ML23024A154
| ML23024A154 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 01/23/2023 |
| From: | James Holloway Dominion Energy South Carolina |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 22-227 | |
| Download: ML23024A154 (1) | |
Text
Dominion Energy South Carolina, Inc.
5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion DominionEnergy.com January 23, 2023 Energy Attn: Document Control Desk Serial No.: 22-227 U.S. Nuclear Regulatory Commission NRANG: RO Washington, DC 20555-0001 Docket No.: 50-395 License No.: NPF-12 DOMINION ENERGY SOUTH CAROLINA VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 PROPOSED REACTOR VESSEL SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE TO SUPPORT POTENTIAL SUBSEQUENT LICENSE RENEWAL ACTIVITY Pursuant to the provisions of Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50), Appendix H, Dominion Energy South Carolina, Inc. (DESC), hereby submits the attached request for approval to add a new surveillance capsule withdrawal schedule in the Virgil C. Summer Nuclear Station (VCSNS) Updated Final Safety Analysis Report (UFSAR) Section 5.4.3.6.2.3, Removal Schedule.
10 CFR 50, Appendix H, Section 111.B.3 requires that changes to the withdrawal schedule be submitted in accordance with 10 CFR 50.4 and that the proposed schedule must be approved by the NRC prior to implementation. Per NRC Administrative Letter 97-04, a proposed change to the withdrawal schedule may be approved without a license amendment if the changes conform to the American Society for Testing and Materials (ASTM) Standard Practice E 185-82. The proposed request complies with ASTM E 185-82 as discussed in the Attachment.
There is no in-vessel capsule at VCSNS currently and the capsules previously withdrawn did not receive fluence greater than the projected peak vessel fluence for an 80-year or 100-year operation. To support the demonstration of reactor vessel integrity through the potential subsequent license renewal(s) at VCSNS, a capsule is planned to be inserted in the reactor vessel with a proposed withdrawal schedule to meet the requirements for a potential 80-year or 100-year operating period.
A detailed description of, and justification for, the proposed activity are provided in the Attachment and Enclosures to this letter, respectively. DESC requests NRC review and approval of this request by January 23, 2024, with a 60-day implementation period for incorporating the VCSNS UFSAR changes.
In accordance with 10 CFR 50.91, a copy of this request, with the Attachment and Enclosures, is being provided to the designated South Carolina State Official.
Serial No.22-227 Docket No. 50-395 Page 2 of 3 Should you have any questions related to this submittal, please contact Yan Gao at (804) 273-2768.
Respectfully, James E. Holloway Vice President - Nuclear Engineering and Fleet Support c ! m : n ~ e t t e r : None.
Attachment:
Enclosures:
- 1. Westinghouse Report WCAP-18728-NP, Rev. 3, "V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Evaluation of Reactor Vessel Integrity Time-Limiting Aging Analysis," August 2022
- 2. Westinghouse Letter CGE-RV000-TM-ME-000004, Rev. 3, "V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Recommended Changes to the Surveillance Capsule Withdrawal Schedule," September 9, 2022
Serial No.22-227 Docket No. 50-395 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 Mr. G. Edward Miller NRG Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, Maryland 20852-2738 NRG Senior Resident Inspector V.C. Summer Nuclear Station Ms. Anuradha Nair-Gimmi Bureau of Environmental Health Services South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Mr. G. J. Lindamood Santee Cooper - Nuclear Coordinator 1 Riverwood Drive Moncks Corner, SC 29461
Serial No.22-227 Docket No. 50-395 Attachment 1 Proposed Revision to VCSNS UFSAR Section 5.4.3.6.2.3, Removal Schedule Virgil C. Summer Nuclear Station (VCSNS) Unit 1 Dominion Energy South Carolina, Inc. (DESC)
Serial No.22-227 Docket No. 50-395 Attachment 1 - Page 1 of 6 TABLE OF CONTENTS
1.0 INTRODUCTION
.............. ......................... .... .................................................................. 2
2.0 PROPOSED CHANGE
S AND TECHNICAL ANALYSIS ........................ .......................... 3
3.0 CONCLUSION
S .............................................................................................................. 5
4.0 REFERENCES
................................................................................................................ 6
Serial No.22-227 Docket No. 50-395 Attachment 1 - Page 2 of 6 Proposed Revision to VCSNS UFSAR Section 5.4.3.6.2.3, Removal Schedule
1.0 INTRODUCTION
NUREG-2191, "Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report" [4.1 ], provides guidance for Subsequent License Renewal (SLR) applicants. The SLR is the renewal of the initial renewed operating license, thereby permitting plants to operate to 80 years. The period of operation between 60 and 80 years is referred to as the subsequent period of extended operation. This guidance states that the Reactor Vessel Material Surveillance Program must meet the requirements of Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50), Appendix H [4.2], and further states that this program includes withdrawal and testing of at least one reactor vessel surveillance capsule addressing the subsequent period of extended operation.
10 CFR 50, Appendix H [4.2] provides the Nuclear Regulatory Commission (NRC) criteria for the design and implementation of reactor vessel material surveillance programs. 10 CFR 50, Appendix H requires the surveillance capsule withdrawal schedule meet the requirements of the edition of the American Society for Testing and Materials (ASTM) E 185 that is current on the issue date of the American Society of Mechanical Engineers (ASME) code to which the reactor vessel was purchased. For reactor vessels purchased after 1982, the design of the reactor vessel material surveillance program and the capsule withdrawal schedule must meet the requirements of ASTM E 185-82 [4.3]. For reactor vessels purchased during or before 1982, later editions of ASTM E 185 may be used, but including only those editions through 1982.
The Virgil C. Summer Nuclear Station (VCSNS), Unit 1 reactor vessel was designed and constructed to ASME Section Ill, 1971 Edition per UFSAR Table 5.2-1 [4.4]. Thus, per 10 CFR 50, Appendix H [4.2], the surveillance capsule program withdrawal schedule may meet the requirements of any version of the ASTM E 185 standard from the 1970 version (the version which was current on the issue date of the ASME Codes to which the reactor vessel was purchased) through the 1982 version. Per WCAP-9234-NP [4.5], the surveillance capsule program was designed to ASTM E 185-73 [4.6], which was the version active at that time the program was developed. Therefore, the requirements of 10 CFR 50, Appendix H [4.2] were met at the time of the design of the VCSNS reactor vessel surveillance program.
NRC Administrative Letter (AL) 97-04 [4.7] summarizes the Commission's decision promulgated in Commission Memorandum and Order CLl-96-13 [4.8]. In this Memorandum and Order [4.8], the Commission found that, while 10 CFR 50, Appendix H
[4.2] section 111.B.3, requires prior NRC approval for all withdrawal schedule changes, only certain changes require the NRC staff to review and approve the changes through the NRC's license amendment process. Specifically, only those changes that are not in conformance with the ASTM standard referenced in 10 CFR 50, Appendix H [4.2], are required to be approved through the license amendment process, whereas changes that
Serial No.22-227 Docket No. 50-395 Attachment 1 - Page 3 of 6 are determined to conform to the ASTM standard require only staff verification of such conformance.
The existing surveillance capsule withdrawal schedule fully satisfies the regulatory requirements for operation through the 60-year license renewal period. However, to preserve the possibility to operate the plant for 80 years and beyond, a capsule must be re-inserted into the reactor vessel to satisfy NUREC-2191 [8.04.1).
2.0 PROPOSED CHANGE
S AN D TECHNICAL ANALYSIS DESC is planning to revise the VCSNS Unit 1 surveillance capsule withdrawal schedule as listed in VCSNS UFSAR [4.4], Section 5.4.3.6.2.3. Specifically, the following changes are being proposed:
2.1 Revise the withdrawal schedule to re-insert Capsule Y during Refueling Outage 30 (Fall 2027) as a standby capsule and withdraw Capsule Y during Refueling Outage 38 (Fall 2039)
To date, VCSNS Unit 1 has tested 5 of 6 capsules initially inserted. Capsule Y is the only remaining untested capsule. The capsules already tested satisfy the requirements under the 60-year license. However, the schedule change is being pursued to meet the 80-year recommendations of NUREG-2191 [4.1] which states one capsule should be withdrawn at an outage with a neutron fluence of between one and two-times the peak reactor vessel wall neutron fluence at the end of the subsequent period of extended operation (SPEO) and tested in accordance with the requirements of ASTM E 185-82 [4.3). The proposed schedule change satisfies the regulatory requirements of the Reactor Surveillance Program for 80 years, provides a mechanism to obtain a neutron fluence to bound 100 years of plant operation, and meets ASTM E 185-82 [4.3] as required by 10 CFR 50, Appendix H [4.2).
GALL-SLR [4.1] Recommendations for Fluence: The 80-year, assumed SPEO of 72 Effective Full Power Year (EFPY), one-times peak reactor vessel wall neutron fluence =
9.06 x 1019 n/cm 2 (which includes a 10% bias on the peripheral and re-entrant corner assembly relative powers) and the two-times the end of SPEO fluence = 1.73 x 1020 n/cm 2 (2 x 8.64 x 1019 n/cm 2 , which conservatively excludes the 10% unbiased). Therefore, the capsule must reach a fluence above 9.06 x 10 19 n/cm 2 and below 1.73 x 1020 n/cm 2 to meet the GALL-SLR [4.1] recommendations.
Capsule Y: Capsule Ywas withdrawn at 17.71 EFPYwith a fluence of 7.01 x 10 19 n/cm 2 .
Therefore, this capsule does not meet the GALL-SLR [4.1] requirements for fluence cited above. Therefore, Capsule Y is recommended to be reinserted and irradiated further. To meet the GALL-SLR [4.1] recommended fluence, Capsule Y must be exposed to a minimum of 4.9 additional EFPY of irradiation to experience one-times the end of SPEO fluence. In order to remain below the two-times the end of SPEO fluence, Capsule Y must be withdrawn before 22.4 EFPY of additional irradiation. Since the minimum 100-year fluence is within the allowable range of one to two-times the 80-year fluence, it is
Serial No.22-227 Docket No. 50-395 Attachment 1 - Page 4 of 6 recommended that Capsule Y be withdrawn after being exposed to a minimum of 10.5 additional EFPY of irradiation in order to experience the minimum 100-year fluence of 1.14 x 1020 n/cm 2 .
Assuming that Capsule Y is reinserted during Refueling Outage 30 in the Fall of 2027, prior to cycle 31 (the final opportunity to reinsert the capsule), and an average fuel cycle length of 1.33 EFPY/cycle, (i.e., ~90% capacity factor), Capsule Y will need 8 fuel cycles to achieve an additional 10.5 EFPY of irradiation (10.5 EFPY) / (1.33 EFPY/cycle), and require removal following Cycle 38 (Fall of 2039). Since the plant has experienced 32.38 EFPY at the end of cycle 26, Capsule Y is projected to be reinserted at 37.7 EFPY and projected to be removed at 48.4 EFPY. These dates are only intended to be an estimate.
All actual dates will be based on withdrawing Capsule Y during the refueling outage nearest to, but following, the capsule fluence achieving the 100-year fluence, which is estimated to be after 10.5 EFPY with Capsule Y re-inserted into the 107° location (or symmetric locations 287° or 343°). Capsule fluence will be used to determine when the capsule is withdrawn, and 10.5 EFPY is an approximation based on the unbiased capsule projections.
2.2 Update the calculated fluence, removal time EFPY (effective full power years), and capsule lead factor values for capsules removed to date.
The values documented in WCAP-18728-NP [4.9] provide the most up to date information for the capsules. Additionally, the updated values verify that the existing pressure temperature limit curves for VCSNS Unit 1 are still bounding.
The following table is a proposed FSAR Section 5.4.3.6.2.3 Surveillance capsule withdrawal schedule for VCSNS Unit 1:
Serial No.22-227 Docket No. 50-395 Attachment 1 - Page 5 of 6 5-4.3.6.2.3 Removal Schedule The schedule for rermval of specimen capsules from the reactor vessel is as follows:
Capsule Calculated Ci12gile LQcatiQO (deg} Lead Factor Rermval time <a> Eluen~e (n/cm2} !b>
u 343 3.04 1.13(EOC1} 6.75 X 10 18 V 107 3.34 2.93(EOC3} 1.54 x 1019 X 287 3.54 5.04 (EOC5} 2.51 X 1019 w 110 3.21 11.21 (EOC 10) 4.63x10 19 z 340 3.10 16.36 (EOC 14) 6.53x10 19 290 3.09 17.71 (EOC 15) 7.01 X 1019 Y{c) 10](c) -3.5 48.2(cJ 1.14x1Q20
( a} Effective full power yeas from plant startup. End of Cycle (EOC) value given in p.=renthesis. Note that core thermal power was uprated from2775 to 2900 MWth stating with operating cyde 10.
(b} Values a-e taken from WCAP-18728-NP Table 2-1
{c} Capsule Y will be reinserted during Refuefing Outage 30 in the Fall of 2027 (prior to Cyde 31) in location 107° (or symnetric locations 287° or 343°), which is projected to occur a 37.7 EFPY. Capsule Y will ochieve the peak 100-yea- fluence, 1.14 x 1()211 n/cml (E > 1.0 MeV}, before removal, which is calculated to require another 10.5 EFPY of opera-ion (37.7 + 10.5 = 48.2). The nea-est outage to the suggested removal of 482 EFPY will occur at EOC 38 after 48.4 EFPY, which is the projected removal time of Capsule Y.
3.0 CONCLUSION
S A strategy has been developed to reinsert and withdraw Capsule Y to support subsequent license renewal. The strategy is based on the revised fluence projections outlined in WCAP-18728-NP [4.9) and Westinghouse Letter CGE-RV000-TM-ME-000004 [4.10].
The proposed change to reinsert Capsule Y during Refueling Outage 30 (Fall 2027) and withdraw Capsule Y during Refueling Outage 38 (Fall 2039) meets the surveillance capsule recommendations of NUREG-2191 [4.1 ], the requirements of ASTM E185-82
[4.3) and 10 CFR 50, Appendix H [4.2).
The proposed change to update the calculated fluence, removal time EFPY (effective full power years), and capsule lead factor values has no impact on the requirements and guidance of ASTM E185-82 [4.3] or 10 CFR 50, Appendix H [4.2].
Serial No.22-227 Docket No. 50-395 Attachment 1 - Page 6 of 6
4.0 REFERENCES
4.1 NUREG-2191, "Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report," Volume 2, July 2017, ADAMS Accession No. ML17187A204 4.2 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements" 4.3 ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," September 1982 4.4 VCSNS UFSAR 4.5 Westinghouse Report, WCAP-9234, Revision 0, "South Carolina Electric and Gas Company Virgil C. Summer Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program," January 1978 4.6 ASTM E185-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials, 1973
- 4. 7 NRG Administrative Letter 97-04, "NRG Staff Approval for Changes to 10 CFR Part 50, Appendix H, Reactor Vessel Surveillance Specimen Withdrawal Schedules,"
September 30, 1997, ADAMS Accession No. ML031210296 4.8 NRG Memorandum and Order CLl-96-13, December 6, 1996, ADAMS Accession No. ML20135F473 4.9 WCAP-18728-NP, Rev. 3, "V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Evaluation of Reactor Vessel Integrity Time-Limited Aging Analysis," August 2022 4.10 CGE-RV000-TM-ME-000004, Rev. 3, "V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal Recommended Changes to the Surveillance Capsule Withdrawal Schedule," September 9, 2022
Serial No.22-227 Docket No. 50-395 Enclosure 1 Westinghouse Report WCAP-18728-NP, Rev. 3, "V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: Ev aluation of Reactor Vessel Integrity Time-Limiting Aging Analysis," A ugust 2022 Virgil C. Summ er Nuclear St ation (VCSNS) Unit 1 Dominion Energy South Carolina, Inc. (DESC)
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 1 of 61 Page 1 of61 Westinghouse Non-Proprietary Class 3 WCAP-18728-NP August2022 Reyision3 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal:
Evaluation of Reactor Vessel Integrity Time-Limited Aging Analyses
@Westinghouse
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 2 of 61
~2ot 61 Westinghouse Non-Proprietazy Class 3 WCAP-18728-:NP Rmsion *3 V.C. Summer Nuclear Station Unit 1 Subsequent License Renewal: EYnluation of Reiar tor Vessel Integrity Time-Limited Aging Analyses Tyler C. Ziegler'"'
RV/CV Design & Analysis D. Brett Lynch'"'
RV/CV Design & .AnaJ.rsis August 2022 Re\-iewen:: Donald M McNntt ill*
RV/CV Dfiign & Analysis Frank M. N edwidek*
N ucleM Operntioos Approved: Lynn A. Patterson*, M.aaager RV/CV Design & An:i.lysis Jesse J. Klingensmith*, Manager Radiation Engineering & Arutlysis
- Electronically approved records are authemicated in the electronic document management sysrun.
Westinghouse Electric Company llC 1000 Westinahouse Drire Cranbcny Tomnbip. PA 16066. *usA C 2022 Westwghouse Electric Company llC All Rights Resez.'ed.
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 3 of 61 Page 3of61 Westinghouse Non-Pr-oprietary Class 3 n RECORD OF REVISION R,u sion 0
Incmporate Dominion's comments on Revision 0.
1 June 2022 es are marked with revision bars.
Incorp0111te Dominion's comments on Revision 1.
2 July2022 es are marked with revision ba.-s.
Inccnporate Dominion's comments on CGE-RV000-1M-ME-000004 Revision 1 on the 3 August 2022 smveillance capsule schedule in Section 7.
es are marked with revision bar's.
WCAP-l&nS-NP Angust 2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 4 of 61 Page4of61 W~.stinghouse N on-Proprietacy Class 3 Ill TABLE OF CO~""TEJ.\"TS LIST OF TABLES***-*****************-********************************************-*******************************************************-******** iv LIST OF FIGURES ..................................................................................................................................... "i EXECUTIVE SU}.11,;{ARY*******************************************************-*******************************************************- **-*** "ii 1 TildE-1..IldITED AGING ANALYSIS ......................................................................................... 1-l 2 CALCUlATED FLUENCE ......................................................................................................... 2-1 3 MATERIALPROPERTYINPUT................................................................................................. 3-l 4 PRESSURIZED THERMAL SHOCK ***************************************************************-************************ 4-1 5 UPPER-SllELF ENERGY ........................................................................................................... 5-1 6 HEATUP AND COOl.DOWN PRESSURE-TEMPERATIJRE LIMIT CURVES ....................... 6-1 6.1 ADJUSTED REFERENCE TEMPERATURES CALCULATION ............................- ...... 6-1 62 P-TLIMITCUR.VESAPPLICABILI'IY .............................................................-*****-****6-14 7 SURVEII.LANCE CAPSULE wrnIDRAWAL SCHEDULES ........-**********-*********-**-*-**-**-** 7-1 8 REFERENCES ****************************************************************-*********************************************************** 8-1 APPENDIX A CREDIBII.lTYEVALUATION OF THE VCSNS UNIT 1 SURVEILLANCE PROGRAM..................................................................................................*-*****************A-1 APPENDIX B EMERGENCY RESPONSE GUIDELINES .................. -*********-******-*********-**- *-**-**-* B-1 WCAP-18728-NP Angus12022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 5 of 61 Page5 of 61 Westinghouse Non-Propriet.uy Class 3 It' LIST OF T.-\BLES Table 1-1 Eval uation ofTime-1.imitedAgingAnalyses Per the Criteria of 10 CFR 54.3 ............... 1-1 Table 2-1 Calculated F ast Neutron (E > 1.0 MeV) Fluence at the Sun.-eillance Capsule Center for VCSNS Unit 1 ***************************************************************************** **-*********- *********************2-2 Table2-2 VCSNS Unit 1 - Maxinmro "):1ast Neutron (E > 1.0 1-feV) Fluence Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline hgions ....................... 2-4 TableJ- 1 Best-Estimate Cu and Ni Weight Percent "1/2lues, Initial RTNi>r Values, and Initial USE Values for the VCSNS Unit 1 RPV Beltline and .Extended Beltline :Materials.. *-***********3-4 Table 3-2 Calculation of Pootio:n 2.1 CF Values for VCSNS Unit 1 Sun-eillance Matcrials........ _3-5 Table 3-3 Summary of the VCSNS Unit 1 RPV Beltline, Extended Beliline, and SurveillaDce Material CF Values based on Regulafofy Guide 1.99, Revision. 2, Position U and Position 2.1 ****************************-*******************************************************************- **- ***************3-6 Table 4-1 RTPTS Calculations for VCSNS Unit 1 at 72 EFPY - *********************-********************************4-2 Table 5-1 Predicted USE 1/4mes at n EFPY for the VCSNS Unit 1 Beltline and &tended Beltline 1.-Iateiials _............................ *................ _..................................*.. -*************- **-***************5-3 Table 6-1 VCSNS Unit 1 Flnence and Fluence Factor Values for the Surntee, 1/4T, and 3/4T Locations at 56 EFPY ...........*-***********************************************************************- ***************6-2 Tab1e6-2 VCSNS Unit 1 Fluence and Fluence F acto r Values for the Surl'ac:e, l /4T, and 3!4T Locations at 71 EFPY ****-******************************** ************************-*****************- *******************6-3 Table6-3 Calculation of the VCSNS Unit l ART Values at the 1/4T Location for the Reactor \1e§el Beltline and&tended Beltline Materials at the End of PEO (56 EFPY) ..*-*******************6-4 Tab1e6-4 Caknlation of the VCSNS Unit 1 ART Values at the 3/4T Loca tion for- the Reactor Vessel Beltline and &tended Beltline Materials at the End of PEO (56 EFPY) ........................ 6-6 Tab1e6-5 Calcnlation of the VCSNS Uni11 ART Values for the Reactor Vessel Extended B eltline Nozzle Materials at the F.nd of PEO (56 EFPY) ..................... *-**--********************************6-S Table 6-6 Cakulation of the VCSNS Unit 1 ART Values at the l /4TLoc:ationfor the Reactor \e.s£cl Beltlwc andF..- nended Beltline Materials at the End ofSPEO (12 EFPY)*- **-*************-6-9 Table 6-7 C-alcnlation of the VCSNS Unit 1 ART Values at the 3/4T Location.for the Reactor Vessel Behline and Extended Beltline Materials at the End ofSPEO [/2 EFPY) ....*-*************6-11 Tab1e6-8 Calculation of the VCSNS Unit 1 ART Values for the React oc Vessel Extended Beltline Nozzle Materials at the End ofSPEO f/2 EFP\') *******************************-************************6-13 Table6-9 Suinmar)' of the Limiting ART Valu.es ................................................ *- *************-*********6-l4 Table7-1 VCSNS Unit 1 Recommended SUITeillance Capsule Withdrawal Sc:hedule...................7-3 Table A-I Regulatory Guide 1.99, Revision 2, Credibility Criteria ................................................ A-1 WCAP-18728-NP Angust2022 RRision3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 6 of 61 P.tge 6of61 Westinghouse Non-Proprietary Class 3 V TableA-2 Calculation of Interim Cbemisuy Factors for the Credibility Evaluation Using VCSNS Unit l Smveillance Capsule Data Only .......................................................................... A-4 Table A-3 VCSNS Unit l Surveillanc-e Capsule Data Scatter about the Best-Fit Line ................... A-5 TableB-1 Evaluation ofVCSNS Unit 1 ERG Limit Ca1egory ....................................................... B-1 WCAP-18728-NP Angust 2022 Rtraion 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 7 of 61 Page7of 6 I Westinghouse Non-Proprietruy Class 3 LIST OF FIGURES Figuce2-1 Axial Boundary of the LOE +17 n/cm.2 Fast Neutron (E > 1.0 MeV) Fluence Threshold in the +Z Direction at 33.75 (end of Cycle 27), 54, and 72 EFPY ********************-**- ******-****2-6 Fignre3-l RPV Schematic for VCSNS Unit 1 ***-*************************************************************************-**3-3 Fignre 5-l Regulatory Guide 1.99, R.e,,ision 2, Position 1.2 & 2.2 Predicted Decrease in Upper-Shelf Eneigy as a Function of Copper and F1uence for VCSNS Unit 1 at fheEnd of SPEO (72 EFPY)*-******-**-****************************************-******************-*****************-************************5-5 Figure 7-1 Original Arrangement of Surveillance Capsules in the VCSNS Unit l Reactor Vessel .. 7-4 WCAP-1872&-NP Angmt2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 8 of 61 Page8 of61 Westinghouse N on-Proprietary Class 3 vii EXECUTIVE
SUMMARY
This report presents the reactor vessel integrity Tune-Limited A ging Analyses (TI.AA) evaluations foe the Vugil. C. Summer Nuclear Station (VCSNS) Unit 1 reactor pressure vessel (RPV) in accordance with the requirements of the License Renewal Rule, 10 CFR Part 54. 1LAAs are calculations that address safety-related aspects of the RPV with.in the bounds of the cwrent 60-ye.ar license. These calrola.tions must also be evaluated to account for an extended period of operation (80 ye111S) al.so tenned Subsequent Period of Extended Operation (SPEO).
VCSNS U nit 1 is currently licensed through 60 years of operation; therefore, with a 20-year license extension, the SPEO is applicable through 80 years of operation. The 60-year TI.AAs evaluated in this report are applicable through 56 Effective Full Power Years (EFPY), which is deemed end of the Period of E.- ucnded Operation (PEO). Similarly, evaluations in this report perfomJed at 80 }-eatS of operation are applicable through 72 EFPY (90% capacity factor of80 ye.us), which is deemed the end ofSPEO. Updated neutron flucnce evaluations were performed and documented in WCAP-18709-NP (Reference 5), as well as in Section 2 of this report. Updated neutr011 fluence eir- aluatiom were used to identify the VCSNS Unit 1 extended be1tline materials and as input to the reactor vessel (RV) integrity evaluations in support of cmrent plant opeiations and subsequent lic61Se renewal In addition lo the RV integtitylLAAevaluations, the VCSNS Unit 1 surveillance data credibility evaluation is contained in Appendix A of this report. While not a TI.AA, Appendix B prcn.i des an Emergency Response Guidelines (ERG) assessment for VCSNS Unit 1 for completeness.
A summ.aiy of remits for the VCSNS Unit 1 Tl.AA evaluation is provided below. B ased on the results of this TI.AA evaluation, it is concluded that the VCSNS Unit 1 RV will continue to meet regulatory requirements through the SPEO.
Flneoce The RV beltline and extended beltline neutron flucnce values applicable to a postulated 20-year license rene~-al period w ere calculated for the VCSNS Unit 1 materials. All transport calculations were carried out using the three-dimensional discrete ordinates code RAPTOR-MJG and the BUGLE-96 cross-section library. The an.al)'llis methodologie~ follow the guidance in Rega.J.atory Guide 1.190 (Reference 2). It i s also consistent with the methodology des cnoed in WCAP-18124-NP-A (Reference 4) that was generically approved by the United States Nuclear Regulatory Commission (USNRC) for calculating exposures of the RPV beltline (i.e., in general, RPV materials opposite the active fuel). See Secti011 2 for m ore details.
Prf'Ssurized Ther mal Shock All of the beliline and extended beltline materials in the VCSNS Unit 1 RV are projected to remain below the RTprs screening criteria values of 2 70°F for base metal and/or longitudinal welds and 300°F for circumferentially oriented welds (per 10 CFR 50.61) through SPEO (72 EFPY). See Section 4 for m ore de1ails_
WC-AP-18n8-NP Angust2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 9 of 61 P,11ll' 9of61 Westinghouse N on-Proprietaiy Class 3 VU1 Upper-Shelf Energy All of the beltline and extended beltline materials in the V CSNS Unit 1 RV are projected to remain above the npper-shelf energy (USE} scaening criterion value of 50 ft-lb (per 10 CTR 50, Appendix G), through SPEO (J2 EFPY). See Section 5 for more details.
A djusted Referen ce Templ'ratures and P -T Limit CUITes Ap plica bility Chl'ck Adjusted Reference Temperatures {ARTs) are calculated for the end of PEO at 56 EFPY and for the end of SPEO ;rt 72 EFPY in ordei- to perf= an applicability check on the existing pressure-temperature (P-1) limit curves for VCSNS Unit 1. With the consideration of TI.AA fluence pcojections, revised Position 2.1 chemimy factor v-alnes, and recalculated initial RTNDT values, the applicability of the VCSNS Unit 1 cylindrical shell P-T limit curves curre:ntly in the Technical Specifications remain applicable through 72 EFPY. The conctu.sion considers the RV inleVoutlet nozzles.
Surnillante Capsull' Withdrawal Schedule With coosidemtion of a 2 0-year license renewal to 8 0 yeais of operation (12 EFPY), Capsule Y, which cnaently resides in the spent fuel pool, mnst be reinserted for additional imldiation. The sun,-eillance capsule withdrawal schedule in Table 7-1 identifies the additional exposure required by the capsule in order to mttf the guidance ofNUREG-2191 (Reference 18) (GAIL-SLR) for a capsule to be- withdrawn and tested between one and two times the pe.ak RV wall neutron flue1ice at the end of SPEO. See Section 7 for more details.
WCAP-1 872&-NP Awrust 2022
~ vision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 10 of 61
~ 10 of 61 Westinghouse N on-Proprietary- Class 3 1-1 I TIME-LL°'IITED AGING Al~ALYSIS Time--Limi.tedAging Analyses (fLAAs) are those licensee calculations that
- 1. Involve systems, stmctmes, and components (SSCs) within the scope of license renewal
- 2. Considtt the effects of aging.
- 3. Invoke time-limited assumptions defined by the current operating term (e.g., 60 years).
- 4. Were determined to be relevant by the licensee in making a safety detcimination.
- 5. Involve conclnsions or provide the basis for conclusions related to the capability of the SSC to pen01D1 its intended functions.
- 6. Are contained or incorporated by reference in the cmrent licensing basis (CI.B).
The potential 1lAAs for the reactor pressure vesse-1 (RPV) are identified in Table 1-1 along with indication of whether or not they meet the six (6) criteria of 10 CFR. 54.3 (Reference 1) for TI..AAs.
Table 1-1 Ernua tion of TDDe-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 Presror~
Presmrized Temp-erature Cak ulated Upper-Sh elf Tune-Limited Aging Analysis Thermal Limits for Fluence Energy Shoc1.4> Heatup and Cooldomi Involves SSC Within the Scope of License Renewal YES YES YES YES ConsidefS the Effects of Aging YES YES YES YES Involves Time-Limited.Assumptions YES YES YES YES Defined by the Cuaeot Operating Tenn Determined to be Relevant by the Licensee in Making a Safety YES YES YES YES Deteonin.ation Involves Conclusions or Pnn,"ides the Basis for Conclusions Related to the Capability of SSC to Perf0m1 Its YES YES YES YES Intended Function Contained or Incoiporated by Reference in the CLB YES YES YES YES Note:
(a) The ~Pn,,surized Thmnal Shock (PIS) nlues arellSl!dto dmnninl! tbe apprq,rim~ Respome Guidelme (ERG)
Lillli:b catepy fer VCSNS Unit 1 ~ the end of ihe potential sub5eqaem license exlmaD period. How~nr, ERG limits ;u,,
aomde the ~ of 10 CFR Part 54.3. ERG limib are ~ in Appendix B.
WCAP-1872&-NP August 2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 11 of 61 Page 11 of 61 Westinghouse N on-Proprietary Class 3 2-1 2 CA.LCULATEDFLUENCE Estimated RPV beltline and extended beltline fast neutron (E > 1.0 MeV) fluenres at the end of 80 yeacs of operation were calculated for VCSNS Unit 1 in WCAP-1 8709-NP. 1be ana1yses methodologies nsed to calculate the VCSNS Unit 1 RPV fluences followed 1he guidance ofRegulatory Guide 1.190, "Calculational and Dosimeliy Methods for Deteonining Pressuce Vessel Neutcon Flnence" (Reference 2). The.se methodologies have been appcoved by the USNRC for the beltline region, i.e., materials directly S1Jttoundiog the core and adjacent materials per 10 CFR 50, Appendix G (Reference 3), which are projected to experience fhe highest flnence. 1be methodologies, along with the NRC safety evaluation, are contained in detail in WCAP-18124-NP-A (Refecence 4). For VCSNS Unit 1, the beltline region has traditionally included the intennediate and lowec shell forgings, and the cireumfecential welds between these components. The traditional beltline and extended beltline materials are identified in Table 2-2 and Figure 2-1.
Materials exceeding a fast neutron (E > 1.0 MeV) flucnce of 1.0 x 1a1 7 o/cm2 at the end of the SPEO are evaluated for changes in fracture taugJmess. RPV materials that are not traditionally plant-limiting because oflow levels of neutron radiation must now be evaluated to detcnnine the a<<:llmlllatcd fl=e at the end ofSPEO. Therefore, fast neutron (E > 1.0 MeV) fluence calcuJations were perl'onned for the VCSNS Unit 1 RPV to detennine where it will exceed a fast neutron (E > 1.0 MeV) floence of 1.0 x ta1 7 n/cm2 at the end ofSPEO. The materials that exceed the 1.0 x 1017 o/cnl fast neutron (E > 1.0 MeV) fluence threshold and we.re not evaluated in pa.st analyses of record as part of the trnditioml be1t1ine, are n-fened to as extended beltline materials in this report and are evaluated to determine the effect of neutron in-adiation embrittlemeot during SPEO.
All the tramport calculations were canied out using the fluee.dimensional discre1e ordinates code RAPTOR-M3G and the BUGLE-96 cross-section library. The BUGLE-96 libr.uy provides a 67-groop coupled neutron-gamma ray cross-section data set produced specifically for light watec reactor applications_
In these analyses, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modeled with an S1& order of angular quadrature. Enagy- and space-dependent core powec distributions, as well as system opernting temperatures, were treated on a :fuel-cycle-specific basis.
The calculations for fuel Cycles 1 through 26 determine the neutron exposure of the pressure vessel and surveillance capsules based on cowpleted fuel cycles. The projection for Cyc1e 27 is based on the actual loading, but yet to be completed, fuel cycle. The projections for Cyl:le 28 and beyond, up to and including the end of PEO (56 EFPY) and the end of SPEO [/2 EFPY), are based on the average core power distnlmtions and reactor operating conditions of Cycles 25, 26, and 27 and are determined both with and without a 10% positive bias on the peripheral and re-entr.mt comer assembly relative powecs.
Table 2-1 gives the VCSNS Unit 1 calcula.ted fast neutron (E > 1.0 MeV) fluences at the capsule loc.atjons including all withdrawn surveillance capsules (Capsules U, V, X, W, Y, and Z).
Table 2-2 presents the fast neutron (E > 1.0 MeV) fluence results for the applicable portions of the pressure vessel from the neutron transport analyses. From Table 2-2 it is observed that outlet nozzles and inlet nozzles ha'll-e fast neutron (E > 1.0 MeV) fluence greater than 1.0 x 1017 nlcm'- at the lowest extent of the nozzle forging to nozzle shell weld al 72 EFPY. All materials loe3ted above the nozzles will remain below 1.0 x 1a17 o/crn2 through 72 EFPY. Table 2-2 indicates that the lo\\"ec shell to lower vessel head WCAP-18728-NP Angust 2022 Revision3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 12 of 61 Page 12 of 61 Westinghouse Non-Proprietary Class 3 2-2 17 circumferential weld, and all materials below it, will remain below 1.0 x 10 n/c:m.2 through SPEO. Figme 17 2 2-1 shows the axial boundary of the 1.0 x 10 n/cm. fluence threshold (at 54 EFPY and 72 EFPY) as a function of azimuthal position (Z versus 8).
All data presented in this section, along with additional details, are presented in WCAP-18709-NP (Reference 5). This includes description of uncertainties and validation of the analytical model based on the measured plant dosimetry.
Table 2-1 Calculated Fast Neutron (E > 1.0 IeV) Flnence at the Surnillance Capsule Center for YCSNS Unit i<->
Cumulath-e Flurnce Cnle Operating (n/em2)
Cyclr u itgth lime (EFP\') 17° 20° (EFPY) 1 1.13 1.13 6.75E+l8"UJ 5.90E+l8 2 0.67 1.80 1.01E+19 8.94E+18 3 1.13 2.93 U 4E+t g(<l l.36E+l9 4 1.16 4.09 2.03E+l9 1.80E+l9 5 0.95 5.04 2.51E+t9CtlJ 2.24E+l9 6 1.17 6.21 3.13E+19 2.79E+l 9 7 1.22 7.43 3.63E+19 3.24E+l 9 8 1.19 8.61 4.13E+l9 3.69E+19 9 1.27 9.89 4.66E+19 4.15E+19 10 132 11.21 5.18E+19 4.63E+t9<*>
11 136 12.56 5.78E+19 5-15E+19 12 137 13.94 6.35E+19 5.66E+19 13 1.09 15.03 6.76E+19 6.02E+19 14 133 16.36 734E+19 6.53E+l9\'I 15 135 17.71 7.88E+19 7.01E+19<1) 16 1.34 19.05 8.42E+19 7.49E+19 17 138 20.43 9.00E+19 8.01E+19 18 1.30 21.73 9.52E+19 8.49E+19 19 131 23.04 1.01E+20 8.97E+19 20 1.36 24.41 1.06E+20 9.46E+19 21 1.29 25.70 1.11E+20 9.93E+19 22 1.29 26.99 1.17E+20 1.04E+20 23 1.34 28.33 l.23E+20 l.09E+20 24 1.32 29.65 l.28E+20 1.14E+20 25 136 31.01 1.34E+20 1.19E+20 26 1.37 32.38 l.39E+20 1.24E+20 27'>) 1.37 33.75 1.45E+20 1.30E+20 No bias 011 ths peripheral mid re,-entrant comu assembl)* relative powers Future\*J - 36.00 1.55E+20 1.38E+20 Future{JJ - 42.00 l.80E+20 1.61E+20 Future\11 - 48.00 2.05E+20 l.83E+20 WCAP-18728-NP Angim 2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 13 of 61 Page 13 of 61 Westinghouse Non-Proprietary Class 3 2-3 Table- 2-1 C:ikufated F ast Neu tron (E > 1.0 Me\) Fluence a t the Sun*eillance C:ip$ule Center for VCS1'-S Unit 1<->
~-ell' Cumulafu-e Flnence ungth Operating (n/cm2)
Cycle Time
([FPY) 17° 20° (EFPY)
Futnre'-' 1 - 54.00 230E+20 2.06E+20 Future" - 60.00 2.55E+20 2.28E+20 Futnre1'1 - 66.00 2.80E+20 2.51E+20 FutureW - 72.00 3.05E+20 2.73E+20
+10% bias on the periplreral and re-entrant comer assembl;y relative Future© - 36.00 1.56E+20 l.39E+20 Future© - 4200 1.83E+20 1.64E+20 Future© - 48.00 2.11E+20 1.88E+20 Future© - 54.00 238E+20 2.13E+20 Future© - 60.00 2.66E+20 2.3 8E+20 Future© - 66.00 2.93E+20 2.63E+20 Future© - 72.00 3.21E+20 2.87E+20 Notes:
(a) In£n11n;rtioo tmn fu>m WCAP-I8709-NP (lWi,n,,:,ce 5).
(b) This '\'ilue is applicable to ~ U.
(c) lhis ~ is ~b!e t o ~ V (d) This '\.uue is ~ b l e to Capsule X (e) This '\'Uue is ~ble to YpSule W.
(f) This 1,-m!i! is ~ b l e to ~ z.
(g) This .,_-;due is appliab!e to Capsule Y.
(b) Cycle 27 was tbe cqnnt oper,,fmgcycle at the tim,, lmSU!DII.IIY ~ mis :mlbmed.
(i) Values beymid Cycle 27 ara based on the . n . ~ cora pcmw distribul:iotlS a:nd reactor apemingcomitiOl!S ofCycles25, 26, md27 md an, d,,tl'Dmll'dbofhmth.w,hrifbo1it a I.I bias oo. the peripbera m d ~ comeJ" :assembly ruaim!J)CJWt'l5.
WCAP-18728-NP Angust2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 14 of 61 Page 14cf61 Wmingbouse NOJ1-Proprietary Class 3 2-4 Tablt 2-2 YC~S Unit 1- i\lmmum Fad l'.tulron (E > 1.0 Mt-\) .Flutnct E:q,tritnctd by tht PttMun Y<=el M:tttrials in tht Beltlint and E:s:ttndNI Btltliat ~on<
Projunons i.1ih no bias on the pmphnal tl1ld re.ffltrarrl comer assmnbly rd,,m,-. JIO>IWS Fast l\tnlron (E > 1.0 Mt\) Flnmrt (n/cm1) l\htuW Loc::ition Mattrial 33.75 .EFP\1,t 36Dl'Y 42 EFPY 48EFPY 5-I EFPY 60EFPY 66 EFPY 72EFPY Ialtt Nozzle to Nozzle Shell Weld 1.41E+17 150E+17 l.75E+17 2.00E+17 2-24E+17 2.49E+17 2.74E+17 2.99E+17 Ocmwt - fl Inltt Nozzle Postulaltd l.68E+16 l.79E+16 209E+16 239E+16 2.68Et16 2.98E+16 3-28E+16 3.58E+1 6l4 l/4T Flaw Outltt Nozzle to Nozzle
&lmded.Belllme SMIIWdd 5.99E+16 639E+l6 7.44E+16 8.49E+16 9.55E+16 1.06E+l7 l.17E+17 127E+17 Mattriah acm-es1 awii)
Omlrt Nonie Postulatld l.67E+l6 l.82E+16(4 8.54£+15 9.11E+15 l.06E+16 121E+16 137E+16 1.52E+16 1/4T F!iw Nozzle Shd)('ol 1.83E+18 1.95E+18 226E+18 2.5SE+18 2.&9E+18 320E+18 3.52E+l8 3.83E+l8 Nozzlt..fo-ln!mnedir.e 1.94E+18 2.07E+18 2 40E+l8 2 74818 3.07E+18 3.40E+18 3.74E+18 4.07E+18 Shell Cimnnfl!fflllw Weld Int~te Shell 4.14E+19 4.40E+l9 5.10E+19 5.81E+19 6.51E+l9 721E+19 7.92E+19 8.62E+19 IntmntdiAlae Shell Lo~todinal Wtld - 1.40E+19 1.49E+19 1.72E+19 1.96E+l9 2.19E+19 2.42E+19 2.66E+19 2.89E+19 45*ms*
Bellline Mmriili In!Oilllldia~ fo-1.ower Shell 4.13E+19 4.40E+19 5.10E+19 5.S0E+l9 6.51E+19 721E+19 7.91E+19 8.62E+19 Cimmifamtial Weld I..cruw Shell 4.1 4E+19 4.40E+l9 5.11E+19 5.81E+l9 6.52E+l9 723E+19 7.93E+19 8.64E+19 urn-er Shell Lonmudinal 1.42E+19 1.51E+19 l.7SE+19 1.98E+19 222E+19 2.46E+19 2.70[+19 2.93E+19 Weld -135°/315° Omsidt ofbcltlme Lem-er Sbdl lo Bottom
!kad Cin:umfermlw 4.69E+15 5.00E+15 5.81E+15 6.62E+15 7.43E+15 82 4E+15 9.05E+15 9.86E+15 region Wdd(<1 WCAP-18728-NP Angust2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 15 of 61 P- 15 of 61 Westingbou,,e Non-Proprietary Class 3 2-5 VCSXS Urut 1- M:mmrnn Fast Neutron (E > 1.0 Me\') Ilumu ~tritnced by the Pres~ure Vessel Materials in th.e Bdtline and Extended Beltline ~ons Proj<<ttoru Mith a+10-/4 bias on 1M peripheral llJltl J'!e-9ltrtmJ coma- imembly n/mn,;, JXl"W:.
Fast Neutron (£> 1.0 ll.Ie\') Fluenc,e (n/cm')
Matuiol Loutio11 Material 33.75 Ill'\vl 36EFP¥ .U EFPY 48EF1'¥ 5-t lll'Y 60 EFPY 66EF1'¥ 7lEFPY lnlet Nozzle to Nozzle Shell l.41E+17 151E+l7 l.77E+17 2.04E+17 230Et-17 2.57E+17 2.83E+l7 3.10E+17 Weld tlowest exfad) lnlet Nozzle Postulated 3.69E+1 6'l l/4T Flaw 1.68£+16 1.80E+16 211E+16 2.43EH6 2.74E+16 3.06E+16 3.38E+16 Outltt Nozzle to Nozzle F.mnded Bellline Shell Weld 5.99E+l6 6.42E+16 7.54EH6 8.66E+l6 9.79E+16 1.09E+17 1.20E+17 132E+17 Materials tlowest _._...-.
Outltt Nozzle Postulated 8.54E+15 9.14E+15 l.07E+16 1.24E+l6 J.40E+16 1.56E+16 1.71E+16 1.88EH6'1 l/4TFlaw NozzieSJiell{1>) 1.83E+18 1.96E+18 2.30E+18 2.64E+18 2.9SE+18 3.32E+18 3.66E+18 4.00E+18 Nome-tG--~te 4.25E+18 1.94E+18 2.08E+18 2.44E+18 2.80E+l8 3.16E+l 8 3.52E+18 3.89E+18 SbeU Cimmlfemltw Weld Intamediate Shrll 4.14EH9 4.42E+19 5.19E+19 5.96E+19 6.73E+19 7.50E+19 827E+l9 9.04E+19 Intmnediate Slid Longitudinal Weld - 1.40E+l 9 150E+19 l .75E+19 2.01E+19 226E+19 2.52E+l9 2.78E+l9 3.03E+19 45*ms*
Behline :Mataials Intmnediare-to-I..ower Shell 4.13E+19 4.42E+19 5.19E+19 5.96E+l9 6.73E+19 7.50E+l9 827E+19 9.04E+19 CircumfeRnlial Weld Lower Shell 4.14E+19 4.42E+19 5.20E+19 5.97E+19 6.74E+19 752E+19 8.29E+19 9.06E+19 1..owu ::ih~
1.42E+19 152E+19 1.78E+19 2.04EH9 230E+19 2.56E+19 2.!P..E+19 3.08E+19 Weld -135"/315° I.mm- Shdl to Bottom Ouwdi, of belllillf Head Circumf,mdial 4.69Hl5 5.03E+15 5.91E+15 6.79£+15 7.67E+15 8.55E+15 9.44E+l5 1.03E+16 r,gion Weld{<l Na!=
(a) V,J,,e limi is Ill< poj<<led EFPY at lhe om of Cycle 27.
(b) E,:poe,n ,"Ulll!J b b n=le , h e l l ~ u-.l.c!, ""'bour.l..tby tbe opc=n,nas Cm- lhe naz:zlo sh!!!l (ah -sbtll).
(c) Mnimmn opo,mo n.1,,r; occm al b RPV oatorDmll>.
(cl) \¥hi!., Ibo fhmice at tm l<<afionis Jess Ihm IB<-17 n'arr, itis idenlmed as o!io:>dod bdliJ:i,, siDct p<ldians o!d>tDCIZZ!e ....,..,.i. 1ht ai!aian.
WCAP-18728-NP August2022 Revisi<m 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 16 of 61 Page 16of6t Westinghouse Non-Proprietuy Class 3 2-6
- -3J .75 EFPY(EOC27) ---*5.f.EFPY --72EFPY 40<} o N~ P ootul td l 4T F!aw
- NllZ21eLoH &tWeld E.'!!tti Nczil.1 Sh.J1 300 {Hm Na Cill2l-2) 0 0 0
--=-'C':--:,.'&_~--=::~--;.___.~.
Noulc.to-I,,rnmfdi&t! Slllll Circ.. U'tld (Hru Xc.:4l't7S4) bilermedia.le Shell lOii @e:n Na.. A9 m.1, LOl\"i!IS!ltll LINHS!llll
-10.) {Hst N.o C9923-l) (He¥ No.. C992J-2)
-300 Lau~ Sl:.!1!-tc-Botlo!!l Had Cir:. We1:1 lH*!d No.3]~ P66) 0 90 m HO 170 31} 3ff.l
-4t4 Figure2-l .-uial Boundary of the l.OE+l7 u/crn.1 Fast:l'\entron (E > 1.0 lie\) Fluence Threshold in the +Z Direction at 33.75 (end of Cycle 27), 5-1, and 72 IIPY WCAP~l872&-NP A.ugust 2021 Re\ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 17 of 61 Page 17 of 61 Westinghouse N on-Proprietary Class 3 3-1 3 MATERIAL PROPERTY INPUT The requirements for RV integrity are specified in 10 CFR 50, Appendix G (Reference 3) and 10 CFR 50.61 (Refemice 6). The beltline region of the RV is defined as the following in 10 CFR 50, Appendix G:
... the region ofthe reactor vessel (shell material incblding welds, heat affected zones, m,d plntes or forgings) that directly sum,unds the effective lteight of the active core and adjacent regions of the reactor vessel that am predicted to experience sufficient neutron mdiah"o11 damage to be considered in the selection ofthe most limiting materlal>>tth regani to radiation damage.
The VCSNS Unit 1 beliline materials coosist of two (2) Intennediate Shell Plates, two (2) Lower Shell Plates, and their associated welds. The VCSNS Unit 1 smveillanre plate material was made from RV Intennediate Shell Plate 11-l, H eat# A91 54-1. The VCSNS Unit 1 RVbeltline welds were fabricated using weld wire Heat # 4P4784, Flux. Type Linde 124, Flux Lot# 3930. The weld material in the VCSNS Unit 1 surveillance program was fabricated with the same material heat, flux type, and :flux lot number.
17 2 Any RV materials that are predicted to experience a n=tron flnence exposure greater than 1.0 x 10 n/cm.
(E > l.O MeV) at the end of the licensed operating period should be considered to experi.ence neutron 7 2 embrittlement. B ased on the results of Section 2 of this report, the materials that exceeded the 1 x 1ot n/cm (E > 1.0 MeV) threshold at 72 EFPY that were not included within the original beltline are considered to be the VCSNS Unit l extended beltline materials and are evaluated to detennine their impact on the proposed SPEO of operatiOll. The VCSNS Unit 1 RV extended beltline contains one (1) Nozzle Shell-to-Intemiediate Shell circumferential weld, hvo (2) Nozzle Shell Plates (also termed upper shell), two (2)
Nozzle Shell longitudinal ,relds, three (3) Inlet Nozzles, three (3) Ootlet Nozzles, and the six (6) Nozzle-7 to-Nozzle Shell welds. Only those materials with a fluence greatec than 1 x 1<>1 u/cnl (E > 1.0 MeV) at the end of SPEO require the effects of embrittlement to be included when evaluating the RV integrity.
The RV forgings/plates and weld materials are shown in Figure 3-1 for VCSNS Unit 1. Used in conjunction with the :fluence data in Table 2-2, and Figure 2-1, the beltline and extended beltline materials are identified as shown in Table 3-1. Note that for RV welds, the tams " girtbn and "circumfecential" are used interchangeably; herein, these welds shall be refelled to as cin:nmfefefltial welds. Similarly, for RV welds, the lem1s "axial" and "longitudimtl" are used interchangeably; herein, these welds shall be referred to as longitudinal welds.
The unimldiated material property inputs used for the RV integrity evaluations herein are contained in PWROG-21037-N P (Refecence 7). P WROG-21037-NP defined or redefined many of the matecial properties and chemistty values using the most up-to-date methodologies and all available data; therefore, the values utiliz.ed herein supersede previously documented vames. The sources and methods used in the detennination of the chemis1ry :fuctOIS and the fracture toughness properties are summarized below.
Ch emical Compositions The best-estimate copper (Cu) and nickel (Ni) chemical compositions for the VCSNS Unit l beltl.ine and elrte.ruied beltline materials are presented in Table 3-1. The best-estimate weight percent copper and nickel WCAP-18728-NP Angmt2022 Re\>ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 18 of 61 P.tge 18 of 61 Westinghouse Non-Proprietary C lass 3 3-2 values for the beltlinr and extended beliline materials were previously reported in PWROG-21037-N P and were included in RY integrity evaluations as part of this TLAA effort.
Fracture Toughn~ PropertiH The most up-to..date initial RTh'Dr and initial USE values ace documented in PWROG-21037-N P for VCSNS Unit 1. The beltline and exten~ beltline material properties of the VCSNS Unit I RV are presented in Table 3-1 hemn.. The differences between the unirnidiated RTNDr values summarized in the FSAR and those determined herein are a result of a change in curve-fitting method (hand-dra\11'11 versus hjperl,olic tangent) used to fit the Chatpy V-notch test data.
Chemil.try F actor \'alues The chemistry fa ctor (CF) values were calculated using Positions 1.1 and 2.1 of Regulatory Guide 1.99, Revision 2 (Reference 8). Position 1.1 uses Tables 1 and 2 from the R.egn!atoty Guide along with the be-St-estimate copper and nickel weight percent values ( contained in Table 3-1 ). Position 2.1 uses the sun-eill.ance capsule data from all capsules tested to date and surveillance data from other plants, as applicable.
Credibility ev-alnations of the VCSNS Unit 1 sunreillance data ar-e prorided in Appendix A of this report.
The calculated capsule flue.nee values ar-e provided in Table 2-1 and are used to detennine the Position 2.1 CFs as shown in Table 3-2. Table 3-3 summarizes the Positions 1.1 and 21 CF values deteanined for the VCSNS Unit 1 RPV belttine and extended beltline materials.
WCAP-18728-NP An.,aust 2022 Re\>ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 19 of 61 P3ge t9 of6t Westinghouse N o11-Proprietary Class 3 3-3 FLAAaE l lGI.HENTS
- F. t~11 00A lDS 18 TlflU 23 REF. l ~llOOA 1--Repr~ents \ eswl ~-eld~ I ,.__ _ __ _ _ Intermediate Shell Figure3-l RPV Schematic for VCSNS Unit I WCAP-18n&-NP Angust2022 Rf\'ision3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 20 of 61 Paoe 20 of61 Westinghouse Non-Proprietary Class 3 3-4 Table 3-1 Best-Estimate Cu and l\1 Weight Percent Values, Initial RTl'\DT Values, and Initial l JSE
\'n.lue-s for the VCSNS Unit I RP\. Brltlin.e and Extended Bl'ltline Interials
lnilial UnilTlldiated OJ laterW Identifkation Wt. % Cu Wt. %?\1 RTN>T USE
( OJ) ffi (ft-lb)
Bdtline Intermediate Shell 11-1 (Heat# A9154-1) 0.10 0.51 21 0 76 Jnlmnediate Shell 11-2 (Heat # A9153-2) 0.09 0.45 -20 0 107 Lower Shell 10-1 (Hl'llt # C'9923-1) 0.08 0.41 5 0 106 Lower Shell 10-2 (Heat# C'9923-2) 0.08 0-41 4 0 92 Intemifdiate Shd.l Long. Weld Seams BC & BD (Heat # 4P4784, Flux Type Lmde 124, Lot # 3930)
Jntezmediate to Lower Shell Cm:. Weld Seam AB 0.05 0.91 -49 0 86 (Heat# 4P4784, Flux Type Linde 124, Lot# 3930)
Lower Shell Long. Weld Seams BA & BB (Heat H4P4784, Flm Type Linde 124, Lot# 3930)
Entnded Bfltline Nozzle Sbdl 12-1 (Heat # C9955-2) 0.13 0.57 9 0 101 Nozzle SMll 12-2 (Heat # C-0123-2) 0.12 0.58 15 0 91 Inlet Nozzle 436B-1 (Heat # Q2Q41W) 0.127<bl 0.76 -20 0 152 Inlet NozzJe 4368-2 (Heat# Q2Q39\ll) 0.127<b) 0.82 0 0 115 Inlet Nozzle 4368-3 (Heat fl Q2Q39W) 0.12'7tl>) 0.82 -20 0 138 Outlet Nozzle437B-1 (}mt# Q2Q40) 0.127<bl 0.85 -10 0 159 Outlet Nozzle 437B-2 (Heat # Q2Q40W) 0.12'7tl>) 0.80 -10 0 165 Outlet Nozzle 437B-3 (Heat # Q2Q44W) 0.12700 0.78 0 0 155 Nozzle to Intennediate Shell Cm:. Weld Seam AC (Heat # 4P4784, Flu:t Type Lioo.e 124, Lot# 3930) 0.05 0.91 -49 0 86 Nozzle Shell Long. Weld Seams BE and BF(<>
o.cw=> 1.01<<) 10C<) rf<) go(<)
Inlet/Outlet Nozzle Foigings to Nozzle Sh.ell Weld Seams l5AIB!C & 16AIB/O<)
Surreillance Material<">
Jntemiediate Shell 11-1 (Heat# A9154-1) - - - - -
Smveillance Weld (Heat# 4P4784, Flt<< Type Linde 124, Lot# 3930) 0.04 0.95 - - -
Nol!!s:
(a) Th! infonnation is mradi>d from PWROG-21037-NP ~ 7). All 1.Mll!S ;a:re based on inf~ extracted from the V.C.
SlUllllll!I" Unit I CMIRs aDdlor TI!S.Sel fabncaliQll IKOrds, ww= noted cihem-ise..
(b) Gensic nlue for SA-508 Cbss 2 IJDZZ!,, fuq;inp fuzn PWR.OG-15109-NP-A ~ 9).
(c) Th! specific hilt mmiber us"'1 in weld 22IDS could not be ide.ntitied. To addnss lhese situations, ..mies "-eel! determine~ on a re1,;m, of all V.C. 5tmllill!I" v.-eld heats used in lhe flbrication of the VCSNS Unit I RV. The.e generic n lue; m!11! defined in PWR.OG-21037-NP ~ e 7).
(d) Sun-e:ill.mce pLm and v.-nd data ~ in WCAP-16298-NP (Re.mmce 10).
WCAP-18728-NP Allgust 2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 21 of 61 Page 2 1 of 61 Westinghouse Non-Proprietary- Cl.ass 3 3-5 Table 3-2 Calc ulation of Position 2.1 CF Values for VCS?\S Unit 1 Sun;eilla nce Iaterial,;
Capsult Flnt c~II) lliamred IF P&R.Tiwr Material Capsule :fP) ART~'> FF1
(:s: IOU o/cm2, ("F)
("F)
E>l.OMe\)
u 0.675 0.890 36.1 32.1 0.792 V 1.54 1.119 532 59.6 1.253 Inlennedi.ate Shell 11-1 X 251 1.247 383 47.8 1.555 (Longitudinal) w 4.63 1387 662 91.8 1.924 z 653 1.451 98.9 143.5 2.106 u 0.675 0.890 14.5 12.9 0.792 V 154 1.119 32.1 35.9 1253 X 251 1247 26.7 33.3 1.555 Intermediate Shell 11-1 (I'ransvene) w 4.63 1387 57.8 802 1.924 z 653 1.451 87.0 1263 2.106 SUM: 663.4 15261 CFn-1 =L(FF
- MlT= ) + I:(FF2) =(3'll..7)+ (7.630) =423Gf Notes:
(a) F1neice taken from Table 2-1.
(b) FF= fluem:e bdir = f."1'1 - 010-~crn_
(c) Musurai MtTi= bkmfnm WCAP-16298-NP ~ 10). The VCSNS Unit l sun-ullmi:e weld musured MlT,mr results m-n, been adj~ by a ratio of 126 ID accomd for chmiistry difFeruices beiv;em the Heat ii 4P4784 sun'2illance weld (CF = 54°F) arid R.V welds (CF= 6S"F). The unadjusted musured MIT>= nlu.es are listed in ~ -
WCAP-18n8-NP August 2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 22 of 61 Page 22 of 61 Westinghouse Non-Propriet.uy Class 3 3-6 Table 3-3 Summary of the VC:SXS Unit I RP\' Beltline, Extended Beldine, and Surnillrulce Material CF Values based on Regufatory Guide 1.99, Rmsion ?,
P osition I.I and Position ?.l Chr ~
- Fac tor Mattrial Drmiption Position Ll <-> Position ?.I (I>)
C°F) (OF)
Btltli111 Intermediate Sht:ll 11-1 (Heat HA9154-1) 65.0 43.5 Intermediate Shell 11-2 (Heat # A9153-2) 58.0 -
Lower Shd.1 10..1 (Heat# C9923-D 51.0 -
Lower Sbell 10..2 /'Heat# C9923-2) 51.0 -
Intmnedi.ate Shell Lomr. Weld Seams BC & BD 68.0 42.3 Infeimediate to Lower Shell Cire. Weld Seam AB 68.0 42.3 Lower Shell Lon2. Weld Seams BA & BB 68.0 42.3 E.umdrd Beltlin, Nozzle Shell 12-1 ffieat # C9955-2) 90.1 -
Nozzle Shell 12-2 l'H""t # C0123-2l 82.6 -
Inlet Nozzle 4368-1 (Heat# 0 2041Wl 92.1 -
Jnlet No.zz:le 436B-2 l'He.at # 02039W) 93.0 -
Inlet Nozzle 436B-3 (H..,.t # 02039W) 93.0 -
Outlet Nazzle 437B-l lHe.at # 02040) 93.0 -
Outlet Nozzle 437B-2 lHP.at # 02040\\') 93.0 -
Outl.etNo2Zle 437B-3 <Heat #02Q44W) 92.6 -
Nome to Intennediate Shell Cire. Weld Seam AC 68.0 42.3 Nozzle Shell Loim. Weld Se.ams BE and BF 82.0 -
Jnlet/Oatlet Nozzle Forl!llll!S to NazzJe Shell Weld 82.0 -
Sun*~1la11c-e M11tmals Inlamediate Shell 11-1 65.0 -
Swveillanre Weld 54.0 -
Notes:
(a) All '\.Ines = based on Tables 1 md 2 o f ~ Guide 1.99, Rerision 2 (Positia:i 1.1) using the ui md Ni ~
pacmtruues git:min Table 3-1 oflhisreport. DrlEi illdicm WWI a at~isnotappmble to the maJ:aial.
(b) Values are from Table 3-2 of this report.
WCAP-18728-NP .Angus12022 R.evision3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 23 of 61 Page 23 of 61
'Westinghouse Non-Proprietary Class 3 4-1 4 PRESSURIZED THERMAL SHOCK A limiting condition on RPV integrity known as Pressurized Themial Shock: (PIS) may occur during a severe system transient such as a loss-of-eoolaot accident (LOCA) or steam line break. Such transients may challeflge the integrity of the RPV under the following conditions: severe overcooling of the inside sncface of the vessel wall followed by high pressurization, significant degradation of vessel material toughness caused by radiation embrittlement, and the presence of a critical-size defect anywhere within the vessel wait In 1985, the USNRC issued a foonal ruling on PTS (10 CFR 50.61 (Reference 6D that established screening criteria on pressurized water reactor (PWR) vessel embrittlement, as measured by the maxinmrn reference nil-ductility transition temperature in the limiting beltline component at the end of license, termed RTPTS.
RTFrS screening values were set by the USNRC for belttine axial welds, forgings or plates, and for beltline cin:umferential weld seams for plant operation to the end of plant license. All domestic PWR. vessels have been required to evaluate vessel embri.ttlemem in accordance with the criteria through the end of license.
The USNRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisions maJre the procedure for calculating the reference temperature for pressurized then:nal shock: (R.Tm) values consistent with the methods given in ~gulatOty Guide 1.99, Revision 2 (Refer-ence 8).
These accepted methods were used with the smface tluence values of Section 2 to calculate the following RTPTS \lalues for the VCSNS Unit 1 RPV materials. The end ofSPEO RTvm caknlations ace presmted in Table 4-1.
PTS Conclu..ion All of the beltline and extended beltline materials in the VCSNS Unit 1 RV are below the Rf= screening criteria values of270"F foe base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds through SPEO (/2 EFPY). These RTprs values are based on the revised initial RTNDr '-'rulles in PWR.00-21037-NP (Reference 7), which are de"v-eloped using ASME Section III (Reference 11) and, if needed, NUREG..0800, BTP 5-3 (Reference 12) methodologies. Limiting tluence values corresponding to the lowest extent of the nozzle welds were med to ca.kn.late fhe RTPTS values foe both the nozzle welds and nozzle forgings.
The VCSNS Unit 1 limiting RTm value for base metal and longitudinal welds at 72 EFPY is 152.5°F (s ee Table 4-1), which c01TeSpon.ds to VCSNS Unit 1 Intermediate Shell 11-1 based on Regulatory Guide 1.99, Position 1.1. Note, that there is sut'\"eillance data available for this material that indicated the ~T:imr will be less than that predicted by RG 1.99, Position 1.1. Howevec, because the surveillance data was detennined to be non-conservative, it is not credited here. Tue VCSNS Unit 1 limiting RfPTS v alue for circnmferentially oriented welds at 72 EFPY is 42.5"F (see Table 4-1), which corresponds to the VCSNS Unit 1 Intennediate to Lower Shell Circumferential Weld Heat # 4P4784 oosed on Regulatory Guide 1.99, Position 2.1 with credible surveillance data. The credible surveillance data for Heat# 4P4784 supersedes the higher RTPIS based onRG 1.99, Position 1.1. Note, both the Position 1.1 an.d2.1 remain below 300°F.
WCAP-18728-NP Angust2022 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 24 of 61 f'a9e24 al GI Werungbonse Non-Propri*t.uy Class 3 4-2 Tabk4--1 RTns C akula tio 1L1 (or \"CSXS Unli 1 at 7l EFPy<,/
R.G. PndKtNI SmFhm,cr RT~(t) M
- u.a1,ru1 1.99, CJ1') (x 101' alr111',
Surf. '11 CJ.1 ART ART~
Rel".?
E > L 01f,\ ')I~
fill) ("F)
("F)I')
C'F) ~ ("F) ("F)
Position Boldinr Malt riah hitmiiedill* Shell l l-1 (Heal # .A9_154-l) 1.1 tit9 .i~ 1.501 21 ?1.5 9.9 J7.0 34.0 151.~
Usinv 1IOIHffdibl* .naw:i//Jmce dattfV 2.1 43.5 9.04 1..501 21 653 0.0 17.0 34.0 1203 lnlmmdillr Shell 11-2 (Hell fJ A!l153-2) I.I 58.0 9.04 1.501 -20 87.0 0.0 17.0 34.0 101.0 l.ow,r Shell 10-1 (Heat II C9923-1) I.I 51.0 9.06 1.501 5 16.6 0.0 17.0 34.0 115.6 1.cm-e, Shell 10-2 (Hell I C9923-2) I.I 51.0 9.06 1.501 4 16.6 0.0 17.0 34.0 114.6 I n t ~ Sbell loog. Weld Seaim BC&BD 3.03 1.293 -49 87.9 0.0 28.0 56.0 94.9 1.1 68.0 (Bnt # 4P4784)
UrinJ? cndibh nm..nJance d,m/11 2.1 423 3.03 1.293 -49 54.7 0.0 14.0 28.0 33.7 I m ~ lo i...n- Shell Cirt. Weld Sem, AB (Hell 11 1.1 68.0 9.04 1.501 -49 1020 0.0 28.0 56.0 109.0 4P4784)
Ustn~ adbh !illr'Milmre datifll 2.1 423 9.04 1.501 -49 635 0.0 14.0 28.0 42.5 Lov,-er Shell Loog. Weld ~ ams BA & BB 1.1 68.0 3.08 1.297 -49 88.2 0.0 28.0 56.0 95.2 (Hnt l 4P4784)
~ crtdib/.e JJDWi11anc11 tbfdll 2.1 423 3.08 1.297 -49 54.9 0.0 14.0 28.0 33.9 utmc M JSfll IJlf Malt lUIS Nozzle Sbell 12-1 (Heat II C9955--2) I.I 90.1 0.400 0.746 9 67.2 0.0 17.0 34.0 110.2 Nozzle Sbell 12-2 (Hffl II C0123-2} 1.1 82.6 0.400 0.146 15 61.6 0.0 17.0 34.0 110.6 lnlrt Nozzle 436B-1 (Hut iJ Q2Q41W) 1.1 92.1 0.0310 0.224 -20 20.6 0.0 103 20.6 21.2 Inld Nozzle 436B-2 (Hat il 02039W) 1.1 93.0 0.0310 0.224 0 20.8 0.0 10-4 20.8 41.6 Inltt Nozzle 436B-3 (He.II fJ Q2Q39W) 1.1 93.0 0.0310 0.224 -20 20.8 0.0 10.4 20.8 21.6 Outlti Nozzle 437B-1 (Heat II Q2Q40) 1.1 93.0 0.0132 0.132 -10 123 0.0 6.1 123 14.5
~ Nozzlt 4378-2 {H5t # Q2Q40W) 1.1 93.0 0.0132 0.132 -10 123 0.0 6.1 123 14.5 ru!el Name 4378-3 ~ II Q2Q44W) 1.1 926 0.0132 0.132 0 122 0.0 6.1 12.2 24.4 NOlZle to Immnediate Shell Cirt:. W.M Semi AC 1.1 68.0 0.425 0.762 -49 51.8 0.0 25.9 51.8 54.7 (Bnt Ii 4P4784}
thing ardibh SJU\wlJmrl dalrlfl 2.1 42.3 0.415 0.762 -49 322 0.0 14.0 28.0 11.2 Nozzle Sbfll Long. Wdd Semi.s BE md BF I.I 820 0.400 0.762 10 61.2 0.0 28.0 56.0 127.2 WCAP-18728-KP ~ 2022 Re\i'iiou 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 25 of 61 P*25al81 Westingbous<<e Noo-Proprietuy Class 3 4-3 Tablt 4-1 RTns Cakalatloo1 for YCSXS Unit I :it 72 £JP\,(,)
R.G. PNdictNI S1uf. Fhlm<< Surf. RTlll1I(C) M *.\RT 1.99, CTI') (x 101' n/cm 1, Cl! 0~
Matuial Rn.? Illll (°I) MtT.sur (°I) fF)CQ ("F) ("F)
PMt.t..
E> l .illt\')(<l ("J)l*l lnlet/Outld Nozzle Forpngs to Nozzle Shell Weld u 820 0.0310 0224 10 18.4 0.0 9.2 18.4 46.7 Seams 15A/BIC& 16A/B/C Nott.:
(a) 'Ju 10 CfR. 50.61 IDl!ffl0do!o11 wu 1llilimd mtbo ~ mtbo RTm nlms_
(l,) a..misaybcbs ara twn &an Tabl.3-3.
(<) ~ lilbD fram T.b!e 2-2 oftbi:, n;,crt (d) FF=fba:ebcmr* _,._.,..,.Oli_
(e) JU",_,lll nme, bnD fram Tm!ti 3-1.
a, (f) P.- 10 CfR. S0.61, the boselDl!lala* =17'1' , d i m ~ cbt, ..-. DOC-<ndiole moot DZl to dolmnim tho CF, ;md b i,,,,, IDl!laJ =8.5"1' -..fm cmlible smmlhm:e dm an used. Also, p,< 10 CfR. 50.61, 1he ,n.ld -.I <JA ~ 2S"F ,.t.,, _ , . ~ dab ""' ""'1<!1!dible craot med lo cl.:mm. tlie CF, am! 1he weld motaI en & l4"F when credible scrnilJmee dab""' used.~. a*nood-o:D!ed 0.5'MIT>= f<<oilm base metils << ..uh, will,<< ffllhoo:I ""'1!ill:mce dab.
W TIM cnd,1,i]ijyenlmtm fu iho ~ Unit l ~ elm m Appe,dix A oflhisn,port clonmiad t1m b VCSNS Umt l smmllma! dab b-tbo ~ Sbell ll-1 (Hul#A915t-l) is d e e m e d ~ ml dio Sar\ Weld (Hmf 41'4784) is wmodawd,.b!a.
WCAP-18723-NP August2022 Reruian 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 26 of 61 Page 26 of ll1 Westinghouse N on-Proprietary Class 3 5-1 5 "UPPER-SHEL F ENERGY The decrease in Cbaqiy upper-shelf energy (USE) is associated with the detttmination of acceptable RPV toughness during the license renewal period when the vessel is exposed to additional inadiation..
The requirements on USE are included in 10 CFR 50, Appcndi,"'t G (Refet"cnre 3). 10 CFR 50, Apperulix G rtqUires utilit~s to submit an analysis at least 3 ye.ars prior to the time that the USE of any RPV material is predicted to drop below 50 ft-lb, as measured by Cluupy V-notch specimen testing.
There are two metho~ that can be used to predict the decrease in USE with irnldiation, depending on the availability of credible smveillance capsnle data as defi.oed in Regulatory Guide 1..99, Revision 2 (Reference 8). Foc vessel beltline materials that are not in the smveillance program or have non-credible data, the Chmpy USE (Position 1.2) is assumed to decrease as a function of flucnce and copper content, as indicated in Regulatory Guide 1.99, Revision 2. When two or m ore credible surveillance sets become available from the reactor, they may be used to detemiine the Chmpy USE of the smveillance material The surveillance data are then used in conjunction with the Regulatory Guide to predict the change in USE (Position 2.2) of the RPV material due to irradiation. Per Regulatory Guide 1.99, Revision 2, when credible data exist, the Position 2.2 projected USE value should be used in preference to the Positioo 1.2 projected USE value. Note, if data from the S[l[Veillance materials is detennined to be non-crechole for determination of AR.Twr by Credibility Criterion 3 of Regulatory Guide 1.99, Revision 2, then "they may be credi"'ble foe determining decrease in uppec-s.helf eneigy if the uppei- shelf can be clearly determined, following the ddinition given inASTM E 185-82.n The 72 EFPY Positioo 1.2 USE values of the vessel materials can be predicted using the cotresponding l /4T fluence projections, the copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2 (see Figure 5-1).
The predicted Position 22 USE values are detemiined for the RV materials that ai-e contained in the surveillance program by ming the reduced plant sunreillance data along with the correspondmg 1/4T tlue:nce projection. The surveillance data was plotted in Regulato:ry Guide 1.99, Revision 2, Figure 2 (i.ee Figure 5-1) using the sun'eillance capsule fluence values documerued in Table 2-1 of this report for VCSNS Unit 1. This data was fitted by drawing a line parnllel to the existing Jines as the upper b ound of all the surveillance data. These reduced lines were used instead of the existing lines to determine the P osition 2.2 end of SPEO USE values.
The projected USE values were calcula.ted to detennine if the VCSNS Unit 1 beltline and extended beltline materials remain above the 50 ft-lb criterion at 72 EFPY (end ofSPEO). These calculations are summarized in Table 5-1.
WCAP-18728-NP Allgust 2022 Re-.ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 27 of 61 Page 27 af61 Westinghouse Non-Proprietary Class 3 5-2 USE Conclusion As shown in Table 5-1, all VCSNS Unit 1 RV materials are projected to remain at or above the USE screening criterion value of 50 ft-lb at 72 EFPY The limiting USE value at 72 EFPY is 63 ft-lb {see Table 5-1); this value corresponds to Intermediate Shell 11-1 using P osition 22. The SU1Veillance data for lnteunediate Shell 11-1 is used despite it being detennined to be non-credible, as the upper shelf can be clearly determined for the swveilwlre specimens (see WCAP-16298-NP). Note, both the Position 1.1 and 2.1 results for Intemiediate Shell 11-1 remain above 50 ft-lb.
WCAP-18728-NP Angust2022 Rfvision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 28 of 61 Page 28 of61 Westinghouse N on-Proprietary Class 3 5-3 Table 5-1 Predicted USE \ ':tlues at 7l EFPY for the \'CSXS Unit 1 Beltline and Extended Beltline Materials Projeded Proj<<ted 1/.ff Flu,nce Unimidiated Wt ~i USE M.iterial (:s: 1019 o/cui1-, USE USE Cu<*> E > 1.0 feY)l'l (ft-lb)C*l Decnase (ft-lb)
(%)
Position 1.2{<)
Jntmnediate Shell 11-1 (Heat# A9154-l) 0.10 5.68 76 29 54 Jnteonediate Shell 11-2 (Heat#A9153-2) 0.09 5.68 107 29 76 Lower Shell 10-1 (Heat# C9923-l) 0.08 5.69 106 29 15 Lower Shell 10-2 (Rm# C9923-2) 0.08 5.69 92 29 65 Jntamediate,Shell l.Dng. Weld SeaJns BC 0.05 1.90 86 22 67
&BD (Heat#4P4784)
Jntennediate to Lower Shell Cm:. Weld 0.05 5.68 86 29 61 Seam AB (Heat H4P4784)
Lower Shell Long. Weld BA&BB 0.05 1.93 86 23 66 c&at # 4P4784)
Nozzle Shell 12-1 <Heat # C9955-2) 0.13 0.251 101 17 84 Nozzle ~ 12-2 lHeat #J C0123-2) 0.12 0.251 91 16 76 Inlet Nozzle 4368-1 (He.at# Q2Q41W) 0.127 0.031()(11) 152 10 137 Inlet Nozzle 4368-2 (Heat# Q2Q39W) 0.127 0.0310Cd) 115 10 104 Inlet Nozzle 4368-3 (Heat# Q2Q39W) 0.127 0.0310Cd) 138 10 124 Outlet Nozzle 437B-1 (Heat# Q2Q40) 0.127 0.0132(d) 159 9 145 Outlet Nozzle 437B-2 (Heat# Q2Q40W) 0.127 0.0132(d) 165 9 150 Outlet Nozzle 437B-3 (Heat # Q2Q44W) 0.127 0.0132(11) 155 9 141 Nozzle to Jntennediaie Shell Ciic. Weld 0.05 02o7 86 14 74 SeamAC(Heat#4P4784)
Nozzle Shell Long. Weld Seams BE & BF 0.06 0.251 80 15 68 ToletfOutlet Nozzle Foigings to Nozzle 0.06 0.031()(d) 80 9 73 Shell Weld Seams 15AJB/C & 16A/B Position 2.2<<>
Intermediate Shell 11-1 (Heat #A9154-l) 0.10 5.68 76 17 63 Intmnediate Shell Long. Weld Seams BC 0.05 1.90 86 9 78
& BD (Heat# 4P4784)
Intermediate to Lower Shell Circ. Weld 0.05 5.68 86 12 76 Seam AB (Heat# 4P4784)
Lower Shell Long. Weld Seams BA & BB 0.05 1.93 86 9 78 (He.at# 4P4784)
Nozzle to Jntennediate Shell Cm:. Weld 0.05 0.2o7 86 6 81 Seam AC (Heat# 4P4784)
Notes coubmed on ml.lo~ p;1ge.
WCAP-18728-NP Angust 2022 Revisio11 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 29 of 61 Page 20of61 Westinghouse N on-Proprietruy Class 3 5-4 Nms:
(a) UlJllm" weight pereat ,-a!nes .and ummdiated USE ,'3lnes 'ITT!ff taken from Tab* 3-1 of this RpOrt. Ifthe base mebl or -nitld Cu
~-eightpercentaps m, below the mmmmn1.-alw! pre;mted inF~ 2 of~Guide 1.99 (0.1 fcr base metal .md 0.05 forU"l!lds),
fbl!D the Cu m!i.&ln~ = c:aasanti,.-el:y rouaded up to the mi:uim:n:nnlue forpmjeded USE deamse determination (b) V;,.Jues tahu mm Table 6-2 oflhi.s report Flumce .aas abol.'1! 1()1 7 n/mr- (E > 1.0 MeV) but below 2 x 10 n!cm. (E > 1.0 17 1 MeV) were l'OllDlkd to 2x 1011 n.fcml {E> 1.0 MeV) mien demmining the% deause beause 2 x 1011 nlmr-is the l ~t f l ~
dispby1!d in Fi~ 2 ofRG 1.99.
(c) Position 12 perceDQge USE ~ , "llhlti l'l'ln cakwmd by pJottmg the lf4T ni...- 1.,wes on RG 1.99, FtgDn! 2 and using the m.m.rw-specific Cu \rt. % n1ues. The puce,t-loss l.ines were atmded into the low £h.wict i1Q of RG 1.99, Figln 2, ie.,
below 1011 nlcm.1, in order to delEmine tbe USE % ~ as ~ Position 22 percm2'1! USE deCJuse nlms ,.,._
deleimilll!d by dmving m upper-bound l i n e ~ to the existing RG 1.99, Figure 2 lines through the applicable smnilhnce chti poim. Thtse n=11s sbao1d be used in prafi!renc-e to the existing gr;qii liDes fa,- Me111i11i11g /lie decrease in USE, becmse the 5111\-eill.anm data is credible.
(d) V~ an, the muimmn fl=e 1.-al!Jei instead of the l/4T fiuence nlues.
WCAP-18728-NP Angust2022 Re\-ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 30 of 61 Page 30<>fl!I
~ ~ finghome Non-Proprietary Class 3
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- r fJuo-Ju:'"'
I t :Ul.d.
I I I I I.II l .ooe+l7 U OE<!8 UU+1' I 11,utton Fluonc t. n/tm1 IE > 1 lhVI Rr;ubtory Guidr 1.99, Rriision 2, Pogtion 1.2 & l.l Pndktrd D<<nase m l ),pfl"-ShrlfEMrcr a, a Tunnion of Copper and Flan~r for ycs.,_..._s Unit l :ir thr End of SPEO (72 EII'Y)
WCAP-1 8728-NP .Au2llSt 20ll
~ moo 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 3 1 of 61 Page 31 of61 Westinghouse Non-Proprictaiy Class 3 6-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LDflT CURVES Heatop and cooldown limit cwves are calculated using the m ost limiting value of RTNDr (reference llil-ductility transition temperature) corresponding to the limiting material in the beltline region of the RPV The most limiting RT= offhe materi.alin the core (beltline) regi..on of the RPV is detennmed byusing the miirradiated RPV material fracture toughness properties and estimating the imtdiation-induced shift (ARTNDT).
Ci.I ADJUSTED REFERENCE 1El'\IPERATURES CALCULATION RTNDr increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactors life, ARl'NDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDr- Using the adjusted reference temperature (ARI) values, pressure-tempera.tore (P-1) limit curves ai-e detemiined in accordance with the requicements of 10 CFR P art 50, Appendix G (Reference 3), as augmented by Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code (Reference 13).
The P-T limit cmv-es far nonnal heatup and cooldown of the p rimary reactor coolant system for VCSNS Unit 1 were previously developed in WCAP-16035-NP (Reference 14). The existing P-T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material Since the development of the cUfVes, the floence values and initial material properties used to cakolateART values have been updated and an applicability check of the current P-T limit co.rres is appropriate.
To confian whether or not the current P-T limit curves will remain valid through the PEO and through the SPEO, updated ART values for the limiting materials w ere computed to account for updated 56 EFPY and 72 EFPY fluence vames, updated Chemimy F actor vames, and updated initial RTNDT values. The Regulatmy Guide 1.99, Revision 2 (Reference 8) methodology was used along with the sum.refluence of Section 2 to calculate ART values, which are summarized in Table 6-3 through Table 6-8. Note, the inlet/outlet nozzle fotgings and associated welds neglect attenuation through the material; thus, ART calcnlations are only needed at one location, i.e., the location ofma.-.wnum fluence. Table 6-1 and Table 6-2 show the sw:face, l /4T, and 3/4T floencc values for 56 EFPY and 72 EFPY, respectively.
ART projections contained herein are based on those projected fluence valoe-S with a 1.1 bias on the pecip.b.eral and re-entrant comef' assembly relative powers.
WCAP-18728-NP August 2022 R.evision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 32 of 61 Page 32of61 Westinghouse Non-Proprietary Class 3 6-2 Table 6-1 \'CS:NS Uoit I Fluence and Fluence- Factor Yalues for the- Surface-, l /4T, and 3/4T Locations at 56 EFPY Surface l/41' Flunce<-> 3/-IT Fluence<->
Flutnct<-> Surface 1/.ff 3/4T Material Description Ff(li) (:x 1ou n/cm2, Ff{b) (:x 101' n/cm2, f"?)
(:i; 101' n/cm1, I '.> 1.0:lli\') &l.O le\)
&l.O. le\)
<h"ue Intennedia1e Shell 11-1 6.99 1.463 439 1376 1.732 1.151 (Heat# A9154-1)
Intennediate Shell 11-2 1.463 439 1376 1.732 1.151 6.99 ffif'llf# A9153-2)
Lower Shell 10-1 1.463 4.40 1376 1.735 1.152
(&it# C9923-1) 7.00 Lower Shell 10-2 7.00 1.463 4.40 1376 1.735 1.152 fHeat # C9923-2)
Intermediate SMl Long. Weld 1.48 1.108 0.582 0.849 235 1231 l'Heat # 4P4784)
Inlennediate to Lower Shell 439 1376 1.732 1.151 Cm:. Weld(Heat# 4P47S4) 6.99 1.463 Lower Shell Long. Weld 1.112 0.592 0.853 239 1235 u o l'Heat # 4P4784) fate11tled Bntline Nozzle Shell 12-1 0309 0.678 0.194 0562 0.0766 0.366 (Heat # C9955-2)
Nozzle Shell 12-2 0.194 0562 0.0766 0.366 0309 0.678 (H~t # C0123-2)
Inlet Nozzle 436B-1 See Note(d)
(Heat #Q2Q41W) 0.0239 0.192 Inlet Nozzle 436B-2 0.192 SeeNote(d) 0.0239 m eat# 02039W>
Inlet Nozzle 436B-3 0.0239 0.192 See Note(d) m eat # 02039\V\
Outlet Nozzle 4378-1 0.0102 0.111 See Note (d)
(Heat# Q2Q40)
Outlet Nozzle 4378-2 See Note(d) 0.0102 0.111 ffiea.H# Q2040WI Outlet Nozzle 4378-3 0.111 SeeNote (d) 0.0102 (Heat# 02044W)
Nozzle to Intermediate Shell 0328 0.693 0.206 0511 0.0813 0.377 Cin:. Weld (Heat # 4P4 7&4)
Nozzle Shell Long. Welds(<) 0309 0.678 0.194 0562 0.0766 0.366 InleVOutlet Nozzle F<<gings to Seenote (d) 0.0239 0.192 Nozzle Shell Welds Notes:
(a) The smface fluence nlues for tbe RV materials= dimmined by iutapohtion from d3ta in Table 2-2 The l/4T and 314T fhEnce nl-. were calcnlm.d from tbe SUlface fhwJce, tbe R.V beltline thi~ (1-75 inclies) .md eqmtiou f= f....r
- e4 24 (*i from
~toiy Guide 1.99, Rrnsiao 2, "-here x =the depih into the TI!Ssel wall (mcliies).
(b) FF= fhmlce fador = II 0.IO'"" Cll)_
(c) Exposure '\~ for the DDzzle shell loogitadmal m!lds all! bounded by tbe exposure nlues fix the DOZZle shell (d) .Analysis of the DOZZle f<<pngs and associah!d ~-elds <ft conser.am:ely permm,,d ~ the maximmn fhwlce through the '\'eSsel wall WCAP-18728-NP Angust 2022 R.e,,ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 33 of 61 Page 33ofCII Westinghouse Non-Proprietary Class 3 6-3 Tablc- 6-2 VCSNS Unit I Flurnce i:tnd Ilutnce Factor Values for th, Surface, l/4T, and 3/4T Locations at 72 EFPY Surfare 1/.ff Flue-nee(-> 3/.ff Flumce<->
Flutnce(-> Surface- 1/.tT (x 101' n/cm2, 3/.ff Material Des<:ription Fftb} (x IOU n/cm2, FF(\) f?)
(x IOU n/em7, E > l.O.Me,') &l.OMe,')
&l.OAM')
Beltfi11e Intennedia.te Shell 11-1 9.04 l.501 5.68 1.427 2240 1218 (Heat# A9154-l)
Intennediate Shell 11-2 9.04 l.501 5-68 1.427 2.240 1218
<Hea.t # A9153-2)
Lower Sheil 10...1 9.06 l.501 5.69 1.427 2.245 1219 (Hmt # C9923-1)
Lower Shell 10...2 9.06 1501 5.69 1.427 2245 1219 (Heat # C9923-2)
Jnt=cdia1e Shell Lang. 3.03 1.293 1.90 1.176 0.751 0920 Weld (Rea.I # 4P4784)
Intennediafe to Lower Shell Cm:. Weld 9.04 1.501 5.68 1.427 2240 1218 (Heat # 4P4784)
Lower Sbell Long. Weld 3.08 1.297 1.93 1.180 0.763 0.924 (Heat # 4P4784) fat.e111Jed Btftli11e Nozzle Shell 12-1 0.400 0.746 0251 0.625 0.0991 0.415
<Heat # C9955-2)
Nozzle Shell 12-2 0.400 0.746 0251 0.625 0.0991 0.415 (Heat # C0123-2)
Inlet Nozzle 436B-l 0.0310 0.224 See Note (d) fR..,.t# O2DtlW\
Inlet Nozzle 436B-2 0.0310 0.224 SeeNote(d)
(Hea.t # Q2Q39W)
Inlet Nozzle 436B-3 0.0310 0.224 SeeNote (d)
(Hea.t # 02039\\-'}
Outlet Nozzle 437B-l 0.0132 0.132 See Note(d)
<Heat #Q2rnO)
Outlet Nozzle 437B-2 0.0132 0.132 SeeNote(d)
(Heat # Q2Q40W)
Outlet Nozzle 437B-3 0.0132 0.132 See Note (d) ffiea.t # 02044W\
Nozzle to Intetmediate Shell 0.425 0.762 0261 0.640 0.105 0.427 Cm:. Weld (Heat# 4P4784)
Nozzle Shell Long. Welds(<) 0.400 0.746 0251 0.625 0.099 0.415 Inlet/Outlet Nozzle Foigings 0.0310 0.224 See note (d) to Nozzle Shell Welds Nolr.:
(:a) The smflice ilw!nce .:alues for the RV materiili "we daemined fnm Table 2-2 lhe l/4T aDd 3/41' :llumc.e .:alues m:re calculated from 1he sud'a~ f1uecce, lhe RV beltlme thiclmess (1.15 inch,,;) mid eqmtion f=f....r
- rl <il from Regal:afmy Guide 1.99, 24 Reruion 2, mlS1! x =th,, depch into Iba .e,;sel "-all (mches).
(b) fF = finence &dor- = -0.la-lottl'II_
(c) &po;ure ,.-ahu,s far the .nozm sbell loogitudin.l "'1ds .ire bounded by the expo~urn .lues fodho! nm:z1,, sh.ll.
(d) Anal}~ of~ IICIIZZl.e fmgin;s and assoc:utl!d '1\-..ld.s ue consentn,oly pmormed usmg the maxim.mi 6.IW!Ce ~ ~ i.~sel
'I\.Jl.
WCAP-18728-NP Augmt2022 Revision3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 34 of 61 Pago 34 ol 61 Wertingbcuse Non-Propriotary Cl.ass 3 6-4 Tabtt6-3 Calculation of the \"CS:'.'i"S Ua.it lART\":ilurs a l th* l/4TLoration for th* Iua~lo r\"~*d &lllint andExtrn.d,d Bel.dint M.:ilfriab at Ibo End of PEO (56EFPY)<>l R.G. ll-lT B n,ure Prtdict,d U 9, l/-lT RT1'DIIC) GI cu M ART hl.altrial Cft'I (:,: 10" D/rm1, Ff('l (°l)!'l ilT1WT ("I) fFf1l ("I) ("I)
Rn-* .? E > 1.0 Mt\)l'l l"'F)
Positioa Btl tlint Maltriah lulmnediate Sb,1111-1 (Heat ii A91 >t--1) 1.1 _65.0 439 1376 ..11 ~ -4 l>:O l],!) .34.0 144.4 Using nor,.m:dibl, sunYilllmu daf,µ) 2.1 435 439 1376 21 59.9 0.0 17.0 34.0 114.9 In!nmtcliate Shell 11-2 (Beat# A9153-2) 1.1 58.0 439 1376 -20 79.8 0.0 17.0 34.0 93.8 Lawer Shell l0-1 (Bul #C9923-l) 1.1 51.0 4.40 1376 5 702 0.0 17.0 34.0 1092 l.ow,r :.at.11 10-21):i,!al# l,YY.U-2) 1.1 51.0 4.40 1376 4 102 0.0 17.0 34.0 108'
~ "Laig. w~ Sums BC& 1W 1.1 68.0 1.48 1.108 -49 153 0.0 28.0 56.0 823
(&at UP4784)
Using cndibl. sun*eiTlmtce dakftl n 4li 1.48 1.108 -49 46.9 0.0 1.fo 28.0 25.9 lulcmtdiote to Lower Shdl Cuc. Weld Seam AB 1.1 68.0 439 1376 -49 93.6 0.0 28.0 56.0 100.6
(&at i14P4784)
- ltitng mdibl. nuwilhmu dat!fl! 2.1
-423 439 1376 -49
-582 0.0 14.0 28.0 372 lower :illel.l I..oag. Wdd Semn HA & HH 68.0 150 1.112 -49 15.6 0.0 28.0 56.0 82.6 1.1
(&at114P4784) -
-423 -- -14.0- 28.0 26.1 Vimg tntlibl.n,,wfllmia w,J 2.1 1.50 1.112 -49 47.1 0.0 utrndrd Bellmlt Matuuh Now ~ 12-1 (Heat,. a>>55-2) 1.1 90.1 0.194 0.563 9 50.7 0.0 17.0 34.0 93.7 No2Zle Sbt.11 12-2 (Heal I! C:0123-2) 1.1 82.6 0.194 0.563 15 465 0.0 17.0 34.0 955 N~tn ln/rnnrdi,.., Shell Clrc. Wdd Se.miAC 1.1 68.0 0206 0577 -49 392 0.0 19.6 392 29.4 (Heat it 4P4'>
z.1mt a mible nu,*d1ltmc;,, dm,t,J 2.1
- 423 0206 0.5T7 -49
-24.4 0.0
-122- 24.4 -02 Nom Sbell Lmg. Wdd Seams BE md BF 1.1 82.0 0.194 0.563 10 46.1 0.0 23.1 46.1 1023
'Nolesmmiood cm~ - -
August 2022 WCAP-18728-NP Rrvision3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 35 of 61 P9 35of81 Westinghouse Non-Propridary Class 3 6-5 Nolo:s:
<*> n . . ~ Gmdo I.99, Kr.moo 2 me&oclo!o1Y was dilmd"' 11,e c.dcub6an or11,e ARTnmes.
(b) Climiisl,yfdor.s m,bbn lnmTable 3-3.
(c) flDllx:a tum 6mt Table 6-l aldii> npcrt (cl) FF*floe,,a,ladll,-* ll-*-,.
<*> Rl'N>T0.1 {U,,imdiDd RI'= ) ,-..'nest.km fnm Tab!, 3-L (0 !'..tho ~ al'~ Gaide l.99, h isi<,o 2, tllo base meblo-.= l'T"F for Pa.ilioll LI md Po,iticn 2.l ' l r i i h ~ ~ <bta, md tllo l,a,emebla,. =8.5"F fir Po,mm,.21 -.rida mdihl<! SlaTI!illancemta. Also,~ Rap,latory G,,;de 1..99, Rfiisiau 2, lhe \ftld mota!o-, s2S"F Ir Pmili<n l.l .md Posilioa 2.l uh DOD-CAdio!o ~
dab, a,d 1be 1'ul .,,.1a1 m
- 14"1' far Posmoa 2.l m!h m,!ib!, smnilbm:, dm. ~ . a.,. need coDt uceed 0.5*,urr..,, Car eilbor m,e ml!t&b or nlds, 11-ilh or mlhaul
.......:n.,,,, elm.
Sbell 11-1 (HutiA91.54-W D,e ~a"21mlia:, b lhe VCSNS Uni! J ..,,,'l!illm<edmin Apps,dix A dtmmined !bat 1h! Va.NS Unit I ~ d > b Cm Ibo ~
l) is d,omodll<llH:nd,1,!e mi lbe Sam,ill,m,o Weld (&,Iii 4P47S4) is doomod Cll!d,l,k_
WCAP-lBn~NP August 2012 Revision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 36 of 61 Page JG of Cl Westinghouse Non-Ptopri.m iy Class 3 6-6 Tablt6-4 C akula tinn nf tht \'CSXS Unit I ART \ '"aluts at tbt J U T Lo<-ation for th e R uctor \'essd Bt ltliu* and Enended Beltlin*
Ma teriah a l th* End of P EO (56EFPY)<-I R.G. JI.ff Fluencr Prtdkttd 1.99, Cf{b) 3/ff RT!M(C) GI Ol I ART Material (i: 10" a/err, FF(I/ ("F)t-l AR.TlllJ'T ("F) ("F)I'> ("F) ("F)
Rl"l".?
Position E >LOMe\ )I') ffi Btltlin* llittrials lnt~te Sbtll 11-1 (Heu II A9154-1) 1.1 65.0 1.73 _1.1~1 21 74.~ 0.0 17.0 34.0 129.8 Umw non--aodible SIDVdllance d1l1tP 21 43.5 1.73 1.151 21 50.1 0.0 17.0 34.0 105.1 lntmnrdim Sbtll 11-2 (Heu# A9153-2) 1.1 58.0 1.73 1.151 -20 66.8 0.0 17.0 34.0 80.8 I..m.-.~ Sbtll 10.1 (Hen II C9923-1) 1.1 51.0 1.73 1.152 5 58.7 0.0 17.0 34.0 97.7 I.m>-.r Sbtll 10.2 (Heat # C9923-2) 1.1 51.0 1.73 1.152 4 58.7 0.0 17.0 34.0 96.7 Inlmnrdim Shell I..cmg. Weld ~ BC &BO 68.0 0.582 0.849 -49 57.7 0.0 28.0 56.0 64.7 1.1 (Htatll 4P-l784) lhing amibls ;ru,wilkmce dalll'"11 21 423 0.582 0.849 -49 35.9
- 0.0 14.0 28.0 14.9 Inlrnmdwt tol..ow,r SMII Circ. WddSeamAB 1.1 68.0 1.73 1.151 -49 783 0.0 28.0 56.0 853 (Heat# 4P4784)
U;;,,g ,;;;;;j,le SIDWillance dafa'II 2.1 423 1.73 1.151 -49 48.7 0.0 14.0 28.0
-- 27.7 I..awu Shell I.a,g. Weld Stams BA & BB 1.1 68.0 0.592 0.853 -49 58.0 0.0 28.0 56.0 65.0 (Heat I! 4P4784)
Using cndibb, nnwi&nce datn'fl 21 ill 0.592 0.853 -49 36.1 0.0 14.0 28.0 15.1 Enmdtd Btldint llitn i.ili Nozzle Shell 12-1 (Heat # C9955-2) 1.1 90.1 0.0767 0366 9 33.0 0.0 16.5 33.0 74.9 Nozzle Sbell 12-2 (He.a II C0123-2) 1.1 826 0.0767 0366 15 302 0.0 15.1 30.2 75.4 Nome to lntmnl'dute Shell Circ. Weld Seam 1.1 68.0 0.0814 03TI -49 25.6 0.0 12.8 25.6 23 AC (Heu ii 4P4789 lhmg cndible Sll1'Wi11tmce daJD'll 2.1 42.3 0.0814 03TI -49 15..9 0.0 8.0 15..9 -17.1 N~ Shell U11!2. Weld Seams BE 211d BF 1.1 820 0.0767 0366 10 30.0 0.0 15.0 30.0 70.0 Nolos CClllbimd "" ~ pi WCAP-18728-NP August2022 Rel.isi.on 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 37 of 61 P"ll'!37of81 We.rtingbouse Non-Pcoprietu y Class 3 6-7 Ncn;:
<*> u.. Rqubhilywdo 1.99, R.\-isim 2 ~ ,ns ulilizadmtmcakubli... oftbeARTnlm. .
(b) Chmislzy fad.,., an takm mm T,blt 3-3.
(c) Thlmce bbs, £mm Table 6-1 of tl,is npcrt.
(cl) fF* flam,ofadar =fl"'" -*-<lll.
(o) R.T,.,,.,lll (UmndialodJlT,.,,.) ,-.Jues tum 6- T>b!o 3-1.
2.1 wilh ll<IIHndibl,o S i l l " \ ~ ma, and ibe base mm! GA=
(f) For Ibo piidazn of~1Dry Gmd,,, 1.99, Rr.ision 2, the base ,ml>! <JA
- l rf £<< Po.ilian 1.1 and Posilioa 1.1 and Posilioa 2.1 witlu*n-<ndible 8.S"F Ear Position 2.1 will, cndio!e ~ dm. Also, per Ra:,,lmy Gmd. 1.99, Rr.ision 2, tbe mid metal a,* 2S"F fer Positim &nitbl!r basemetils or -id,., ,rim surnilh:a<,, ma, :mil tbe lOOdlllllal a*
- 14"1' £or Pasilim 2.1 nh cndto!e S'\l1\""1lmce dab. B'oln!s,,,-, a DeedDDt - - 1 0.5*AR.T,.,T ar 1"ithmt ""'1!illazice dab.
VCSNS Ucif 1 s= ~ elm fur tbe la!mmdim Shell 11-1 (Rut ii (v ll,e cndibilityr.-.luli a:, htbt VCSNS Unit 1 sun...U-.. elm in Appendix A deJamined thattbe A91S4-l) is dNmod DOIKffl!l>!a and the S..,.'Oillai,ce Weld (Hut# 4Jl.l7S4) is dtemd<ffdi' August2022 WCAP-1872S-NP Re.ision3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 38 of 61 Page38of6I
\\'estinghonse Non-Propridaty Cl= 3 Tabk6-5 Cakulation of 1hr \"CS!'IS Unit l ART ,*aln** for thr Rrador YHwl Extrndrd B,lclinr ~oul* Mnt,riab at tbr End of PEO (56 EFPY)(Il R.G. Muima.m PrNfirttd 1.99, CJ1') n uttmi Mn RI1'M'(t) ~TllDY GI cu M ART Marr rial Red (x 10" n/cnr, Ill" ("F) I*) ('F) ("I)(' ("I) ("F)
("I)
Position E>LOMe\')(<l l.aletNo:z:zle436B--l (Hnt/#Q2Q41W) 1.1 92.1 0.0239 0.192 -20 17.7 0.0 8.8 17.7 lS.4 Inlet Nomr 4368--2 (}lttt# Q2Q39W) 1.1 93.0 0.0239 0.192 0 17.8 0.0 8.9 17.8 35.7 IDlrl r.om e 4368--3 ltlt.1111 Q2Q39W) 1.1 93.0 0.0239 0.192 -20 17.8 0.0 89 17.8 lS.7 OU!letr,;om,, 437B-l ltie.11# 1,l~ J 1.1 93.0 0.0102 0.111 -10 103 0.0 52 103 10.6 Outlet Nozzle 437B-2 (Heat ii Q~ W) 1.1 93.0 0.0102 0.111 -10 103 0.0 52 103 10.6 Outltt No:z:zle 437B-3 (Hnl ll Q1Q44W) 1.1 92.6 0.0102 0.111 0 103 0.0 5.1 103 20.5 w!dlClutld Nozzle Fmgiugs to N02Zle ~ Weld 1.1 82.0 0.0239 0.192 10 lS.7 0.0 19 15.7 41.5 Semis l5A/B/C& l6A/BIC Nol.s:
(>) Tho L su]m,,y Guido 1.99, Rnision 2 mo!hodology,.... utilized mibo ~ 0 1 1 alll>t AR.Tn!no,..
(b) Om,;st,y fadm, .,. bbD mm T>h!e :n (c) Fhma l>kmlhm T*bl* 6-1 ald:is npart (d) fl'= lhe,ce fado< = .*-d-tfll.
{e) IU',...,..,,, (Ucrradi;md RI...,.) n!nes bkmmm Table '.i-1.
(f) Per tbo pida,,ca oCRquhlay Guida 1.99, Rms;m2, tl:a bz.e mml <>4= 17"F rc.-Pmm"" l .l a Posilim 2.l m!h DOI><ndiba ~ dzb, mi ibo base - 1 ru=
Posilim 2.1 ~-ilh DOCH:ft<hl,Je B.5"1' mr Posi!ico 2.1 l:rilh cncib!e smnill-. dm..Also, p e r ~ Guide 1.99, IIR<-is:ioa 2, i1>o -id am.I a,.=2S"F fir Pomiau. 1.1 ml .,,..,-uis, wuh
~ mta, a,d Ibo -i.lmml "* = 14'F lot PO<ition 2.1 ..-:oh uodihlt ~ clala. llawe\w, a*need ootsceed 0.5*AR:r,.,,. b--1>.- mot.ls
.,..Ti!lx,,t ~ .i.i..
August 2022 WCAP-1872&-NP Ra'ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 39 of 61 1'3go 39of81 Weslingbo~ Noo.-Propririaty Class 3 6-9 Table 6-6 Cal(ubtioo of tbt VCSXS Un.it I ART Val ts a t tbt l/4T Location for th* Rra(tor Y~sfl Btltlin* and h t,odtd B,Jtlin*
Materials at tht End of Sl'EO (72 EFP\ )<->
1 R.G. 1/.ff flltfll(f Prrdicted 1/.fT a~ M ART Mattrlal l .99, Rt,*.2 CFO> (x 101' nlcml, E>LOAir\')(.,
('F)
{°I) C°I) IO ("F) {°I)
Position Bf!dmt llittrlals Intmnodiate :.!lffl 11-1 (Httt #A9154-l) 1.1 65.0 :;.68 1.427 21 92.7 0.0 17.0 34.0 147.7 Um,g non-cmiibl, sun"tillanc. dati# 2.1 43.5 5.68 1.427 21 62.l 0.0 17.0 34.0 117.1 Snai 11-2 tHUl # A9153-2) 1.1 58.0 5.68 1.427 -20 82.7 0.0 17.0 34.0 96.7 Lower Shtll 10-1 (Httt II C99l3-1) 1.1 51.0 5.69 1.427 5 72.8 0.0 17.0 34.0 111.8 um-..- Shell 10-2 (Rm I C9923-2) 1.1 51.0 5.69 1.427 4 72.8 0.0 17.0 34_0 110.8 I n l ~ Shell L1RJ!- Wtld Swns BC &: BD 68.0 1.90 1.176 -49 so.o 0.0 28.0 56.0 87.0 (Hm li 4P4784) 1.1 t&tng aimbt. ~ datiftl 2.1 423 1.90 1.176 -49 49.7 0.0 14.0 28.0 28.7 tn:moNWe to Lcm;er =-~ Circ. w ad seam AH 1.1 68.0 5.68 1.427 -49 97.0 0.0 28 .0 56.0 lOtO (Heat 41'47&4) 603
- 0.0 14.0 28.0 393 lGing a-edible sunYilltmos daJdtl 2.1 423 5.68 1.427 -49 Lower Shell Lang. Weld Stnm BA & BB 68.0 1.93 1.180 -49 803 0.0 28.0 56.0 873 1.1 (Htat f 4P4184) lGing mJi/,/,: sun-.lilmla datifl/ 2.1 423 1.93 1.180 -49 49.9 0.0 14.0 28.0
--28.9 Extmded Beltbt Matu:uls Nome~ 12-1 (Heat i C99S5-2) 1.1 90.1 0.251 0.625 9 563 0.0 17.0 34.0 993 Nozzle Shell 12-2 (Hui Ii C0123-2) 1.1 82.6 0.251 0.625 15 51.6 0.0 17.0 34.0 100.6 NOZ21e to ~ ShellCirc. Wl'ld SeamAC 1.1 68.0 0267 0.640 -49 43.6 0.0 21.8 43.6 38.I (Httt II 4P47_!9 "tbi,,g a-t:tfibl6 ninYillana dttf(lt} 2.1 423
-0267
- 0.640 -49 27.1 0.0 13.5 27.1 52 Nozz.lr Sbdl l.mg. W~ Sftlm BE md BF I .I 82.0 0.251 0.625 10 513 0.0 25.6 513 1125 Ncm ~ c m mllol.-Eg pa."1!.
WCAP-1872&-NP August 2022 Rew.ion 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 40 of 61 Page40ol61 Westinghouse Non-P.-oprietary Class 3 6-10 NDIH:
(a) The ~ Guido 1.99, Ra\IDl:m 2 mtlbodolo,r""' olllmd. mthe akolatioa.oftheART nlua (b) Cbmiisby 6.c:im, ..-. tlbn mm Ttbl. 3-3.
(c) n-:. md FlneDoo FildlD bbn lioon T>blot 6-2 of this npcrt (d) fFz Ouo,,a,ad,ar a H-O'°"'lil'll.
(e) RT,_...., (Ummclimd R.T....,.) n.'ms tabn th>m Tab?. 3-l.
base mo!a! G.\=8.51' (I) l'lr lbt gmdmce cfRep,lm,ryQiide 1.99, R.oruioo 2, the base ..,.b} Ill* 17°F fir Po,itian l.l mdl'osilm2.l "'ithDOD-<ndib!ur<1ilm ce dab, mdfho
£co-Position 2.1 ,.;;i,anb!a ~ elm. Also, por R,p,b!myGuid.t 1.99, R.-~ 2, the ..-.!dlDl!bl GA a 2.&"F f<< Pomon l.l mdPosilion 2.1 m lhl!Oll-<ndible ~
elm, am the ,...?,I ..-1 G.\ = 147 far Posmaa 2.1 will> ~11I* ~ e b b . ~ . . . _ need DOt eia:aed. O.S*ARTmT r~ oitba- base mmls or ~
,rid, er "-itbout
""""1lm:,e ebb-(!;) The m!di.luty orumion iortbe VCSNS Umt l ""'-eillm,e elm in Appa,d:iJ; A dotmninod tlm the VCSNS Umt l ~ d3!a f<<the Imm,,,di,n Sholl l l-1 (&at 1!,A9l54--
l) is <lem,odnm-cndibl,, ar.d tbe Surm1m,co Weld {Hui I 4P47&4) is doomed~
WCAP-18728-NP Angust2022 R£\-ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 41 of 61 l'3ge41 ol81 Westi!lgbouse Non-Propriebsy Ct.us 3 6-11 Tab!, 6-7 C:tlnilation of th* YCS:XS Unit l ART \ *aluts at th* 3/4T Location for th* Rn<tor Ytsstl BelrliM and Enmdtd Beltline Matuials at tb End of SPEO (72 En'Y)<->
RG.
llitnial 1.99, Cfl>I 3/.ff lhlmr*
(x loJJn/cm 1, 3/-IT RT~ (t)
Pttdicttd 4R.Tsr 01 .... :u ART Ru.? E > 1.0 M,\)('l FFt4 (°F)l-l
("F)
("F) ~ ("F) ("F)
Position Bdtli.a* !bttrbls IDla:mtdia:e ~ 11-1 (Hm # A9l54-1) 1.1 65.0 224 1.218 21 792 ~ 11_0 34.0 1342 Umw non-a"'1ibh SUJ'\Qlkmc. daiJi 2.1 43.5 2.24 1.218 21 53.0 0.0 17.0 34.0 108.0 Jn!mnediate SbdJ 11-2 (Hm # A9153-2) 1.1 58.0 2.24 1.218 -20 70.7 0.0 17.0 34.0 84.7 Lowa Shell 10-1 (Rm# C9923-1) 1.1 51.0 2..25 1.219 5 622 0.0 17.0 34.0 101.2 Loin*.- Slid) 10-2 (Rm# C9923-2) 1.1 51.0 2.2S 1.219 4 622 0.0 17.0 34.0 1002 JntrrrnrdimSbelll.mg. Weld Sums BC&BD 1.1 68.0 0.751 0.920 -49 62.5 0.0 28.0 56.0 69.5 (Heat ii 41'4784)
Usmg crftb1'16 svrmllanee daJdll 2.1 42.3 0.751 0.920 -49 38.9 0.0 14.0 28.0 17.9
~ lo l.cmw Shell Circ. Weld~ AB 224 1.218 -49 82.9 0.0 28.0 56.0 89.9 1.1 68.0 Qhat # 4P4 784)
Usmg cmt,1'16 SJm¥illm,<< daJ.,p) 2.1 42.3 2.24 1.218 -49 51.5 0.0 14.0 28.0 30.5 Lower Sbdl l.mg. Weld Seams BA & BB 1.1 68.0 0.763 0.924 -49 62.8 0.0 28.0 .56.0 69.8 (Hm# 4P_4784) r rm.., cri,dibl* no-mllana, daftllJ 2.1 42.3 0.763 0.924 -49 39.1 0.0 14.0 28.0 18.1 utrDllfd Bd!luu ~ tuim Nozzle Shell 12-1 (Heat # C99SS-2) 1.1 90.1 0.0991 0.415 9 37.4 0.0 17.0 34.0 80.4 NozzkShdl 12-2(Heat# C0123-2) 1.1 82.6 0.0991 0.415 JS 343 0.0 17.0 34.0 833 Nc,zm to Jnlmnedia SbeJI Circ. Wdd ~AC 1.1 68.0 0.105 0.427 -49 29.1 0.0 14.5 29.1 9.1 (lkat # 4P4784)
UsiM ,mlibl* mrmllm,ce daJJf/J 2.1 42.3 0.105 0.427 -49 IS.I 0.0 9.0 18.1 -fiii N~Sholllm2.. Wdd Smm BE ml BF 1.1 82.0 0.0991 0.415 10 34.0 0.0 17.0 34.0 78.1 Noh:, ca::t>imd cm lilllo,rmg pap,.
WCAP-18728-NP Angust 2022 Rmsioo 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 42 of 61 Pq 42 ol 61
\\'fflinghouse Non-Proprietruy Class 3 6-12 Not.:5:
(*) The R.p,l;,!myGuim l..99, R.-'im> 2 molbodolou ,nslllilmdm lhe ~ oftba ART nlms.
(b) Cbomislry fadms an tum frtm Ti t,! . 3-3.
(<) flimrt blom ha T*lu 6-2 of dm ,_t.
(d) fFs llm,,co fimr= P"- *.,.,.dll_
(e) RI""'" (Urii,ndir,,d Rimr) ~ lam ha Tob!, 3-1.
(f) P..-tbopidm<e of:R.p,bbyGwdt l..99, Rmsioo2,1habosemeblm=l7"F lc< PO<iticm I.I ml l'osi6an2.I -.rilhDOD-<nd,l>!esg\-.illmotc!m,atha b....a.bl aa=SSF
£..-Fmitian 2.1 wiihcnclil,J, =-.illmeeclm. Also,pu R...,.blay Guide 1.99,~-isian 2, lho,nld moalaAs28"F r<< Posilian 1.l amP~2.1 ..-ith~l>le .....-.iJbnc,,
c!m, md the ,nJd mob! a. s 14"F for Pcsiiim. 2.1 wrilh cmil,le sun-oilJ.mct dm. ~ ,,. no.d DOt aa,ed O.S*MIT,...,y fa: oibo- base mmls or "Rids, with<< 'ilboal mn-.lbaco elm.
(v The cn<hl,i]jtyn-.!mliao hthe VCSNS Uflit I surmn_,., dabio..A_,di,tA dmmimedlhat the VCSNS Unit l ~ elm for the bnnmdim Sbtll 11-1 (Ent ll A9154-
- 1) is deemed ~ ml the ~ Weld(Hut 4P47S4) is doomed cndil>!e.
\\ CAP-18728-NP Augus12022 Re\-ision 3
Serial No.22-227 Docket No. 50-395 Enclosure 1: Page 43 of 61 F'3Qe 43 of 61
'\\'estulgbouse Non-Proprietary Class 3 6-13 Tablr 6--8 CakuL,tion of th* YCS~S Unit I ARI" \.alu~ for th* Rrtttor Y~w l En,ndrd Bdtlin* ~ozzle Maltriah at the End of SPEO (72 En-'}">
R.G. Mllimum RTllllr(q Pfflliclfll CJ1 1.99, Flut11<< Max 1\1 .-\RT 1\latnial R,,*.2 Cfl'l (x 101' n/cm 1, fT("I ("F) (oJ .un:
- l T'F fai- Positicm 1.1 md Positiai, 2.1 ,m1, mixndib!a .....-.illa,,a elm, md Ibo base mml .,_..
- 2. U.S. Nnclear Regulatory Commiwon. Office of Nuclear Regulatory Research, Regulatory Gui.de 1.190. "Cakulafiona1 aru1 Dosimmy Methods for Detennining PresSlllC ,csset New-on Fluence," March 2001. [A.gcncy,dde Documents Access and Management S}'Stem (ADAMS) Accession Number MI.010890301]
- 3. Code ofFedenil Regulations, 10 CFR 50, Appendix G, ~Fracture Toughness Requirements," U.S.
- 4. Westinghouse Repon WCAP-1812~ NP-A Revision 0, °Fluence Determination with RAPTOR-M3G and FERRET," July 2018.
- 5. Westinghouse Report WCAP-18709-NP, ~'ision 1, "V.C. Summer Nuclear Station Unit l SlJbsequeot license Renewal: Reactor Pressure Vessel Extended Beltline Neutton Exposure Emuation." F ebmaiy 2022.
- 6. Code ofFederal Regulations, 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Presmrized Theuna1 Shod: faents," U.S. Nuclear R.egulntory ~ '> ion, Federal Register, November 29, 2019.
- 7. Westinghouse PWROO Report PWROG-21037-NP, Re11ision 1, "Determination ofUnilradiated RT:sDT :ind Uppet-ShelfEnergy Values of the V.C. Summer Ullit 1 Reactor Vesrel Materials," May 2022.
- 8. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatoiy Re.s:ean:h, Regulatory Guide 1.99, Revision 2, '-'Radiation .Embrittlemem of Reactor Vessel :Materials," May 1988.
- 9. WC$tinghouse PWROG Report PWROG-15109-NP-A, Revision 0, '-PWRPressure Vessel Nozzle Appendi:t G Evaluation," JannaIY 2020. [ADAMS Accession Number ML20024E573]
- 10. Westinghouse Report WCAP-1629&--NP, Revision 0, "Analysis of Capsule Z from the South Carolirui Electric & Gas Company V. C. Smnmer Reactor Vessel Radiation Snr."eillanre Progr.un,"
- 12. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 of LWR Editi= Branch Technical Position 5-3, '-"Fracture Toughness Requirements," Re1;isfon 4, U.S. 11."uclear Regulatory- Commis-sion, :March WU>. [ADk\lS Accession Num ber ML18338A516]
- 13. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Di\1-ision 1, "Fracture Toughness Criteria foe Protection Against Failure.~
- 14. Wb-tinghouse Report WCAP-16305-NP, R.En.ision 0, "'v. C. Summer Heatup and CooldownLimit Curves for Nonna! Operation,,,,. August 2004.
- 15. AS1M E185-82, "Standard PractiC'e for Conducting Surveillance Tests fOT light-Water Cooled Nuclear Power Reactor Vessels,= American Sociely for Testing and Materia1s, 1982.
- 16. Code of Federal Regulations, 10 CFR 50, .Appendi:t H, "Reactor Vessel Material Sutveillance ProgramR.equirem.en:ts,"U.S. Nuclear Regulatoiy Commi.ssion, Federal Register, October 2, 2020.
- 17. NUREG-1801, Revision 2, "Generic Aging Lessons Leamed (GAIL) Report,." u_s_ Nncle.ar Regnlatoiy Commission, December 2010_ [ADAMS Accession Number MI.1034900.fl]
- 18. NUREG-2191, Volume 2, 'ueneric Aging Lessons Learned ror Subsequent license Renewal (GALL-SLR) Report." U.S. Nuclear Regulatory Commission, Inly 2017. [ADA..JJS Accession Number .MI.17187.t1204]
- 19. \Vb-tinghous-e Rfport WCAP-9214, =South Carolina Electric and Gas Company Virgil C. Summer Nuclear Plant Unit No. 1 Reactor Ves..ooel Radiation Smveillance Program," January 1978.
- 20. ASTM E185-73, "Standard Ref-nrorneoded Practice for Snn~lance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials, 1973.
- 21. Virgil C. Summer 1'.'uclear Station Unit 1 Updated Final Safety Analysis Report(UFSAR), M.ay 31, 2018.
- Intennediate Shell 11-1 (Heat#A9154-1)
- Inteanediate Shell 11-2 (Heat #A9153-2)
- Lower Shell 10-1 (Heat # C9923-2)
- I.owerShell 10-2 (Heat# C9923-l)
- Lower Shell Long. Weld Seams BA & BB (Heat # 4P4784, Flux Type Linde 124, Lot#
- 55 of 61 Westinghouse Non-Proprietary Class 3 A-4 Table A-2 Cakofation oflnterbn Chemistry Factors for tht Credibility E nluatio Using YCSNS Unit I Sornillance Capsole Data Only Capsule Measured FJum.ceC-> Ff(b) FPART:rmr FF1 Ahttriru Capsule (x 101f n/cm1, ARTNDi<*> ("F)
- 6RTNDt) + 1:(IF) = (663.4) + (15.261) = 43.5"F u 0.675 0.890 '12.1 20.2 0.792 V 154 1.119 47.0 52.6 1.253 X 2.51 1247 '12.1 283 1.555 Smveillance Weld w 4.63 1387 43.5 603 1.924 z 6.53 1.451 65.2 94.6 2.106 SUM: 256.1 7.630 CFS=.l!.old =1:(FF
- t\RTwr) + UFF2) =(256.1) +O.630)= 33.6°F Note:.:
- 59 of 61 V.'e$tinghouse Non-Propriemy Class 3 B-1 APPENDIXB El\'IERGENCY RESPONSE GUIDELINES The Emergency R.esp-Onse Guideline (ERG) limits were developed to establish guidance fot operator action in the event of an emergency situation, such as a PIS event (Reference B-1). Generic categories of limits were de\>-eloped for the guidelines based on the limiting inside surface RTNDr. These generic categories were conservatively generated for the Westinghouse Owners Group (WOO) to be applicable to all Westinghouse plants.
Subject:
V.C_ Swrtmer Nuclear Station Unit 1 Sllhsequent lie~ ~wal: Recommended Changes to the Surreillmce CaJ>f,u le Withdrawal Schedule Attachment A Recommended V.C_ Summer Nudea:r Station Unit 1 Sunreillaru:e Capsule Withdrawal Schedule This letter transmits the changes recommended to the V.C. Summer Nucleai- Station Umt 1 capsule withdrawal sclrMnle. Attachment A provides. technical justification and demonstrates scliedule compli:mce ,vith the applicable vetsion(s) of American Society for Testing and Materials (ASTM) E185. Note that these recommended schednles support the Vugil C. Summer Nuclear Station Unit 1 proposed 80-year license.
Do not h esitate to contact the uodersigned with any questions regarding the contents of this letter.
The putp<>se of.Rfiision 1 w as to incOll)O.rate Dominion* s comments on Re,'isioo 0.
The purpose ofRe-..ision 2 is to update the schedule consistent with the remsertion of Capsule Yin the F all of 2027.
The rei.ision also corrects all error- in the stated EFPY for EOC 26. Thls correction does extend the last opportunity to reinsert Capsule Y by one cycle; howei.-er, the fPCOOlrnendation for the reinsertion in the Fall of20l7 is still maintained in ot"der to provide extra tim.e fur analysis and licensing.
The putp0se of Revision 3 is to coaect a typographical error. Cha:oges are marked with revision bars.
EI..ECTRONICAll.Y APPROVEIY. Verified: ELECTRONICAILY APPROVED1 Tyler c_ Ziegler & D_Brett L}iich Donald M Mcl'."'utt ill RV/CV Design & .Analy3is RV/CV De~gn & Analysis Approved: ELECTRONICALl...Y APPROVED1 L}'Illl ,A_ Patterson, Manager RV/CV Design & Analysis 1
Eleetro11icaJl.f awrm-rrl ucords are autl1e11ticated in tlie electronic tlocw11e11t ma11ageme11t system.
©2022 W ~ Elec:tric Company llC All Rights R.esaved.
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 2 of 14 Page 2 of 14 Westinghouse Non-Propri~taty Class 3 CGE-RVOO0-TM-ME-000004, Rev. 3 Attachment A Page 1 of 12 Attachment A Recommended V.C. Summer Nuclear Station Unit 1 Surveillance Capsule Withdrawal Schedule
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 3 of 14 Page3of 14 Westinghouse Non-Proprietary Class 3 CGE-RVOO0-TM-ME-000004, Rev. 3 Attachment A Page 2 of 12 This attachment provides the recomme11ded capsule withdrawal :.chedule for V.C. SummH Nu clear Station (VCSNS) Unit 1. This includes demonstration of the schedule's compliance with ASTM E185-82 [Ref 1],
as prescribed by 10 CFR 50, Appendix H (Ref. 2] with consideration ofNUREG-1801 (GALL) [Ref 11]
andNUREG-219 1 (GAIL-SLR) (Ref. 3].
10 CFR 50, Appendix H [Ref 2] states:
The design of the surveillance program and the wit/1d1*a1*,al schedule must meet tl1e requirements of the edition ofAS'IM E 185 that is current 011 the issue date of the ASME Code to which the reactor W!sscl was purchased; f or reactor vessels p urchased ofter 1982, tire design of lite su.rveilltmce program and the withdrawal schedule must meet the requirements ofASTM E 185-82.
For reactor vessels purchased in or befo re 1982, later editions ofASTM E 185 may be used, but including onzy those editions through 1982. For each capsule withdrawat the test procedures and reporting requirements must meet the requirements ofASTM E 185 to the extent practicable for tire configuration ofthe specimens in the capsule.
The reactor vessel was designed and constmcted to ASME Section III, 1971 Edition pee FSAR. Table 5.2-1
[Ref 4). Tons, per 10 CFR 50, AppendiK H, the sun.-ei.Uance program withdrawal schedule may meet the requirements of any venion of the AS1M E185 s1andard from the 1970 vasion (the vasion which was etment on the issue date of the ASME Codes to which the reactor vessel was purchased) through the 1982 version.
Per WCAP-9234-NP [Ref 5], the surveillance capsule pcogram was designed to AS1M E1 85-73 [Ref 9],
which was the veision active at that time the program was developed. Therefore, the requirements of 10 CFR. 50, Appendix H were met at the time of the design of the reactor vessel surveillance program.
ASTM El85-82 Schedule Recommendations To d:r.te, VCSNS Uni1 l has tested 5 of6 capsules initially inserted. Capsule Y is the only m:uafoinguntested capsule. The capsule-S tested satisfy the requirements under the 60-year license, as discussed later in the letter, therefore, this letter addcesses the schedule changes necessary to support 80 years of operation.
Table 1 of this letter contams the recommended surveillance capsule withdrawal schedule for 11pdating the FSAR Section 5.4.3.6.2.3 in support of 80 years of operation. The schedule satisfies the regnlatory reqnirements of the Reactor Surveillance Program for 80 years, and meets ASTM El85-82 [Ref 1] as required by 10 CFR 50, Appendix H [Ref. 2].
PerNUREG-2191 [Ref 3],one capsule should be withdrawn at anoutage \\"itha neutron fluence ofbetween one Mld two times the peak reactor v-essel w all neutron flnence at the end of the SPEO and tes1ed in accordance with the requirements of ASTM E185-82 [Ref. 1]. Assuming an end ofSPEO of 72 EFPY; the endofSPEO fluence = 9.06 x 1019 o/cm.2 {which includes a 10% bias on the peripheral and re-entrant comer 20 2 assembly relati\-e powen) and two times the end of SPEO tlnence of 1.73 x 10 o/cm. (2 x 19 8.64 x 10 o/cni2, which conse.rvatively e.-tcludes the 10% unbiased). In order to optimize the value of this S1JITeillance capsule, it is recommended that a potential 100-year operating period also be conndered.
Extrapolating from the 66 EFPY and 72 EFPY fluence results from WCAP-18709-NP yields a projected 90 EFPY p eak reactor vessel fluence of 1.14 x 102° o/c:ni2. Capsule Y was pulled at 17.71 EFPY with a fluence of7.0l x 1019 o/cm2
- Therefore, this capsule cunently fails to meet the GALL-SLR. requirements.
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 4 of 14 Page 4 of 14 Westinghouse N on-Proprietacy Class 3 CGE-RVOO0-TM-}.£000004, R ev . 3 Attachment A Page3ofl2 Capsule Y should be reinserted into a 17° octagonally symmetric location (the highest lead factor locations:
107'>, 28 7", or 343°) to be irradiated further. Using the unblased data in Table 4, Capmle Y should be exposed to a minimum of 4.9 EFPY of operation in order to experience the end of SPEO fluence, but it should be withdrawn before 22.4 EFPY to remain below 2 x the 80-y~ar flueace. Howenr, it is recommended that Capsule Y be exposed to~ roioinmm of 10.5 EFPY of operation to experience the peak 100-year iluence ofl.14 x l oN n!cm.2*
To prevent fhe RV .from experiencing a fluence greater than the a\.-ailable test da1a (6.53 x 1019 nlcni base-d on the exposure of Capmle Z), Capsule Y should be reinserted 105 EFPY prior to the RV reaching 9 1 6.53 x 1lr n/cm , which is projected to occur at 52.4 EFPY per Table 2-2 of WCAP-1 8728-NP (Ref 6].
HoweYer, this does not acc-OWif fur ~ psule resting. Currently 10 CFR. 50, Appendix H requires that results of capsule testing be provided to the NR.C staff no later than 18 m onths (1.5 yem ) after the capsule is withoom-n.. 'lbe:refore, it is RCommended that Capsule Y be reinserted prior to 40 EFPY (:::: 52.4 EFPY -
10.5 EFPY - 1.5 EFPY). Based ~ C)"Cle 26 (the la,t fully completed cycle) being canplcted in the Fall of2021 at 3238 EFPY, there is 7.62 EFPY remaining. D epending on capacity factor, this is approximately 8 to 8.5 years. This u,jll pnt 40 EFPY at the F all of2029 or Spring of 2030. Theref ore. Refueling Outage 31 (Spring of 2029) will provide 1he final opportunity to reinsert the capsule. It is recommended to reinsert Capsul.c Yin.Refueling Outage 30 (Fall of2027) in order to proi.i de extra time for analysis and licensing.
Assuming that Capsule Y is reinserted during Refueling Outage 30 in the Fall of2027, prior to Cycle 31 0 and an average fuel cycle length of 1.33 EFPY/cycle, i.e. - 90"/4 capacity factor, Capsule Y \liill need 8 cycles to achiere an additional 10.5 EFPY of.imldiation (105 EFPY} / (1.33 EFPY/cycle), which m eans the remo\'al of Capsule Y following Cycle 38 (Fall of 2039). Sim::e the plant has experienced 3238 EFPY at EOC 26, Capsule Y will be rejmerted at 37.7 EFPY andiemo'l.*ed at 48.4 EFPY.
These dates a.re only intended to be an em.mate. All commifment3 should be based 0 11 t1t"ithdcawing Capsule Y at the outage n~arest to, but follo-wing, the capsule fluence achieving the 100-year flue-nee, which is estimated to be after 10.5 EFPY after Capsnle Y is re-inserted into the 107° location {or symmetric locations 287° or 343°).
It is noted that the cap5Dle fluence should be used to detennine when the capsule is withdrav.u, and the EFPY is an approximation based on the unbiased capsule projections. The capsule should be withdrawn at the outage nearest to, but following, when the capsule fluence is met.
E,aluation of Compliance with *.\STM I:185-82 The first step in determining the surveillance capmle withdrawal schedule compliance with ASTM El85-82 is to de1emiiae the minimum number of caprules to withdraw and/or rest. ASTM El85-82 bases the number of capsules on the ma.'rinmm ARTwT projected at the ,.--essel surface for all reactor vessel materials. Per Table 2 of this letter, the maximum AR.TNDT value at the end ofSPEO is 102.0°F. Sinre the maximnm ARTwy values are projecled to be abm..-e 100°F, but below 200"F fonr capsules are required to be pulled 0
per Table 1 ofASTM El85-82. To date, VCSNS Unit l bas *withdrawn and tested 5 of the capsules. Beamse u
AS1ME185-82 based on plant operati011 during the original 40-year license term. the requi:remems are supplmien1cd 1l3ing NUREG-1801 {GAll.) and NUREG-2191 (GALL-SLR) [Ref. 3].
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 5 of 14 Page 5 of 14 Westinghouse N on-Proprietaiy Class 3 CGE-RVOOO-TM-ME-000004, Rev. 3 Attachment A Page 4 of12 The next step is to demonstrate that fhe condition of the capsules rested, or scheduled to be tested, meet fhe requirements of ASTM E 185-82. This comparison is performed in Table 3. This tible demonstrates that the recommended schedule mee.ts the recommendations of ASTM E185-82 as reqwred by 10 CFR. 50, Appendix H and the guidance in the GALL and GALL-SLR..
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 6 of 14 Page6of 14 Westinghouse Non-Proprietary Class 3 CGE-RVOO0-lM-ME-000004 Re\*. 3 Attachment A Page5of12 Rl'fnence.s
- 1. ASTME185-82, "Sta:ndardPracticeforCon.ducting Snn'eill:!:nceTestsforlight-Watec Coo1edNuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.
- 2. Code ofF~ Regulations 10 CFR. 50. Appendix IL "Reactor Vess.el M aterial S1J11.--e:illa.nce Program Requirements" U.S. Nuclear Regulatoty Commission, Fede.rai Register, October 2, 2020.
- 3. U.S. Nuclear Regulabny Co~on, "Generic Aging Lessons I.earned for Subsequent License Renewal (GALL-SLR) Report," NUREG-2191, Vol 2, Inly 2017. [Age>>C}wide Documents A ccess and Management System (ADA...\lS) Accession Number MLl 7187A204]
- 4. Virgil C. Summer NucleM Station Unit 1 Updated Final Safety Analysis Report, R.e.,"ision 20.01, 1n1y2021,
- 5. Westinghouse Report, WCAP-9234, Revision 0, "South Carolin.a Electric and Gas Company Virgil C.
Summer Nuclear Plant Unit No. l Reactor Vessel Radiation Sun--e.illance Program," Januuy 1978.
- 6. Westinghouse Report, WCAP-1 8728-NP, Revision 3, "V.C. Summer Nuclear Station Unit 1 Subseqllellt License Renewal: Evaluation of"Rnctor Vessel Integrity Time-limited Aging Analyses,"
August 2022.
- 7. PWROG Report, PWR.OG-21037-NP, Revision 1, Determination of Uninadiated R.Twr and Upper-Shelf Energy Values of the V.C. Summer Unit 1 Reactor Vessel Materials," May 2022.
- 8. Presmrized Water R.ea~or Owners Group (PWROG) Report. PWR.OG-15109-NP-A. Re\iision 0, "PWR. Pressure Vessel Nozzle .Appendix G Evaluation."' January 2020. (,ll)A...\ £S Accession N umber
!--1I.20024EJ73]
- 9. ASTM E1 85-73, "'Standard Recommended Pmctice for Surveillance Tests for Nuclear Reactor Vessels." American Society for Tes1ing and Materials, 1973.
- 10. U.S. Noclear R.egulatoty Commissi~ Office of Nudear R.egulatoiy Research., Regulatory Guide 1.99, R.e\,"ision 2. '-"Radiation Embrittlement of Reactor Vessel Materials,.. fay 1988. [ADAMS Ace -ssion Numbe.r MI.00374028,fJ
- 11. NUREG-1801, R.el..-ision 2, ~eric Aging Lessons Leamed (GALL) Report" U.S. Nuclear R.egulaiory Commission, December 2010. [ADAMS Accession Number MI.103490041J
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 7 of 14 Page 7 of 14 Westinghouse N on-Proprietary Class 3 CGE-RVOO0-TM-ME-000004, Rev. 3 A ttachment A Page6of12 T ab le I Recommended Sun*eillance Capsule \Vithdr:nral Schedule<*>
Location Capsule Lead Remonl TuneOO Cafsule Flue ce Capmle (deg) Factor (EFPY) (n/cm , E>l.O Me\) (1,)
u 343 3.04 l.13(EOC 1) 6.75 X 1011 V 107 3.34 2.93 (EOC 3) 1.54 X 1()111 X 287 3.54 5.04(EOC5) 2.51 X 1019 w 110 3.21 11.21 (EOC 10) 4.63 X 10111 z 340 3.10 1636 (EOC 14) 6.53 X 1<>1!1 y(c) 290 3.09 17.71 (EOC 15) 7.0} X 10111 101<*> - 3.5 48.2(<) 1.14 X 1()20 Notesfor Table l:
(a) E&'edm! fiill po11,w year. from plant startup. End of Cycle (E(X;) ,.me P'\"1!11 i n ~- Note tb.t C'On!
thmnal pov.wwas uprated m>m m s to 2900 MWa. stutmg with operating cycle 10.
(b) Valoes ill!! taken from WCAP-lS-TIS-NP {Rei 6] Table 2-l (c) Capsule Y will be reinserted clmmg lW'u.eling Oubge 30 in the Fill of2fl1.7 (prior to Cycle 31) in location 107" (orS)'Dlllll!'lric~ons 28T or343j, which is project~to occurat 37.7 EFPY. Capsule Y mllachir..-.! thepeak 100-year ~ . 1.14 x l 010 n!an2 (E > l .0 ~leV), befon remoi.-al, miicb is ~ to n,qain? :mother 10.5 EFPY of operation (37.7 + 10.5 = 48.2). The Ill!m!St ou12ge to the ~ remw.11 of 482 EFPY will occur at EOC 38 after 48.4 EfPY, which is the project~ mma..d time of Capsule Y.
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 8 of 14 Page 8of 14 Westinghouse N on-Proprietary Class 3 CGE--RVOO0-TM-ME-000004, Rev. 3 Attachment A Page7of12 TablPl ARTNDT C alculations fo r thl' Y.C. Summl'r Unit I RPaetor Yessel at 72 EFPY Surlael' PndictPd RG l.99 Wt. *,t, Wt. % CFII>> Flume~> Ff{,I)
RHrtor Vessel Ma tnial ARTm~>
Position Cu<*> Ni<> ("F) (x IOlt nlrm.1,
("F)
E>LOMPV)
Beltti.ne l'tlaterials fulel:mcdiate Shell 11-1 1.1 0.10 0.51 65.0 9.04 1.501 975
@at l#_A9!?+1) ___
Using non-aedib/a SU1Willm1ce ------2.1-- ---- ---------- ........... .......... ... .. ..... .. .. ... .*-*-*-*.... - - -- -- -
datdfJ - - 435 9.04 1.501 65.3 Intennediate Sbell 11-2
<Heat# A9153-2) u 0.09 0.45 58.0 9.04 1.501 87.0 Lower Shell 10-1 1.1 0.08 0.41 51.0 9.06 1.501 76.6 tHeat # C9923-D Lower Shell 10-2 fHPM# C9923-2) 1.1 0.08 0.41 RO 9.06 1.501 76.6 Intermediate Shell Long.
Weld Seams BC & BD 1.1 0.05 0.91 68.0 3.03 1.293 87.9
@ at# 4P4784)
.Usini*cret1tbb;;;;;:Ji1nn"ce "Jtiia'.fJ . 2.1 -------
- ***423 "" 3.03 1.293 54.7 Intenned.iate to Lower Shell Circ.
Weld Seam AB 1.1 0.05 0.91 68.0 9.04 1.501 102.0
...... @~t~~f:1-J~ L .... ... .I Usinll credible SILT'l'eillance dnlrf!J - * --*----
2.1 *--------* * * * * ** * * *
--*423 ** ******1ro.r***-- ***uoc* **** 63_5**--
l ower Shell Long.
Weld Seams BA & BB 1.1 0.05 0.91 68.0 3.08 1.297 882
.....*... .Qieat# 4P4784L ......*. . ------ -* -* -------
Usin2 aediblB sun*eillmtce dnlrf.lJ 2.1 - - 423 3.08 1.297 54.9 Extended & ltlin.t> Mattrials Nozzle Shell 12-1 1.1 0.13 0.57 90.1 0.400 0.746 612 tHeat# C9955-2)
Nozzle Shell 12-2 1.1 0.12 0.58 826 0.400 0.746 61.6 fHeat # C0123-2)
Nozzle to Intmnediate Shell Circ.
WeldSeam AC 1.1 0.05 0.91 68.0 0.425 0.762 51.8 (Ht~!Jt ~f~}S§)
Using credible Sllr'fflllance dalrfll 2.1 --------
- ***423 ** ******o.425 ***-- *--0_16i ** ***. 32.2***
- Nozzle Shell Long. 1.1 0.06 1.01 820 0.400 0.746 61.2 Weld Seams BE and BF Inlet Nozzle 436B-1 1.1 0.127 0.76 92.1 0.0310 0.224 20.6
<HP.at # 0 20.tlWI Inlet Nozzle 436B-2 1.1 0.127 0.82 93.0 0.0310 0.224 20.8
<Heat# Q2Q39Wl Inlet Nozzle 436B-3 1.1 0.127C!> 0.82 93.0 0.0310 0.224 20.8
<Heat# O2O39Wl Outlet Nozzle 437B-1 1.1 0.127C!> 0.85 93.0 0.0132 0.132 123 ffieat # 02040)
Outlet Nozzle 437B-2 0.127{!)
rHPat #1-,,.-,...--,w, 1.1 0.80 93.0 0.0132 0.132 123 Outlet Nozzle 437B-3 (Heat# Q2Q44W) 1.1 0.12100 0.78 92.6 0.0132 0.132 12.2
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 9 of 14 Page 9of 14 Westinghouse Non-Proprietary Class 3 C-GE-RVOOO-TM-ME-000004, Rev. 3 Attachment A Page8 of 12 InlctlOutlct Nozzle Forgings to Nozzle Shell 1.1 0.06 1.01 82.0 0.031 0.224 18.4 Weld Seams 15AIBJC & 16AIB/C Notes for Table 2:
(a) Chemimy 1,alt11!s were tlbn from t\ CAP-18728-NP [Re£ 6], UDless ocherniseDOted.
(b) Regulatmy Guidi! (R.G) 1.99 [Re£ IO], Position 1.1 Oenisby Fart= (CFs) .Ill! based on Tublei l and 2 of Regulattxy Guidi! 1.99, Rerisian 2 using thl! Cu and N i ~ peta!llt , -ahies. Regubtmy Guide (RG) 1.99, Position 21 CFs an c:alomted based on anilab e sur.-eillaoce dab in WCAP-18728-NP [Re£ 6].
(c) FJn,,,,a, bbn man \JlCAP-187>-8-NP pw: 6] and include a 10"/4 bias on the periphery.
(d) FF= fl:umce f.ctor = f111.2&- cno-q (I))_
(e) ARI= = CF :;;. FF per R.G 1.99.
(f) Th! cndilnlity e..iluation fbr 1he VCSNS Unit I sun~ data in.Appem!ix A ofWCAP-18728-NP ~ that tru-VCSNS Unit I son.-ei:llma data fix- tru- Inle:nMdiate Shell ll-1 (Heat # A9154-l ) is deaned non-credible and the S ~ W21.d (Heat Ii 4P4784) is deemed Cil!dio le.
(g) ~ , .tlue £or SA-508 Class 2 name £,xgiDgs &om PWR.OG-15109-NP-A [Re£ S].
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 10 of 14 We.--Jingllome Non-Proprim,y Class 3 CGE-RVOOO-nl-ME-000004, Rev. 3 Altadmem A Pagr9ofl2 T ahlt 3 Compari>oa of dit Y.C. S n t.'air l Cap,lllt Uitlulr:m~ Sd1tdlllt ffidi ASill EIS.~, dit G.U.L, nd da* GAlL-SLR.
R,q-.in111ttl for Capo \Titlulmnl S<htdlllt S"'Zltmt*lal Gllid.uct i! pu ASill £18-"-S? Tahlt I l<I from All ud G.ll.L-SI.R CoDU11tJ1uO'l 3 EFPY or al thr time when the ~ omtrOll ~ U was 1ri!bdl:n-n al EOC I, "'i lh c ""P""'R of 1.13 EFPY md 1 1hmn of the capsule =<eds .S " 10" 1J!ari, ar at ~ -6.7 5x 1011 .,1cn,l.
I thr time ooim ibe ~pr,di<t,d &rr,,w of all N!A
~ 11111mm is ~ j()'f, AS'IME185-112is m r t ~ the czpsnk ,ns 'WillldJN,n wiJ:h a tm.nc.
- hicli<\..,-.,.,.,.,_ fint_ pal,r fhm j X 1011 ll'a!r.
Capowo V wn 'lritlwlln-n al EOC3, wilh m "'P"S'ft of 2.93 EFl'Y and 1 f!amte = J.54x lO" n'r:ml. ThepH): l t4T &.noeatEOL (.....aed lobe 36 ElJ'Y, 4..42 X I0" 1>'c nr)ll 2.Jt X 10" 1>'c:m2. (lj 6 EFPY ar at the time i1t.m ibe acamiab.~ Dflllnm s mce c apml,. V flumc,n mnch ltss 11w, 278 x 10'* n1cm1, Caps,,kx flomce oftheap,ule c ~ to 1hr will be <oD>id<rld. ClplU!e X W111 i1iihdn,m at EOC .S, wilh m l!llpC>Sln 2 NIA
- pprulIIIII!< mkf. U - (EOL) llnonce at 1hr of.S.04 EFPY m:I I lluma, of251 x 10" D!c:m2.
ruttor ,*Cl$Cl l/4T loc:alica, wlli<hr= m-. fir ASlM EIS5-ll isme! . thr lm!,no, iJ \~ - thr lt-iT llnosic, md the 6EFPY amia. ha addition, at the time of wilhdmnl of llli capsw.
the EFPY un Jibly bo<ed m diffued flumu ~ and posm>ly a cliffi!nm EFPY tttm..
Cqm1f W in.s wilhmnua1EOC 10, *ilh m apo=e oflUI EFPY 15 EFPY or al ibt timt wh,ri theattlm!lll!ltd w h Jlumce --4.63 x 10'1 1>'mr. The clad~ m,o}flumce at EOL 3 l:N1llll llxD::e of the ~ com,pon,:b to the (36Ffl'Y} WH"Jlll'~H2 x 10'1n/cm1.
~ EOL ~ al 1llf md<< ..,.,:d imfr WA
..-.n locali,on, ~ <<:mOS lim. AS1ME185--ll2 is mrt ~ the c..,ie fluma,.,.. pater Ihm the t!uair:t of the cbd/ba\e mtbL up,uleZ wu1'ilbfhwDII EOC 14, 1'illl ~ of1636EFPY md * ~ = 6.53 x lO" nlmi. ~ cbd-1,ue mttal flnentt at 1hr fDd Pe1 lbt GALL, me capsule is of PEO (56 EFPY)\QS waJimr,dy 7.00 x 10 nlcml. Ho-,.fl; lbt 11 EOL arnot las ihlll COC1' or n,all'f Ihm micf sch<dw* imp!,mmlfd.h liameRlle'i<II md app=ei by the NRC pw EOL'--' ~ . Thismaybe mocUtd ca -.rill>:ll:n11 md ~ at a fhlaxe of 4 -lfz lhmo,n or gutt<> lh1ll *" -iJl:in !he Safety E-.-walicn P,q,cn (SEJ!) fur 1hr 6().J~u lic<me :rmeITTI 1hr bosis of[,R"ious lnts. This apmlf may~ bold wu based on a fiutu:e of 6.-10 x 10" Dlcml a.t 54 EFPY. The "'JlO'UI""
wilhom liS!ing follcming wilh<hm-al. twice the pw e::d of PEO \"eZd
!h,c,a. m, also ....n ,.ilhm lbt 20'/4 unamintyrequirrd by Rr!"l1lo<YGuide LllJO.
ASlM El8~ md OAIJ.. mi<l>In are a:mida-td met GAIL-SLR. one capsule is mtblmm Addi6.oml data is n,:p,ir,d IO meet thu aitmaa; 1llmmre, Cap<Ul,, Y mi tes'.td at a Jln!!lce ofnot 1... lh,u s NIA
<m<e or p na: 6w, twic.~the pw should be ~ md iindwrd as :mv.-n in Tobit I.
<lld ofSPEOve=ltlutix:e.
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 11 of 14 Pago 11 of 14 Westinghouse Non-Proprietaiy Cl&U 3 CGE-RVOOO.TM-ME-000004, Rev. 3 Attac:hmentA P1ge 10 of 12 No<.; bDJ,'s:l:
(I) lbo ap,u-nbo.<J-d.mm,b,,s, mdEfPY= ' - ... bbn&.m T.ble 4 oflhis loaw.
(b) 11io 1!4T -.nlM was ~ - tho ,.,bc,e ll,m,oe, tbt n12<tor nssel behliae ~ (l.7S inc:!-,), md ~ £= t.r
- __.,."' mm RG 1.99,:Rmsiool, - . , x= ll:a doplh
. tho ns,ol ,nll (m:u:).
(o) VCSNSmom tho ~ oC.4 cap,,L'-mthdmnl:d,ad,:!t za danoedi:i ASI'Ml!l8~sincetl,e limilmg = =-ispull!rlhao. lOO"F md loss d= 200"F.
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 12 of 14 Page 12 of 14 Westinghouse N on-Proprietary Class 3 CGE-RVOOO-TM-ME-000004, Rev. 3 A ttachment A Page 11 of12 T able 4 SuIYrilllince C apsule Proj r rted Floeate (n/cm1, E >1.0 Mel) for V.C. Summrr Unit 1<->
Cumubiin R o,n.cr (n/cm2)
Cydt Cydl' Lfngth Operating (EFPY) 17" 20° Timr (EFPY) 1 1.13 1.13 6.75E+tg(bl 5.90E+18 2 0.67 1.80 l.01E+19 8.94E+l8 3 1.13 2-93 l.54E+t!){<l 136E+19 4 1.16 4.09 2.03E+19 1.80E+19 5 0.95 5.04 2.51E+1 9<,I) 224E+19 6 1.17 6.21 3.13E+l9 2.79E+l9 7 1.22 7.43 3.63E+l9 3.24E+l9 8 1.19 8.61 4.13E+19 3.69E+19 9 1.27 9.89 4.66E+19 4.15E+19 10 1-32 11.21 5.18E+19 4.63E+t g(*l 11 1.36 12.56 5.78E+19 5.15E+19 12 1.37 13.94 6.35E+19 5.66E+19 13 1.09 15.03 6.76E+19 6.02E+19 14 1.33 16.36 7.34E+19 6.53E+t9© 15 1.35 17.71 7.88E+l9 7.0lE+tg(!)
16 1.34 19.05 8.42E+19 7.49E+l9 17 1.38 20.43 9.00E+19 8.01E+19 18 1.30 21.73 9.52E+19 8.49E+l9 19 1.31 23.04 l.01E+20 8.97E+l9 20 1.36 24.41 l.06E+20 9.46E+19 21 1.29 25.70 l.11E+20 9.93E+19 22 1.29 26.99 1.17E+20 1.04E+20 23 1.34 28.33 l.23E+20 l.09E+20 24 1.32 29.65 1.28E+20 1.14E+20 25 1.36 31.01 l.34E+20 1.19E+20 26 1.37 32.38 1.39E+20 124E+20 27(/1) 1.37 33.75 l.45E+20 1.30E+20 No bias on t!ie p eripheral a11d re-e:nlnmt comer assembly relo.tive powers Future@ - 36.00 l.55E+20 138E+20 Future - 42.00 1.80E+20 1.61E+20 Future@ - 48.00 2.05E+20 l.83E+20 Future© - 54.00 2.30E+20 2.06E+20 Future© - 60.00 2.55E+20 228E+20 Future© - 66.00 2.80E+20 2.51E+20 Future - 72.00 3.05E+20 2.73E+20
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 13 of 14 Page 13 of 14 Westinghome N on-Propriet31Y Class 3 CGE-RV000-1M-ME-000004, Rev. 3 Attachment A Page 12 of12 Cumufatin Flutnce (n/C"m2)
Cydt u ngth Cyd t Operating (EFPY) 17° 20" Time (EFPY) 10% bias on the peripheral and re-entrant corner assembly relatil'e powers Future© - 36.00 1.56E+20 139E+20 Future(i) - 42.00 1.83E+20 1.64E+20 Future© - 48.00 2.11E+20 1.88E+20 Future© - 54.00 2.38E+20 2.1 3E+20 Future(i) - 60.00 2.66E+20 23 8E+20 Future© - 66.00 2.93E+20 2.63E+20 Future(i) - 72.00 3.21E+20 2.87E+20 Notes foc T b e 4:
(a) Infonmtion bl.en from WCAP-18728-NP [Re£ 6].
(b) This v.uue is applic.ble to Capmle U.
{c) This .alm, is applicable to Capmle V.
(d) This r.iiue is applic.ble to Capmle X (e) This r.!lue is applicable to Capsule W.
(f) This r.!lue is applic.ble to Capsule Z.
(g) This nlue is applicable to Capmle Y.
(h) Cycle 27 was &e currat openting cycle at the time the lhlence analysis was perl'mmed.
(i) Values beyuod Cycle 27 are based on the .nuage an power d:istnlrotims and rudoc ~
CODditions of cycles 25, 26, and 27 and are de_1Emi:Ded bo1h lrilh and mfbout a 1.1 bias on 1he periphel'.lll and re--entr.mt <<1rI1B .wembly i:ehtm powers.
Serial No.22-227 Docket No. 50-395 Enclosure 2: Page 14 of 14 Pag@ 14 of 14 CGE-ffVOOC)-TIHlf-GOOOO( Rt tlOn :S N 011-f>ropltabry ~ 3 ancl 511311 not b f ~ l'I Ille P* numbeMg Df U- doclzmnl. **
Approval Information Author Approval Lynch Donald Sep-09-2022 13:41 :22 Author Approval Ziegler Tyler Sep-12-2022 08:49:48 Verifier Approval McNutt Don Sep-12-2022 09:15:01 Manager Approval Patterson Lynn Sep-12-2022 09:27:50