Text

Modified Rt PTS Values and Reactor Vessel Surveillance Capsule Withdrawal Schedule
	ML17284A195
Person / Time
Site:	Beaver Valley
Issue date:	10/06/2017
From:	Bologna R; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-17-292
	Download: ML17284A195 (143)
	v • d • e

FENOC' Beaver Valley Power Station P.O. Box4 RrstEnergy Nuclear Operating Company Shippingport, PA 15077 Richard D. Bologna 724-682-5234 Site Vice President Fax: 724-643-8069 October 6, 2017 10 CFR 50.61(b}(1}

L-17-292 10 CFR 50.61 (c)(3) 10 CFR 50 Appendix H ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Modified RTPTS Values and Reactor Vessel Surveillance Capsule Withdrawal Schedule Pursuant to 10 CFR 50.61(b)(1} and (c)(3}, FirstEnergy Nuclear Operating Company (FENOC) hereby requests approval of modified pressurized thermal shock reference temperature (RTPrs) values for Beaver Valley Power Station, Unit No. 1 (BVPS-1),

reactor vessel beltline and extended beltline region materials. This submittal also requests approval of a modified reactor vessel surveillance capsule withdrawal schedule pursuant to 10 CFR 50, Appendix H, Section 111, paragraph 8.3.

A description of the pressurized thermal shock evaluation, and results, that support the request for approval of RTPTS values is provided in Enclosure A. The updated RTPTS values incorporate Capsule X fluence analysis results, sister plant surveillance capsule test results, and revised unirradiated nil-ductility reference temperature (RTNDT(U>) values for each of the four reactor vessel beltline plate materials.

The FENOC evaluation of the proposed changes to the reactor vessel surveillance capsule withdrawal schedule is provided in Enclosure B.

The report WCAP-18102-NP, Revision 0, 11 Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," is provided in Enclosure C. This report contains the pressurized thermal shock evaluation for BVPS-1 reactor vessel beltline and extended beltline region materials at the end of 50 effective full power years. The BVPS-1 reactor vessel surveillance capsule withdrawal schedule was modified to be consistent with the recommendations included in Appendix G, "Surveillance Capsule Withdrawal Schedule," of the report.

FENOC requests approval of the revised RTPTs values and reactor vessel surveillance capsule withdrawal schedule by October 10, 2018.

Beaver Valley Power Station, Unit No. 1 L-17-292 Page 2 There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at 330-315-6810.

Sincerely,

/&--.--

Richard D. Bologna

Enclosures:

A. Request for Approval of Modified RT PTs Values B. Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule C. WCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation" cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Enclosure A L-17-292 Request for Approval of Modified RT Prs Values (9 Pages Follow)

Request for Approval of Modified RT PTs Values Page 1 of 9 Pursuant to 10 CFR 50.61(b)(1) and (c)(3), FirstEnergy Nuclear Operating Company (FENOC) requests approval of modified pressurized thermal shock reference temperature (RT PTs) values for Beaver Valley Power Station, Unit No. 1 (BVPS-1) reactor vessel beltline and extended beltline region materials presented in Appendix E, "Pressurized Thermal Shock Evaluation," ofWCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation."

Using the prescribed methodology identified in 10 CFR 50.61(c), theWCAP-18102-NP report provides RT PTS values for reactor vessel beltline and extended beltline region materials (with predicted fluence values greater than 1.0 x 10 17 neutrons per square centimeter or n/cm2, at energies greater than 1.0 million electron volts or MeV) at the end of 60 years of operation (that is, 50 effective full power years of operation or EFPY, and end of license extension). The updated RT PTs values incorporate Capsule X fluence analysis results, sister plant surveillance capsule test results, and revised unirradiated nil-ductility reference temperature (RT NDT(U)) values for each of the four reactor vessel beltline plate materials.

Using the original material properties for the reactor vessel plate materials, the RT NDT(U) values for the four beltline plate materials were revised using a hyperbolic tangent curve fit instead of a hand-drawn curve, as previously reported in a FENOC letter dated September 20, 2016 (Accession No. ML16265A047). The hyperbolic tangent curve graphs and associated Charpy V-notch data tables for each beltline plate material are provided on pages 2 through 9. Incorporation of the revised RTNDT(U) values caused a significant change in the 50 EFPY limiting material RT PTS value for BVPS-1.

The previous RT PTs values for 50 EFPY met the 10 CFR 50.61 pressurized thermal shock (PTS) screening criteria for beltline and extended beltline materials, with the exception of the limiting material (lower shell plate 86903-1, heat C6317-1). Previously, it was expected that the PTS screening limit of 270 degrees Fahrenheit for lower shell plate 86903-1 would be reached at 39.6 EFPY. BVPS-1 would not have been able to operate to the end of the license extension period without implementing the requirements of 10 CFR 50.61a, "Alternate fracture toughness requirements for protection against pressurized thermal shock events," or a flux reduction measure to manage PTS.

As a result of incorporating the revised RTNDT(U) values for the four beltline plate materials, the updated BVPS-1 RT PTS values contained inWCAP-18102-NP, Revision 0, meet the 10 CFR 50.61 PTS screening criteria for beltline and extended beltline materials through the license extension period (50 EFPY). The updated limiting RT PTs value for base metal or longitudinal weld materials at 50 EFPY is 258.1 degrees Fahrenheit, which corresponds to lower shell plate 86903-1 (using Position 1.1). The updated limiting RT PTs value for circumferentially oriented welds at 50 EFPY is 206.3 degrees Fahrenheit, which corresponds to the upper shell to intermediate shell girth weld 10-714 (Heat# 305414, using position 2.1). Appendix E ofWCAP-18102-NP documents the updated RT PTS values for 50 EFPY and the changes that were made from the previous analysis of record for BVPS-1 PTS, which were contained in

Request for Approval of Modified RT PTS Values Page 2 of 9 WCAP-15571 Supplement 1, Revision 2, "Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program" (Accession No. ML13151A059).

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Enclosure B L-17-292 Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule (6 Pages Follow)

Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule Page 1 of 6 Table of Contents 1.0

SUMMARY

DESCRIPTION 2.0 REQUIREMENTS

3.0 PROPOSED CHANGE

S

4.0 TECHNICAL EVALUATION

5.0 CONCLUSION

Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule Page 2 of 6 1.0

SUMMARY

DESCRIPTION Pursuant to the surveillance program criteria of 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Section II1.B.3, FENOC is requesting approval of proposed changes to the Beaver Valley Power Station, Unit No. 1, (BVPS-1) reactor vessel material irradiation surveillance capsule withdrawal schedule. The request is also submitted to satisfy BVPS-1 renewed operating license condition 2.H "Capsule Withdrawal Schedule," that states:

For the renewed operating license term, all capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials (ASTM) E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation.

The proposed BVPS-1 reactor vessel surveillance capsule withdrawal schedule is consistent with the recommendations included in Appendix G, "Surveillance Capsule Withdrawal Schedule," of WCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation." The proposed changes to the reactor vessel surveillance capsule withdrawal schedules are consistent with the recommendations specified in ASTM standard E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels,"

and the end-of-life capsule withdrawal requirement in NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report,"Section XI.M31, "Reactor Vessel Surveillance."

A coordinated pressurized water reactor (PWR) vessel surveillance program has been developed and is documented in the Electric Power Research Institute technical report, "Materials Reliability Program: Coordinated PWR Reactor Vessel Surveillance Program (CRVSP) Guidelines (MRP-326)," dated 2011. The purpose of the CRVSP is to increase the fluence levels of future surveillance capsules at withdrawal while maintaining compliance with 10 CFR 50 Appendix H and consistency with the license renewal guidance of NUREG-1801. The CRVSP will help generate high-fluence PWR surveillance data in support of extended life operations.

2.0 REQUIREMENTS 10 CFR 50, Appendix H, requires nuclear power plant licensees to implement reactor vessel surveillance programs to "monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region . . . which result from exposure of these materials to neutron irradiation and the thermal environment." 10 CFR 50, Appendix H, Section II1.B.1 states in part that the design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of the ASTM E 185 that is current on the issue date of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code to which the reactor vessel was purchased.

Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule Page 3 of 6 The rule permits the use of later editions of ASTM E 185, but including only those editions through 1982 (that is, ASTM E 185-82).

10 CFR 50, Appendix H, Section 111.B.3, requires NRC approval of a proposed reactor vessel surveillance capsule withdrawal schedule prior to implementation. NRC Administrative Letter 97-04, "NRC Staff Approval for Changes to 10 CFR 50, Appendix H, Reactor Vessel Surveillance Specimen Withdrawal Schedules," dated September 30, 1997, specifies that changes to reactor vessel surveillance capsule withdrawal schedules that do not conform to ASTM E 185 require approval by the license amendment process, whereas changes that conform to the ASTM standard require only staff verification of such conformance. The proposed change to the BVPS-1 surveillance capsule withdrawal schedule conforms to ASTM E 185-82, and therefore, a license amendment is not required for this proposed change.

3.0 PROPOSED CHANGE

S The current surveillance capsule withdrawal schedule and the proposed modified schedule are provided below.

The current BVPS-1 Updated Final Safety Analysis Report (UFSAR) Table 4.5-3, "Reactor Vessel Material Irradiation Surveillance Schedule," is as follows:

Current (Original) Lead Withdrawal Fluence, f(a)

Capsule Capsule Location Factor(a) EFPY(b) [n/cm2 , E > 1.0 MeV]

V 165 ° 1.61 1.16 2.99 X 10 18 u 65 ° 1.06 3.59 6.04 X 10 18 w 245 ° 1.11 5.89 9.30 X 10 18 y 295 ° 1.2 14.29 2.05 X 10 1 9 X(c) 285 ° 1.72 26.5(c) 5.01 X 10 19(c)

T(d) 65 ° (55 ° ) 0.99 Standby(d) ---

° ° S(e) 295 (45 ) 0.64 Standby(e) ---

Z(f)

° 165 (305 ) ° 1.24 36.6(1) ---

Notes:

a) Actual lead factor and fluence from WCAP-15571, Supplement 1, Revision 2.

b) Effective full power years (EFPY) from plant startup. Changes to this column will require prior NRC approval (except to indicate that capsules have been removed as specified in Section 111.B.3, Appendix H of 10 CFR 50).

c) Capsule X is planned to be withdrawn at the end of Cycle 22, which corresponds to 26.5 EFPY. This capsule will meet the requirements of ASTM E 185-82 for the fifth capsule to be withdrawn for the 40-year end-of-life (EOL).

Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule Page 4 of 6 d) Capsule T was moved to the Capsule U location at the end of Cycle 10. In order to achieve higher fluence data for this capsule, Capsule T should be relocated to the current Capsule Z location when Capsule Z is withdrawn from the vessel [see footnote (f)].

e) Capsule S was moved to the Capsule Y location at the end of Cycle 19. In order to achieve higher fluence data for this capsule, Capsule S should be relocated to the Capsule X location when Capsule X is withdrawn from the vessel at 26.5 EFPY.

f) Capsule Z was moved to the original Capsule V location at the end of Cycle 10.

Based on the current information, Capsule Z should be withdrawn after 36.6 EFPY, which corresponds to the peak vessel fluence at 60-year EOL (50 EFPY),

5.58x10 19 n/ cm2 (E > 1.0 MeV).

The following proposed BVPS-1 surveillance capsule withdrawal schedule is consistent with WCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," and meets the requirements of ASTM E 185-82.

Capsule Capsule Withdrawal Capsule Fluence(c)

Capsule Status(a) Lead Location EFPY(b) (n/cm2 , E > 1.0 MeV)

Factor(a)

Withdrawn V 165 ° 1.47 1.2 2.97 X 1018 (EOG 1) u 65° Withdrawn (EOG 4) 1.00 3.6 6.18 X 1018 Withdrawn w 245° (EOG 6) 1.05 5.9 9.52 X 1018 Withdrawn y 295 ° 1.14 14.3 2.10x10 19 (EOG 13)

Withdrawn X 285 ° 1.57 26.6 4.99 X 10 19 (EOG 22) 285 ° S(d) In Reactor 0.74(d) Note (d) 2.58 X 10 19 (d)

(45 ° /295 ° )

T(e) 65 ° (55 ° ) In Reactor 0.94(e) Note (e) 3.28 X 10 19 (e)

Z(f) 165 ° (305 ° ) In Reactor 1.20(f) Note (f) 4.1 8 X 101 9 (f)

Notes:

(a) Updated in WCAP-18102-NP, Revision O; Table 2-12.

(b) EFPY from plant startup.

(c) Updated in WCAP-18102-NP, Revision O; Table 2-11.

(d) Capsule S was moved to the Capsule Y location at the end of cycle 19, and then moved to the Capsule X location at the end of cycle 22. Reported fluence value and lead factor are accumulated through the end of cycle 24. Capsule S should

Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule Page 5 of 6 remain in the reactor. If additional metallurgical data is needed for Beaver Valley Unit 1, such as in support of a second license renewal to 80 total years of operation, withdrawal and testing of Capsule S should be considered.

(e) Capsule T was moved to the Capsule U location at the end of cycle 10. Reported fluence value and lead factor are accumulated through the end of cycle 24.

Capsule T should remain in the reactor and continue to accrue irradiation for potential future testing, if needed.

(f) Capsule Z was moved to the original Capsule V location at the end of cycle 10.

Reported fluence value and lead factor are accumulated through the end of cycle 24. Based on the current information, Capsule Z should be withdrawn after 39 EFPY, which corresponds to the peak vessel fluence at end-of-license extension (50 EFPY), 5.89 x 10 19 n/cm2 (E > 1.0 MeV).

4.0 TECHNICAL EVALUATION

This request proposes to modify the BVPS-1 reactor vessel surveillance capsule withdrawal schedule to account for updated neutron fluence projections based on the results of the latest surveillance Capsule X evaluation and sister plant surveillance data (St. Lucie Unit 1, Fort Calhoun, and Millstone Unit 2). Calculations were performed for end-of-license extension at 50 EFPY. The surveillance capsules are used to monitor changes in the toughness properties of ferritic materials in the reactor vessel beltline.

The surveillance capsules are located closer to the core than the reactor vessel beltline materials, so that fracture toughness testing can be used to determine the nil-ductility transition temperature of the vessels at a later time in life.

The current BVPS-1 pressure-temperature (P-T) limit curves were developed in accordance with 10 CFR 50, Appendix G, "Fracture Toughness Requirements," through 50 EFPY. These limits are based on the limiting beltline material (that is, lower shell plate 86903-1) adjusted reference temperature (ART) values in WCAP-18102-NP, Revision 0.

The proposed change to the surveillance capsule schedule is based on the requirements of the 1982 Edition of ASTM E 185 as provided in 10 CFR 50, Appendix H. Since the reference temperature nil-ductility transition shift (LlRT Nor) projected at end-of-license extension is greater than 200 °F for a few reactor vessel beltline materials, ASTM E 185-82 requires the withdrawal of a minimum of five capsules. Five of the eight BVPS-1 reactor vessel surveillance capsules (V, U, W, Y, and X) have been withdrawn and tested. Capsule X was withdrawn at the end of Cycle 22, meeting the surveillance capsule withdrawal requirements for the original 40 year EOL that are defined in the 1982 edition of ASTM E 185.

A sixth surveillance capsule (Capsule Z) is recommended to be withdrawn after 39 EFPY, which corresponds to the peak vessel fluence at end-of-license extension (50 EFPY), 5.89 x 10 19 n/cm2 (E > 1.0 MeV). This meets the end-of-life capsule withdrawal requirement contained in NUREG-1801,Section XI.M31.

Evaluation of Proposed Changes to Beaver Valley Power Station, Unit No. 1, Reactor Vessel Surveillance Capsule Withdrawal Schedule Page 6 of 6 Capsule T and Capsule S will remain in the vessel to accrue irradiation and be utilized for 80-year EOL data if determined to be necessary.

5.0 CONCLUSION

The proposed surveillance capsule withdrawal schedule meets the reactor vessel capsule withdrawal schedule criteria in ASTM E 185-82 as required by 10 CFR 50, Appendix H. Additionally, the proposed withdrawal schedule satisfies the end-of-life capsule withdrawal requirement provided in Section XI.M31 of NUREG-1801.

Enclosure C L-17-292 WCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation" (123 Pages Follow)

Westinghouse Non-Proprietary Class 3 WCAP-18102-NP June 2017 Revision 0 Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-18102-NP Revision 0 Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Benjamin E. Mays*

Materials Center of Excellence Alexandria M. Carolan*

Piping Analysis and Fracture Mechanics Alex J. Markivich

Containment and Radiological Analysis June 2017 Reviewers: Elaine M. Ruminski* Approved: David B. Love*, Manager Materials Center of Excellence Materials Center of Excellence Anees Udyawar* Benjamin A. Leber*, Manager Piping Analysis and Fracture Mechanics Piping Analysis and Fracture Mechanics Jesse J. Klingensmith* Laurent P. Houssay*, Manager Nuclear Operations & Radiation Analysis Nuclear Operations & Radiation Analysis

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

Westinghouse Non-Proprietary Class 3 ii RECORD OF REVISION Revision 0: Original Issue WCAP-18102-NP June 2017 Revision 0

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

SUMMARY

.......................................................................................................................... ix INTRODUCTION ........................................................................................................................ 1-1 2 CALCULATED NEUTRON FLUENCE ..................................................................................... 2-1

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 2-l 2.3 CALCULATIONAL UNCERTAINTIES ........................................................................ 2-3 3 FRACTURE TOUGHNESS PROPERTIES ................................................................................. 3-1 4 SURVEILLANCE DATA ............................................................................................................. 4-1 5 CHEMISTRY FACTORS ............................................................................................................. 5-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ................ 6-1 6.1 O VERALL APPROACH ................................................................................................. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ............................................................................................................ 6-1 6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS ........................................... 6-5 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .......................................... 7-1 8 HEATUP AND COOLDOW N PRESSURE-TEMPERATURE LIMIT CURVES....................... 8-1 9 REFERENCES ............................................................................................................................. 9-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (K1t)......................................... A-l APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES ................................ B-1 APPENDIX C OTHER REACTOR COOLANT PRESSURE BOUNDARY FERRITIC COMPONENTS .......................................................................................................................... C-1 APPENDIX D BEAVER VALLEY UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION ........................................................................................................................... D-1 APPENDIX E PRESSURIZED THERMAL SHOCK EVALUATION .................................... E-1 WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 iv APPENDIX F VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ...................................................................F-1 APPENDIXG SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE .........................G-1 WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 V LIST OF TABLES Table 2-1 Pressure Vessel Material Weld Axial Locations............................................................... 2-5 Table 2-2 Reactor Core Thermal Power Level for Beaver Valley Unit 1 ........................................ 2-5 Table 2-3 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface................................................................................................ 2-6 Table 2-3 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface................................................................................................ 2-7 Table 2-4 Calculated Iron Displacements per Atom at the Pressure Vessel Clad/Base Metal Interface

......................................................................................................................................... 2-8 Table 2-5 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Plates............... 2-9 Table 2-6 Calculated Iron Displacements per Atom at the Pressure Vessel Plates ........................ 2-10 Table 2-7 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds ....................................................................... 2-11 Table 2-8 Calculated Iron Displacements per Atom at the Pressure Vessel Outlet Nozzle, Inlet Nozzle, and Circumferential Welds ............................................................................... 2-12 Table 2-9 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Longitudinal Welds ............................................................................................................................. 2-13 Table 2-10 Calculated Iron Displacements per Atom at the Pressure Vessel Longitudinal Welds .. 2-14 Table 2-11 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of the Surveillance Capsules ......................................................................................................................... 2-15 Table 2-12 Summary of Calculated Surveillance Capsule Lead Factors ......................................... 2-16 Table 2-13 Calculational Uncertainties ............................................................................................ 2-17 Table 3-1 Summary of Beaver Valley Unit 1 Reactor Vessel Base Metal Material Initial RT NDT Determination Methodologies ......................................................................................... 3-2 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 Materials ............................................................... 3-3 Table 3-3 Summary of Beaver Valley Unit 1 Replacement Reactor Vesel Closure Head and Vessel Flange Initial RT NDT Values ............................................................................................. 3-4 Table 4-1 Beaver Valley Unit 1 Surveillance Capsule Data.............................................................4-2 Table 4-2 St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data for Weld Heat # 90136

......................................................................................................................................... 4-3 Table 5-1 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Lower Shell Plate B6903-1 Using Surveillance Capsule Data...................................................................... 5-2 Table 5-2 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat # 305424 Using Surveillance Capsule Data.................................................................................... 5-2 WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 vi Table 5-3 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat # 90136 Using Surveillance 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, l/4T and 3/4T Locations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY ...................................... 7-3 Table 7-2 Adjusted Reference Temperature Evaluation for the Beaver Valley Unit 1 Reactor Vessel Beltline Materials through 50 EFPY at the l/4T Location .............................................. 7-4 Table 7-3 Adjusted Reference Temperature 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 LimitingART Values Used in the Generation of the Beaver Valley Unit 1 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 Methodology (w/ K 1c, w/ Flange Notch, and w/o Margins for Instrumentation 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 Methodology (w/ K1c, w/ Flange Notch, and w/o Margins for Instrumentation Errors) ................................................................................ 8-7 TableA-1 K1t Values for Beaver Valley Unit I at 50 EFPY 100° F/hr Heatup Curves (w/ Flange Requirements and w/o Margins for Instrument Errors) ..................................................A-2 TableA-2 Kit Values for Beaver Valley Unit 1 at 50 EFPY 100° F/hr Cooldown Curves (w/ Flange Requirements and w/o Margins for Instrument Errors) ..................................................A-3 Table B-1 ART Calculations for the Beaver Valley Unit 1 Reactor Vessel Nozzle Materials at 50 EFPY............................................................................................................................... B-3 Table B-2 Summary of the Limiting ART Values for the Beaver Valley Unit I Inlet and Outlet Nozzle Materials ............................................................................................................. B-4 Table D-1 Mean Chemical Composition and Operating Temperature for St. Lucie Unit I and Millstone Unit 2 .............................................................................................................. D-4 Table D-2 Operating Temperature Adjustments for the St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data .............................................................................................. D-5 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 ..... D-5 Table D-4 Best-Fit Evaluation for Surveillance Weld Metal Heat# 90136 Using St. Lucie Unit 1 and Millstone Unit 2 Data ..................................................................................................... D-6 Table D-5 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Beaver Valley Unit 1 Surveillance Capsule Data ........................................................................ D-7 Table D-6 Beaver Valley Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line.......... D-8 WCAP-18102-NP June 2017 Revision 0

WestinghouseN on-Proprietary Class 3 vii Table E-1 RT PTs Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY ......

........................................................................................................................................ E-3 Table F-1 Nuclear Parameters Used in the Evaluation ofNeutron Sensors .................................. F-10 Table F-2 Monthly Thermal Generation during the First 24 Fuel Cycles ......................................F-11 Table F-3 Calculated Fast Neutron (E > 1.0 MeV ) Fluence Rate and Cj Factors at the Surveillance Capsule Center, Core Midplane Elevation .................................................................... F-16 Table F-4a Measured Sensor Activities and Reaction Rates of Surveillance Capsule V ................ F-18 Table F-4b Measured Sensor Activities and Reaction Rates of Surveillance Capsule U ................ F-19 Table F-4c Measured Sensor Activities and Reaction Rates of Surveillance Capsule W ............... F-20 Table F-4d Measured Sensor Activities and Reaction Rates of Surveillance Capsule Y ................ F-21 Table F-4e Measured Sensor Activities and Reaction Rates of Surveillance Capsule X ................ F-22 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center ......................................................................................... F-23 Table F-6 Comparison of Calculated and Best-Estimate Exposure Rates at the Surveillance Capsule Center............................................................................................................................ F-26 Table F-7 Comparison of Measured/Calculated (M/C) Sensor Reaction Rate Ratios Including all Fast Neutron Threshold Reactions ................................................................................ F-27 Table F-8 Comparison of Best-Estimate/Calculated (BE/C) Exposure Rate Ratios ..................... F-27 Table G-1 Surveillance Capsule Withdrawal Schedule ................................................................... G-1 WCAP-18102-NP June2017 Revision 0

Westinghouse Non-Proprietary Class 3 Vlll LIST OF FIGURES Figure 2-1 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Midplane; Octant with No Surveillance Capsules ....................................................................................................2-18 Figure 2-2 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Midplane; Octant with Surveillance Capsules ....................................................................................................2-19 Figure 2-3 Beaver Valley Unit 1 r,z Reactor Geometry ................................................................... 2-20 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 Methodology (w/ K1c) ...................................................................................................... 8-3 Figure 8-2 Beaver Valley Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -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 Methodology (w/ K1c) .......................................................................... 8-4 Figure B-1 Comparison of Beaver Valley Unit 1 Beltline P-T Limits to Inlet Nozzle Limits .......... B-6 Figure B-2 Comparison of Beaver Valley Unit 1 Beltline P-T Limits to Outlet Nozzle Limits ....... B-7 WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 IX EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature (P-T) limit curves for normal operation of the Beaver Valley Unit I reactor vessel. The heatup and cooldown P-T limit curves were generated using the limiting Adjusted R eference Temperature (ART) values for Beaver Valley Unit I increased by a small margin for conservatism. The limiting ART values were those of Lower Shell Plate B6903-1 (Position 1.1) at both 1/4 thickness (l /4T) and 3/4 thickness (3/4T) locations. The P-T limit curves were generated using the K 1c methodology detailed in the 1998 Edition through the 2000 Addenda of the ASME (American Society of Mechanical Engineers)

Code,Section XI, Appendix G. The P-T limit curve generation methodology is consistent with the NRC approved methodology documented in WCAP-14040-A, Revision 4.

The P-T limit curves were generated for 50 effective full-power years (EFPY) using heatup rates of 60 and 100 ° F/hr, and cooldown rates of 0, -20, -40, -60, and -100°F/hr. The curves were developed with the flange requirements and without margins for instrumentation errors. They can be found in Figures 8-1 and 8-2.

Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 50 EFPY.

Appendix B contains a P-T limit evaluation of the reactor vessel inlet and outlet nozzles based on a flaw postulated at the inside surface of the reactor vessel nozzle corner. As discussed in Appendix B, the P-T limit curves generated based on the limiting cylindrical beltline material (Lower Shell Plate B6903-1) bound the P-T limit curves for the reactor vessel inlet and outlet nozzles for Beaver Valley Unit 1 at 50 EFPY.

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

Appendix D contains an updated credibility evaluation for Beaver Valley Unit 1 considering all applicable sister plant surveillance program data and updated Beaver Valley Unit 1 surveillance capsule fluence values.

Appendix E contains a Pressurized Thermal Shock (PTS) evaluation for the reactor vessel materials at 50 EFPY for Beaver Valley Unit 1. In the previous Beaver Valley Unit 1 analysis of record, the limiting reactor vessel plate material, Lower Shell Plate B6903-1, was predicted to exceed the RTPTs screening criteria of 270 °F for plates at 39.6 EFPY of plant operation. However, as discussed in Appendix E, this material, while still the limiting material, is now predicted to remain under the RTPTS screening limit through 50 EFPY (end of license extension [EOLE]).

Appendix F contains an evaluation of the neutron dosimetry contained in the Beaver Valley Unit 1 surveillance capsules withdrawn to date.

Appendix G contains an updated surveillance capsule withdrawal schedule for Beaver Valley Unit 1.

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Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION Heatup and cooldown P-T limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced LiRTNDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NOTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F.

RTNDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, L'.iRTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (RTNDT(u))- The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The U.S. Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. 1]. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values (RTNDT(U) + L'.iRTNDT + margins for uncertainties) at the l/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown P-T limit curves documented in this report were generated using the most limiting ART values (increased by a small margin for conservatism) and the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [Ref. 2]. Specifically, the Krc methodology of the 1998 Edition through the 2000 Addenda of ASME Code,Section XI, Appendix G [Ref. 3] was used. The K rc curve is a lower bound static fracture toughness curve obtained from test data gathered from several different heats of pressure vessel steel. The limiting material is indexed to the Krc curve so that allowable stress intensity factors can be obtained for the material as a function of temperature. Allowable operating limits are then determined using the allowable stress intensity factors.

The following statement excludes the fluence method. For the purpose of this plant-specific evaluation, the P-T limit curve generation method of WCAP-14040-A, Revision 4 is identical to the P-T limit curve generation method of WCAP-14040-NP-A, Revision 2 [Ref. 21] with the addition of the allowance for Beaver Valley Unit 1 to also utilize ASME Code Case N-640 and ASME Code Section XI, Appendix G

( 1995 Edition through 1996 Addenda). The fluence method utilized is detailed in Section 2.

The purpose of this report is to present the calculations and the development of the Beaver Valley Unit 1 heatup and cooldown P-T limit curves for 50 EFPY. This report documents the calculated ART values and the development of the P-T limit curves for normal operation. The calculated ART values for 50 EFPY are documented in Section 7 of this report. The fluence projections used in calculation of the ART values are provided in Section 2 of this report.

The P-T limit curves herein were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [Ref. 4] have been incorporated 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. Discussion of the other ferritic RCPB components relative to P-T limits is contained in Appendix C.

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Westinghouse Non-Proprietary Class 3 1-2 Appendix D contains an updated credibility evaluation for surveillance data applicable to Beaver Valley Unit I, and Appendix E contains a Pressurized Thermal Shock evaluation for the Beaver Valley Unit 1 reactor vessel materials at 50 EFPY.

Appendix F contains an evaluation of the neutron dosimetry contained in the Beaver Valley Unit 1 surveillance capsules withdrawn to date.

Appendix G contains an updated surveillance capsule withdrawal schedule for Beaver Valley Unit I.

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Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

A discrete ordinates (SN) transport analysis was performed for the Beaver Valley Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis. An evaluation of the dosimetry sensor sets from the first five surveillance capsules is provided in Appendix F. The dosimetry analysis shows that the +/-20% (1a) acceptance criteria specified in Regulatory Guide 1.190 [Ref. 6] is met. The validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 EFPY.

All of the calculations described in this section and in Appendix F were based on nuclear cross-section data derived from the Evaluated Nuclear Data File (ENDF) database (Specifically, ENDF/B-VI).

Furthermore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 6]. Additionally, the fluence calculations herein were performed in accordance with WCAP-14040-A, Revision 4 [Ref. 2], which is a topical report having an NRC approved method that complies with NRC Regulatory Guide 1.190. The method used for the fluence calculations is the same as that employed during the analysis of Beaver Valley Unit 1 Capsule Y documented in WCAP-15571, Revision 0 [Ref. 22].

2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Beaver Valley Unit 1 reactor vessel, a series of fuel-cycle-specific forward transport calculations were completed using the following three-dimensional fluence rate synthesis technique:

cp(r, z) cp(r, 0, z) = cp(r, 0) x cp(r) where cp(r, 0, z) is the synthesized three-dimensional neutron fluence rate distribution, cp(r, 0) is the transport solution in r,0 geometry, cp(r, z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and cp(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation.

This synthesis procedure was completed for each operating cycle at Beaver Valley Unit 1.

Plan views of the r,0 geometry of the Beaver Valley Unit 1 reactor at the core midplane are shown in Figures 2-1 and 2-2. In each of these figures, a single octant is depicted showing the arrangement of surveillance capsules, where Figure 2-1 shows an octant with no surveillance capsules, and Figure 2-2 shows an octant with surveillance capsules. The maximum exposure of the pressure vessel occurs in octants with no surveillance capsules. In developing these analytical models, nominal design dimensions were employed for the various structural components. Likewise, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full power operating conditions. The coolant densities were treated on a fuel-cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core WCAP-18102-NP June 2017 Revision 0

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 consisted of 189 radial by 7 5 azimuthal intervals. Mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence 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 calculations is shown in Figure 2-3. The model extends radially from the centerline of the reactor core out to a location interior to the neutron shield tank and over an axial span from an elevation approximately five feet below to five feet above the active fuel. As in the case of the r,0 models, nominal design dimensions and full-power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent 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 explicitly included in the model. The r,z geometric mesh description of this reactor model consisted of 189 radial by 223 axial intervals. As in the case of the r,0 calculations, mesh sizes were chosen to ensure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r,z calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 189 radial mesh intervals included in the r,z model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The data utilized for the core power distributions in plant-specific transport analyses included cycle dependent fuel assembly initial enrichments, bumups, and axial power distributions. This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle-averaged neutron fluence rate, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and bumup history of individual fuel assemblies.

From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the DORT discrete ordinates code [Ref. 7] and the BUGLE-96 cross-section library [Ref. 8]. The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P5 Legendre expansion and angular discretization was modeled with an S 16 order of angular quadrature.

Energy- and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.

In Table 2-1, axial locations of the Beaver Valley Unit 1 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 corresponds to the midplane of the active fuel stack.

Cycle-specific calculations were performed for Cycles 1 through 24, with core thermal powers given in Table 2-2.

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Westinghouse Non-Proprietary 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 fluence rate and fluence (E > 1.0 MeV ) and iron displacements 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 terms of the maximum exposure experienced 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.

Calculated fast neutron fluence (E > 1.0 MeV) and dpa for the pressure vessel plates are provided in Tables 2-5 and 2-6, respectively. Calculated fast neutron fluence (E > 1.0 MeV ) and dpa for the pressure vessel circumferential welds are provided in Tables 2-7 and 2-8, while the equivalent data for the longitudinal welds are provided in Tables 2-9 and 2-10, respectively.

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

Projections were based on the bumup 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 consistent with these assumptions.

In Table 2-11, calculated fast neutron fluence (E > 1.0 MeV ) for the surveillance capsule for the Beaver Valley Unit 1 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 calculated neutron exposure of the Beaver Valley Unit 1 reactor pressure vessel materials is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H. B. Robinson power reactor benchmark experiment.

3. An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments.

4. Comparisons of the plant-specific calculations with all available dosimetry results from the Beaver V alley Unit 1 surveillance program.

The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques 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 comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.

The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods' approximations as well as to a lack of WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-4 knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Beaver Valley Unit 1 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment ( comparisons with Beaver Valley Unit 1 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule or pressure vessel neutron exposures.

Table 2-13 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 2. The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Appendix F support these uncertainty assessments for Beaver Valley Unit 1.

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Westinghouse Non-Proprietary Class 3 2-5 Table 2-1 Pressure Vessel Material Weld Axial Locations Axial Location<a)

Material Inches cm Lower Shell to Lower Closure Head Weld -121.10 -307.59 Lower Shell to Intermediate Shell Weld -20.50 -52.07 Intermediate 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 Core Power Cycle (MWt) 1 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.

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Westinghouse Non-Proprietary Class 3 2-6 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 Rate (E > 1.0 MeV)

Operating (n/cm2 -s) Elevation(a)

Cycle Time (cm)

(EFPY) oo 15 ° 30° 45° Maximum I 1.2 5.51E+10 2.53E+l0 l.42E+ 10 9.20E+09 5.51E+l0 -54.00 2 1.9 5.79E+10 2.69E+10 l.54E+10 l.OIE+IO 5.79E+l0 104.00 3 2.7 6.28E+10 2.87E+10 l.58E+IO l.OOE+lO 6.28E+l0 -54.00 4 3.6 4.70E+10 2.21E+l0 l.21E+IO 7.72E+09 4.70E+10 -54.00 5 4.8 4.56E+10 2.15E+10 1.18E+IO 7.79E+09 4.56E+IO -54.00 6 5.9 3.52E+10 l.91E+10 l.19E+10 7.65E+09 3.52E+10 -54.00 7 7.1 4.30E+l0 2.14E+10 1.16E+10 7.67E+09 4.30E+IO -60.00 8 8.2 4.23E+10 2.17E+10 119E+10

. 7.52E+09 4.23E+IO -54.00 9 9.6 3.79E+IO l.99E+10 1.19E+IO 8.25E+09 3.79E+10 -54.00 10 10.8 2.96E+10 l.56E+10 l.07E+10 7.49E+09 2.96E+10 -60.00 11 11.8 2.97E+10 I.SOE+10 1.1 IE+10 8.IOE+09 2.97E+l0 -54.00 12 12.9 3.08E+10 l.63E+10 l.16E+10 7.23E+09 3.08E+l0 -54.00 13 14.3 3.21E+IO 1.65E+l0 1.llE+IO 7.49E+09 3.21E+l0 -58.00 14 15.6 3.25E+10 l.44E+10 8.45E+09 5.86E+09 3.25E+10 -60.00 15 16.9 2.77E+IO l.38E+10 9.34E+09 6.49E+09 2.77E+10 -118.00 16 18.4 3.24E+10 1.63E+10 9.60E+09 6.72E+09 3.24E+10 -54.00 17 19.6 3.18E+10 1.57E+10 8.99E+09 5.99E+09 3.18E+10 -60.00 18 21.0 3.71E+10 1.78E+10 9.87E+09 6.51E+09 3.71E+10 -54.00 19 22.5 3.22E+l0 l.68E+10 l.OIE+lO 7.29E+09 3.22E+IO -54.00 20 23.8 3.78E+IO 1.80E+l0 l.02E+IO 7.19E+09 3.78E+10 50.00 21 25.2 4.02E+l0 l.81E+10 9.69E+09 6.61E+09 4.02E+l0 50.00 22 26.6 3.74E+10 l.73E+10 l.OOE+IO 7.20E+09 3.74E+10 50.00 23 28.0 3.93E+l0 l.81E+10 1.0IE+lO 7.05E+09 3.93E+IO -54.00 24 29.3 3.44E+IO l.69E+10 9.71E+09 6.84E+09 3.44E+10 50.00 Note:

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

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary 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/cm2) Elevation<a)

Cycle T ime (cm)

(EFPY) oo 15 ° 30° 45 ° Maximum 1 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+l7 5.47E+17 3.23E+18 -54.00 3 2.7 4.79E+18 2.21E+18 1.23E+18 7.96E+17 4.79E+18 -54.00 4 3.6 6.16E+18 2.85E+18 1.58E+18 1.02E+18 6.16E+18 -54.00 5 4.8 7.87E+18 3.66E+18 2.03E+18 1.3lE+18 7.87E+18 -54.00 6 5.9 9.10E+18 4.33E+18 2.44E+18 1.58E+18 9.10E+18 -54.00 7 7.1 1.08E+19 5.17E+18 2.90E+18 1.88E+18 1.08E+19 -54.00 8 8.2 l.23E+19 5.92E+18 3.31E+18 2.15E+18 1.23E+19 -54.00 9 9.6 1.39E+19 6.79E+18 3.83E+18 2.50E+18 1.39E+19 -54.00 10 10.8 1.50E+19 7.37E+18 4.24E+18 2.79E+18 1.50E+19 -54.00 11 11.8 1.59E+19 7.83E+18 4.58E+18 3.03E+18 1.59E+19 -54.00 12 12.9 1.70E+19 8.41E+18 4.99E+18 3.29E+18 1.70E+19 -54.00 13 14.3 1.84E+19 9.12E+18 5.47E+18 3.62E+18 1.84E+19 -54.00 14 15.6 1.98E+19 9.72E+18 5.82E+18 3.86E+18 1.98E+19 -54.00 15 16.9 2.09E+19 1.03E+19 6.21E+18 4.13E+18 2.09E+19 -54.00 16 18.4 2.24E+19 1.10E+19 6.65E+18 4.43E+18 2.24E+19 -54.00 17 19.6 2.36E+19 1.16E+19 6.99E+18 4.67E+18 2.36E+19 -54.00 18 21.0 2.53E+19 l.24E+19 7.43E+18 4.95E+18 2.53E+19 -54.00 19 22.5 2.68E+l9 1.32E+19 7.90E+18 5.29E+18 2.68E+19 -54.00 20 23.8 2.84E+19 1.40E+19 8.33E+18 5.60E+18 2.84E+19 -54.00 21 25.2 3.01E+19 1.48E+19 8.75E+18 5.89E+18 3.01E+19 -54.00 22 26.6 3.17E+19 1.55E+19 9.19E+18 6.20E+18 3.17E+19 -54.00 23 28.0 3.34E+19 1.63E+19 9.62E+18 6.50E+18 3.34E+19 -54.00 24 29.3 3.48E+19 1.70E+19 1.00E+19 6.78E+18 3.48E+l 9 -54.00 32.0 3.80E+19 1.85E+19 1.09E+19 7.38E+18 3.80E+19 -54.00 36.0 4.27E+19 2.07E+19 1.21E+19 8.27E+18 4.27E+19 -54.00 40.0 4.73E+19 2.29E+19 1.34E+19 9.15E+18 4.73E+19 -54.00 Future 48.0 5.66E+19 2.72E+19 1.59E+19 1.09E+19 5.66E+19 -54.00 50.0 5.89E+19 2.83E+19 1.65E+19 1.14E+19 5.89E+19 -54.00 60.0 7.06E+19 3.38E+19 1.96E+19 1.36E+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-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-8 Table 2-4 Calculated Iron Displacements per Atom at the Pressure Vessel Clad/Base Metal Interface Cumulative Maximum Iron Displacements per Atom Operating (dpa) Elevation(a)

Cycle Time (cm)

(EFPY) oo 15° 30° 45° Maximum 1 1.2 3.23E-03 l.50E-03 8.29E-04 5.40E-04 3.23E-03 -54.00 2 1.9 5.18E-03 2.42E-03 l.34E-03 8.77E-04 5.18E-03 -54.00 3 2.7 7.69E-03 3.58E-03 l.97E-03 l.28E-03 7.69E-03 -54.00 4 3.6 9.88E-03 4.62E-03 2.53E-03 l.64E-03 9.88E-03 -54.00 5 4.8 l.26E-02 5.93E-03 3.24E-03 2.llE-03 l.26E-02 -54.00 6 5.9 l.46E-02 7.0lE-03 3.91E-03 2.54E-03 l.46E-02 -54.00 7 7.1 l.73E-02 8.37E-03 4.64E-03 3.02E-03 l.73E-02 -54.00 8 8.2 l.97E-02 9.59E-03 5.3lE-03 3.44E-03 l.97E-02 -54.00 9 9.6 2.23E-02 l.l0E-02 6.14E-03 4.02E-03 2.23E-02 -54.00 10 10.8 2.41E-02 l.19E-02 6.78E-03 4.47E-03 2.41E-02 -54.00 11 11.8 2.56E-02 l.27E-02 7.32E-03 4.86E-03 2.56E-02 -54.00 12 12.9 2.73E-02 l.36E-02 7.98E-03 5.28E-03 2.73E-02 -54.00 13 14.3 2.95E-02 l.48E-02 8.75E-03 5.80E-03 2.95E-02 -54.00 14 15.6 3.l7E-02 l.57E-02 9.3lE-03 6.l9E-03 3.l7E-02 -54.00 15 16.9 3.36E-02 1.67E-02 9.93E-03 6.62E-03 3.36E-02 -54.00 16 18.4 3.59E-02 l.79E-02 l.06E-02 7.llE-03 3.59E-02 -54.00 17 19.6 3.79E-02 l.88E-02 l.12E-02 7.48E-03 3.79E-02 -54.00 18 21.0 4.05E-02 2.0lE-02 l.19E-02 7.94E-03 4.05E-02 -54.00 19 22.5 4.29E-02 2.14E-02 l.26E-02 8.48E-03 4.29E-02 -54.00 20 23.8 4.55E-02 2.26E-02 l.33E-02 8.97E-03 4.55E-02 -54.00 21 25.2 4.83E-02 2.39E-02 l.40E-02 9.44E-03 4.83E-02 -54.00 22 26.6 5.09E-02 2.51E-02 l.47E-02 9.94E-03 5.09E-02 -54.00 23 28.0 5.36E-02 2.64E-02 l.54E-02 l.04E-02 5.36E-02 -54.00 24 29.3 5.59E-02 2.75E-02 l.60E-02 l.09E-02 5.59E-02 -54.00 32.0 6.09E-02 2.99E-02 l.74E-02 l.l8E-02 6.09E-02 -54.00 36.0 6.84E-02 3.35E-02 1.94E-02 l.33E-02 6.84E-02 -54.00 40.0 7.59E-02 3.70E-02 2.14E-02 l.47E-02 7.59E-02 -54.00 Future 48.0 9.08E-02 4.41E-02 2.54E-02 1.75E-02 9.08E-02 -54.00 50.0 9.45E-02 4.59E-02 2.64E-02 l.82E-02 9.45E-02 -54.00 60.0 l.l3E-01 5.47E-02 3.14E-02 2.17E-02 l.13E-0l -54.00 Note:

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

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary 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 (E> 1.0 MeV)

Operating (n/cm2)

Cycle Time (EFPY) Upper Intermediate Lower Shell Shell Shell 1 1.2 l.79E+l7 2.0IE+l8 2.02E+l8 2 1.9 3.87E+l7 3.22E+l8 3.23E+18 3 2.7 5.48E+l7 4.78E+l8 4.79E+18 4 3.6 6.83E+17 6.14E+l8 6.16E+18 5 4.8 8.81E+17 7.85E+l8 7.87E+18 6 5.9 9.89E+17 9.09E+l8 9.10E+l8 7 7.1 l.16E+l8 l.08E+l9 l.08E+l9 8 8.2 l.28E+l8 l.22E+l9 l.23E+19 9 9.6 l.44E+l8 l.39E+19 l.39E+19 10 10.8 l.54E+l8 l.50E+l9 1.50E+19 11 11.8 l.63E+l8 l.59E+l9 l.59E+l9 12 12.9 l.73E+l8 l.70E+l9 l.70E+l9 13 14.3 l.86E+l8 l.84E+19 l.84E+19 14 15.6 2.06E+18 l.97E+19 l.98E+l9 15 16.9 2.24E+18 2.09E+19 2.09E+19 16 18.4 2.45E+18 2.24E+l9 2.24E+19 17 19.6 2.64E+l8 2.36E+l9 2.36E+l9 18 21.0 2.84E+18 2.52E+l9 2.53E+19 19 22.5 3.04E+18 2.67E+l9 2.68E+19 20 23.8 3.24E+18 2.83E+19 2.84E+19 21 25.2 3.47E+l8 3.01E+l9 3.01E+l9 22 26.6 3.67E+18 3.17E+19 3.17E+19 23 28.0 3.88E+18 3.34E+l9 3.34E+l9 24 29.3 4.07E+18 3.48E+19 3.48E+l9 32.0 4.48E+18 3.79E+l9 3.80E+19 36.0 5.08E+18 4.26E+19 4.27E+19 40.0 5.68E+18 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+l9 60.0 8.68E+18 7.04E+l9 7.06E+l9 WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary 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.l7E-03 5.l8E-03 3 2.7 9.llE-04 7.68E-03 7.69E-03 4 3.6 l.14E-03 9.86E-03 9.88E-03 5 4.8 1.47E-03 l.26E-02 l.26E-02 6 5.9 l.65E-03 1.46E-02 1.46E-02 7 7.1 l.92E-03 l.73E-02 l.73E-02 8 8.2 2.14E-03 l.96E-02 l.97E-02 9 9.6 2.41E-03 2.23E-02 2.23E-02 10 10.8 2.57E-03 2.41E-02 2.41E-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.l IE-03 2.95E-02 2.95E-02 14 15.6 3.43E-03 3.16E-02 3.l7E-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 4.0SE-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 7.45E-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 l.14E-02 9.06E-02 9.08E-02 50.0 l.19E-02 9.43E-02 9.45E-02 60.0 l.44E-02 l.13E-0l 1.13E-0l WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary 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 Maximum Fast Neutron Fluence (E > 1.0 MeV)

Cumulative (n/cm2)

Operating Lower Shell Lower Shell Cycle Intermediate Time Outlet Inlet to to Lower (EFPY) Shell to Nozzle(a) Nozzle<a> Intermediate Closure Upper Shell Shell Head 1 1.2 3.29E+l5 4.36E+15 l.79E+l7 2.01E+l8 4.53E+l4 2 1.9 6.99E+l5 9.29E+l5 3.87E+17 3.22E+l8 7.35E+l4 3 2.7 9.78E+l5 l.30E+l6 5.48E+l7 4.78E+l8 l.13E+l5 4 3.6 l.24E+16 l.65E+16 6.83E+l7 6.14E+l8 l.47E+l5 5 4.8 l.64E+16 2.l7E+l6 8.81E+l7 7.85E+l8 l.86E+l5 6 5.9 l.85E+l6 2.45E+l6 9.89E+l7 9.09E+l8 2.16E+15 7 7.1 2.18E+l6 2.89E+l6 l.16E+l8 l.08E+l9 2.60E+l5 8 8.2 2.44E+l6 3.23E+l6 l.28E+l8 l.22E+l9 2.95E+l5 9 9.6 2.76E+l6 3.64E+l6 l.44E+l8 l.39E+l9 3.35E+ 15 10 10.8 2.98E+l6 3.93E+16 l.54E+l8 l.50E+l9 3.63E+l5 11 11.8 3. l7E+l6 4.18E+l6 l.63E+l8 1.59E+19 3.85E+l5 12 12.9 3.39E+l6 4.47E+l6 l.73E+l8 l.70E+l9 4.11E+l5 13 14.3 3.68E+l6 4.84E+l6 l.86E+l8 l.84E+l9 4.45E+l5 14 15.6 4.15E+l6 5.46E+l6 2.06E+l8 l.97E+l9 4.84E+l5 15 16.9 4.60E+l6 6.05E+l6 2.24E+l8 2.09E+l9 5.19E+l5 16 18.4 5.12E+16 6.73E+l6 2.45E+18 2.24E+l9 5.61E+ 15 17 19.6 5.59E+16 7.33E+16 2.64E+l8 2.36E+l9 5.98E+l5 18 21.0 6.04E+l6 7.92E+l6 2.84E+l8 2.52E+l9 6.41E+l5 19 22.5 6.51E+l6 8.53E+l6 3.04E+l8 2.67E+l9 6.80E+l5 20 23.8 6.97E+l6 9.14E+l6 3.24E+l8 2.83E+l9 7.21E+l5 21 25.2 7.49E+l6 9.80E+l6 3.47E+l8 3.01E+l9 7.67E+l5 22 26.6 7.96E+l6 l.04E+l7 3.67E+18 3.17E+l9 8.08E+l5 23 28.0 8.44E+16 l.10E+l7 3.88E+l8 3.34E+l9 8.52E+l5 24 29.3 8.89E+l6 l.16E+17 4.07E+l8 3.48E+l9 8.91E+l5 32.0 9.83E+16 l.29E+l7 4.48E+l8 3.79E+l9 9.74E+l5 36.0 l.12E+l 7 l.47E+l7 5.08E+l8 4.26E+l9 l.10E+l6 40.0 l.26E+l7 l.65E+l7 5.68E+l8 4.72E+l9 l.22E+l6 Future 48.0 l.54E+17 2.01E+l7 6.88E+l8 5.65E+l9 1.46E+16 50.0 l.61E+l7 2.l0E+17 7.18E+18 5.88E+l9 l.53E+16 60.0 l.95E+l7 2.55E+17 8.68E+l8 7.04E+l9 1.83E+l6 Note:

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

WCAP-18102-NP June 2017 Revision 0

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

Operating Lower Shell Lower Shell Cycle Intermediate Time Outlet Inlet to to Lower (EFPY) Shell to Nozzle<a> Nozzle<a> Intermediate Closure Upper Shell Shell Head 1 1.2 1.0lE-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 9.1lE-04 7.68E-03 7.76E-06 4 3.6 3.70E-05 4.33E-05 1.14E-03 9.86E-03 1.0lE-05 5 4.8 4.76E-05 5.57E-05 1.47E-03 l.26E-02 l.27E-05 6 5.9 5.45E-05 6.36E-05 1.65E-03 1.46E-02 1.48E-05 7 7.1 6.39E-05 7.47E-05 l.92E-03 l.73E-02 1.78E-05 8 8.2 7.18E-05 8.39E-05 2.14E-03 l.96E-02 2.02E-05 9 9.6 8.16E-05 9.53E-05 2.41E-03 2.23E-02 2.30E-05 10 10.8 8.76E-05 1.02E-04 2.57E-03 2.41E-02 2.49E-05 11 11.8 9.28E-05 1.08E-04 2.72E-03 2.55E-02 2.64E-05 12 12.9 9.89E-05 l.15E-04 2.89E-03 2.73E-02 2.82E-05 13 14.3 1.07E-04 l.25E-04 3.llE-03 2.95E-02 3.05E-05 14 15.6 l.16E-04 1.35E-04 3.43E-03 3.16E-02 3.30E-05 15 16.9 l.24E-04 l.45E-04 3.73E-03 3.35E-02 3.53E-05 16 18.4 1.34E-04 l.57E-04 4.08E-03 3.58E-02 3.82E-05 17 19.6 1.43E-04 l.68E-04 4.40E-03 3.78E-02 4.07E-05 18 21.0 1.54E-04 1.80E-04 4.73E-03 4.04E-02 4.35E-05 19 22.5 l.64E-04 1.92E-04 5.05E-03 4.28E-02 4.62E-05 20 23.8 l.75E-04 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 l.98E-04 2.31E-04 6.IOE-03 5.08E-02 5.48E-05 23 28.0 2.09E-04 2.44E-04 6.46E-03 5.35E-02 5.78E-05 24 29.3 2.19E-04 2.56E-04 6.77E-03 5.58E-02 6.04E-05 32.0 2.41E-04 2.81E-04 7.45E-03 6.08E-02 6.60E-05 36.0 2.72E-04 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 8.25E-05 Future 48.0 3.68E-04 4.30E-04 l.14E-02 9.06E-02 9.90E-05 50.0 3.83E-04 4.48E-04 l.19E-02 9.43E-02 l.03E-04 60.0 4.63E-04 5.41E-04 1.44E-02 1.13E-01 1.24E-04 Note:

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

W CAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary 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 (n/cm2)

Cycle T ime (EFPY) 45° Intermediate 45° Lower Shell Shell 1 1.2 3.36E+17 3.37E+l 7 2 1.9 5.46E+l 7 5.47E+l 7 3 2.7 7.95E+17 7.96E+l 7 4 3.6 1.02E+18 1.02E+18 5 4.8 1.31E+18 1.31E+18 6 5.9 1.58E+18 1.58E+18 7 7.1 1.88E+18 1.88E+18 8 8.2 2.14E+18 2.15E+18 9 9.6 2.50E+18 2.50E+18 10 10.8 2.78E+18 2.79E+18 11 11.8 3.03E+18 3.03E+18 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+18 16 18.4 4.42E+18 4.43E+18 17 19.6 4.66E+18 4.67E+18 18 21.0 4.94E+18 4.95E+18 19 22.5 5.28E+18 5.29E+18 20 23.8 5.58E+18 5.60E+18 21 25.2 5.87E+18 5.89E+18 22 26.6 6.18E+18 6.20E+18 23 28.0 6.48E+18 6.50E+18 24 29.3 6.77E+18 6.78E+18 32.0 7.37E+18 7.38E+18 36.0 8.25E+18 8.27E+18 40.0 9.13E+18 9.15E+18 Future 48.0 1.09E+19 1.09E+19 50.0 1.13E+19 1.14E+19 60.0 1.35E+19 1.36E+19 WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-14 Table 2-10 Calculated Iron Displacements per Atom at the Pressure Vessel Longitudinal Welds Maximum Iron Displacements per Cumulative Atom Operating (dpa)

Cycle Time EFPY 45° Intermediate 45° Lower

( )

Shell Shell 1 1.2 5.39E-04 5.40E-04 2 1.9 8.76E-04 8.77E-04 3 2.7 l.28E-03 l.28E-03 4 3.6 l.63E-03 l.64E-03 5 4.8 2.l0E-03 2.lIE-03 6 5.9 2.53E-03 2.54E-03 7 7.1 3.02E-03 3.02E-03 8 8.2 3.44E-03 3.44E-03 9 9.6 4.0IE-03 4.02E-03 IO 10.8 4.46E-03 4.47E-03 11 11.8 4.85E-03 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 6.61E-03 6.62E-03 16 18.4 7.lOE-03 7.llE-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 9.44E-03 22 26.6 9.92E-03 9.94E-03 23 28.0 l.04E-02 1.04E-02 24 29.3 1.09E-02 l.09E-02 32.0 l.18E-02 1.18E-02 36.0 1.32E-02 1.33E-02 40.0 1.46E-02 1.47E-02 Future 48.0 l.75E-02 l.75E-02 50.0 1.82E-02 l.82E-02 60.0 2.17E-02 2.17E-02 WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-15 Table 2-11 Calculated Fast Neutron Fluence (E > 1.0 MeV) at the Center of the Surveillance Capsules Cumulative Neutron (E > 1.0 MeV) Fluence (n/cm2)

Operating Time V u w y X s<a) T(bl z(c)

Cycle (15°) °)

(25° ) °) °/15 °)

(EFPY) (25 (25 (15°) °/

(45 25 (35°/25° ) (35°/15°)

1 1.2 2.97E+l8 1.99E+18 l.99E+l8 1.99E+18 2.97E+l8 1.03E+l8 1.32E+l8 l.32E+l8 2 1.9 -- 3.26E+18 3.26E+l8 3.26E+18 4.84E+18 l.69E+18 2.l7E+18 2.17E+18 3 2.7 -- 4.81E+18 4.81E+18 4.81E+ 18 7.17E+18 2.46E+18 3.19E+18 3.19E+18 4 3.6 -- 6.18E+18 6.18E+18 6.18E+18 9.25E+l8 3.15E+18 4.08E+18 4.08E+18 5 4.8 -- -- 7.90E+18 7.90E+18 l.19E+19 4.04E+18 5.22E+18 5.22E+18 6 5.9 -- -- 9.52E+18 9.52E+18 l.40E+19 4.86E+ 18 6.29E+18 6.29E+18 7 7.1 -- -- -- 1.13E+19 l.67E+19 5.77E+18 7.44E+18 7.44E+l8 8 8.2 -- -- -- l.29E+19 l.91E+l9 6.56E+ 18 8.47E+18 8.47E+l8 9 9.6 -- -- -- l.49E+19 2.19E+19 7.68E+18 9.85E+18 9.85E+18 10 10.8 -- -- -- l.64E+19 2.38E+19 8.52E+18 l.09E+19 l.09E+19 11 11.8 -- -- -- l.76E+l9 2.52E+19 9.28E+18 l.21E+19 1.24E+19 12 12.9 -- -- -- l.92E+19 2.71E+19 l.01E+l9 l.37E+19 1.42E+19 13 14.3 -- -- -- 2.IOE+19 2.93E+19 1.lOE+19 1.55E+19 1.65E+19 14 15.6 -- -- -- -- 3.12E+19 1.18E+19 1.68E+l9 l.84E+l9 15 16.9 -- -- -- -- 3.30E+19 1.26E+19 1.83E+19 2.02E+19 16 18.4 -- -- -- -- 3.54E+l9 l.35E+19 2.00E+19 2.26E+19 17 19.6 -- -- -- -- 3.74E+19 l.42E+19 2.13E+l9 2.45E+19 18 21.0 -- -- -- -- 3.99E+19 l.51E+19 2.29E+19 2.70E+19 19 22.5 -- -- -- -- 4.24E+19 l.61E+19 2.47E+19 2.95E+19 20 23.8 -- -- -- -- 4.49E+19 1.78E+19 2.63E+19 3.20E+19 21 25.2 -- -- -- -- 4.74E+19 l.94E+19 2.80E+19 3.46E+19 22 26.6 -- -- -- -- 4.99E+19 2.11E+19 2.96E+19 3.70E+19 23 28.0 -- -- -- -- -- 2.36E+19 3.13E+19 3.95E+19 24 29.3 -- -- -- -- -- 2.58E+l9 3.28E+19 4.18E+19 32.0 -- -- -- -- -- 3.06E+19 3.60E+19 4.65E+19 36.0 -- -- -- -- -- 3.77E+19 4.07E+19 5.36E+19 40.0 -- -- -- -- -- 4.48E+19 4.54E+19 6.07E+19 Future 48.0 -- -- -- -- -- 5.90E+19 5.49E+19 7.49E+19 50.0 -- -- -- -- -- 6.25E+ 19 5.72E+19 7.84E+19 60.0 -- -- -- -- -- 8.02E+19 6.91E+19 9.61E+19 Notes:

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

(b) Capsule Twas moved to a 25° location after Cycle 10.

(c) Capsule Z was moved to a 15 ° location after Cycle 10.

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Westinghouse Non-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 1 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°la) In Reactor 0_74(d)

° ° b T (55 /65 i ) In Reactor 0_94(d)

Z (305 °/165 °lc) In Reactor 1.20<d)

Notes:

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

(b) Capsule Twas moved to the 65 ° location after Cycle 10.

(c) Capsule Zwas moved to the 165 ° location after Cycle 10.

(d) The lead factors for the capsules remaining in the reactor are calculated based on End of Cycle (EOC) 24, the last completed operating cycle.

WCAP-18102-NP June 2017 Revision 0

WestinghouseN on-Proprietary Class 3 2-17 Table 2-13 Calculational Uncertainties Uncertainty Description Vessel Inner Capsule Radius PCA Comparisons 3% 3%

H.B. Robinson Comparisons 3% 3%

Analytical Sensitivity Studies 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

WCAP-18102-NP June2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-18 c.. - ShlHulSIM oQ 9

~-

t:;-

oQ 33,8 67.5 101.2 135.0 168.8 202.5 236.2 270.0

[cm)

Figure 2-1 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Midplane; Octant with No Surveillance Capsules WCAP-18102-N P June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-19

- c.

- CnanSIMI g,jaj Tcrt Wair OCI UI 0

101.2 135.0 168.8 202.5 236.2 270.0

[cm]

Figure 2-2 Beaver Valley Unit 1 r,0 Reactor Geometry at the Core Midplane; Octant with Surveillance Capsules WCAP-18102 -NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 2-20

~ 7======= =~ -

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

110 rJ co

~in E,,;

21 0,

c,;

ID I

ID rJ C) 7 en

~

N 1-------- --

I 59.6 119.2 178.8 238.4 298.o R

[cm]

Figure 2-3 Beaver Valley Unit 1 r,z Reactor Geometry WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The requirements for P-T limit curve develop ment are specifie d in 10 CFR 50, Append ix G [Ref. 4]. The beltline region of the reactor vessel is defined as the followin g in 10 CFR 50, Append ix G:

"the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficien t neutron radiation damage to be considered in the selection of the most limiting materia l with regard to radiation damage. "

The Beaver Valley Unit 1 beltline materials tradition ally included the intermed iate and lower shell plate and weld materials; however, as described in NRC Regulat ory Issue Summar y (RIS) 2014-11 [Ref. 9],

any reactor vessel materia ls that are predicted to experien ce a neutron fluence exposur e greater than 1.0 x 10 17 n/cm 2 (E > 1.0 MeV) at the end of the licensed operatin g period should be conside red in the development of P-T limit curves. The materials that exceed this fluence threshol d are referred to as extended beltline material s 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 extende d beltline material s include the upper shell forging, upper to interme diate shell girth weld, the nozzle to upper shell welds, and the nozzle forging materials. The inlet and outlet nozzles are conside red a part of the extende d beltline, as the exposur e at the nozzle welds is conserv atively used to represen t the exposur e at the nozzles.

Per NRC RIS 2014-11, the nozzle materials must be evaluate d for their potential effect on P-T limit curves regardless of exposur e

- See Append ix B for more details.

As part of this P-T limit curve develop ment effort, the methodo logy and evaluati ons used to determin e the initial RT NOT values for the Beaver Valley Unit 1 reactor vessel beltline and extended beltline base metal materials were reviewe d and updated, as appropr iate. Table 3-1 contains a summar y of these methodologies. Summa ry of the best-estimate copper (Cu) and nickel (Ni) contents, in units of weight percent (wt. %), as well as initial RT NOT values for the reactor vessel beltline and extende d beltline materials are provide d in Table 3-2 for Beaver Valley Unit 1. Table 3-3 contains a summar y of the initial RT NOT values of the reactor vessel flange and replacem ent reactor vessel closure head. These values serve as input to the P-T limit curves "flange- notch" per Append ix G of 10 CFR 50 - See Section 6.3 for details.

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of Beaver Valley Unit 1 Reactor Vessel Base Metal Material Initial RT NDT Determination Methodologies Reactor Vessel Material Methodology Upper Shell Forging BTP 5-3, Paragraph B 1.1 (3ia> [Ref. IO]

Intermediate and Lower Shell Plates ASME Code,Section III, Subsection NB-2300Cb) [Ref. I1]

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

Notes:

(a) The Beaver Valley Unit 1 Certified Material Test Report (CMTR) does not list the orientation of the Charpy V Notch tests results for the upper shell forging material. However, even though BTP 5-3, Paragraph B1.1 (3) methodology for SA-508, Class 2 material must be used due to this lack of Charpy V-Notch orientation, the initial RTNDT for this material remains drop-weight limited and is confirmed (See Table 3-2) due to excellent Charpy V-notch test results (in the assumed strong-orientation).

(b) The reactor vessel beltline plate material initial RTNDT values were determined in accordance with the methodology of ASME Code,Section III, Subsection NB-2300 [Ref. 11] utilizing CVGraph, Version 6.02 as documented in Westinghouse Letter MCOE-LTR-15-15-NP, Revision 1 [Ref. 12].

(c) The initial RTNDT values of the Beaver Valley Unit 1 inlet and outlet nozzles were determined in accordance with BWRVIP-173-A [Ref. 13] See Appendix B for more details.

WCAP-18102-NP June 2017 Revision 0

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 Materials<a)

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

Cu Ni (OF)

Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-l<b) 0.14 0.62 26.8 Intermediate Shell Plate B6607-2 C4381-ib) 0.14 0.62 53.6 Lower Shell Plate B6903-1 C6317-1(bl 0.21 0.54 13.1 Lower Shell Plate B7203-2 C6293-ibl 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&B Lower Shell Longitudinal Welds 305414 0.34 0.61 -56 20-714 A&B Reactor Vessel Extended Beltline Materials Upper Shell Forging B6604 123V339VA1 0.12 0.68 40

(

305414 d) 0.34 0.61 -56 (3951 & 3958)

AOFJ 0.03 0.93 10 Upper Shell to Intermediate Shell Girth Weld 10-714 FOIJ 0.03 0.94 10 EODJ 0.02 1.04 10 HOCJ 0.02 0.93 10 Inlet Nozzle B6608-l 95443-1 0.10 0.82 48.5 Inlet Nozzle B6608-2 95460-1 0.10 0.82 -15.2 Inlet Nozzle B6608-3 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 10 Inlet Nozzle Welds DBIJ 0.02 0.97 10 1-717B, 1-717D, 1-717F EOEJ 0.01 1.03 10 ICJJ 0.03 0.99 10 JACJ 0.04 0.97 10 Outlet Nozzle B6605-1 95415-1 0.13 0.77 -26.2 Outlet Nozzle B6605-2 95415-2 0.13 0.77 3.3 Outlet Nozzle B6605-3 95444-1 0.09 0.79 10.1 ICJJ 0.03 0.99 10 Outlet Nozzle Welds IOBJ 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-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 3-4 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<a>

Fracture Chemical Toughness Reactor Vessel Material Composition Heat Number Property and Identification Number Wt.% Wt.% Initial RTNDT(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 IO Surveillance Weld Data<e>

Beaver Valley Unit I 305424 0.26 0.61 ---

St. Lucie Unit I 0.23 0.07 ---

90136 Millstone Unit 2 0.30 0.06 ---

Fort Calhoun 305414 0.35 0.60 ---

Notes:

(a) All values originally documented in WCAP-15571, Supplement 1, Revision 2 [Ref. 14], unless otherwise noted.

(b) The reactor vessel beltline plate material heat numbers were taken from the Beaver Valley Unit 1 CMTRs.

(c) The initial RTNDT values for all the reactor vessel welds are generic. The initial RTNDT values for the base metal materials were updated or confirmed as discussed in Table 3-1.

(d) The chemistry values for weld Heat# 305414, as reported in WCAP-15571, Supplement 1, Revision 2 [Ref. 14], were rounded up for consistency with the reactor vessel beltline weld material that shares this same heat number.

(e) Surveillance data exists for weld Heat# 90136, # 305424, and# 305414 from multiple sources; see Section 4 for more details. The data for Beaver Valley Unit 1 weld metal Heat# 90136 was taken from WCAP-17896-NP [Ref. 5]. The data for St. Lucie Unit 1 weld metal Heat# 90136 was taken from the St. Lucie Unit 1 License Amendment Request for Extended Power Uprates, Attachment 5, Table 2.1.2-4 [Ref. 15]. The data for Millstone Unit 2 weld metal Heat #

90136 was taken from Table 4-1 of WCAP-16012 [Ref. 16]. The data for Fort Calhoun weld metal Heat# 305414 was taken from Table 5.2-4b of the 2012 Beaver Valley Unit 1 P-T Limits revision report [Ref. 17].

Table 3-3 Summary of Beaver Valley Unit 1 Replacement Reactor Vessel Closure Head and Vessel Flange Initial RTNDT Values Initial RTNDT Reactor Vessel Material Methodology

{°F)

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

[Ref. 11]

BWRVIP-173-A, Alternate Approach 2 Vessel Flange I o<b)

[Ref. 12]

Notes:

(a) The initial RTNDT value of the replacement reactor vessel (RV) closure head was taken from WCAP-16799-NP, Revision 1 [Ref. 18] and was determined in accordance with the methodology of ASME Code,Section III, Subsection NB-2300 [Ref. 1 1].

(b) The initial RTNDT value of the vessel flange was updated, utilizing the methodology of BWRVIP-173-A, Alternate Approach 2 [Ref. 13], from the value documented in WCAP-16799-NP, Revision 1 [Ref. 18].

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary 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 surveillance program. In addition to the plant-specific surveillance data, data from surveillance 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 surveillance 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-l . The surveillance capsule weld material for Beaver Valley Unit 1 is Heat# 305424, which is applicable to the intermediate shell longitudinal welds. Table 4-1 summarizes the Beaver Valley Unit 1 surveillance 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 surveillance capsule, Capsule X, were documented in WCAP-17896-NP [Ref. 5]. Appendix D concludes that the Beaver Valley Unit 1 surveillance plate and weld (Heat # 305424) material are non-credible; therefore, a full margin term will be utilized in the ART calculations contained in Section 7.

The Beaver Valley Unit 1 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 1 and Millstone Unit 2 surveillance programs. Thus, the St. Lucie Unit 1 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 summarizes the applicable surveillance capsule data pertaining to weld Heat # 90136. The combined surveillance data is deemed credible per Appendix D; however, as a result of the Millstone Unit 2 surveillance data including both weld Heat # 90136 and 10137, the Position 2.1 chemistry factor calculations for weld Heat# 90136 will utilize a full margin term for conservatism. See Appendix D for details.

The Beaver Valley Unit 1 reactor vessel upper shell to intermediate 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 surveillance program. Thus, in WCAP-15571, Supplement 1, 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-17896-NP [Ref. 5] concluded that the weld Heat# 305414 data is non-credible; therefore, a full margin term will be utilized in the ART calculations contained in Section 7.

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 Beaver Valley Unit 1 Surveillance Capsule Data Capsule Fluence<a> Measured 30 ft-lb Transition Material Capsule (x 10 19 n/cm2 , E > 1.0 MeV) Temperature Shift<h> {°F)

V 0.297 127.9 u 0.618 118.3 Lower Shell Plate B6903-1 (Longitudinal) w 0.952 147.7 y 2.10 141.7 X 4.99 175.8 V 0.297 138.0 u 0.618 132.1 Lower Shell Plate B6903-1 (Transverse) w 0.952 180.2 y 2.10 166.9 X 4.99 179.0 V 0.297 159.8 u 0.618 164.9 Surveillance Weld Material (Heat# 305424) w 0.952 186.3 y 2.10 178.5 X 4.99 237.8 Notes:

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

(b) Data was taken from Table 5-10 of WCAP-17896-NP [Ref. 5].

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 St. Lucie Unit 1 and Millstone Unit 2 Surveillance Capsule Data for Weld Heat# 90136 Capsule Fluence<a) Inlet Temperature (x 10 19 n/cm2, E > 1.0 Measured 30 ft-lb Transition Material Capsule (a) Temperature(b) Adjustment<c)

MeV) Temperature Shift<a) (°F)

(OF) (OF) 97 ° 0.5174 72.34 541 -1.7 St. Lucie Unit I 104 ° 0.7885 67.4 544.6 1.9 Data

' 284 ° 1.243 68.0 546.3 3.6 97 ° 0.324 65.93 544.3 1.6 Millstone Unit 2 104 ° 0.949 52.12 547.6 4.9 Data 830 1.74 56.09 548.0 5.3 Notes:

(a) For surveillance weld heat# 90136, data pertaining to the St. Lucie Unit l were taken from the St. Lucie Unit l License Amendment Request for Extended Power Uprates, Attachment 5, Table 2.1.1-3 [Ref. 15]. Data pertaining to Millstone Unit 2 were taken from Table 5-10 ofWCAP-16012 [Ref. 16].

(b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal.

(c) Temperature adjustment = 1.0*(T capsule - Tptant), where Tpt ant = 542.7°F for Beaver Valley Unit 1. 542.7°F is the cycle-by-cycle average downcomer temperature for Beaver Valley Unit 1 for Cycle 1 through Cycle 22. The temperature adjustment procedure is applied to the weld RTNDT data for each of the St. Lucie Unit 1 and Millstone Unit 2 capsules in the Position 2.1 chemistry factor calculation - See Section 5 for more details.

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary Class 3 5-1 5 CHEMISTRY FACTORS The chemistry factors (CFs) were calculated using Regulatory Guide 1.99, Revision 2, Positions 1.1 and 2.1. Position 1.1 chemistry factors for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of Regulatory Guide 1.99, Revision

2. The best-estimate copper and nickel weight percent values for the Beaver Valley Unit 1 reactor vessel materials are provided in Table 3-2 of this report.

The Position 2.1 chemistry factors are calculated for the materials that have available surveillance program results. The calculation is performed using the method described in Regulatory Guide 1.99, Revision 2. The Beaver Valley Unit 1 surveillance data as well as the applicable sister plant data was summarized in Section 4 of this report, and will be utilized in the Position 2.1 chemistry factor calculations in this Section.

The Position 2.1 chemistry factor calculations are presented in Tables 5-1 through 5-3 for Beaver Valley Unit 1 reactor vessel materials that have associated surveillance data. These values were calculated using the surveillance data summarized in Section 4 of this report. Note that the Position 2.1 chemistry factor for weld Heat # 305414 was previously reported in WCAP-15571, Supplement 1, Revision 2 [Ref. 14]

and is therefore not recalculated. All of the surveillance data is adjusted for irradiation temperature and chemical composition differences in accordance 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 calculations in Section 7 according to the conclusions of the credibility evaluation for each of the surveillance materials, as documented in Section 4.

The Position 1.1 chemistry factors are summarized along with the Position 2.1 chemistry factors in Table 5-4 for Beaver Valley Unit 1.

WCAP-18102-NP June 2017 Revision 0

Westinghouse Non-Proprietary C lass 3 5-2 Table 5-1 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Lower Shell Plate B6903-1 Using Surveillance Capsule Data Capsule ra>

LS Plate B6903-1 ARTNDT(c) FF*ARTNDT Capsule (x 10 19 n/cm2 , E > 1.0 FF FF2 Data (OF) (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 Orientation 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 Transverse Orientation 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 CFLSPlateB6903-l = L(FF

LiRTNDT) + L(FF2) = (1585.86) 7 (11.]54) = 142.2° F Notes:
(a) f = fluence.
(b) FF = fluence factor= 1° 28 - 0* io*Iog f).
(c) L\RTNDr values are the measured 30 ft-lb shift 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 ra>
Weld Metal ARTNDT(c) FF*ARTNDT Capsule (x 10 19 n/cm 2, E > 1.0 FF(b) FF2 Heat # 305424 {°F) {°F)
MeV)
V 0.297 0.6677 169.4 (159.8) 113.10 0.446 u 0.618 0.8652 174.8 (164.9) 151.23 0.749 Beaver Valley Unit 1 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 252.1 (237.8) 353.39 1.965 SUM: 1039.87 5.577 CFweld Heat# 305424 = L(FF
LiRTNDT) + L(FF2) = (1039.87) + (5.577) = 186.5°F Notes:
(a) f = fluence.
(b) FF = fluence factor = 1° 28 - 0* 1 O*Iog f).
(c) L\RTNDr values are the measured 30 ft-lb shift values. The L\RTNDr values are adjusted using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 4-1 of this report). Ratio applied to the Beaver Valley Unit 1 surveillance data = CFvessel weld I CFsurv Weld = 191. 7 °F I 181.6 °F = 1.06.
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Westinghouse Non-Proprietary Class 3 5-3 Table 5-3 Calculation of Beaver Valley Unit 1 Chemistry Factor Value for Weld Heat# 90136 Using Surveillance Capsule Data Weld Metal Capsule ra> ARTNDT(c) FF*ARTNDT Capsule 19 FF FF2 Heat# 90136 (x 10 n/cm2 , E > 1.0 MeV) {°F) {°F) 97 ° 0.5174 0.8160 82.6 (72.34) 67.44 0.666 St. Lucie Unit I ° 104 0.7885 0.9333 81.1 (67.4) 75.68 0.871 Data 284° 1.243 1.0606 83.8 (68.0) 88.85 1.125 97° 0.324 0.6902 67.5 (65.93) 46.61 0.476 Millstone Unit 2 d 104 ° 0.949 0.9853 57.0 (52.12) 56.18 0.971 Data< )
830 1.74 1.1523 61.4 (56.09) 70.74 1.328 SUM: 405.50 5.437 CfweldHeat#90!36 = L(FF
LlRTNoT) + L(FF2) = (405.50) + (5.437) = 74.6°F Notes:
(a) f = fluence.
(b) FF = fluence factor = fO 28 - O I O* og f).
l (c) RTNDT values are the measured 30 ft-lb shift values. The RTNDT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 4-2 of this report). The temperature adjustments are listed in Table 4-2. Ratio applied to the St. Lucie Unit 1 surveillance data
= CFv essel Weld/ CFsurv Weld = 124.3°F / 106.6 °F 1.17. A ratio of 1.00 was conservatively applied to the Millstone Unit 2 surveillance data, since CFvessel Weld < CF surv Weld*
(d) Millstone Unit 2 surveillance data contains specimens from both weld Heat# 90136 and weld Heat# 10137. However, this inclusion of an additional heat is not expected to negatively impact the subsequent reactor vessel integrity calculation results, as additional conservatisms are in place. See Appendix D for more details.
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Westinghouse Non-Proprietary 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)
Heat Number and Identification Number Position 1.1 <a ) Position 2.1 Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-l C4381-l 100.5 ---
Intermediate Shell Plate B6607-2 C4381-2 100.5 ---
Lower Shell Plate B6903-l C63l7-l 147.2 142.ib)
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 Welds 20-714 A&B 305414 210.5 216.9(e)
Reactor Vessel Extended Beltline Materials Upper Shell Forging B6604 123V339VA1 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-l 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 l-717B, l-717D, l-717F DBIJ 27.0 ---
EOEJ 20.0 ---
ICJJ 41.0 ---
JACJ 54.0 ---
Outlet Nozzle B6605-l 95415-1 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 l-717A, l-717C, l-717E JACJ 54.0 ---
HOCJ 27.0 ---
EODJ 27.0 ---
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Westinghouse Non-Proprietary Class 3 5-5 Table 5-4 Summary of Beaver Valley Unit 1 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (° F)
Heat Number and Identification Number Position 1.1<a> Position 2.1 Outlet Nozzle Welds 1-717A, 1-717C, 1-717E ---
FOIJ 41.0 (continued)
Surveillance Weld Data Beaver Valley Unit 1 305424 181.6 ---
St. Lucie Unit 1 106.6 ---
90136 Millstone Unit 2 135.5 ---
Fort Calhoun 305414 212.0 ---
Notes:
(a) Position I.1 chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-2 of this report and Tables 1 and 2 of Regulatory 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 this report. As discussed in Section 4, the surveillance 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 report. As discussed in Section 4, the surveillance weld data for Heat# 305424 is not credible.
(e) Position 2.1 chemistry factor was taken from WCAP-15571, Supplement 1, Revision 2 [Ref. 14]. As discussed in Section 4, the surveillance weld data for Heat# 3 05414 is not credible.
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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, K 1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K1c, for the metal temperature at that time. K 1c is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 3]. The K 1c curve is given by the following equation:
K Jc =33 .2+ 20 . 734 *e 02 <T-RT,vm )]
[O (1)
where, K 1c (ksi,/in.) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K 1c curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.
6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
(2)
where, Kim stress intensity factor caused by membrane (pressure) stress K 1t 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 RTNDT 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 June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-2 For membrane tension, the corresponding K 1 for the postulated defect is:
K1m = Mmx (pR;/t) ( 3) where, M m for an inside axial surface flaw is given by:
Mm = 1.85 for .Ji < 2, Mm = 0.926 .Ji for 2 s .Ji s 3.464, Mm = 3.21 for .Ji > 3.464 and, Mm for an outside axial surface flaw is given by:
Mm 1.77 for .Ji < 2, Mm 0.893 .Ji for 2 S Ji S 3 .464 ,
Mm 3.09 for .Ji > 3.464 Similarly, Mn for an inside or an outside circumferential surface flaw is given by:
Mn 0.89 for .Ji < 2, Mn 0.443 .Ji for 2 S Ji S 3 .464 ,
Mm 1.53 for .Ji > 3 .464 where:
p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in) .
For bending stress, the corresponding K 1 for the postulated axial or circumferential defect is:
K 1b = Mb* Maximum Bending Stress, where Mb is two-thirds of M m (4)
The maximum K 1 produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:
K 11 = 0.953x10*3 x CR x t2*5 ( 5) where CR is the cooldown rate in °F/hr., or for a postulated axial or circumferential outside surface defect K it= 0.753xl0-3 x HU x t2*5 (6) where HU is the heatup rate in °F/hr.
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Westinghouse Non-Proprietary Class 3 6-3 The through-wall temperature difference associated with the maximum thermal K1 can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code,Section XI, Appendix G, Fig. G-2214-2 for the maximum thermal K 1 *
(a) The maximum thermal K 1 relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).
(b) Alternatively, the K 1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 1/4-thickness axial or circumferential inside surface defect using the relationship:
K1t = (1.0359Co + 0.6322C1 + 0.4753C2 + 0.3855C3) * & (7) or similarly, K 1t during heatup for a 1/4-thickness outside axial or circumferential surface defect using the relationship:
K11 = (1.043 Co+ 0.630 C1 + 0.481 C 2+ 0.401 C3) * & (8) where the coefficients C0, C 1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
3 o-(x) =Co+ C1(x I a)+ C2(x I a)2 + C 3(X I a) (9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).
Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 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-14040-A, Revision 4 (equation 2.6.1-1). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2 /hr at 70° F and 0.379 ft2/hr at 550° F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2 -° F.
At any time during the heatup or cooldown transient, K 1 c is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code,Section XI, paragraph G-2120), the appropriate value for RTNDT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 6-4 intensity factors, Kit, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained, and from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T 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 T (temperature) across the vessel wall developed during cooldown results in a higher value of K1c at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K 1c exceeds K1t, 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 1/4T location, and therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1c for the inside 1/4T flaw during heatup is lower than the K1c for the flaw during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K 1c values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The third portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T 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-NP June 2017 Revision 0
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-temperature curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point 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 HEADNESSEL FLANGE REQUIREMENTS 10 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 RTNDT 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 RTNDT of l 0 °F is associated with the vessel flange of the Beaver Valley Unit 1 vessel, so the minimum allowable temperature of this region is l 30 °F at pressures greater than 621 psig (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.
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Westinghouse Non-Proprietary Class 3 7-1 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART =Initial RTNDT + dRTNDT + Margin (10)
Initial RTNDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [Ref. 11]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided if there are sufficient test results to establish a mean and standard deviation for the class.
dRTNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
logf) dRTNDT =CF* f(0.28-0.!0 (11)
To calculate dRTNDT 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,(depth x) -
- Lsurf: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 dRTNDT at the specific depth.
The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2].
Table 7-1 contains the surface fluence values at 50 EFPY, which were used for the development of the P T limit curves contained in this report. Table 7-1 also contains the 1/4T and 3/4T calculated fluence values and fluence factors, per Regulatory 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 materials.
Margin is calculated as M =2 CT ; + CT . The standard deviation for the initial RTNDT margin term ( cr1) is 0 °F when the initial RTNDT is a measured value, and l 7° F when a generic value is available. The standard deviation for the dRTNDT margin term, cr, is l 7°F for plates or forgings when surveillance data is not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data is used. For welds, cr is equal to 28°F when surveillance capsule data is not used or is non-credible, and is 14°F (half the value) when credible surveillance capsule data is used. The value for cr need not exceed 0.5 times the mean value ofdRTNDT
Contained in Tables 7-2 and 7-3 are the 50 EFPY ART calculations at the I/4T and 3/4T locations for generation ofthe Beaver Valley Unit I heatup and cooldown curves.
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Westinghouse Non-Proprietary Class 3 7-2 The inlet and outlet nozzle forging materials for Beaver Valley Unit 1 have projected fluence values that exceed the 1 x 10 17 n/cm2 fluence threshold at 50 EFPY per Table 2-7; therefore, per NRC RIS 2014-11
[Ref. 9], neutron radiation embrittlement must be considered herein for these materials. The nozzle ART calculations conservatively utilize the maximum fluence value for each nozzle material, as documented in Appendix B. Thus, ART calculations for the inlet and outlet nozzle forging materials utilizing the 1/4T and 3/4T fluence values are excluded from Tables 7-2 and 7-3, respectively. Finally, the second conclusion of TLR-RES/DE/CIB-2013-01 [Ref. 20] states that if RTNDT is calculated to be less than 25° F, then embrittlement need not be considered. This conclusion was applied, as necessary, to the ART calculations documented in Tables 7-2 and 7-3.
The limiting ART values for Beaver Valley Unit 1 to be used in the generation of the P-T limit curves are based on Lower Shell Plate B6903-1 (Position 1.1 ). For conservatism, 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" methodology, for Lower Shell Plate B6903-1 are summarized in Table 7-4.
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Westinghouse Non-Proprietary Class 3 7-3 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 Surface 1/4Tf 3/4Tf Fluence, (a) 1/4T 3/4T Reactor Vessel Region (n/cm2 , (n/cm2 ,
(n/cm2, FF FF E> 1.0MeV) E>l.OMeV)
E> 1.0MeV)
Reactor Vessel Beltline Materials Intermediate Shell Plates 5.88 X 10 19 3.666 X 10 19 1.3370 1.425 X 10 19 1.0982 Lower Shell Plates 5.89 X 10 19 3.672 X 10 19 1.3374 1.427 X 10 19 1.0987 Intermediate to Lower Shell 5.88 X 10 19 3.666 X 10 19 1.3370 1.425 X 10 19 1.0982 Girth Weld Intermediate Shell J.13 X 10 19 7.04 X 10 18 0.9018 2.74 X 10 18 0.6469 Longitudinal Welds Lower Shell Longitudinal 1.14 X 10 19 7.11 X 10 18 0.9042 2.76 X 10 18 0.6492 Welds Reactor Vessel Extended Beltline Materials Upper Shell Forging 7.18 X 10 18 4.48 X 10 18 0.7764 1.74 X 10 18 0.5366 Upper to Intermediate Shell 7.18xl0 18 4.48 X 10 18 0.7764 1.74 X 10 18 0.5366 Girth Weld Inlet Nozzle to Upper Shell 2.10 X 10 17(b) 1.31 X 10 17 0.1313 5.09 X 10 1 6 0.0679 Weld - Lowest Extent Outlet Nozzle to Upper 1.61 X 10 17(b) 1.00 X 10 17 0.1099 3.90 X 10 16 0.0556 Shell Weld - Lowest Extent Notes:
(a) 50 EFPY fluence values were 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 for the inlet and outlet nozzle materials. The actual nozzle forging fluence values, at the location of a postulated flaw along the nozzle corner region, are expected to be lower since they are further away from the active core.
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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 the l/4T Location Reactor Vessel Material and ID CF l/4T Fluence l/4T RTNDT(U)(a) ARTNDT(b) a.Ca)
O'!!,.
(c)
Margin ART<d>
Heat Number Number (OF) (n/cm2, E > 1.0 MeV) FF (OF) (°F) (OF) (OF) (OF) (OF)
Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-l 100.5 3.666 X 10 19 1.3370 26.8 134.4 0 17 34.0 195.2 Intermediate Shell Plate B6607-2 C4381-2 100.5 3.666 X 10 19 1.3370 53.6 134.4 0 17 34.0 222.0 Lower Shell Plate B6903-l C63 l 7-l 147.2 3.672 X 10 19 1.3374 13.1 196.9 0 17 34.0 244.0 Using Beaver Valley Unit 1 C6317-l 142.2 3.672 X 10 19 1.3374 13.1 190.2 0 17 34.0 237.3 surveillance data Lower Shell Plate B7203-2 C6293-2 98.7 3.672 X 10 19 1.3374 0.4 132.0 0 17 34.0 166.4 Intermediate to Lower Shell Girth 90136 124.3 3.666 X 10 19 1.3370 -56 166.2 17 28 65.5 175.7 Weld 11-714 Using St. Lucie Unit 1 and Millstone 90136 74.6 3.666 X 10 19 1.3370 -56 99.7 17 28 65.5 109.3 Unit 2 surveillance data Intermediate Shell Longitudinal 305424 191.7 0.704 X 10 19 0.9018 -56 172.9 17 28 65.5 182.4 Welds 19-714 A&B Using Beaver Valley Unit 1 305424 186.5 0.704 X 10 19 0.9018 -56 168.2 17 28 65.5 177.7 surveillance data Lower Shell Longitudinal Welds 305414 210.5 0.711 X 10 19 0.9042 -56 190.3 17 28 65.5 199.9 20-714 A&B Using Fort Calhoun surveillance data 305414 216.9 0.711 X 10 19 0.9042 -56 196.1 17 28 65.5 205.6 Reactor Vessel Extended Beltline Materials<e>
Upper Shell Forging B6604 123V339VA1 84.2 0.448 X 10 19 0.7764 40 65.4 0 17 34.0 139.4 Upper Shell to Intermediate Shell 305414 210.5 0.448 X 10 19 0.7764 -56 163.4 17 28 65.5 172.9 Girth Weld 10-714 (3951 & 3958) 305414 Using Fort Calhoun surveillance data 216.9 0.448 X 10 19 0.7764 -56 168.4 17 28 65.5 177.9 (3951 & 3958)
AOFJ 41.0 0.448 X 10 19 0.7764 10 31.8 17 15.9 46.6 88.4 Upper Shell to Intermediate Shell FOIJ 41.0 0.448 X 10 19 0.7764 10 31.8 17 15.9 46.6 88.4 Girth Weld 10-714 ( continued) EODJ 27.0 0.448 X 10 19 0.7764 10 0.0 (21.0) 17 0 34.0 44.0 HOCJ 27.0 0.448 X 10 19 0.7764 10 0.0 (21.0) 17 0 34.0 44.0 WCAP-18102-NP June 2017 Revision 0
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 the 1/4T Location Reactor Vessel Material and ID CF l/4T Fluence l/4T RTNDT(U)(a) ARTNDT(b) a,<a) GA(c) Margin ART<d>
Heat Number Number (OF) (n/cm2, E > 1.0 MeV) FF (OF) (OF) (OF) (OF) (OF) (OF)
EODJ 27.0 0.0131 X 10 19 0.1313 10 0.0(3.5) 17 0 34.0 44.0 FOIJ 41.0 0.0131 X 10 19 0.1313 10 0.0(5.4) 17 0 34.0 44.0 HOCJ 27.0 0.0131 X 10 19 0.1313 IO 0.0(3.5) 17 0 34.0 44.0 Inlet Nozzle Welds 1-717B, 1-717D, DBIJ 27.0 0.0131 X 10 19 0.1313 10 0.0(3.5) 17 0 34.0 44.0 l-717F EOEJ 20.0 0.0131 X 10 19 0.1313 IO 0.0(2.6) 17 0 34.0 44.0 ICJJ 41. 0 0.0131 X 10 19 0.1313 IO 0.0(5.4) 17 0 34.0 44.0 JACJ 54.0 0.0131 X 10 19 0.1313 10 0.0(7.1) 17 0 34.0 44.0 19 ICJJ 41.0 0.0100 X 10 0.1099 10 0.0(4.5) 17 0 34.0 44.0 IOBJ 27.0 0.0100 X 10 19 0.1099 10 0.0(3.0) 17 0 34.0 44.0 19 Outlet Nozzle Welds 1-717A, l-717C, JACJ 54.0 0.0100 X I 0 0.1099 10 0.0(5.9) 17 0 34.0 44.0 1-717E HOCJ 27.0 0.0100 X 10 19 0.1099 10 0.0(3.0) 17 0 34.0 44.0 EODJ 27.0 0.0100 X 10 19 0.1099 10 0.0(3.0) 17 0 34.0 44.0 FOIJ 41.0 0.0100 X 10 19 0.1099 10 0.0(4.5) 17 0 34.0 44.0 Notes:
(a) The plate and forging material initial RTNDT values are measured values. The initial RTNDT values for all of the reactor vessel welds are generic; hence cr1 = 17° F for all reactor vessel welds.
(b) As discussed in Section 7, calculated L'\RTNDT values less than 25° F have been reduced to zero per TLR-RES/DE/CIB-2013-01 [Ref. 20]. Actual calculated L'\RTNoT values are listed in parentheses.
(c) As discussed in Section 4, the surveillance plate and weld Heat# 305414 and# 305424 data were deemed non-credible. The surveillance weld data for Heat# 90136 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. 1], the base metal cr" = 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal cr" = 28 °F for Position 1.1 and 2.1 with non-credible surveillance data.
Since a full margin term will be used for Heat# 90136, cr,., = 28 °F with credible surveillance data for Position 2.1 for this weld heat. However, cr,., need not exceed 0.5*.t\RTNDT*
(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART= RTNDT(U) + RTNDT + Margin.
(e) As discussed in Section 7, the inlet and outlet nozzle forging material ART calculations utilizing a l /4T fluence value are excluded from this table because the nozzle material ART values necessary to perform the fracture mechanics evaluations on these materials are those calculated utilizing a maximum fluence value for each material.
These maximum ART values are documented in Table B-1.
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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 3/4T Location Reactor Vessel Material and ID CF 3/4T Fluence RTNDT(U)(a) ARTNDT(b) a?> ( )
G!!,,. c Margin ART(d)
Heat Number 3/4T FF Number (°F) (n/cm2, E > 1.0 MeV) (OF) (OF) (OF) (OF) (OF) (OF)
Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-l C4381-l 100.5 1.425 X 10 19 1.0982 26.8 110.4 0 17 34.0 171.2 Intermediate Shell Plate B6607-2 C4381'"2 100.5 1.425 X 10 19 1.0982 53.6 110.4 0 17 34.0 198.0 Lower Shell Plate B6903-l C63l 7-l 147.2 1.427 X 10 19 1.0987 13.1 161.7 0 17 34.0 208.8 Using Beaver Valley Unit I C6317-l 142.2 1.427 X 10 19 1.0987 13.1 156.2 0 17 34.0 203.3 surveillance data Lower Shell Plate B7203-2 C6293-2 98.7 1.427 X 10 19 1.0987 0.4 108.4 0 17 34.0 142.8 Intermediate to Lower Shell Girth 90136 124.3 1.425 X 10 19 1.0982 -56 136.5 17 28 65.5 146.0 Weld 11-714 Using St. Lucie Unit 1 and Millstone 90136 74.6 1.425 X 10 19 1.0982 -56 81.9 17 28 65.5 91.4 Unit 2 surveillance data Intermediate Shell Longitudinal 305424 191.7 0.274 X 10 19 0.6469 -56 124.0 17 28 65.5 133.5 Welds 19-714 A&B Using Beaver Valley Unit I 305424 186.5 0.274 X 10 19 0.6469 -56 120.6 17 28 65.5 130.2 surveillance data Lower Shell Longitudinal Weld 305414 210.5 0.276 X 10 19 0.6492 -56 136.6 17 28 65.5 146.2 20-714 A&B Using Fort Calhoun surveillance data 305414 216.9 0.276 X 10 19 0.6492 -56 140.8 17 28 65.5 150.3 Reactor Vessel Extended Beltline Materials(e)
Upper Shell Forging B6604 123V339VA1 84.2 0.174 X 10 19 0.5366 40 45.2 0 17 34.0 119.2 Upper Shell to Intermediate Shell 305414 210.5 0.174 X 10 19 0.5366 -56 113.0 17 28 65.5 122.5 Girth Weld 10-714 (3951 & 3958) 305414 Using Fort Calhoun surveillance data 216.9 0.174 X 10 19 0.5366 -56 116.4 17 28 65.5 125.9 (3951 & 3958)
AOFJ 41.0 0.174 X 10 19 0.5366 10 0.0 (22.0) 17 0 34.0 44.0 Upper Shell to Intermediate Shell FOIJ 41.0 0.174 X 10 19 0.5366 10 0.0 (22.0) 17 0 34.0 44.0 Girth Weld 10-714 ( continued) EODJ 27.0 0.174xl0 19 0.5366 10 0.0 (14.5) 17 0 34.0 44.0 HOCJ 27.0 0.174 X 10 19 0.5366 10 0.0 (14.5) 17 0 34.0 44.0 WCAP-18102-NP June 2017 Revision 0
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 3/4T Location Reactor Vessel Material and ID CF 3/4T Fluence RTNDT(U)(a) ARTNDT(b) a ?) CJA(
c)
Margin ART(d)
Heat Number 3/4T FF (OF) (OF) (OF) (OF) (OF) (OF) (OF) 2 Number (n/cm , E > 1.0 MeV)
EODJ 27.0 0.00509 X 10 19 0.0679 10 0.0 (1.8) 17 0 34.0 44.0 FOIJ 41.0 0.00509 X 10 19 0.0679 10 0.0 (2.8) 17 0 34.0 44.0 HOCJ 27.0 0.00509 X 10 19 0.0679 10 0.0 (1.8) 17 0 34.0 44.0 Inlet Nozzle Welds l-717B, 1-717D, DBIJ 27.0 0.00509 X 10 19 0.0679 10 0.0 (1.8) 17 0 34.0 44.0 1-717F EOEJ 20.0 0.00509 X 10 19 0.0679 10 0.0 (1.4) 17 0 34.0 44.0 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 X 10 19 0.0679 10 0.0 (3.7) 17 0 34.0 44.0 ICJJ 41.0 0.00390 X 10 19 0.0556 10 0.0 (2.3) 17 0 34.0 44.0 IOBJ 27.0 0.00390 X 10 19 0.0556 10 0.0 (1.5) 17 0 34.0 44.0 Outlet Nozzle Welds 1-717A, JACJ 54.0 0.00390 X 10 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 44.0 EODJ 27.0 0.00390 X 10 19 0.0556 10 0.0 (1.5) 17 0 34.0 44.0 FOIJ 41.0 0.00390 X 10 19 0.0556 10 0.0 (2.3) 17 0 34.0 44.0 Notes:
(a) The plate and forging material initial RTNDT values are measured values. The initial RTNDT values for all of the reactor vessel welds are generic; hence er1 = 17°F for all reactor vessel welds.
(b) As discussed in Section 7, calculated LiRTNoTvalues less than 25° F have been reduced to zero per TLR-RES/DE/CIB-2013-01 [Ref. 20]. Actual calculated LiRTNmvalues are listed in parentheses.
(c) As discussed in Section 4, the surveillance plate and weld Heat # 305414 and # 305424 data were deemed non-credible. The surveillance weld data for Heat # 90136 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 1.99, Revision 2 [Ref. 1], the base metal er = l 7° F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal er = 28 °F for Position 1.1 and 2.1 with non-credible surveillance data.
Since a full margin term will be used for Heat # 90136, er = 28 ° F with credible surveillance data for Position 2.1 for this weld heat. However, er need not exceed 0.5*.!iRTNDT*
(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNDT(U) + LiRTNDT + Margin.
(e) As discussed in Section 7, the inlet and outlet nozzle forging material ART calculations utilizing a 3/4T fluence value are excluded from this table because the nozzle material ART values necessary to perform the fracture mechanics evaluations on these materials are those calculated utilizing a maximum fluence value for each material.
These maximum ART values are documented in Table B-1.
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Westinghouse Non-Proprietary Class 3 7-8 Table 7-4 Summary of the Limiting ART Values Used in the Generation of the Beaver Valley Unit 1 Heatup and Cooldown Curves at 50 EFPY 1/4T Limiting ART<a> 3/4T Limiting ART<a>
244.5°F 209.5°F Lower Shell Plate B6903-l (Position 1.1)
Note:
(a) The ART values used for P-T limit curve development in this report are the limiting l/4T and 3/4T ART values calculated in Tables 7-2 and 7-3 rounded to the nearest whole number, then increased by 0.5 °F to add additional conservatism.
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Westinghouse Non-Proprietary Class 3 8-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 6 and 7 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4.
Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100° F/hr applicable for 50 EFPY, with the flange requirements and using the "Axial Flaw" methodology. Figure 8-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of -0, -20, -40, -60, and -100° F/hr applicable for 50 EFPY, with the flange requirements and using the "Axial Flaw" methodology. The heatup and cooldown curves were generated using the 1998 Edition through the 2000 Addenda ASME Code Section XI, Appendix G.
Allowable combinations of temperature and pressure for specific temperature 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 discussed in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 8-1 (heatup curve only). The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 Edition through the 2000 Addenda ASME Code Section XI, Appendix G as follows:
1.5 Krm < K1 c (13)
where, K 1m is the stress intensity factor covered by membrane (pressure) stress, 2
K1 c = 33.2 + 20.734 e [o.o (T-RTNDT>l, T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.
The criticality limit curve specifies pressure-temperature limits for core operation in order to provide additional margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40° F higher than the minimum permissible temperature in the corresponding pressure temperature curve for heatup and coo]down calculated as described in Section 6 of this report. For the heatup and coo]down curves without margins for instrumentation errors, the minimum temperature for the inservice hydrostatic leak tests for the Beaver Valley Unit 1 reactor vessel at 50 EFPY is 301° F; this temperature value is calculated based on Equation (13). The vertical line drawn from these points on the WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 8-2 pressure-temperature curve, intersecting a curve 40 °F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 8-1 and 8-2 define all of the above limits for ensuring prevention of non-ductile failure for the Beaver Valley Unit 1 reactor vessel for 50 EFPY with the flange requirements and without instrumentation uncertainties. The data points used for developing the heatup and cooldown P-T limit curves shown in Figures 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 cylindrical beltline and extended beltline reactor vessel materials. These ART values were slightly increased to add additional margin; this approach is conservative. As discussed in Appendix B, the P-T limits developed for the cylindrical beltline region bound the P-T limits for the reactor vessel inlet and outlet nozzles for Beaver Valley Unit 1 at 50 EFPY.
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Westinghouse Non-Proprietary Class 3 8-3 MATERIAL PROPERTY BA SIS LIMITING MATERIAL: Lower Shell Plate B6903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY: 114T, 244.5° F (Axial Flaw) 3/4T, 209.5° F (Axial Flaw) 2500 i;:::=======================:;.----.
!Operlim Version:5.4 Run: 19454 Operlim xlsm Version: 5. 4 ! --.--------i 2250 2000 1750 1500 Cl) 1250 Cl) 1000 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 Methodology (w/ K1c)
WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 8-4 MATERIAL PROPERTY BA SIS LIMITING MATERIAL: Lower Shell Plate B6903-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 50 EFPY: 1/4T, 244.5°F (Axial Flaw) 3/4T, 209.5°F (Axial Flaw) 2500 .-----_-_-_-_-_-_-_-_-_-_-_-_-_-_-_-_-_-_-_-_-_----------
loperlim Version:5.4 Run:19454 Operlim.xlsm Version: 5.41[
2250 2000 1750 1500 Cl) 1250 ti) ti)
Cl)
"C Cl) 1000 750 Cooldown Rates
°F/Hr Steady-State 500 -20
-40
-60
-100 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (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) Applicable 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-18102-NP June 2017 Revision 0
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 Methodology (w/
K1c, w/ Flange Notch, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100°F/hr Heatup 100°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 110 603 350 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 June 2017 Revision 0
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 Methodology (w/
K1c, w/ Flange Notch, and w/o Margins for Instrumentation Errors) 60 °F/hr Heatup 60°F/hr Criticality 100° 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 June 2017 Revision 0
Westinghouse Non-Proprietary 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 Methodology (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-18102 -NP June2017 Revision0
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 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) 210 840 210 808 215 802 215 775 215 725 215 861 215 831 220 830 220 805 220 761 220 884 220 856 225 860 225 838 225 801 225 910 225 884 230 894 230 875 230 846 230 938 230 915 235 931 235 916 235 895 235 970 235 949 240 973 240 961 240 949 240 1004 240 987 245 1018 245 1011 245 1010 245 1043 245 1029 250 1069 250 1067 250 1067 250 1085 250 1075 255 1125 255 1125 255 1125 255 1132 255 1127 260 1183 260 1183 260 1183 260 1184 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 1851 305 1979 305 1979 305 1979 305 1979 305 1979 310 2120 310 2120 310 2120 310 2120 310 2120 315 2275 315 2275 315 2275 315 2275 315 2275 320 2448 320 2448 320 2448 320 2448 320 2448 WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES
1. 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 1, "Fracture Toughness Criteria for Protection Against Failure."
4. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.
5. Westinghouse Report WCAP-17896-NP, Revision 0, "Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 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-Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998.
8. RSICC Data Library Collection DLC-185, "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," 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 (ADAMS)
Accession Number ML14149A165}
10. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements,"
Revision 2, U.S. Nuclear Regulatory Commission, March 2007.
11. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, Section NB-2300, "Fracture Toughness Requirements for Material."
12. Westinghouse Letter MCOE-LTR-15-15-NP, Revision 1, "Determination of Unirradiated RTNoT Values of the Four Beaver Valley Unit 1 Reactor Vessel Beltline Plate Materials Using a Hyperbolic Tangent Curve Fit," dated July 6, 2015.
13. BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data/or BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.
14. Westinghouse Report WCAP-15571 Supplement 1, Revision 2, "Analysis of Capsule Y from the Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," September 2011.
15. Florida Power & Light Letter L-2010-078, Attachment 5, "License Amendment Request Extended Power Uprates Licensing Report Florida Power & Light St. Lucie Nuclear Plant, Unit 1," April 2010.
[ADAMS Accession Number MLJOl 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-NP June 2017 Revision 0
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-NP, Revision 1, "Beaver Valley Power Station Unit 1 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 MLJ 10070570]
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 ML14318Al 77]
21. Westinghouse Report WCAP-14040-NP-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 1 Reactor Vessel Radiation Surveillance Program," November 2000.
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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (K1t)
Tables A-1 and A-2 contain the thennal stress intensity factors (Kit) for the maximum heatup and cooldown rates at 50 EFPY for Beaver Valley Unit 1. The reactor vessel cylindrical 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-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-2 TableA-1 K1t Values for Beaver Valley Unit 1 at 50 EFPY 100°F/hr Heatup Curves (w/ Flange Requirements and w/o Margins for Instrument Errors)
Water Vessel Temperature 1/4T Thermal Stress Vessel Temperature 3/4T Thermal Stress Temp. at 1/4T 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 -8.531 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 7.418 130 112.433 -10.118 93.898 7.612 135 117.114 -10.341 98.171 7.785 140 121.829 -10.535 102.523 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 -11.151 120.529 8.423 165 145.773 -11.275 125.142 8.519 170 150.613 -11.385 129.788 8.606 175 155.467 -11.491 134.462 8.687 180 160.330 -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 9.077 210 189.666 -12.064 167.713 9.131 WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 A-3 TableA-2 K1t 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 1/4T -100°F/hr Cooldown Temp. Location for-100° F/hr l/4T Thermal Stress (OF) Cooldown (°F) Intensity Factor (ksi v'in.)
210 232.426 13.510 205 227.352 13.454 200 222.278 13.398 195 217.204 13.342 190 212.131 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 12.512 115 136.031 12.457 110 130.958 12.403 105 125.886 12.349 100 120.813 12.295 95 115.741 12.240 90 110.669 12.187 85 105.597 12.133 80 100.526 12.079 75 95.454 12.025 70 90.382 11.972 65 85.311 11.919 60 80.241 11.865 WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES As described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. B-1], reactor vessel non-beltline materials may define pressure-temperature (P-T) limit curves that are more limiting than those calculated for the reactor vessel cylindrical shell beltline materials. Reactor vessel nozzles, penetrations, and other discontinuities have complex geometries that can exhibit significantly higher stresses than those for the reactor vessel beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the reference temperatures (RTNDT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.
The methodology contained in WCAP-14040-A, Revision 4 [Ref. B-2] was used in the main body of this report to develop P-T limit curves for the limiting Beaver Valley Unit 1 cylindrical shell beltline material; however, WCAP-14040-A, Revision 4 does not consider ferritic materials in the area adjacent to the beltline, specifically the stressed inlet and outlet nozzles. Due to the geometric discontinuity, the inside comer regions of these nozzles are the most highly stressed ferritic component outside the beltline region of the reactor vessel; therefore, these components are analyzed in this Appendix. P-T limit curves are determined for the reactor vessel nozzle comer region for Beaver Valley Unit 1 and compared to the P-T limit curves for the reactor vessel traditional beltline region in order to determine if the nozzles can be more limiting than the reactor vessel beltline as the plant ages and the vessel accumulates more neutron fluence. The increase in neutron fluence as the plant ages causes a concern for embrittlement of the reactor vessel above the beltline region. Therefore, the P-T limit curves are developed for the nozzle inside comer region since the geometric discontinuity results in high stresses due to internal pressure and the cooldown transient. The cooldown transient is analyzed as it results in tensile stresses at the inside surface of the nozzle comer.
An axial flaw is postulated at the inside surface of the reactor vessel nozzle comer and stress intensity factors are determined based on the rounded curvature of the nozzle geometry. The allowable pressure is then calculated based on the fracture toughness of the nozzle material and the stress intensity factors for the flaw. A discussion of the flaw depth is located in Appendix B.2.
B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (Krc) 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 Krc fracture toughness curve is dependent on the Adjusted Reference Temperature (ART) value for irradiated materials. The ART values for the inlet and outlet nozzle materials are determined using the methodology contained in Regulatory Guide 1.99, Revision 2 [Ref. B-3], which is described in Section 7 of this report, and weight percent (wt.
%) copper (Cu) and nickel (Ni), initial RTNDT 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 B-1 and a summary of the limiting inlet and outlet nozzle ART values for Beaver Valley Unit 1 is presented in Table B-2.
Nozzle Material Properties Copper and nickel weight percent values and the subsequent Position 1.1 CF values, were previously document in WCAP-15571, Supplement 1, Revision 2 [Ref. B-4] and are taken directly from this source for this analysis. The initial RTNDT values were determined for each of the Beaver Valley Unit 1 reactor WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-2 vessel inlet and outlet nozzle forging materials using the BWRV IP-173-A [Ref. B-5] Alternative Approach 2 methodology, contained in Appendix B of that report. For all six of the Beaver Valley Unit 1 inlet and outlet nozzle materials, CVGraph Version 6.02 was utilized to plot the material-specific Charpy V -Notch impact energy data from the Certified Material Test Reports (CMTRs) to determine the transition temperatures at 35 ft-lb and 50 ft-lb, as specified in the Alternative Approach 2 methodology.
The 35 ft-lb and 50 ft-lb temperatures were then evaluated per the Alternative Approach 2 methodology presented in BWRVIP-173-A, to determine the initial RTNDT 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 conservatively assumed that the orientation was the "strong direction" for each nozzle forging. Therefore, the 50 ft-lb transition temperatures for the inlet nozzles were increased by 30° F to provide conservative estimates for specimens oriented in the weak direction per the Alternative Approach 2 methodology in BWRVIP-173-A.
The material properties of the Beaver Valley Unit 1 inlet and outlet nozzle forging materials are documented in Table B-1.
Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > 1 MeV) exposure of the Beaver V alley Unit 1 reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculations were calculated at 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 conservatism.
Per Table 2-7, the inlet nozzles are determined to receive a projected maximum fluence of 2.10 x 10 17 n/cm2 (E > 1 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/cm2 (E > 1 MeV) at the lowest extent of the nozzles at 50 EFPY. Per NRC RIS 2014-11 [Ref. B-1], embrittlement of reactor vessel materials, with projected fluence values greater than 1 x 10 17 n/cm2, must be considered. However, the second conclusion of TLR-RES/DE/CIB-2013-01 [Ref. B-6] states that if RTNDT is calculated to be less than 25° F, then embrittlement 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 calculations for the Beaver Valley Unit 1 inlet and outlet nozzle forging materials are summarized in Table B-1.
The use of the embrittlement conclusion of TLR-RES/DE/CIB-2013-01 [Ref. B-6], and thus the limiting ART values summarized 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 10 17 n/cm2 (E > 1.0 MeV) and 4.10 x 10 17 n/cm2 (E > 1.0 MeV), respectively. If these fluence values are reached, embrittlement 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 Materials at 50 EFPY Fluence at Lowest Wt.% Wt.% CF<a> RTNDT(U)(c) ARTNDT(d) au (jA(e) Margin ART Reactor Vessel Material Extent of Nozzle(b) FF(b) cu<a) Ni<a> (OF) (OF) (OF) (OF) (OF) (OF) (OF)
(n/cm2 , E > 1.0 MeV)
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 Inlet Nozzle B6608-2 0.10 0.82 67.0 0.0210 X 10 19 0.1773 -15.2 0 (11.9) 0 0 0.0 -15.2 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 19 Outlet Nozzle B6605-l 0.13 0.77 95.3 0.0161 X 10 0.1501 -26.2 0 (14.3) 0 0 0.0 -26.2 Outlet Nozzle B6605-2 0.13 0.77 95.3 0.0161 X 10 19 0.1501 3.3 0 (14.3) 0 0 0.0 3.3 Outlet Nozzle B6605-3 0.09 0.79 58.0 0.0161 X 10 19 0.1501 10.1 0 (8.7) 0 0 0.0 10.1 Notes:
(a) Cu and Ni wt.% values, as well as CF values were obtained from WCAP-15571, Supplement 1, Revision 2 [Ref. B-4].
(b) Fluence values conservatively correspond to 50 EFPY fluence values at the lowest extent ofthe nozzle weld. FF values were calculated using Regulatory Guide 1.99, Revision 2.
(c) RTNDT(U) values were determined using the Alternative Approach 2 methodology as described in Appendix B ofBWRVIP-173-A.
(d) Calculated RTNDT values less than 25° F have been reduced to zero per TLR-RES/DE/CIB-2013-01 [Ref. B-6]. Actual calculated RTNDT values are listed in parentheses for these materials.
(e) Per Regulatory Guide 1.99, Revision 2, the base metal nozzle forging materials aA = l 7 ° F for Position 1.1 without surveillance data. However, aA need not exceed 0.5*RTNDT* This conclusion is consistent with footnote (d) since embrittlement is not considered for any ofthe six Beaver Valley Unit 1 nozzles; hence aA = 0.
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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- l 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 1 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 corners are provided in ORNL study, ORNL/TM-2010/246 [Ref. B-7], and are consistent with ASME PVP2011-57015 [Ref. B-8]. The methodology includes postulating an inside surface 1/4T nozzle corner flaw, and calculating through-wall nozzle corner 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 1/4T circular comer 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 1/4T 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 1/5T flaw is only used for the Inlet Nozzle B6608-1, and all other nozzles use the 1/4T 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 form:
where, cr = through-wall stress distribution x = through-wall distance from inside surface A0, A 1, A 2, A3 = coefficients of polynomial fit for the third-order polynomial, used in the stress intensity factor expression discussed below WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-5 The stress intensity factors generated for a rounded nozzle comer for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010/246. The stress intensity factor expression for a rounded comer is:
where, K1 stress intensity factor for a circular comer crack on a nozzle with a rounded inner radius comer a crack depth at the nozzle comer, for use with 1/4T (25% of the wall thickness)
The stress intensity factors for the inlet nozzle 1/ST 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 1/4T and 1/ST flaw bounds the potential non-conservatisms of evaluating K 1 at only the deepest point.
The Beaver Valley Unit 1 reactor vessel inlet and outlet nozzle P-T limit curves are shown in Figures B-1 and B-2, respectively, based on the stress intensity factor expression discussed above; also shown in these figures are the traditional beltline cooldown P-T limit curves from Figure 8-2. The nozzle P-T limit curves are provided for a cooldown rate of 100°F /hr, along with a steady-state curve.
An outside surface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve is also not provided since it would be less limiting than the cooldown nozzle P-T limit 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 comer.
Conclusion Based on the results shown in Figures B-1 and B-2, it is concluded that the nozzle P-T limits are bounded by the traditional beltline curves. Therefore, the P-T limits provided in Section 8 for 50 EFPY remain limiting for the beltline and non-beltline reactor vessel components.
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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 i-------
ti)
Q.
~
, 1250 I
ti) ti) a, 0..
E Sti) 1000 en
~
C:
<<s 0 750 0
u... Cooldown
....0
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Rates ca °F/Hr a, Steady-State a:: 500 -20
-40
-60
-100 Boltup 250 Temperature 0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperatu re {°F}
Figure B-1 Comparison of Beaver Valley Unit 1 Beltline P-T Limits to Inlet Nozzle Limits WCAP-18102 -NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-7 2500 . . . . - - - - - - -- - - - - - - - - - - - - - - - - - - - - - ,
2250 Outlet Nozzle Cooldown
-100 °F/hr 2000 Outlet Nozzle Steady State 1750 r ______...,._ -- -('**--*~ ~---- -
1500
~
fl)
C.
f
~ 1250 -
fl) f 0..
E
~ 1000 -
~
en C:
co 0 750 ---- - ., .. - -
I 0
...0
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...co
(.)
Rates
°F/Hr (1)
Steady-State a::: 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-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 B-8 B.3 REFERENCES B-1 NRC Regulatory Issue Summary 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," U.S.
Nuclear Regulatory Commission, October 14, 2014. [ADAMS Accession Number ML14149Al65}
B-2 Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.
B-3 Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.
Nuclear Regulatory Commission, May 1988.
B-4 Westinghouse Report WCAP-15571 Supplement 1, Revision 2, "Analysis of Capsule Y from the Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," September 2011.
B-5 BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.
B-6 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 ML14318A177]
B-7 Oak Ridge National Laboratory Report, ORNL/TM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles -
Revision 1," June 2012.
B-8 ASME PVP2011-57015, "Additional Improvements to Appendix G of ASME Section XI Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerville, July 2011.
B-9 Appendix G to the 2001 Edition through the 2003 Addenda of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."
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Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C OTHER REACTOR COOLANT PRESSURE BOUNDARY FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [Ref. C-1], requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature requirement for all RCPB components, which is specified in NB-2332(b) of the Section III ASME Code, is the relevant requirement that would affect the pressure-temperature (P-T) limits. The lowest service temperature (LST) requirement ofNB-2332(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 requirements ofNB-2332(b) are not applicable to the Beaver Valley Unit 1 P-T limits and the only ferritic RCPB components that are not part of the reactor vessel beltline or extended beltline consist of the replacement steam generators, the replacement reactor vessel closure head and the pressurizer.
The replacement steam generators (RSG) were designed and evaluated to the 1989 Edition Section III ASME Code and met all applicable requirements at the time of construction. Furthermore, the RSGs have not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for these components with regards to P-T limits.
The replacement reactor vessel closure head materials have been considered in the development of the P-T limits, see Section 6.3 of this report for further detail. The replacement reactor vessel closure head was constructed to the 1989 Edition Section III ASME Code and met all applicable requirements at the time of construction. Furthermore, the replacement reactor vessel closure head has not undergone neutron embrittlement that would affect P-T limits.
The pressurizer was constructed to the 1965 Edition through 1966 Winter Addenda Section III ASME Code and ASME Code Case-1401 and met all applicable requirements at the time of construction and is original to the plant. Furthermore, the pressurizer has not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for this component with regards to P-T limits.
C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"
U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, December 19, 1995.
C-2 ASME B&PV Code Section III, Division I, NB-2332, "Material for Piping Pumps, and Valves, Excluding Bolting Material."
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Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D BE AVER VALLEY UNIT 1 SURVEILL ANCE PROGR AM CREDIBILITY EVALU ATION D.1 INTRODUCTION Regulatory Guide 1.99, Revision 2 [Ref. D-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position 2.1 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature of reactor vessel beltline materials using surveillance capsule data. The methods of Position 2.1 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
The credibility of all surveillance program data previously applicable to the Beaver Valley Unit 1 reactor vessel was assessed in WCAP-17896-NP [Ref. D-2]. However, since this evaluation, additional weld Heat# 90136 surveillance capsule data from the Millstone Unit 2 surveillance program has been deemed applicable, and the Beaver Valley Unit I surveillance capsule fluence values have been updated. Thus, this Appendix documents the necessary updates to the credibility evaluation of surveillance program data applicable to Beaver Valley Unit 1.
The Millstone Unit 2 surveillance program includes two distinct welds, Heat# 90136 and Heat# 10137.
In previous analyses, this weld surveillance 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 individually deposited. It cannot be determined with full confidence how much of the overall surveillance weld is which weld metal heat and, furthermore, exactly which weld heat specimens are contained in which surveillance capsules in the Millstone Unit 2 program.
The Millstone Unit 2 ( combined) surveillance weld data met the second and third credibility criteria of Regulatory Guide 1.99, Revision 2 [Ref. D-1]. Additionally, Table D-2 of WCAP-16012 [Ref. D-3]
indicates that all of the measured weld RTNDT values were within the I -sigma scatter band; therefore, suggesting that there is good agreement between the measured capsule data and the embrittlement correlations. 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 differently for the two separate welds with different, as-measured, copper and nickel contents. However, since the (combined) weld material already passes the Regulatory Guide 1.99, Revision 2 credibility analysis, a re-evaluation of the material (as two separate heats) is not expected to significantly change the overall results of the subsequent reactor vessel integrity analyses. Thus, the surveillance weld metal will be considered to be only Heat# 90136 for the evaluations contained herein. All currently determined input data for Position 2.1 chemistry factor determination (See Section 5) and surveillance data credibility assessment documented in this Appendix will be used "as-is," as documented in the Millstone Unit 2 surveillance capsule analyses of record.
For conservatism, 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 remaining credible (see Section D.2). Additionally, the Beaver Valley Unit 1 intermediate to lower shell girth weld seam 11-714 (Heat# 90136) was assigned the most limiting calculated ART and RTPTs values during the evaluations contained in Section 7 and Appendix E, respectively. Thus, since the values determined using Position 2.1 are less conservative than the values determined using Position 1.1, the more conservative WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 D-2 Position 1.1 values were used. However, despite these additional conservatisms, 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 1 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 1 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 ofWCAP-17896-NP [Ref. D-3]. Note also that the credibility assessment of Heat# 305414 data remains valid as documented in Appendix D ofWCAP-17896-NP [Ref. D-2].
Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of RTNDT 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 E185-82 [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 RTNDT 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 recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 [Ref. D-6]. At this meeting, the NRC presented five cases. Of the five cases, Case 5 ("Surveillance Data from Other Sources Only") most closely represents the situation for the Beaver Valley Unit 1 reactor vessel intermediate to lower shell girth weld seam 11-714 (Heat # 90136) as described below. Of the five cases, Case 1
("Surveillance data available from plant but no other source") most closely represents the situation for the Beaver Valley Unit 1 surveillance materials as described below.
Heat # 90136 (Case 5) - This weld heat pertains to the intermediate to lower shell girth weld seam 11-714 in the Beaver Valley Unit 1 reactor vessel. This weld heat is not contained in the Beaver Valley Unit 1 surveillance program. However, it is contained in the St. Lucie Unit 1 and Millstone Unit 2 surveillance programs. NRC Case 5 per Reference D-6 is entitled "Surveillance Data from Other Sources Only" and most closely represents the situation for Beaver Valley Unit 1 weld Heat# 90136.
Lower Shell Plate B6903-1 (Case 1) - This plate material will be evaluated using the NRC Case 1 guidelines as described above.
Weld Heat # 305424 (Case 1) - This weld heat pertains to the intermediate shell longitudinal welds in the Beaver Valley Unit 1 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 1 weld Heat# 305424.
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Westinghouse Non-Proprietary Class 3 D-3 Credibility Assessment Case 5: Weld Heat# 90136 (St. Lucie Unit 1 Data Only)
Following the NRC Case 5 guidelines, the St. Lucie Unit 1 and Millstone Unit 2 surveillance weld metal (Heat# 90136) will be evaluated for credibility. Weld Heat # 90136 pertains to Beaver Valley Unit 1 reactor vessel intermediate to lower shell girth weld seam 11-714, but is not contained in the Beaver Valley Unit 1 surveillance program.
In accordance with the NRC Case 5 guidelines, the data from only St. Lucie Unit 1 will be analyzed first, since the irradiation environment for St. Lucie Unit I is judged closer to that of Beaver Valley Unit 1 as evidenced by the temperature adjustments documented in Table 4-2. This assessment was performed in Appendix D of WCAP-17896-NP [Ref. D-2] and concluded that the surveillance data for Heat# 90136 from St. Lucie Unit 1 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.
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Westinghouse Non-Proprietary Class 3 D-4 Credibility Assessment Case 5: Weld Heat# 90136 (All data)
In accordance 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 temperature 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 Temperature during Material Capsule Wt.%<a) Wt.%<a) Period of Irradiation (°F)'b) 97° 541 Weld Metal Heat# 90136 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) 830 548.0 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 1 and Millstone Unit 2 surveillance capsule data will be adjusted to the mean chemical composition and operating temperature calculated in Table D-1.
St. Lucie Unit 1 data CF Mean 121.2 ° F (calculated per Table 1 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. Weld (St. Lucie Unit I) 106.6 ° F (from Table 5-4)
Ratio = 121.2 + 106.6 = 1.14 (applied to St. Lucie Unit 1 surveillance data for weld Heat# 90136 in the credibility evaluation)
Millstone Unit 2 data CFl\.tean 12 l.2 ° F CF Surv. Weld (Millstone 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 temperature adjustments are as shown in Table D-2.
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Westinghouse Non-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 Temperature during Mean Operating Temperature Material Capsule Period of Irradiation (°F) Temperature (°F) Adjustment (°F) 97° 541 -4.3 Weld Metal Heat# 90136 ° 104 544.6 -0.7 (St. Lucie Unit 1 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) 0 83 548.0 2.7 Using the chemical composition and operating temperature adjustments described and calculated above, an interim chemistry factor is calculated for weld Heat # 90136 using the St. Lucie Unit 1 and Millstone Unit 2 data. This calculation is shown in Table D-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 Capsule ra> ARTNDT(c) FF*ARTNDT Material Capsule FF<h> FF2 (x 1019 n/cm2 , E > 1.0 MeV) {°F) (OF)
Weld Metal Heat# 97° 0.5174 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 1 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 0 1.74 1.1523 52.3 (56.09) 60.29 1.328 SUM: 365.59 5.437 CFHeat# 90136 = :I:(FF
RTNDT) + :I:(FF2) = (365.59) + (5.437) = 67.2 °F Notes:
(a) f = fluence.
O 28 - 01 O*Iog f)
(b) FF = fl uence factor = f .
(c) RTNDT values are the measured 30 ft-lb shift values. Each RTNDT value has first been adjusted according to the temperature adjustments summarized in Table D-2. Then, the RTNDT values for each surveillance weld data point are adjusted by the ratios determined previously for weld Heat # 90136 (pre-adjusted values are listed in parentheses and were taken from Table 4-2).
The scatter of RT NDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. D-1] is presented in Table D-4.
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Westinghouse Non-Proprietary Class 3 D-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 (Slopebest-r.t) (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 1 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 RTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. 0-1], should be less than 28 ° F for weld metal. Table 0-4 indicates that 100% (six out of six) of the surveillance data points fall within the +/- 1 cr of 28 ° F scatter band for surveillance weld materials. Therefore, the surveillance weld material (Heat # 90136) is deemed "credible" per the third criterion when all available data is considered.
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Westinghouse Non-Proprietary Class 3 D-7 Credibility Assessment Case I: Lower Shell Plate B6903- 1 and Weld Heat # 305424 Following the NRC Case I guidelines, the Beaver Valley Unit I surveillance plate and weld metal (Heat #
305424) will be evaluated for credibility. Note that when evaluating the credibility of the surveillance weld data, the measured LiRT NDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values. The chemistry factors for the Beaver Valley Unit 1 surveillance plate and weld material contained in the surveillance program were calculated in accordance with Regulatory Guide 1.99, Revision 2, Position 2.1 and are presented in Table D-5. The scatter of L\RTNDT values about the functional form of a best-fit line drawn as described in Regulatory 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 1 Surveillance Capsule Data Capsule ,<a>
ARTNDT(c) FF*ARTNDT Material Capsule (x 10 19 n/cm 2 , E FF<h> FF2 (OF) (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- l w 0.952 0.9862 147.7 145.66 0.973 (Longitudinal) 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
( Transverse) 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 B6903-1 = L(FF
ARTNDT) + 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: 981.01 5.577 CF Surv. Weld I{FF
ARTNDT) + I{FF2) (981.01) + (5.577) = ]75.9° F

Notes:
(a) f = fluence*
(b) FF = fluene factor = t'0 28 - 0* 1 o*Jog f).
(c) RTNDT values are the measured 30 ft-lb shift values.
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Westinghouse Non-Proprietary 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-tit) FF ARTNDT ARTNDT ARTNDT (Base Metal)
(x 10 19 n/cm2 ,
(OF) (OF) (OF) {°F) <28°F (Weld)
E>l.OMeV)
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 B6903-1 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 B6903-1 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, +/- 1 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 +/- 1 cr of l 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 +/- 1 cr 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 1 reactor vessel weld. The Position 2.1 chemistry factor calculation, as applicable to the Beaver Valley Unit 1 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 rTs calculations contained in Appendix E. However, consistent with the discussion in Section 0.1 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-17896-NP [Ref. D-2].
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Westinghouse Non-Proprietary 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 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.
D-3 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.
D-4 Code of Federal Regulations, IO CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.
D-5 ASTM El 85-82, "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 92-0 I and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, I 998.
[ADAMS Accession Number MLJ 10070570]
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Westinghouse Non-Proprietary Class 3 E-1 APPENDIX E PRESSURIZED THERMAL SHOCK EVALUATION E.1 PRESSURIZED THERMAL SHOCK CALCULATIONS Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break. Such transients may challenge the integrity of the reactor pressure vessel (RPV) under the following conditions: severe overcooling of the inside surface of the vessel wall followed by high repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect anywhere within the vessel wall.
In 1985, the U.S. NRC issued a formal ruling on PTS (10 CFR 50.61 [Ref. E-1]) that established screening criteria on PWR vessel embrittlement, as measured by the maximum reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTS* RTPTs screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end of license. The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RTPTs) 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 I RPV materials at 50 EFPY (EOLE).
The EOLE RTPTs calculations are summarized in Table E-1. The following changes and updates to the analysis of record for PTS at Beaver Valley Unit 1, WCAP-15571 Supplement 1, Revision 2 [Ref. E-3],
have been incorporated into the calculations contained in Table E-1 of this letter report.
1. Incorporation of the Capsule X results as documented in WCAP-17896-NP, Revision O [Ref. E-4], updated reactor vessel fluence values (See Section 2), surveillance capsule irradiated material testing results for Lower Shell Plate B6903-1 and Intermediate Shell Longitudinal Welds 19-714 A&B (Heat# 305424) (See Sections 4 and 5), and revised credibility conclusions (See Appendix D).
2. Incorporation of sister plant surveillance capsule test results for weld Heat # 90136 from the Millstone Unit 2 reactor vessel surveillance capsule program (See Sections 4, 5 and Appendix D).
Due to the uncertainty in the incorporation of the surveillance data from Millstone Unit 2 (two wire heats were used in the Millstone 2 surveillance weld, with some specimens being Heat #
90136 and others from another weld wire [Heat# 10137]), a full-margin term was used for this material in the RTPTs calculations contained in Table 1, even though the revised credibility analysis confirmed that Heat# 90136 remained credible.
3. Incorporation of revised initial reference nil-ductility transition temperature (RTNDT(u)) values for the four Beaver Valley Unit I reactor vessel plate materials as documented in Westinghouse Letter MCOE-LTR-15-15-NP, Revision I [Ref. E-5] (See Section 3). It was also concluded that the upper shell forging material for the Beaver Valley Unit 1 reactor vessel has an appropriate RTNDT(U) value even though Branch Technical Position (BTP) 5-3, Paragraph B1.1 (3) [Ref. E-6]
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WestinghouseN on-Proprietary Class 3 E-2 methodology for SA-508, Class 2 forging material must be used, due to lack ofclear definition of Charpy V-notch orientation. The initial RTNDT value ofthis 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 ofBWRVIP-173-A [Ref. E-7] to redefine the initial RTNDT values of the six Beaver Valley Unit I nozzle forging materials (See Section 3 and Appendix B).
5. Utilization ofthe following two conclusions from Section 4 ofTLR-RES/DE/CIB-2013-01 [Ref.
E-8], as appropriate:
1. The beltline is defined as the region of the RPV adjacent to the reactor core that is projected to receive a neutronfluence level of lxl017 nlcm2 (E > 1.0 MeV) or higher at the end of the licensed operating period.
2. Embrittlement effects may be neglected for any region of the RPV if either of the following conditions are met: (1) neutronfluence is less than lx10 17 n/cm2 (E > 1.0 MeV) at EOL, or (2) the mean value of AT30 estimated using an ETC acceptable to the staff is less than 25 °F at EOL. The estimate of AT30 at EOL shall be made using best-estimate chemistry values.
Therefore, embrittlement of reactor vessel materials with AT30 (which is equivalent to ARTNDT) values less than 25° F need not be considered in the subsequent RTPTs calculations documented in Table E-1.
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Westinghouse Non-Proprietary Class 3 E-3 Table E-1 RT PTS Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY Reactor Vessel Material and ID CF<a> EOLE Fluence<a) RTNDT(lJ}(a) ARTNDT(b) O"u O'A(c) Margin RTPTS(d)
Heat Number FF Number (OF) (n/cm2 , E > 1.0 MeV) (OF) (OF) (OF) (OF) (OF) {°F)
Reactor Vessel Beltline Materials Intermediate Shell Plate B6607-1 C4381-1 100.5 5.88 X 10 19 1.4330 26.8 144.0 0 17 34.0 204.8 Intermediate Shell Plate B6607-2 C4381-2 100.5 5.88 X 10 19 1.4330 53.6 144.0 0 17 34.0 231.6 Lower Shell Plate B6903-1 C6317-1 147.2 5.89 X 10 19 1.4333 13.1 211.0 0 17 34.0 258.1 Using Beaver Valley Unit I C6317-1 142.2 5.89 X 10 19 1.4333 13.1 203.8 0 17 34.0 250.9 surveillance data Lower Shell Plate B7203-2 C6293-2 98.7 5.89 X 10 19 1.4333 0.4 141.5 0 17 34.0 175.9 Intermediate to Lower Shell Girth 90136 124.3 5.88 X 10 19 1.4330 -56 178.1 17 28 65.5 187.6 Weld 11-714 Using St. Lucie Unit I and Millstone 90136 74.6 5.88 X 10 19 1.4330 -56 106.9 17 28 65.5 116.4 Unit 2 surveillance data Intermediate Shell Longitudinal Welds 305424 191.7 1.13 X 10 19 1.0341 -56 198.2 17 28 65.5 207.8 19-714 A&B Using Beaver Valley Unit I 305424 186.5 1.13 X 10 19 1.0341 -56 192.9 17 28 65.5 202.4 surveillance data Lower Shell Longitudinal Welds 305414 210.5 1.14 X 10 19 1.0366 -56 218.2 17 28 65.5 227.7 20-714 A&B Usinf; Fort Calhoun surveillance data 305414 216.9 1.14 X 10 19 1.0366 -56 224.8 17 28 65.5 234.4 Reactor Vessel Extended Beltline Materials Upper Shell Forging B6604 123V339VA1 84.2 0.718 X 10 19 0.9071 40 76.4 0 17 34.0 150.4 Upper Shell to Intermediate Shell 305414 210.5 0.718 X 10 19 0.9071 -56 190.9 17 28 65.5 200.5 Girth Weld 10-714 (3951 & 3958) 305414 Using Fort Calhoun surveillance data 216.9 0.718 X 10 19 0.9071 -56 196.7 17 28 65.5 206.3 (3951 & 3958)
AOFJ 41.0 0.718 X 10 19 0.9071 10 37.2 17 18.6 50.4 97.6 Upper Shell to Intermediate Shell FOIJ 41.0 0.718 X 10 19 0.9071 10 37.2 17 18.6 50.4 97.6 Girth Weld 10-714 (continued) EODJ 27.0 0.718 X 10 19 0.9071 10 0.0 (24.5) 17 0 34.0 44.0 HOCJ 27.0 0.718 X 10 19 0.9071 10 0.0 (24.5) 17 0 34.0 44.0 WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 E-4 Table E-1 RT PTs Calculations for the Beaver Valley Unit 1 Reactor Vessel Materials at 50 EFPY Reactor Vessel Material and ID CF<a> EOLE Fluence<a) RTNDT(U)(a) ARTNDT(b) au O't._(
c)
Margin RTPTS(d)
Heat Number FF Number (OF) (n/cm2, E > 1.0 MeV) (OF) (OF) (OF) ( OF ) (OF) (OF)
Inlet Nozzle B6608-1 95443-1 67.0 0.0210 X 10 19 0.1773 48.5 0.0(11.9) 0 0 0.0 48.5 Inlet Nozzle B6608-2 95460-1 67.0 0.0210 X 10 19 0.1773 -15.2 0.0(11.9) 0 0 0.0 -15.2 Inlet Nozzle B6608-3 95712-1 51.0 0.0210 X 10 19 0.1773 11.4 0.0(9.0) 0 0 0.0 11.4 EODJ 27.0 0.0210 X 10 19 0.1773 10 0.0(4.8) 17 0 34.0 44.0 FOIJ 41.0 0.0210 X 10 19 0.1773 10 0.0(7.3) 17 0 34.0 44.0 HOCJ 27.0 0.0210 X 1019 0.1773 10 0.0(4.8) 17 0 34.0 44.0 Inlet Nozzle Welds 1-717B, 1-717D, DBIJ 27.0 0.0210 X 1019 0.1773 10 0.0 (4.8) 17 0 34.0 44.0 1-717F EOEJ 20.0 0.0210 X 10 19 0.1773 10 0.0(3.5) 17 0 34.0 44.0 ICJJ 41.0 0.0210 X 1019 0.1773 10 0.0(7.3) 17 0 34.0 44.0 JACJ 54.0 0.0210 X 1019 0.1773 10 0.0 (9.6) 17 0 34.0 44.0 Outlet Nozzle B6605-1 95415-1 95.3 0.0161 X 10 19 0.1501 -26.2 0.0(14.3) 0 0 0.0 -26.2 Outlet Nozzle B6605-2 95415-2 95.3 0.0161 X 10 19 0.1501 3.3 0.0 (14.3) 0 0 0.0 3.3 Outlet Nozzle B6605-3 95444-1 58.0 0.0161 X 10 19 0.1501 IO.I 0.0(8.7) 0 0 0.0 10.1 ICJJ 41.0 0.0161 X 10 19 0.1501 10 0.0(6.2) 17 0 34.0 44.0 IOBJ 27.0 0.0161 X 1019 0.1501 10 0.0(4.1) 17 0 34.0 44.0 Outlet Nozzle Welds 1-717A, 1-7 l 7C, JACJ 54.0 0.0161 X 10 19 0.1501 10 0.0(8.1) 17 0 34.0 44.0 1-717E HOCJ 27.0 0.0161 X 10 19 0.1501 10 0.0(4.1) 17 0 34.0 44.0 EODJ 27.0 0.0161 X 10 19 0.1501 10 0.0(4.1) 17 0 34.0 44.0 FOIJ 41.0 0.0161 X 10 19 0.1501 10 0.0 (6.2) 17 0 34.0 44.0 Notes:
(a) CF values were taken from Table 5-4, fluence values were taken from Tables 2-5, 2-7, and 2-9, and RTNDT(U) values were taken from Table 3-2.
(b) Calculated L\RTNDT values less than 25°F have been reduced to zero per TLR-RES/DE/CIB-2013-01 [Ref. E-8]. Actual calculated L\RTNDT values are listed in parentheses for these materials.
(c) As discussed in Section 4, the surveillance plate and weld Heat # 305414 and # 305424 data were deemed non-credible. The surveillance weld data for Heat# 90136 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of 10 CFR 50.61 [Ref. E-1], the base metal er,.. = 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and the weld metal er,.. = 28°F for Position 1.1 and 2.1 with non-credible surveillance data. Since a full margin term will be used for Heat# 90136, er,..= 28°F with credible surveillance data for Position 2.1 for this weld heat. However, er,.. need not exceed 0.5*L\RTNDT*
(d) The 10 CFR 50.61 [Ref. E-1] methodology was utilized in the calculation of the PTS values.
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Westinghouse Non-Proprietary Class 3 E-5 E.2 PRESSURIZED THERMAL SHOCK CONCLUSIONS The Beaver Valley Unit 1 limiting RTPTS value for base metal or longitudinal weld materials at 50 EFPY is 258.1 ° F (see Table E-1), which corresponds to Lower Shell Plate B6903-1 (using Position 1.1). The Beaver Valley Unit 1 limiting RTPTs value for circumferentially oriented welds at 50 EFPY is 206.3 °F (see Table E-1), which corresponds to the Upper Shell to Intermediate Shell Girth Weld 10-714 (Heat # 305414, using Position 2.1).
Therefore, all of the beltline and extended beltline materials in the Beaver Valley Unit 1 reactor vessel are below the RTPTs screening criteria of 270 °F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds through EOLE (50 EFPY).
In the PTS analysis of record for Beaver Valley Unit 1, WCAP-15571 Supplement 1, Revision 2 [Ref. E-3], the limiting reactor vessel plate material, Lower Shell Plate B6903-1, was predicted to exceed the RTPTs screening criteria of 270 °F for plates at 39.6 EFPY of plant operation. However, with the reevaluation of the Beaver Valley Unit 1 reactor vessel beltline plate material initial RTNDT values [Ref.
E-5], along with incorporation of the Capsule X results [Ref. E-4], this material, while still the limiting material, is now predicted to remain under the RTPTs screening limit through EOLE. With consideration of the revised initial RTNDT value for Lower Shell Plate B6903- l, along with incorporation of the Capsule X results, this material is now predicted to remain under the RTPTS screening limit through a minimum of 80 EFPY of plant operation.
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Westinghouse Non-Proprietary Class 3 E-6 E.3 REFERENCES E-1 Code of Federal Regulations, 10 CFR 50.61, "Fracture Toughness Requirements 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 Westinghouse Report WCAP-15571 Supplement 1, Revision 2, "Analysis of Capsule Y from the Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," September 2011.
E-4 Westinghouse Report WCAP-17896-NP, Revision 0, "Analysis of Capsule X from the FirstEnergy Nuclear Operating Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," September 2014.
E-5 Westinghouse Letter MCOE-LTR-15-15-NP, Revision 1, "Determination of Unirradiated RTNDT Values of the Four Beaver Valley Unit 1 Reactor Vessel Beltline Plate Materials Using a Hyperbolic Tangent Curve Fit," dated July 6, 2015.
E-6 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements," Revision 2, U.S. Nuclear Regulatory Commission, March 2007 E-7 BWRVIP-173-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835.
E-8 U.S. NRC Technical Letter Report TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels," Office of Nuclear Regulatory Research [RES],
dated November 14, 2014. [ADAMS Accession Number ML14318A177}
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Westinghouse Non-Proprietary Class 3 F-1 APPENDIX F VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS F.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn from service to date at Beaver V alley 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, "Calculational 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 measurements 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 1 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>l.O MeV) Displacements Capsule ID Location Time [EFPY] [n/cm2] [dpa]
V 15° End of Cycle 1 1.2 2.97E+l8 4.93E-03 u 25° End of Cycle 4 3.6 6.18E+l8 l.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.IOE+19 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-Proprietary Class 3 F-2 Reaction of Sensor Material Interest Capsule V Capsule U Capsule W Capsule Y Capsule X Copper-63 6 3Cu(n,a)6°Co X X X X X 54 54 Iron-54 Fe(n,p) Mn X X X X X Nickel-58 58 Ni(n,p)58Co X X X X X Uranium-238 238U(n,f)131Cs X X LOST X X Neptunium-237 237Np(n,f) mes X X X X X Cobalt-Aluminum* 59 Co(n;y)6°co X X X X X
The cobalt-aluminum measurements for this plant include both bare and cadmium-covered wire sensors.
Pertinent physical and nuclear characteristics 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 fluence rate at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron fluence rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron fluence rate level incident on the various monitors may be derived from the activation measurements 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 characteristics 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-aluminum 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 1 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 representation 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 Xis 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:
R = --------------- A n p No F YL -J C1 [J-e-At;J [e-Atd,J J i=I Pref where:
R Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Prer(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.
Average core power level during irradiation periodj (MW).
Pref Maximum or reference power level of the reactor (MW).
CJ Calculated ratio of <j>(E > 1.0 MeV) during irradiation period j to the time weighted average <j>(E > 1.0 Me V) over the entire irradiation period.
A Decay constant of the product isotope (1/sec).
tJ Length of irradiation periodj (sec).
Decay time following irradiation periodj (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 ]/[Pred accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. T he 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, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-4 employing low-leakage fuel management, the additional Cj 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-leakage 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-specific neutron fluence rate values along with the computed values for Cj are listed in Table F-3. These fluence rate values represent the cycle-dependent 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 1 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 1 sensor reaction rates are summarized as follows:
Correction Factor Capsule V Capsule U Capsule W Capsule Y Capsule X 63Cu(n,a) Radial Gradient 0.956 0.956 0.956 0.956 0.956 54Fe(n,p) Radial Gradient 1.050 1.051 1.051 1.051 1.050 58Ni(n,p) Radial Gradient 1.158 1.163 1.163 1.163 1.158 238U(n,f) Radial Gradient 1.000 1.000 NIA 1.000 1.000 237Np(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 238U and 237Np sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the Beaver Valley Unit 1 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 NIA 0.806 0.712 23sU(y,f) 0.955 0.960 NIA 0.960 0.956 Net 238U Correction Factor 0.834 0.826 NIA 0.774 0.681 z31Np(y,f) 0.982 0.983 0.984 0.984 0.983 WCAP-18102-NP June 2017 Revision 0
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 238U impurities, plutonium build-in, and gamma-ray-induced 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 (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:
Ri +/- 8R i = )crig +/- 8 0-)(cp g +/- 8 <p) g relates a set of measured reaction rates, Ri, to a single neutron spectrum, <pg, through the multigroup dosimeter reaction cross section, <Jig, each with an uncertainty 8. 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 1 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
( (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:
1. 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 multiple foil set.
3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.
For the Beaver Valley Unit 1 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 1 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 6
3Cu(n,a)6°Co 5%
54 54 Fe(n,p) Mn 5%
58 Ni(n,p)58Co 5%
23sU(n,t)131Cs 10%
231Np(n,t)131Cs 10%
59 Co(n,y)6°Co 5%
These uncertainties are given at the 1cr 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 MeV neutron sources.
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Westinghouse Non-Proprietary Class 3 F-7 For sensors included in the Beaver Valley Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross sections are provided in the SNLRML documentation package.
Reaction Uncertainty 6
3Cu(n,a)6°Co 4.08-4.16%
54 Fe(n,p) 54Mn 3.05-3.11 %
58 58 Ni(n,p) Co 4.49-4.56%
23sU(n,f)131Cs 0.54-0.64%
237Np(n,f)mCs 10.32-10.97%
6° 59 Co(n,y) 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-squares 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:
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, 2
H = (g-g') 2 2y WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-8 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations 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 Normalization Uncertainty (Rn) 15%
Fluuence Rate Group Uncertainties (Rg, Rg)*
(E > 0.0055 MeV) 15%
(0.68 eV < E < 0.0055 MeV) 25%
(E < 0.68 eV) 50%
Short Range Correlation (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 Correlation 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-squares evaluations of the dosimetry from the Beaver Valley Unit 1 surveillance capsules withdrawn to date are provided in Tables F-5 and F-6. In Table F-5, measured, calculated, and best-estimate 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-squares adjusted reaction rates.
These ratios of MIC and measured-to-best-estimate (M/BE) illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. In Table F-6, comparison of the calculated and best-estimate values of neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the best-estimate-to-calculated (BE /C) ratios observed for each of the capsules.
The data comparisons provided in Tables F-5 and F-6 show that the adjustments to the calculated spectra are relatively small and well within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross sections. Further, these results indicate that the use of the least-squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance 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 displacements at the surveillance capsule locations is specified as 12% at the 1cr level. From Table F-6, it is noted that the corresponding uncertainties associated with the least-squares adjusted exposure parameters have been reduced to 6% for neutron fluence rate (E > 1.0 MeV) and 6-7% for iron atom displacement rate. Again, the uncertainties from the least-squares evaluation are at the I cr level.
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Westinghouse Non-Proprietary Class 3 F-9 Further 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 evaluation 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 1 data is 1.01 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 MeV) 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 1 reactor pressure vessel.
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Westinghouse Non-Proprietary Class 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 (MeVia> Half-life (%)
63 Copper-63 Cu (n,a) 0.6917 5.0-11.9 5.272 y n/a 54 Iron-54 Fe (n,p) 0.0585 2.2 8.5 312.11 d n/a 58 Nickel-58 Ni (n,p) 0.6808 1.7-8.4 70.82 d n/a Uranium-238 238U (n,t) 1.0000 1.4 7.2 30.07 y 6.02 Neptunium-237 231Np (n,t) 1.0000 0.4-4.8 30.07 y 6.17 59 Cobalt-Aluminum Co (n,y) 0.0015 non-threshold 5.272 y n/a Note:
(a) The 90% response range is defined such that, in the neutron spectrum characteristic of the Beaver Valley Unit 1 surveillance capsules, approximately 90% of the sensor response is due to neutrons in the energy range specified with approximately 5% of the total response due to neutrons with energies below the lower limit and 5% of the total response due to neutrons with energies above the upper limit.
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Westinghouse Non-Proprietary Class 3 F-11 Table F-2 Monthly T hermal 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 June 2017 Revision 0
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 1941312 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 1689361 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 1546013 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 1903021 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 June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-13 Table F-2 Monthly T hermal 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-01 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-01 1684701 Nov-03 1817339 Sep-96 1901466 Feb-99 1019933 Jul-01 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 June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-14 Table F-2 Monthly T hermal 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-10 1924250 Sep-12 2084237 Jul-05 1998525 Dec-07 2155422 May-IO 2153565 Oct-12 2153783 Aug-05 1998289 Jan-08 2154241 Jun-IO 2084317 Nov-12 2087475 Sep-05 1932792 Feb-08 1925350 Jul-IO 2152557 Dec-12 2154578 Oct-05 2000893 Mar-08 2106637 Aug-IO 2153379 Jan-13 2154191 Nov-05 1933689 Apr-08 2084532 Sep-IO 2083825 Feb-13 1758987 Dec-05 1998375 May-08 2154156 Oct-IO 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-II 2154831 Jun-13 2084869 Apr-06 643633 Sep-08 2084264 Feb-II 1944444 Ju l-13 2154058 May-06 1833886 Oct-08 2154327 Mar-II 2151515 Aug-13 2153987 Jun-06 1934547 Nov-08 2087649 Apr-II 1752956 Sep-13 1935163 Jul-06 1998137 Dec-08 2154107 May-II 2154209 Oct-13 0 Aug-06 1666197 Jan-09 2154018 Jun-I 1 2084557 Nov-13 1397606 Sep-06 1864587 Feb-09 1945788 Jul-II 2147118 Dec-13 2155010 Oct-06 2062087 Mar-09 2150315 Aug-II 2154308 Jan-14 528851 Nov-06 1991907 Apr-09 1236557 Sep-II 2084682 Feb-14 1891737 Dec-06 2058622 May-09 661421 Oct-11 2152575 Mar-14 2151295 Jan-07 2058939 Ju n-09 2085427 Nov-II 2087006 Apr-14 2084711 Feb-07 1828585 Jul-09 2155082 Dec-II 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 2139605 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-IO 2153807 Jun-12 2085729 Nov-14 2087509 Sep-07 1514542 Feb-IO 1944710 Jul-12 2151237 Dec-14 2153981 WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-15 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-15 2154143 Feb-15 1945558 Mar-15 2151145 Apr-15 1245394 May-15 473968 Jun-15 1922472 Jul-15 2155674 Aug-15 2155704 Sep-15 2085592 Oct-15 2154801 Nov-15 2088157 Dec-15 2155053 Jan-16 2154801 Feb-16 2015642 Mar-16 2151916 Apr-16 2084943 May-16 2154636 Jun-16 2084938 Jul-16 2153963 Aug-16 2149558 Sep-16 1365500 WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Propriet ary Class 3 F-16 Table F-3 Calculated Fast Neutron (E > 1.0 MeV) Fluence Rate and Cj Factors at the Surveillance Capsule Center, Core Midplane Elevation Cycle <p(E > 1.0 MeV) (n/cm2-s]
Length Fuel Cycle (EFPS) Capsule V Capsule U CapsuleW Capsule Y Capsule X 1 3.66E+07 8.l0E+lO 5.43E+l0 5.43E+10 5.43E+l0 8.lOE+l0 2 2.26E+07 5.62E+l0 5.62E+10 5.62E+l0 8.27E+l0 3 2.49E+07 - 6.21E+l0 6.21E+10 6.21E+l0 9.38E+l0 4 2.91E+07 - 4.73E+l0 4.73E+10 4.73E+l0 7.14E+l0 5 3.76E+07 - 4.56E+10 4.56E+l0 6.94E+10 6 3.51E+07 - 4.63E+10 4.63E+l0 6.15E+l0 7 3.95E+07 - - - 4.47E+l0 6.80E+l0 8 3.48E+07 - - - 4.70E+l0 6.98E+l0 9 4.35E+07 - - - 4.60E+l0 6.46E+10 10 3.77E+07 - 3.86E+l0 4.86E+10 11 3.05E+07 - - - 4.06E+l0 4.78E+l0 12 3.58E+07 - - - 4.45E+l0 5.18E+10 13 4.31E+07 - - - 4.17E+10 5.23E+10 14 4.16E+07 - - - 4.54E+10 15 4.19E+07 - - - - 4.30E+10 16 4.56E+07 - - - - 5.25E+l0 17 3.89E+07 - - - - 5.02E+l0 18 4.39E+07 - - - 5.69E+l0 19 4.65E+07 - - - - 5.42E+10 20 4.27E+07 - - - 5.82E+l0 21 4.44E+07 - - - - 5.82E+l0 22 4.33E+07 - - - - 5.55E+10 TimeWeighted Average Fluence Rate 8.lOE+lO 5.46E+10 5.12E+ 10 4.66E+l0 5.94E+l0 WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-17 Table F-3 Calculated Fast Neutron (E > 1.0 MeV) Fluence Rate and Cj Factors at the Surveillance Capsule Center, Core Midplane Elevation Cycle cj Length Fuel Cycle [EFPS] Capsule V Capsule U Capsule W Capsule Y Capsule X 1 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 4.31E+07 - - - 0.895 0.881 14 4.16E+07 - - - - 0.765 15 4.19E+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-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-18 Table F-4a Measured Sensor Activities and Reaction Rates of Surveillance Capsule V Adjusted Reaction Measured Activity<a) Saturated Activity Rate <b)
Reaction Location (dps/g) (dps/g) (rps/atom) 6 3 Cu (n,a) 6 °Co Top-Mid 4.24£+04 3.68£+05 5.62£-17 Middle 4.42£+04 3.84£+05 5.86£-17 Bot-Mid 4.28£+04 3.72£+05 5.67£-17 A verage 5.72E-17 54 54 Fe (n,p) Mn Top 5.84£+05 3.94£+06 6.25£-15 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 5.70£-15 Bottom 5.37£+05 3.62£+06 5.75£-15 Average 5.89E-15 58 58 Ni (n,p) Co Top-Mid 9.57E+05 5.64£+07 8.07E-15 Middle 9.75E+05 5.74E+07 8.22E-l5 Bot-Mid 9.06E+05 5.34£+07 7.64£-15 Average 7.98E-15 8
23 U (n,t) 137Cs (Cd) Middle 1.36£+05 5.39E+06 3.54£-14 5 9 Including 23 U, 23 Pu, and y fission corrections: 2.95E-14 237 Np (n,t) 137Cs (Cd) Middle 9.70E+05 3.84£+07 2.45£-13 Including y fission corrections: 2.41E-13 59 6 Co (n,y) °Co Top 7.24E+06 6.30E+07 4.l IE-12 Bottom 7.24£+06 6.30E+07 4.IIE-12 Average 4.llE-12 59 6 Co (n,y) °Co (Cd) Top 2.59E+06 2.72£+07 l.78E-12 Bottom 2.55E+06 2.68E+07 1.75£-12 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 1 rated reactor power of 2652 MWt.
WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-19 Table F-4b Measured Sensor Activities and Reaction Rates of Surveillance Capsule U Adjusted Reaction Measured Activity<a) Saturated Activity Rate<h)
Reaction Location (dps/g) (dps/g) (rps/atom) 63 Cu (n,a) 6°Co Top-Mid 9.44E+04 3.01E+05 4.60E-l7 Middle l.01E+05 3.22E+05 4.92E-l7 Bot-Mid 9.30E+04 2.97E+05 4.53E-l7 Average 4.68E-17 54 54 Fe (n,p) Mn Top l.21E+06 2.82E+06 4.47E-15 Top-Mid l.13E+06 2.63E+06 4.l7E-15 Middle l.22E+06 2.84E+06 4.50E-15 Bot-Mid l.16E+06 2.70E+06 4.28E-15 Bottom l.14E+06 2.65E+06 4.21E-15 Average 4.32E-15 58 Ni (n,p) 58Co Top-Mid 4.81E+06 3.98E+07 5.70E-15 Middle 5.22E+06 4.32E+07 6.19E-15 Bot-Mid 4.92E+06 4.08E+07 5.84E-15 Average 5.91E-15 238U (n,f) mes (Cd) Middle 2.89E+05 3.81E+06 2.50E-14 235 239 Including U, Pu, and y fission corrections: 2.06E-14 237 Np (n,f) mes (Cd) Middle 2.14E+06 2.82E+07 l.80E-13 Including y fission corrections: 1.77E-13 59 Co (n,y) 6°Co Top l.17E+07 3.74E+07 2.44E-12 Bottom l.16E+07 3.71E+07 2.42E-12 Average 2.43E-12 59 Co (n,y) 6°Co (Cd) Top 4.27E+06 l.65E+07 l.08E-12 Bottom 4.18E+06 l.61E+07 l.05E-12 Average 1.06E-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 June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-20 Table F-4c Measured Sensor Activities and Reaction Rates of Surveillance Capsule W Adjusted Reaction Measured Activity<a) Saturated Activity Rate <h)
Reaction Location (dps/g) (dps/g) (rps/atom) 63 6 Cu (n,a) °Co Top-Mid 1.14E+05 2.65E+05 4.03E-17 Middle 1.20E+05 2.78E+05 4.25E-17 Bot-Mid 1.18E+05 2.74E+05 4.18E-17 Average 4.15E-17 54Fe (n,p) S4Mn Top 1.00E+06 2.55E+06 4.05E-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 Bottom 9.19E+05 2.35E+06 3.72E-15 Average 3.73E-15 58Ni (n,p) 58Co Top-Mid 2.17E+06 3.47E+07 4.96E-15 Bot-Mid 2.20E+06 3.51E+07 5.03E-15 Average 5.00E-15 237 Np (n,f) 137 Cs (Cd) Middle 2.57E+06 2.14E+07 1.37E-13 Including y fission corrections: 1.34E-13 59Co (n,y) 6°Co Top 1.36E+07 3.16E+07 2.06E-12 Bottom 1.40E+07 3.25E+07 2.12E-12 Average 2.09E-12 59Co (n,y) 6°Co (Cd) Top 4.93E+06 1.38E+07 9.03E-13 Bottom 4.99E+06 1.40E+07 9.14E-13 Average 9.0SE-13 Notes:
(a) Measured specific activities are indexed to a counting date of August 10, 1988.
(b) Reaction rates are referenced to the Cycles 1-6 average rated reactor power of 2652 MWt.
WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-21 Table F-4d Measured Sensor Activities and Reaction Rates of Surveillance Capsule Y Adjusted Reaction Measured Activity<a) Saturated Activity Rate<h)
Reaction Location (dps/g) (dps/g) (rps/atom) 63Cu (n,a) 6°Co Top-Mid l.56E+05 2.50E+05 3.81E-17 Middle l.67E+05 2.67E+05 4.08E-l7 Bot-Mid l.56E+05 2.50E+05 3.81E-l 7 Average 3.90E-17 54Fe (n,p) 54Mn Top l.42E+06 2.40E+06 3.81E-15 Top-Mid l.35E+06 2.29E+06 3.63£-15 Middle l.43E+06 2.42E+06 3.84£-15 Bot-Mid l.37E+06 2.32E+06 3.68£-15 Bottom l.33E+06 2.25E+06 3.57£-15 Average 3.71E-15 58Ni (n,p) 58Co Top-Mid l.71E+07 3.47E+07 4.97£-15 Middle l.84E+07 3.74E+07 5.35£-15 Bot-Mid l.73E+07 3.51E+07 5.03£-15 Average 5.llE-15 238 U (n,f) mes (Cd) Middle 7.79E+05 3.03E+06 1.99£-14 239 Including 235U, Pu, and y fission corrections: 1.54E-14 237Np (n,f) mes (Cd) Middle 5.39E+06 2.10E+07 1.34£-13 Including y fission corrections: 1.31E-13 s9Co (n,y) 6oCo Top l.81E+07 2.90E+07 1.89£-12 Bottom l.61E+07 2.58E+07 1.68£-12 Average 1.79E-12 59Co n,y) 6°Co (Cd) Top 6.78E+06 l.31E+07 8.57£-13
(
Bottom 6.79E+06 l.32E+07 8.58£-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 1-13 average rated reactor power of 2652 MWt.
WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-22 Table F-4e Measured Sensor Activities and Reaction Rates of Surveillance Capsule X Adjusted Reaction Measured Activity(a) Saturated Activity Rate(b)
Reaction Location (dps/g) (dps/g) (rps/atom) 6 6 3Cu (n,a) °Co Top-Mid 2.18E+05 2.75E+05 4.l9E-l7 Middle 2.31E+05 2.91E+05 4.44E-l7 Bot-Mid 2.20E+05 2.77E+05 4.23E-l7 Average 4.28E-17 54 Fe (n,p) 54Mn Top l.53E+06 2.67E+06 4.24E-15 Top-Mid l.46E+06 2.55E+06 4.04E-15 Middle l.59E+06 2.77E+06 4.40E-15 Bot-Mid l.55E+06 2.70E+06 4.29E-15 Bottom l.64E+06 2.86E+06 4.54E-15 Average 4.30E-15 58 Ni (n,p) 58Co Top-Mid 5.73E+06 4.36E+07 6.25E-15 Middle 6.10E+06 4.65E+07 6.65E-15 Bot-Mid 5.87E+06 4.47E+07 6.40E-15 Average 6.43E-15 8
23 U (n,f) mes (Cd) Middle 1.87E+06 4.47E+06 2.94E-14 5
Including 23 U, 239Pu, and y fission corrections: 2.00E-14 7
23 Np (n,f) mes (Cd) Middle l.09E+07 2.61E+07 l.66E-13 Including y fission corrections: 1.63E-13 59 Co (n,y) 6°Co Top 2.82E+07 3.56E+07 2.32E-12 Bottom 3.01E+07 3.80E+07 2.48E-12 Average 2.40E-12 59 Co (n,y) 6°Co (Cd) Top l.84E+07 2.81E+07 1.83E-12 Bottom l.24E+07 l.89E+07 1.23E-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.
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Westinghouse Non-Proprietary Class 3 F-23 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsule V Reaction Rate [rps/atom]
Reaction Measured Calculated Best-Estimate MIC M/BE 63 Cu(n,cx)6°Co 5.72£-17 5.50£-17 5.60£-17 1.04 1.02 S4Fe(n,p)s4Mn 5.89£-15 6.02£-15 5.98£-15 0.98 0.99 58 58 Ni(n,p} Co 7.98£-15 8.24£-15 8.15£-15 0.97 0.98 z3 sU(n,f) 131Cs (Cd) 2.95£-14 2.87£-14 2.88£-14 1.03 1.02 37 z Np(n,f)131Cs (Cd) 2.41£-13 2.14£-13 2.28£-13 1.13 1.05 59Co(n,y)6° Co 4.lIE-12 4.47£-12 4.12£-12 0.92 1.00 59 Co(n,y}6°Co (Cd) 1.76£-12 1.72£-12 1.76£-12 1.03 1.00 Note:
See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.
Capsule U Reaction Rate [rps/atom)
Reaction Measured Calculated Best-Estimate MIC M/BE 63 Cu(n,cx)6°Co 4.68£-17 4.30£-17 4.53£-17 1.09 1.03 54 54 Fe(n,p) Mn 4.32£-15 4.41£-15 4.46£-15 0.98 0.97 58 Ni(n,p}58 Co 5.91£-15 5.99£-15 6.05£-15 0.99 0.98 238U(n,f) 137 Cs (Cd) 2.06£-14 1.99£-14 2.05£-14 1.04 1.01 237Np(n,f) 137 Cs (Cd) 1.77£-13 1.40£-13 1.61£-13 1.27 1.10 59 Co(n,y}6° Co 2.43£-12 2.59£-12 2.44£-12 0.94 1.00 59Co(n,y}6°Co (Cd) 1.06£-12 l.0IE-12 1.06£-12 1.06 1.00 Note:
See Section F. I .2 for details describing the Best-Estimate (BE) reaction rates.
WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-24 Table F-5 Comparison of Measured, Calculated, and Best-Estimate Reaction Rates at the Surveillance Capsule Center Capsule W Reaction Rate [rps/atom]
Reaction Measured Calculated Best-Estimate MIC M/BE 6
3Cu(n,a)6° Co 4.l 5E-l 7 4.09E-17 4.00E-17 1.02 1.04 54 54 Fe(n,p) Mn 3.73E-15 4.16E-15 3.81E-15 0.90 0.98 58 Ni(n,p)58 Co 5.00E-15 5.64E-15 5.14E-15 0.89 0.97 237Np(n,f)131Cs (Cd) l.34E-13 l.3IE-13 l.27E-13 1.03 1.05 59 Co(n,y)6°Co 2.09E-12 2.42E-12 2.IOE-12 0.86 1.00 59 6° Co(n,y) Co (Cd) 9.08E-13 9.41E-13 9.07E-13 0.97 1.00 Note:
See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.
Capsule Y Reaction Rate [rps/atom]
Reaction Measured Calculated Best-Estimate MIC M/BE 6
3Cu(n,a)6°Co 3.90E-l7 3.78E-17 3.84E-17 1.03 1.02 54 Fe(n,p)54Mn 3.71E-15 3.81E-15 3.75E-15 0.97 0.99 58 58 Ni(n,p) Co 5.1IE-15 5.l7E-15 5.1OE-15 0.99 1.00 238U(n,f)137 Cs (Cd) l.54E-14 l.71E-14 l.67E-14 0.90 0.92 231Np(n,f)131Cs (Cd) l.3IE-13 1.19E-13 l.24E-13 1.11 1.06 59 Co(n,y)6°Co l.79E-12 2.19E-12 1.80E-12 0.82 0.99 59 Co(n,y)6°Co (Cd) 8.58E-13 8.5 IE-13 8.54E-13 1.01 1.00 Note:
See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.
W CAP-18102-NP June 2017 Revision 0
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 [rps/atom]
Reaction Measured Calculated Best-Estimate MIC M/BE 6
3Cu(n,a ) 6°Co 4.28E-17 4.34E-l 7 4.28E-17 0.99 1.00 54 54 Fe(n,p ) Mn 4.30E-15 4.56E-15 4.45E-15 0.94 0.97 58 Ni(n,p ) 58Co 6.43E-15 6.22E-15 6.21E-15 1.03 1.03 23sU(n,f)131Cs (Cd) 2.00E-14 2.13E-14 2.l0E-14 0.94 0.95 231Np(n,f)131Cs (Cd) l.63E-13 l.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 59 Co(n,y)6°Co (Cd) l.53E-12 1.22E-12 1.S0E-12 1.26 1.02 Note:
See Section F.1.2 for details describing the Best-Estimate (BE) reaction rates.
WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-26 Table F-6 Comparison of Calculated and Best-Estimate Exposure Rates at the Surveillance Capsule Center cp(E > 1.0 MeV) [n/cm2-s]
Calculated Best-Estimate Best-Estimate BE/C Capsule ID Uncertainty (la')
V 8.12E+IO 8.l 9E+IO 6 1.00 u 5.47E+10 5.68E+IO 6 1.03 w 5.13E+IO 4.69E+IO 6 0.91 y 4.67E+10 4.58E+IO 6 0.98 X 5.95E+IO 5.89E+IO 6 0.99 Note:
See Section F.1.2 for details describing the Best-Estimate (BE) exposure rates.
Iron Atom Displacement Rate [dpa/s]
Calculated Best-Estimate Best-Estimate BE/C Capsule ID Uncertainty (lo')
V l.33E-10 l.35E-10 7 1.01 u 8.70E-ll 9.17E-ll 7 1.05 w 8.16E-ll 7.60E-l1 7 0.93 y 7.42E-11 7.33E-1l 6 0.98 X 9.67E-ll 9.60E-ll 7 0.99 Note:
See Section F.1.2 for details describing the Best-Estimate (BE) exposure rates.
WCAP-18102-NP June 2017 Revision 0
Westinghouse Non-Proprietary Class 3 F-27 Table F-7 Comparison of Measured/Calculated (MIC) Sensor Reaction Rate Ratios Including all Fast Neutron T hreshold Reactions MIC Rations Capsule Capsule Capsule Capsule Capsule Average % Std Dev Reaction V u w y X 63 Cu(n,a)6°Co 1.04 1.09 1.02 1.03 0.99 1.03 3.5 54 54 Fe(n,p) Mn 0.98 0.98 0.90 0.97 0.94 0.95 3.6 58 Ni(n,p)58Co 0.97 0.99 0.89 0.99 1.03 0.97 5.3 23sU(n,f) 131Cs (Cd) 1.03 1.04 - 0.90 0.94 0.98 7.0 237Np(n,f) 137Cs (Cd) 1.13 1.27 1.03 1.11 1.05 1.12 8.4 Note:
The overall average M/C ratio for the set of24 sensor measurements is 1.01 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 (j>(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-NP June 2017 Revision 0
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 1, Revision 2, Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program, September 2011.
F-6 WCAP-17896-NP, Revision 0, Analysis of Capsule X from 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
£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 1 reactor vessel.
Table G-1 Surveillance Capsule Withdrawal Schedule Capsule Capsule Lead Withdrawal Capsule Fluence Capsule Status(a)
Location Factor(a) EFPY(b,c) (n/cm 2, E > 1.0 MeVic>
Withdrawn V 165° 1.47 1.2 2.97E+l8 (EOC 1)
Withdrawn u 65° (EOC 4) 1.00 3.6 6.18E+18 Withdrawn w 245° (EOC 6) 1.05 5.9 9.52E+18 Withdrawn y 295° 1.14 14.3 2.10E+19 (EOC 13)
Withdrawn X 285° 1.57 26.6 4.99E+19 (EOC 22) 285 ° s(d> In Reactor 0_74(d) Note (d) 2.58E+19<<l>
(45°/295°)
T<e> 65° (55° ) In Reactor 0.94(e) Note (e) 3.28E+19(e>
z<0 165° (305° ) In Reactor 1.20<0 Note (t) 4.18E+19<0 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 fluence value and lead factor are accumulated through EOC 24. Capsule S should remain in the reactor. If additional metallurgical data is needed for Beaver Valley Unit 1, such as in support of a second license renewal to 80 total years of operation, withdrawal and testing of Capsule S should be considered.
(e) Capsule T was moved to the Capsule U location at the EOC 10. Reported fluence 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 10. Reported fluence 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/cm2 (E > 1.0 MeV).
G.1 REFERENCES G-1 ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," ASTM, July 1982.
G-2 Code of Federal Regulations 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Federal Register, Volume 60, No. 243, dated December 19, 1995.
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v t e Letter Sequence Request
EPID:L-2017-LLR-0131, Modified Rt PTS Values and Reactor Vessel Surveillance Capsule Withdrawal Schedule (Approved, Closed) EPID:L-2017-LLR-0137, Modified Rt PTS Values and Reactor Vessel Surveillance Capsule Withdrawal Schedule (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Administration Questions 18 December 2017 19 December 2017 Answers 5 January 2018 Results Approval Other:
MONTHYEAR2017-12-19 [Table View]

L-17-292, Modified Rt PTS Values and Reactor Vessel Surveillance Capsule Withdrawal Schedule

Text

SUBJECT:

Enclosures:

SUMMARY

3.0 PROPOSED CHANGE

4.0 TECHNICAL EVALUATION

5.0 CONCLUSION

SUMMARY

3.0 PROPOSED CHANGE

4.0 TECHNICAL EVALUATION

5.0 CONCLUSION

SUMMARY

2.1 INTRODUCTION

SUMMARY

2.1 INTRODUCTION

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L-17-292, Modified Rt PTS Values and Reactor Vessel Surveillance Capsule Withdrawal Schedule

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SUBJECT:

Enclosures:

SUMMARY

3.0 PROPOSED CHANGE

4.0 TECHNICAL EVALUATION

5.0 CONCLUSION

SUMMARY

3.0 PROPOSED CHANGE

4.0 TECHNICAL EVALUATION

5.0 CONCLUSION

SUMMARY

2.1 INTRODUCTION

SUMMARY

2.1 INTRODUCTION

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