 - - 0.904 0.993 1.035 7 3.95E+07 - - - 0.960 1.145 8 3.48E+07 - - - 1.008 1.175 9 4.35E+07 - - - 0.987 1.088 10 3.77E+07 - - - 0.828 0.818 11 3.05E+07 - - - 0.871 0.805 12 3.58E+07 - - - 0.955 0.872 13 4.3 IE+07 - - - 0.895 0.881 14 4.16E+07 - - - - 0.765 15 4. I 9E+07 - - - - 0.724 16 4.56E+07 - - - - 0.883 17 3.89E+07 - - - - 0.845 18 4 .39E+07 - - - - 0.959 19 4.65E+07 - - - - 0.913 20 4.27E+07 - - - - 0.980 21 4.44E+07 - - - - 0.981 22 4.33E+07 - - - - 0.935 Average 1.00 1.00 1.00 1.00 1.00 WCAP-18102-N P February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 F-18 Table F-4a Measured Sensor Activities and Reaction Rates of Surveillance Capsule V Adjusted Reaction Measured Activity(*> Saturated Activity Rate(b>

Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,a) 6°Co Top-Mid 4.24E+04 3.68E+05 5.62E-17 Middle 4.42E+04 3.84E+05 5.86E-17 Bot-Mid 4.28E+04 3.72E+05 5.67E-17 Average 5.72E-17 S4Fe (n,p) s4Mn Top 5.84E+05 3.94E+06 6.25E-l 5 Top-Mid 5.35£+05 3.61E+06 5.73£-15 Middle 5.62E+05 3.79E+06 6.02E-15 Bot-Mid 5.32E+05 3.59E+06 5.70E-15 Bottom 5.37E+05 3.62E+06 5.75E-15 Average 5.89E-15 58 58 Ni (n,p) Co Top-Mid 9.57E+05 5.64E+07 8.07E-15 Middle 9.75E+05 5.74E+07 8.22E-15 Bot-Mid 9.06E+05 5.34E+07 7.64E-15 Average 7.98E-15 238 137 U (n,f) Cs (Cd) Middle l .36E+05 5.39£+06 3.54E-14 235 239 Including U, Pu, and y fission corrections: 2.95E-14 23 7 137 Np (n,f) Cs (Cd) Middle 9.70E+05 3.84E+07 2.45E- l 3 Including y fission corrections: 2.41E-13 59 6 Co (n ,y) °Co Top 7.24£+06 6.30E+07 4.l lE-12 Bottom 7.24E+06 6.30E+07 4.I IE-12 Average 4.11E-12 59 6 Co (n,y) °Co (Cd) Top 2.59E+06 2.72£+07 I.78E-12 Bottom 2.55E+06 2.68E+07 I.75E-l 2 Average 1.76E-12 Notes:

(a) Measured specific activities are indexed to a counting date of September 16, 1980.

(b) Reaction rates are referenced to the Cycle I rated reactor power of2652 MWt.

WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse No n-Proprietary Class 3 F-19 Table F-4b Measured Sensor Activities and Reaction Rates of Surveillance Capsule U Adjusted Reaction Measured Activity<"> Saturated Activity Rate<bl Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,a) 6°Co Top-Mid 9.44E+04 3.0I E+05 4.60E-17 Middle 1.0I E+05 3.22E+05 4.92E-l 7 Bot-Mid 9.30E+04 2.97E+05 4.53E-17 Average 4.68E-17 54 54 Fe (n,p) Mn Top l.21E+06 2.82E+06 4.47E-15 Top-Mid 1.13E+06 2.63E+06 4.17E-15 Middle 1.22E+06 2.84E+06 4.S0E-15 Bot-Mid 1.16E+06 2.70E+06 4.28E-15 Bottom 1.14E+06 2.65E+06 4.21E-15 Average 4.32E-15 58 58 Ni (n,p) Co Top-Mid 4.81E+06 3.98E+07 5.70E-1 5 Middle 5.22E+06 4.32E+07 6.19E-15 Bot-Mid 4.92E+06 4.08E+07 5.84E-15 Average S.91E-15 238 137 U (n,t) Cs (Cd) Middle 2.89E+05 3.8 1E+06 2.50E-14 235 239 Including U, Pu, and y fi ssion corrections: 2.06E-14 23 7 137 Np (n,t) Cs (Cd) Middle 2.14E+06 2. 82E+07 l.80E-13 Including y fission corrections: l.77E-13 59 6 Co (n,y) °Co Top l.17E+07 3.74E+07 2.44E-12 Bottom l.16 E+07 3.71E+07 2.42E-1 2 Average 2.43E-12 59 6 Co (n,y) °Co (Cd) Top 4.27E+06 l.65E+07 l.08E-l 2 Bottom 4.18E+06 1.61 E+07 1.05E-12 Average l.06E-12 Notes:

(a) Measured specific activities are indexed to a counting date of April 3, 1985 .

(b) Reaction rates are referenced to the Cycles 1-4 average rated reactor power of 2652 MWt.

WCAP-18102-NP February 20 I 8 Revision l

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-20 Table F-4c Measured Sensor Activities and Reaction Rates of Surveillance Capsule W Adjusted Reaction Measured Activity<") Saturated Activity Rate<h)

Reaction Location (dps/g) (dps/g) (rps/atom) 63 6 Cu (n ,a) °Co Top-Mid l.14E+05 2.65E+05 4 .03E-17 Middle 1.20E+05 2.78E+05 4.25E-17 Bot-Mid I. I 8E+05 2.74E+05 4. 18E-17 Average 4.ISE-17 S4Fe (n,p) S4Mn Top I.00E+06 2 .55E+06 4.0SE-15 Top-Mid 9.20E+05 2.35E+06 3.73E-15 Top-Mid 8.80E+05 2.25E+06 3.56E-15 Middle 9.44E+05 2.41E+06 3.82E-15 Bot-Mid 8.85E+05 2.26E+06 3.58E-15 Bot-Mid 8.92E+05 2.28E+06 3.61E-15 Botto m 9.19E+05 2.35E+06 3.72E-15 Average 3.73E-15 58 58 Ni (n,p) Co Top-Mid 2.17 E+06 3.47E+07 4.96E-15 Bot-Mid 2.20 E+06 3.51E+07 5.03E-15 Average 5.00E-15 23 7 137 Np (n,f) Cs (Cd) Middl e 2.57E+06 2.14E+07 1.37E-1 3 Including y fission corrections: I.34E-13 59 6 Co (n,y) °Co Top 1.36E+07 3. 16E+07 2.06E-12 Bottom I .40E+07 3 .25E+07 2. 12E-12 Average 2.09E-12 59 6 Co (n ,y) °Co (Cd) Top 4 .93 E+06 1.38E+07 9.03E-13 Bottom 4.99E+06 I .40E+07 9. 14E-1 3 Average 9.08E-13 Notes:

(a) Measured specifi c acti vities are indexed to a counting date of August I 0, 1988.

(b) Reaction rates are referenced to the Cycles 1-6 average rated reactor power of 2652 MWt.

WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprieta ry Class 3 F-21 Table F-4d Measured Sensor Activities and Reaction Rates of Surveillance Capsule Y Adjusted Reaction Measured Activity'*) Saturated Activity Rate(b)

Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n ,a) 6°Co Top-Mid l.56E+05 2.50E+05 3.81E-17 Middle l .67E+05 2.67E+05 4.08E-l 7 Bot-Mid l.56 E+05 2.50E+05 3.81 E-17 Average 3.90E-17 54Fe (n,p) 54Mn Top I .42E+06 2.40£+06 3.81E-15 Top-Mid l.35E+06 2.29£+06 3.63£-15 Middle 1.43E+06 2.42E+06 3.84E-15 Bot-Mid l.37E+06 2 .32E+06 3.68E-15 Bottom l.3 3E+06 2.25E+06 3.57£-15 Average 3.71E-15 58Ni (n,p) 58 Co Top-Mid l.71 E+07 3.47E+07 4.97E-15 Middle l.84E+07 3.74E+07 5.35E- I 5 Bot-Mid l.73E+07 3.51E+07 5.03E-15 Average 5.11 E-15 238 U (n,f) m es (Cd) Middle 7.79£+05 3.03E+06 l.99E-l4 235 239 Including U, Pu, and y fission corrections: 1.54E-14 237 Np (n,f) m es (Cd) Middle 5.39E+06 2. 10£+07 l.34E-13 Including y fission corrections: 1.31E-13 s9Co (n,y) 6oCo Top 1.81 E+07 2.90£+07 I .89E-12 Bottom l.6I E+07 2.58E+07 l.68E- I 2 Average 1.79E-12 59Co (n,y) 6

°Co (Cd) Top 6.78E+06 l.3 IE+07 8.57E-13 Bottom 6.79E+06 l.32E+07 8.58E-13 Average 8.58E-13 Notes:

(a) Measured specific activities are indexed to a counting date of March 24, 2000.

(b) Reaction rates are referenced to the Cycles 1-1 3 average rated reactor power of2652 MWt.

WCAP-18102 -NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-22 Table F-4e Measured Sensor Activities and Reaction Rates of Surveillance Capsule X Adjusted Reaction Measured Activity<"> Saturated Activity Rate<bl Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,a) 6°Co Top-Mid 2.18E+05 2.75£+05 4.19£-17 Middle 2.3 IE+05 2.91£+05 4.44E-17 Bot-Mid 2.20£+05 2.77£+05 4.23£-17 Average 4.28E-17 54Fe (n,p) 54Mn Top 1.53£+06 2.67£+06 4.24£-15 Top-Mid 1.46£+06 2.55£+06 4.04£-15 Middle 1.59£+06 2.77£+06 4.40£-15 Bot-Mid 1.55£+06 2.70£+06 4.29E-15 Bottom 1.64£+06 2.86£+06 4.54E-15 Average 4.30E-15 58Ni (n,p) 58Co Top-Mid 5.73£+06 4.36£+07 6.25£-15 Middle 6.10£+06 4.65£+07 6.65£-15 Bot-Mid 5.87£+06 4.47£+07 6.40£-15 Average 6.43E- 15 238 U (n,f) m es (Cd) Middle 1.87£+06 4.47£+06 2.94£-14 235 239 Including U, Pu, and y fission corrections: 2.00E-14 23 7 Np (n,f) m es (Cd) Middle 1.09£+07 2.61£+07 1.66£-13 Including y fission corrections: 1.63E-13 59 Co (n ,y) 6°Co Top 2.82£+07 3.56£+07 2.32£-12 Bottom 3.01£+07 3.80£+07 2.48£-12 Average 2.40E-12 59Co (n ,y) 6°Co (Cd)

Top 1.84E+07 2.81£+07 1.83£-12 Bottom I .24E+07 l.89E+07 1.23£-12 Average 1.53E-12 Notes:

(a) Measured specific activities are indexed to a counting date of April I, 2014.

(b) Reaction rates are referenced to the Cycles 1-22 average rated reactor power of 2718 MWt.

WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36 :20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-23 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsu le V Reaction Rate lrps/atoml Reaction Measured Ca lculated Best-Estimate M/C M/BE 63 Cu(n ,a) 6°Co 5.72E-I 7 5.50E-17 5.60E-I 7 1.04 1.02 S4Fe(n,p)s4Mn 5.89E-15 6.02E-I 5 5.98E-15 0.98 0.99 58 Ni(n ,p )5 8Co 7.98E-15 8.24E-15 8.15E-15 0.97 0.98 23sU(n,f)1 31Cs (Cd) 2.95E- 14 2.87E- 14 2.88 E-14 1.03 1.02 237Np(n,f)1 31Cs (Cd) 2.41 E- I 3 2.14E-1 3 2.28E-13 1.13 1.05 59Co(n,y)6°Co

4. I IE-1 2 4.47E-12 4.12E- 12 0.92 1.00 59Co(n,y)6°co (Cd) l.76E-12 l.72E-12 I.76E- l 2 1.03 1.00 Note:

See Section F. 1.2 for details describing the Best-Estimate (BE) reaction rates.

Ca psule U Reaction Rate lrps/atoml Reaction Measured Ca lculated Best-Estimate M/C M/BE 63 Cu(n,a)6°Co 4.68E-l 7 4.30E- I 7 4.53E-l 7 1.09 1.03 54 Fe(n,p)5 4Mn 4.32E-15 4.4 IE- 15 4.46E-15 0.98 0.97 58 Ni(n,p)58 Co 5.9 1E-1 5 5.99E- 15 6.05E-15 0.99 0.98 238 U(n,f) 137 Cs (Cd) 2.06 E- 14 I.99E- I 4 2.05E- 14 1.04 1.01 237Np(n,f) 137 Cs (Cd) I.77E- I 3 I.40E- I 3 1.6 I E-1 3 1.27 1.10 59 Co(n,y)6°Co 2.43E-1 2 2.59E-1 2 2.44E- 12 0.94 1.00 59Co(n,y)6°Co (Cd)

I .06 E-1 2 1.0I E- 12 l .06 E- 12 1.06 1.00 Note:

See Section F. 1.2 for details describing the Best-Estimate (BE) reaction rates.

WCAP-18102-NP February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary C lass 3 F-24 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsule W Reaction Rate lrps/atoml Reaction Measured Calculated Best-Estimate MIC M/BE 63 Cu(n,a)6°Co 4.15E-17 4.09E-l 7 4.00E-17 1.02 1.04 54Fe(n,p)54Mn 3.73E-15 4.16E-l 5 3.81 E-15 0 .90 0.98 58 Ni(n,p 8

)5 Co 5.00E-15 5.64E-l 5 5.14E-15 0.89 0.97 z31N p(n,f) 131Cs (Cd) I .34E-1 3 1.31 E-1 3 l.27E- I 3 1.03 1.05 59 Co(n,y)6°co 2.09E-1 2 2.42E- 12 2. I0 E- 12 0 .86 1.00 59 Co(n ,y)6°co (Cd) 9.08E-13 9.41 E-13 9.07E-13 0.97 1.00 Note:

See Section F. 1.2 for deta ils describing the Best-Estimate (BE) reaction rates.

Capsule Y Reaction Rate !rps/atom]

Reaction Measured Calculated Best-Estimate MIC M/BE 63 Cu(n,a)6°co 3 .90E-17 3.78E-17 3.84E-17 1.03 1.02 54 Fe(n,p )5 4Mn 3.7 1E-15 3.8 IE-1 5 3.75E-15 0 .97 0.99 58 N i( n,p 8

)5 Co 5.11 E-15 5.17E-15 5.I0E-15 0 .99 1.00 z3s U( n,f) 1J1Cs (Cd) l .54E- 14 J.71E-l 4 1.67E-14 0.90 0 .92 z31N p(n,f)1 31Cs (Cd) 1.3 I E- 13 1.19E-1 3 1.24E- I 3 I. 11 1.06 59Co(n,y) 6

°Co l.79E- 12 2. 19E-12 l.80 E- 12 0 .82 0.99 59 Co(n,y)6°Co (Cd) 8.58E-1 3 8.51E-13 8.54E-13 1.01 1.00 Note:

See Section F. 1.2 fo r details descri bing the Best-Estimate (BE) reaction rates.

WCAP-18102-N P February 2018 Revision I

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-25 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsule X Reaction Rate lrps/atoml Reaction Measured Calculated Best-Estimate MIC M/BE 63 Cu(n,a) 6°Co 4.28E-17 4.34E-17 4.28E-l 7 0.99 1.00 54Fe(n,p)54Mn 4.3 0E-15 4 .56E-l 5 4.45 E- I 5 0.94 0.97 58Ni(n,p)58 Co 6.43E-15 6.22E-l 5 6.2 1E-15 1.03 1.03 238 U( 37 n,tY Cs (Cd) 2.00E- 14 2.13E-14 2.I0 E-14 0.94 0.95 237Np(n,t) 137 Cs (Cd) l.63E- I 3 1.55E- 13 l.59E-13 1.05 1.03 59 Co(n,y) 6°Co 2.40E- 12 3. 18E- 12 2.45E-12 0.76 0.98 59Co(n,y)6°co (Cd) I .53E-1 2 l.22E- I 2 I .50E-12 1.26 1.02 Note:

See Section F. 1.2 for details describing the Best-Estimate (BE) reaction rates.

WCAP-18102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( Thi s statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 F-26 Table F-6 Comparison of Calculated and Best-Estimate Exposure Rates at the Surveillance Capsule Center q>( E > 1.0 MeV) ln/cm 2-sl Calculated Best-Estimate Best-Estimate BE/C Capsule ID Uncertainty (lcr)

V 8. 12£+ 10 8. 19£+ 10 6 1.00 u 5.47E+ I0 5.68E+ I0 6 1.03 w 5.13E+ I0 4.69E+ IO 6 0.91 y 4.67 E+ I0 4.5 8E+ I0 6 0.98 X 5.95E+ I0 5.89E+ I0 6 0.99 Note:

See Section F. 1.2 for details describing the Best-Estimate (B E) exposure rates.

Iron Atom Displacement Rate ldpa/sl Calculated Best-Estimate Best-Estimate BE/C Capsule ID Uncertainty (lcr)

V 1.33E-I0 l.35E- I0 7 1.01 u 8.70E-II 9.17E-1 I 7 1.05 w 8.16E-I I 7.60E-II 7 0.93 y 7.42E-11 7.33E- II 6 0.98 X 9.67E- I I 9.60E-I I 7 0.99 Note:

See Section F. 1.2 for detai ls describing the Best-Estimate (BE) exposure rates.

WCAP-18102-NP February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM. ( This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary C lass 3 F-27 Table F- 7 Comparison of Measured/Calculated (MIC) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions M/C Rations Capsule Capsule Ca psule Capsule Capsule Reaction V u w y Average  % Std Dev X

63 C u(n

,a) 6°Co 1.04 1.09 1.02 1.03 0.99 1.03 3.5 54Fe(n,p)s4 Mn 0.98 0 .98 0.90 0.97 0.94 0.95 3.6 58 58 N i(n ,p) Co 0.97 0.99 0.89 0 .99 1.03 0.97 5.3 2JsU( n,f) 131Cs (Cd) 1.03 1.04 - 0.90 0 .94 0.98 7 .0 2J1N p(n ,f)1 31Cs (Cd) 1.13 1.27 1.03 I.I I 1.05 1.12 8.4 Note:

The overall average M/C ratio for the set of24 sensor measurements is I .01 with an associated standard deviation of 8.2%.

Table F-8 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios BE/C Ratio Capsule ID cp(E > 1.0 MeV) dpa/s V 1.00 1.01 u 1.03 1.05 w 0.91 0 .93 y 0.98 0.98 X 0.99 0.99 Average 0.98 0 .99

% Standard Deviation 4.5 4.4 WCAP-18102-N P February 2018 Revision I

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Westinghouse Non-Proprietary Class 3 F-28 F.2 REFERENCES F-1 U. S. Nuclear Regulatory Commission, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

F-2 WCAP-9860, Revision 0, Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit No. 1 Reactor Vessel Radiation Surveillance Program, January 1981.

F-3 WCAP-10867, Revision 0, Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 1985.

F-4 WCAP-12005, Revision 0, Analysis of Capsule W from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, November 1988.

F-5 WCAP-15571 Supplement I, Revision 2, Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2011.

F-6 WCAP-17896-N P, Revision 0, Analysis of Capsule Xfrom the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2014.

F-7 F. Schmittroth, FERRET Data Analysis Code, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland , WA, September 1979.

F-8 RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross Section Compendium, July 1994.

F-9 ASTM Standard E 944-13, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (!IA), 2013.

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Westinghouse Non-Proprietary Class 3 G-1 APPENDIX G SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE The following surveillance capsule removal schedule (Table G-1) meets the requirements of ASTM E 185-82 [Ref. G-1] as required by 10 CFR 50, Appendix H [Ref. G-2]. Note that it is recommended for future capsule(s) to be removed from the Beaver Valley Unit I reactor vessel.

Table G-1 Surveillance Capsule Withdrawal Schedule Capsule Capsule Lead Withdrawal Capsule Fluence Capsule Status<")

Location Factor<") EFPY(b,c) (n/cm2, E > 1.0 Mevt>

Withdrawn V 165° 1.47 1.2 2.97E+18 (EOC I)

Withdrawn u 65 ° 1.00 3.6 6.18E+ l8 (EOC 4)

Withdrawn w 245 ° 1.05 5.9 9.52E+ 18 (EOC 6) y Withdrawn 295 ° 1.14 14.3 2. I0E+ l9 (EOC 13)

Withdrawn X 285 ° 1.57 26.6 4.99E+ l9 (EOC 22)

S(d) 285 ° ln Reactor 0.74(d) Note (d) 2.58E+ l9(d)

( 45 °/295 °)

T<el 65° (55°) In Reactor 0.94(e) Note (e) 3.28E+ 19<el z <fl 165 ° (305 °) In Reactor 1.20<fl Note (f) 4.18E+ J9<D Notes:

(a) Updated in Section 2: see Table 2-12.

(b) EFPY from plant startup.

(c) Updated in Section 2; see Table 2-11.

(d) Capsule S was moved to the Capsule Y location at the End of Cycle (EOC) 19, and then moved to the Capsule X location at the EOC 22. Reported tluence value and lead factor are accumulated through EOC 24. Capsule S should remain in the reactor. If additional metallurgical data is needed for Beave r Valley Unit I, such as in support of a second license renewal to 80 total years ofoperation, withdrawal and testing of Capsule S should be considered.

(e) Capsule T was moved to the Capsule U location at the EOC I 0. Reported tluence value and lead factor are accumulated through EOC 24. Capsule T should remain in the reactor and continue to accrue irradiation for potential future testing, if needed.

(t) Capsule Z was moved to the original Capsule V location at the EOC I 0. Reported tluence value and lead factor are accumulated through EOC 24. Based on the current information, Capsule Z should be withdrawn after 39 EFPY, which corresponds to the peak vessel fluence at EOLE (50 EFPY), 5.89 x 10 19 n/cm 2 (E > 1.0 MeY).

G.1 REFERENCES G-1 ASTM E 185-82, " Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," ASTM , July 1982.

G-2 Code of Federal Regulations IO CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Federal Register, Volume 60, No. 243 , dated December 19, 1995.

WCAP-18102-N P February 2018 Revision 1

• • • This record was final approved on 2/27/2018 8:36:20 AM . ( This statement was added by the PRIME system upon its validation)